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Contents


11.09.2017 16:15 Keynote Michel Giot

Invited lectures - 101

Quo Vadis Europa?

Michel Giot

SCK.CEN, Av. Herrmann Debrouxlaan 40, 1160 Brussels, Belgium

michel.giot@uclouvain.be

 

Without well-educated and trained personnel, industrial activities are impossible. This is obviously true for the nuclear industry where a large set of skills, competences and attitudes are needed in all fields of technology, not only nuclear. However, training of specialists in nuclear sciences and engineering is of outmost importance. This is why, during the last decades, in Europe, considerable efforts have been made by the higher education institutions supported by the states, the European Union and the industry to offer young students updated curricula internationally recognized, embedded in high quality research environment. These efforts took place in a difficult context that will be described in the first part of this paper, a context whose lessons need to be considered for the present and the future.
Since this difficult context is linked to the reluctance of a significant fraction of European Society to recognize the benefits of fission nuclear energy in the de-carbonated energy mix, the second part of this paper is devoted to the necessary dialog about sustainability with the general public and the public authorities. The analysis includes all three aspects of sustainability of nuclear energy, namely ecological, economical and social, as well as some key issues for gaining in trustworthiness. Communication about nuclear research appears to be one of these key issues, a challenge for all of us.






12.09.2017 08:30 Invited Uwe Stoll

Invited lectures - 103

Nuclear Safety Research: a GRS Perspective

Uwe Stoll

Gesellschaft für Anlagen- und Reaktorsicherheit gGmbH, Boltzmannstr. 14, 85748 Garching bei München , Germany

uwe.stoll@grs.de

 

Major reactor accidents such in Japan 2011 tend to shift reactor safety programmes toward low-probability, high-consequence scenarios and rare phenomena associated with core melt. One reason for this is that closing knowledge gaps in severe accident behaviour and adding (mitigative) safety features is widely perceived as an effective approach to increase safety. Although important, this approach has somehow put back the fact that there are also operational effects not sufficiently understood or analysed such as increased corrosion of fuel rod cladding, neutron flux fluctuations in PWRs and fuel assembly bowing. Germany’s central Technical Support Organisation GRS has made sure that in its reactor safety research programme both is addressed appropriately: (severe) accident related topics and operational issues. The presentation addresses the current research status of the aforementioned operational issues and points out further work to be done in these fields. In addition, two examples of research activities concerning accidents are discussed. One is the estimation of radioactive materials releases during severe accidents in the spent fuel pool. The results gained from this analysis can be used e.g. for the Fast Source Term Prognosis tool FaSTPro developed by GRS for decision-making during emergency situations. Another topic described in the presentation is a new method of statistical assessment of loss-of-coolant accidents, answering a recommendation of the German Reactor Safety Commission.






12.09.2017 09:10 Thermal-hydraulics - plenary

Thermal-hydraulics - 220

Development of RCS-Containment Coupled Analysis Model and Evaluation of LBLOCA for APR-1400 NPPs

Jeongyun Kim1, Young Seok Bang2

1Korea Institute of Nuclear Safety , 62 Gwahak-ro, Yuseong-gu, Daejeon, 34142, South Korea

2Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

jykim1025@kins.re.kr

 

Containment is an air-tight building, which contains a nuclear reactor and its pressurizer, reactor coolant pumps, steam generator, and other equipment or piping that might otherwise release fission products to the atmosphere in the event of an accident. In case of LOCA, containment back pressure is an important factor to determine the core behavior during reflood phase and ultimately a performance of ECCS. It is necessarily required to analyze the ECCS performance with a coupled method of RCS and containment for a realistic estimate.
The ECCS performance is influenced by the recovery rate of reactor core water level, which is called the reflood rate. This reflood rate could be explained with the capacity that ECC water extrude saturated/supersaturated steam to containment, and containment pressure is one of major factor to affect the reflood rate. If containment pressure is remained low during reflood phase, specific volume of steam generated from the reactor core increases and friction between steam and structure also increases. In this circumstance, the amount of steam pushed out to containment decreases and the reflood rate also decreases, which means ECCS performance is degraded consequently.
In this study, for the analysis of the ECCS performance and containment pressure behavior, the RCS-containment coupled analysis model is developed and verified using MARS-KS and CONTEMPT4 computation code. With this model, audit calculation and sensitivity analysis of the ECCS performance are conducted for the Shinkori unit 3, 4 NPPs which represent APR-1400 design. CONTEMPT4 is a digital computer code that describes the response of containment systems subjected to postulated LOCA conditions and MARS-KS is a code designed for realistic analysis tools based on best estimate modeling for application in the thermal hydraulic analyses of nuclear reactor systems.
MARS-KS and CONTEMPT4 have been coupled using the method of dynamic-link-library(DLL) technique. Overall configuration of the code system is designed so that MARS-KS will be a main driver program which use CONTEMPT4 as associated routines. With this system, the data for the discharge rate of mass and energy and containment back pressure are exchanged at each time step.
To identify an influence of containment parameters to the ECCS performance, a sensitivity analyses are conducted for containment parameters with the RCS-containment coupled analysis model. Those parameters are selected from the CONTEMP4 input, which are initial temperature, initial pressure, a free volume, a mass flow of spray, surface areas of a passive heat sink and a condensation heat transfer coefficient.
For the containment analysis, assumptions to calculate minimum containment pressure are directly opposite from those for peak containment pressure. Therefore, for the each containment parameter, the sensitivity analysis is conducted for three conditions, which are for peak containment pressure, for minimum containment pressure and a central value from those maximum and minimum value.






12.09.2017 09:30 Thermal-hydraulics - plenary

Thermal-hydraulics - 201

Computational Fluid Dynamics Study of Pressurized Thermal Shock Transients in the Reactor Pressure Vessel

Michal Jaros, Nathan Lafferty, Guian Qian, Bojan Ničeno, Markus Niffenegger

Paul Scherrer Institut Nuclear Energy and Safety Research Department, Reaktorstrasse, CH-5232 Villigen, Switzerland

michal.jaros@psi.ch

 

During a loss-of-coolant accident (LOCA), emergency core cooling system (ECCS) injects water into the hot reactor pressure vessel (RPV) through the cold legs. In combination with high pressure, high stresses in the RPV wall are developed due to high thermal gradients caused by the water flowing over hot vessel walls. The accurate prediction of pressurized thermal shock (PTS) is particularly important in life-management of aged reactors – susceptibility to brittle fracture is higher due to material ageing and increased neutron irradiation.
In this work, a thermal-hydraulic study of PTS conducted with computational fluid dynamics (CFD) simulations in a prototypical RPV is presented. It is a part of an ongoing research project, whose focus is the investigation of probabilistic fracture mechanics. A large break LOCA (LB-LOCA) with a postulated guillotine break in one of the hot legs is analysed. The final purpose of this study is to supply the stress analysis software with temperature field in order to examine stresses resulting from thermal gradients. The influence of pressure in LB-LOCA scenarios is relatively weak, since the RPV depressurizes very fast. Previous results from RELAP5 reveal that during the LB-LOCA flow in the RPV is two-phase. Therefore, simulations are performed with two-phase volume-of-fluid (VOF) approach with sharp interface formulation. Turbulence is modelled with standard k-? model. The transient CFD simulation begins at the start of the emergency core cooling (ECC) injection when virtually all water from the RPV has evaporated, and continues to follow the fill-up with cold water from ECCS.
The transient boundary conditions for CFD simulations are obtained from RELAP5. User-defined functions read mass flows and temperatures at the inlet of ECCS, and insert them in the corresponding planes in the CFD model. Simulations are started with lower plenum filled with water and the rest of the RPV with steam, both at saturation temperature. The water from ECC injection starts to flow in the cold legs and reaches the downcomer, influencing the volume fraction and the temperature field in the RPV. Simulations show that the amount of cold water is initially small and the flow is attached to the RPV wall in the vicinity of the inlet nozzle. No significant thermal gradients occur at this time. When the amount of cold water becomes slightly larger, it flows virtually only behind the neutron shield – attached to the core barrel. Subsequently, the injected water starts to splash on the core barrel wall. Backflows and rapid changes in volume fraction are present. Furthermore, formation of the cold plume in the RPV wall begins. Below the cold legs’ inlet nozzles, significant thermal gradients occur. These regions are particularly susceptible to high thermal stresses. Later in the transient, the cold water flows in the whole volume of the downcomer. Moreover, ECC injections from both loops begin to interfere and multidirectional movement of the cold plume affects the temperature field in the whole RPV wall – significant thermal gradients appear below the hot legs. Finally, there is a significant increase in water level in the vessel, which compares well with precursory RELAP5 results.
Through the CFD simulation details of the two-phase flow are obtained. Three-dimensional results confirm inadequacy of one-dimensional RELAP5 calculations to predict thermal loads caused by ECC injection during LB-LOCA scenario, and show the crucial importance of CFD for simulation of the PTS phenomena.






12.09.2017 09:50 Thermal-hydraulics - plenary

Thermal-hydraulics - 202

An Euler-Euler Multiphysics Solver for the Analysis of the Helium Bubbling System in the MSFR

Eric Cervi, Stefano Lorenzi, Antonio Cammi, Lelio Luzzi

Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

eric.cervi@polimi.it

 

The Molten Salt Fast Reactor (MSFR) developed in the framework of the H2020 SAMOFAR Project (http://samofar.eu/) is a circulating fuel nuclear reactor in which a mixture of molten thorium and uranium fluorides acts as fuel and coolant simultaneously. From a computational point of view, the simulation of nuclear reactor dynamics is a complex task, needing accurate solution for both neutronics and thermal-hydraulics and considering the coupling between them. This is even more important in circulating fuel nuclear reactors, in which the velocity field of the fuel salt mixture has a significant impact on the distribution of the precursors, affecting the reactor kinetics. In addition, a bubbling system is envisaged in the MSFR to efficiently remove the gaseous fission products in the salt. The gas bubble injection can be also envisaged as a possible option for the reactivity control of the MSFR, exploiting the void reactivity feedback of the air bubbles in the fuel mixture. This design choice rises new challenging issues that must be addressed carefully in order to evaluate the feasibility of this option. Currently, there is a lack in the literature about the effect of the bubbling systems on the neutronics and the reactor dynamics as well. As for the fluid dynamics, the salt mixture compressibility can have a relevant impact on the dynamics of the system, especially in fast transients in which the fuel temperature and density fields are not in thermodynamic equilibrium. In this regard, compressibility of gas bubbles may also induce delays in the reactivity feedback mechanisms.
In the light of the previous considerations, in this work we present a multiphysics OpenFOAM solver for the MSFR dynamics, which is aimed at addressing the above issues. A multi-group diffusive model is adopted for the neutronics, using constant-group cross sections calculated in Montecarlo simulations. The problem of bubbling is treated by considering the fuel salt mixture and the air as two continuous and impenetrable fluids, each described by a phase fraction and by averaged conservation laws (Euler-Euler approach). Both the molten salt and the gas bubbles are treated as compressible fluids, to evaluate the effect of compressibility on the reactor transients. In addition, the transport equations of precursors in the liquid phase are implemented.
The solver is tested by means of a simplified model of the MSFR. Different power transients are also simulated, highlighting the contribution of the bubbling system to the core reactivity. The void reactivity feedback coefficient is evaluated on the basis of the average void fraction and is compared to Montecarlo simulations. The effect of the real bubble distribution on the core reactivity is assessed, pointing out the differences with respect to simulations carried out with uniform void fractions. The outcomes of this analysis constitute the starting point for further research on the MSFR dynamics and transient analysis, with a particular focus on the design of the reactivity control systems and the optimization of the reactor as well.






12.09.2017 10:10 Poster session - RED

Thermal-hydraulics - 206

New model for the shape of a critical liquid wave in vertical churn flow

Matej Tekavčič1, Boštjan Končar1, Ivo Kljenak2

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 Ljubljana, Slovenia

matej.tekavcic@ijs.si

 

In vertical churn flow of gas and liquid in a cylindrical pipe, one of characteristic phenomena observed are periodic liquid waves of large amplitude travelling upwards along the wall. The mechanisms of such flow can be associated with the onset of flooding phenomena or counter-current flow limitation encountered during a hypothetical loss-of-coolant accident in pressurized water nuclear reactors. The prediction of the onset of flooding is still very uncertain even for the simplest cases and a thorough understanding of the triggering mechanisms is required.
In representative experiments from the literature, liquid wave that develops near the porous wall liquid inlet travels downwards at first due to the force of gravity. As the wave grows in its amplitude, the resulting reduction in the area for the gas flow increases the upward pressure force. Eventually, the wave reaches its critical size, where the two opposing forces are balanced and the wave is almost stationary and approximately symmetrical for a brief moment. Subsequently such critical wave reverses its course and is carried upward by the gas flow.
In the present paper, we propose a new model for the shape of such critical liquid wave in vertical churn flow regime of gas and liquid. The proposed model is based on the Lorentz function and offers better agreement with measured wave shapes than simpler hemispherical or sinusoidal based models. At the same time, the accuracy of the new model is comparable to the recently proposed Gaussian model, which is analytically more complex.






12.09.2017 10:10 Poster session - RED

Thermal-hydraulics - 209

Numerical investigation of the eddy-viscosity models ability to predict heat transfer within the dry storage for nuclear waste

Remache Amel1, Yacine Addad2

1Université des sciences et de la technologie d'Oran MB, El Mnaouar, BP 1505, Bir El Djir 31000, Oran, Algeria

2Khalifa University of Science Technology and Research, PO.Box 127788, Abu Dhabi, United Arab Emirates

remmache.ammel1990@gmail.com

 

The purpose of this work is to investigate numerically the eddy-viscosity models ability to predict heat transfer within the dry storage for nuclear waste. The commercial computational fluid (CFD) code, Star-CCM+ is used and the effect of the turbulence model is studied to determine the temperature distribution in the VSC-17 system. The turbulence models tested are the low-Re k-? model, the elliptic blendingk-? model, and the algebraic turbulent heat fluxeddy-viscositymodel (AFM). The first two mentioned models adopt the standard isotropic eddy-diffusivity type model(also known as the simple gradient diffusionhypothesis, SGDH) to model the turbulent heat flux. I.e., in those models the turbulent heat flux is modelled as , with the turbulent Prandtl number set to a constant value, = 0.9. On the other hand, the algebraic turbulent heat flux model (AFM) uses a truncated from of second-moment representation of the heat flux vector. The Numerical predictions are compared with experimental data of Mckinnon et al. to assess the models ability in mimicking the correct flow dynamics.In addition, it is known that both convective and radiation heat transfer modes are equally important in thestorage cask passive cooling process, hence the modeling of these two dominant modesis assessed in the current study.

Keywords: Star-CCM+, The VSC-17, algebraic heat flux model.






12.09.2017 10:10 Poster session - RED

Thermal-hydraulics - 210

Direct Numerical Simulations of Thermal Fluctuations in a Flow Over a Backward Facing Step With Solid Walls

Jure Oder, Iztok Tiselj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

jure.oder@ijs.si

 

In this paper we present the direct numerical simulations of thermal fluctuations in walls and in turbulent flow of liquid metal flowing past a backward-facing step (BFS) with finite dimensions and solid walls. The BFS geometry can be visualised as a channel where one of the walls has a shape of a step. The flow is lowing from the narrower part to the wider part. The simulations are performed in three dimensions.
The temperature field is a passive scalar in our simulations. This means that the temperature differences in the flow do not influence the flow. With this approximation, natural convection cannot be simulated.
For the inflow boundary condition over the BFS, a fully developed turbulent velocity field with constant temperature is used. To obtain this inflow, a recycling boundary condition is used. We take the values for velocity and temperature from a plane parallel and downstream from the inflow and we impose them as the inflow boundary condition. The strem wise component of velocity is scaled at this operation to ensure a constant mass flow rate.
Simulations are performed with the NEK5000 code. The most notable feature of this code is the use of spectral elements to solve for velocity, temperature and any other passive scalar. It is an open source code developed by the Argonne National Laboratory.
Spectral element method is a hybrid method between finite element method and a collocation spectral method. The method divides the computational domain into finite elements, within which a spectral method is used to solve for variables. This method allows for the use of spectral method in irregularly shaped geometries and to perform direct numerical simulations in such geometries.
This work is part of work that is performed within the SESAME project of Horizon2020 research programme and is a continuation of research at our department.






12.09.2017 10:10 Poster session - RED

Thermal-hydraulics - 211

Investigation of Partial Coolant Flow Blockage in a Sodium Fast Reactor Assembly with Coarse-mesh Methodologies

Stefan Radman1, Carlo Fiorina2, Andreas Pautz1

1Swiss Federal Institute of Technology (EPFL), Station 3, Lausanne, Switzerland

2Politecnico di Milano - department of energy, Via La Masa 34, 20156 Milano, Italy

stefan.radman@epfl.ch

 

An open source multi-physics platform for the steady state and transient analysis of nuclear reactors is currently being developed at the Laboratory for Reactor Physics and Systems Behaviour at the EPFL. With regard to the thermal-hydraulics, the application of coarse-mesh methodologies based on a porous medium approach is envisioned to have a number of advantages. First, it provides substantial benefits in terms of reduction of the computational load of the simulations compared to standard CFD approaches. Secondly, it provides some 3-D modelling capabilities when compared to 1-D legacy codes. The ultimate goal is the treatment of both single and multiphase flows with said methodologies, though only single-phase capabilities are currently implemented.
Sodium Fast Reactors are selected in this work as an interesting application of coarse-mesh methodologies due to the significant 3-D effects that are expected to arise from the high thermal conductivity of the coolant. In particular, the case under investigation consists in the partial coolant flow blockage in a Sodium Fast Reactor assembly at various reactor power and flow levels. The system is modeled as a bundle of 7 assemblies, namely a central assembly subject to partial flow blockage and unblocked surrounding assemblies. The assemblies are separated by an inter-assembly gap.
The impact of blockage is assessed for various reactor power and flow levels based on fuel and coolant temperatures, and on the onset of boiling. The effect of coolant flow in the inter-assembly gap is investigated parametrically as a function of unblocked assembly coolant flow velocity.






12.09.2017 10:10 Poster session - RED

Thermal-hydraulics - 212

Quantification of the discontinuity of the temperature variance dissipation rate at a fluid-solid interface: wall-resolved Large Eddy Simulation of turbulent channel flow with conjugate heat transfer.

Cedric Flageul1, Sofiane Benhamadouche2, Iztok Tiselj3, Martin Ferrand4

1Institut Jožef Stefan, Jamova cesta 39, 1000 Ljubljana, Slovenia

2EDF R&D, Fluid Mechanics, Energy and Environment Dept., 6 Quai Wattier, 78401 Chatou, France

3Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

4Electricite de France, Research and Development Division, Avenue les Renardieres, Ecuelles, 77818 Moret sur Loing Cedex, France

cedric.flageul@ijs.si

 

Conjugate heat transfer represents the actual thermal coupling between a fluid and a solid part. It is of prime importance in nuclear industrial applications where fluctuating thermal stresses are a concern, e.g. in case of a severe emergency cooling (Pressurized Thermal Shock) or long-term ageing of materials such as thermal striping occurring in T-junctions. For such complex applications, numerical investigations often rely on Reynolds Averaged Navier Stokes (RANS) or wall-modelled Large Eddy Simulation (LES) approaches. The present article deals with the improvement of refined RANS approaches for conjugate heat transfer.
RANS models for conjugate heat transfer are relatively recent (Craft et al., Journal of turbulence, Vol. 11, 2010). Using Direct Numerical Simulation (DNS), the authors of the present abstract have recently established that the dissipation rate (??) associated with the temperature variance (?2) is discontinuous at the fluid-solid interface in case of conjugate heat transfer (Flageul et al., International Journal of Heat and Mass Transfer, Vol. 111, 2017). Actually, there is currently no coupled RANS model for conjugate heat transfer taking this discontinuity into account. As a result, from an industrial perspective, LES remains the best option for thermal fatigue prediction but needs refinement at the wall, which makes it very expensive if not unaffordable, at high Reynolds numbers.
Experimental data on conjugate heat transfer are scarce, expensive and difficult to carry out in order to obtain fine quantities such as ??. This makes DNS and wall-resolved LES the only solutions to obtain such data. In this paper, we will assess the ability of wall-resolved Large Eddy Simulation to estimate this discontinuity of ?? on channel flows using Code_Saturne, Électricité de France in-house and open-source CFD software. This is a step forward towards a rich validation database for future RANS models adapted to conjugate heat transfer.






12.09.2017 10:10 Poster session - RED

Thermal-hydraulics - 213

On the capability of URANS modelling of multiple impinging jets

Martin Draksler, Boštjan Končar

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

martin.draksler@ijs.si

 

Though the main features of the jet impingement flow can be reproduced by the time-averaged models, the transient phenomena play a decisive role at the heat transfer prediction on the impingement surface. These have been rather accurately simulated in our previous studies using the Large Eddy Simulation (LES). But this approach requires high amount of computational resources that are hardly affordable, especially in the case of multiple jets. The need for less expensive but still transient turbulence models is therefore obvious.
In this paper, the predictive capability of Unsteady Raynolds Averaged (URANS) modelling approach for simulation of multiple jet impingement is studied. The numerical model will combine the open-source CFD code OpenFOAM and two-equation Eddy-viscosity type of turbulence model. The selection of the specific turbulence model will be based on the a-priory steady-state analysis and grid refinement study. The transient simulation results will be averaged and compared with the experimental data and time-averaged results of a well-resolved LES. The purpose of this study is to analyze and evaluate the potential of URANS approach in terms of accuracy and computational cost. The study will focus on evaluation of second order flow statistics and turbulence budgets.






12.09.2017 10:10 Poster session - RED

Materials, integrity and life management - 302

Investigation of the paramagnetic centers of neutron - irradiated nanocrystalline silicon carbide (3C-SiC) particles

Elchin Huseynov

Institute of Radiation Problems National Academy of Science of Azerbaijan, F. Aghayev 9, AZ 1143 Baku, Azerbaijan

elchin.h@yahoo.com

 

Cubic modification silicon carbide nano particles used at the present experiment, which has 120 m2/g specific surface area (SSA), 18nm particle size and 0.03g/cm3 bulk density (true density is 3.216 g/cm3) (US Research Nanomaterials, Inc., TX, USA). The samples used during experiment were irradiated by neutron flux ( 2x1013 n/cm2s) at full power (250kW) in the channel A1 of TRIGA Mark II light water pool type research reactor in the "Reactor Center" of Institute Jozef Stefan (IJS) in Ljubljana, Slovenia. Neutron irradiation and determination of characteristic neutron flux parameters were conducted according to the methodics known in the literature [1,2].
Paramagnetic centers and their nature in the nanocrystalline silicon carbide (3C-SiC) particles have been investigated comparatively before and after neutron irradiation. The Electron Paramagnetic Resonance (EPR) measurements were performed over the broad range of magnetic field from 0.05 to 0.55 T (500 to 5500 Gauss) in order to detect different paramagnetic centers, which could be presented in the samples and more precisely in the region of 0.3270 to 0.3370 T, i.e. in the region where the most paramagnetic centers appears including free radicals. Neutron irradiation effects on the newly formed concentration of and vacancies has been investigated. The number of paramagnetic centers for different values of g-factor has been calculated appropriate to local cases existed around 3300G. After neutron irradiation creation mechanism of paramagnetic centers in the nanocrystalline silicon carbide has investigated. Effect of neutron flux on the 3C-SiC nanoparticles has been investigated various papers [3-5]. A strong signal has been observed at g = 2.006 in the nanocrystalline 3C-SiC particles. The increase in existed signal intensity and new signals have been observed after neutron radiation. The formation of additional 29Si or 13C isotopes has been observed in the nanocrystalline 3C-SiC particles after neutron irradiation. Simultaneously, formation of anisotropic and isotropic HF structured Si ( ) and C ( ) vacancies after neutron irradiation. Due to the increasing concentration of and vacancies and newly formed isotopes has lead to new paramagnetic centers. Therefore, the total number of paramagnetic centers at different values of g factor increased from 1.5x1020 center/g to 2.7x1020 center/g (approximately twice). Moreover, the number of centers appropriate to free electrons (g = 2.006) two times increase (from 1.03x1018 center/g to 1.9x1018 center/g) after neutron irradiation.

References

1. Gasper Zerovnik, Manca Podvratnik, Luka Snoj, “On normalization of fluxes and reaction rates in MCNP criticality calculations”, Ann. Nucl. Energy 63, 126–128 (2014)
2. Zerovnik, G et al., “Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers”, Applied Radiation and Isotopes, 96, 27-35 (2015)
3. Elchin M. Huseynov "Permittivity-frequency dependencies study of neutron-irradiated nanocrystalline silicon carbide (3C-SiC)" NANO 12, No. 5, 1750068, 2017
4. Elchin M. Huseynov "Investigation of the agglomeration and amorphous transformation effects of neutron irradiation on the nanocrystalline silicon carbide (3C-SiC) using TEM and SEM methods" Physica B: Condensed Matter 510, 99–103, 2017
5. Elchin Huseynov "Neutron irradiation and frequency effects on the electrical conductivity of nanocrystalline silicon carbide (3C-SiC)" Physics Letters A 380/38, 3086-3091, 2016






12.09.2017 10:10 Poster session - RED

Materials, integrity and life management - 303

Medium and Low Voltage Cable Measurements - TD, PD, LIRA

Josip Ceovic, Matko Širola

ELMONT d.o.o., Cesta krških žrtev 135e, 8270 Krško, Slovenia

josip.ceovic@elmont-kk.si

 

Elmont d.o.o. Krško – Following the world trend of LTE (Life Time Extension) in power plants we added new path in our scope of services (beside of electrical maintenance, upgrade of systems, quality control and testing). After we successfully implemented cable testing in Medium Voltage area few years ago, we decided to spread our scope to Low Voltage area also, since there is much more low voltage cables that are very important for main safety systems.
Scope of work – We are identifying potential downgraded conditions for safety and operational important cables in special areas (heat, water, radiation). Our main scope is visual control, and testing with analysis.
For low voltage cables we are using Line Resonance Analysis (LIRA) method. LIRA technology is based on the transmission line theory, through the estimation and analysis of the complex line impedance as a function of the applied signal frequency. We can monitor the global, progressive degradation of the cable insulation due to harsh environment conditions (high temperature, humidity, radiation) and detect local degradation of the insulation material due to mechanical impacts or local abnormal environmental conditions. With LIRA we can detect local degradation with localization error average less than 0.3 % of total cable length.
For medium voltage cables we are using new methods with a power generator that uses Very Low Frequency – 0,1 Hz (VLF). The main reason for this is that the measurement unit needs 500 times less energy than the unit which uses 50 Hz frequency (50/0,1=500). With this power source we are performing dielectric loss measurements – Tan delta (TD) and Partial discharge measurements (PD).
TD measurements show the severity of Water treeing in the measured cable. Water trees mainly come from moisture and are therefore present in cables that lie in manholes filled with water or they submerged in any other way.
PD measurements show the severity of voids or other types of defects in cable insulation. These defects can arise during the manufacturing of the cable or they can arise during the installation of the cable or from an accident with the cable during the operational time.






12.09.2017 10:10 Poster session - RED

Materials, integrity and life management - 304

The effect of hydrogen content on the embrittlement of E110 and E110G alloys

Tamas Novotny, Erzsébet Perezné Feró

MTA Centre for Energy Research, Konkoly Thege ut 29-33, Budapest-1121, Hungary

novotny.tamas@energia.mta.hu

 

During normal operation in light water nuclear reactors, a part of the hydrogen, which is generated by the corrosion processes and the radiolysis, is incorporated into the zirconium fuel cladding. This effect may change the mechanical properties of the cladding.
Steam oxidation measurements at high temperatures have previously indicated if oxidation is followed by significant hydrogen uptake, alloy’s embrittlement is much faster as if the material does not absorb any hydrogen. However, these measurements could only be concluded from the combined effect of oxidation and hydrogen uptake. In order to evaluate the separate effect of hydrogen we aimed to carry out experiments charging hydrogen into the metal alloy without steam oxidation.
Hydrogen charging of the zirconium alloy can be performed by electrolysis or in hydrogen atmosphere at high temperature. The high temperature hydrogenation (mostly above 300 °C) is carried out in a closed system, in the presence of pure hydrogen gas or in inert gas atmosphere with hydrogen.
For the hydrogenation of E110 and E110G cladding samples, the latter method was chosen because the absorption of hydrogen in zirconium at 300 °C is a rather slow process. Our experiments were carried out at 600 °C.
Based on the measurements of E110 and E110G alloys, it is possible to verify whether the cladding is still ductile while containing the maximum permissible hydrogen (400 ppm) specified by the fuel manufacturer. It is also possible to determine the hydrogen content of the ductile-to-brittle transition. To determine the ductile-to-brittle transition of the alloys, mechanical tests are required with known hydrogen-content of cladding samples.
Our experimental work consisted of the following subtasks:
1. Upload E110 and E110G samples (rings) with different amount of hydrogen,
2. Ring compression tests of hydrogen-containing samples,
3. Measuring the substantial hydrogen content of the rings,
4. Determination of the ductile-to-brittle transition of E110 and E110G alloys as a function of the hydrogen content.
The experiments showed the ductile-brittle transition of the E110G cladding is between 3200 ppm and 4200 ppm and for the E110 cladding at 4000 ppm.
In this poster, I summarize and compare the results of the high temperature hydrogenation of the E110 and E110G samples and the subsequent mechanical studies.






12.09.2017 10:10 Poster session - RED

Severe accidents - 412

Analysis of Stratified Steam Explosion Duration Considering Recent SES-S1 Test

Vasilij Centrih, Matjaž Leskovar, Mitja Uršič

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

matjaz.leskovar@ijs.si

 

A steam explosion is an energetic fuel-coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. An important condition for the occurrence of a steam explosion is the initial coarse premixing of the melt and the water. In nuclear reactor safety analyses steam explosions are primarily considered in the melt jet-water pool configuration, where due to the melt jet fragmentation the required premixture is efficiently produced. It was long believed that in the stratified melt-water configurations no sufficient premixture is formed which could produce strong steam explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with corium simulant materials revealed that strong steam explosions may develop spontaneously also in stratified melt-water configuration. To better understand the stratified steam explosion phenomena, especially the characteristics of the premixture formation, the SES-S1 test has been performed within the frame of the EU SAFEST project. Recent investigations brought up the question whether the apparent additional premixture is formed just before the explosion or also during the explosion propagation, and one of the external indicators of the way the premixture is formed is the explosion duration.
In the paper, the stratified steam explosion duration is studied by the analysis of experimental observations and by computer simulations with the MC3D code in the SES-S1 experimental conditions. The explosion duration characteristics are analysed firstly by varying the available melt mass in the premixed layer. A parametric analysis is performed varying some material properties and experimental conditions to see, what influences the explosion duration the most, and to answer the question, why such a different explosion duration is observed for two similar referential material compositions used in the PULiMS and SES tests, WO3/ZrO2 and WO3/Bi2O3.






12.09.2017 10:10 Poster session - RED

Severe accidents - 413

Simulation of local radiation and core degradation effects with the code system AC2 during a severe accident

Liviusz Lovasz1, Sebastian Weber2

1Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Forschungszentrum, Boltzmannstr. 14 , 85748 Garching, Germany

2Gesellschaft für Reaktorsicherheit (GRS), Schwertnergasse 1, 50667 Köln, Germany

liviusz.lovasz@grs.de

 

The accident in Fukushima pointed out the importance of severe accidents simulations. In the future experiments have to be conducted and simulation tools have to be developed, in order to better understand the phenomena during a severe accident and to make better predictions about the evaluation of severe accidents.
The module ATHLET-CD (Analysis of THermal-hydraulics of LEaks and Transients with Core Degradation), which is part of the system code AC2 (ATHLET, ATHLET-CD, COCOSYS) is designed to describe the reactor coolant system thermal-hydraulic response during severe accidents, including core damage progression as well as fission product and aerosol behaviour, to calculate the source term for containment analyses, and to evaluate accident management measures. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development.
Currently, the program divides the core into concentric rings, similar to other severe accident codes. This approach works very well if the accident phenomena are symmetric. But with this approach it is impossible to simulate local effects in the core, like asymmetric power distribution or local cooling shortages, which could have a strong impact on the evaluation of the severe accident.
In order to be able to simulate these local effects the core nodaliziation has to be changed. The change in the core nodalization also means a departure from the cylindrically symmetric configuration, which makes the radiative heat transfer calculations complex.
A new model was developed for the radiative heat transfer calculations. This basically means the determination of view factors, three dimensionally and with taking shadowing effects into account. This new model was implemented in ATHLET-CD, uniquely in this field.
To demonstrate the huge impact of the new nodalization and the new radiation model a hypothetical severe accident scenario with strong local characteristics was simulated and analysed.






12.09.2017 10:10 Poster session - RED

Severe accidents - 414

Simulation of THAI Hydrogen Deflagration Experiments using ASTEC Severe Accident Code

Ivo Kljenak

Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 Ljubljana, Slovenia

ivo.kljenak@ijs.si

 

The issue of hydrogen combustion during a severe accident in a nuclear power plant (NPP) came to prominence after the accident at the Three Mile Island (USA) NPP in 1979, and has received new attention since the accident at the Fukushima Daiichi (Japan) NPP in 2011. The Fukushima accidents also highlighted that both in-depth understanding of severe accident sequences and development or improvement of adequate severe accident management measures are essential in order to further increase the safety of NPPs operated in Europe. The CESAM (Code for European Severe Accident Management) was a research and development project within the 7th Framework Programme of the European Commission that lasted from 2013 to 2017. The specific purpose of the project was the improvement and further development of the European reference code ASTEC towards use in severe accident management analyses for NPPs.
Within assessment of the hydrogen combustion modelling in the ASTEC code, 29 (twenty-nine) experiments on hydrogen combustion, which were performed in the THAI experimental facility, were simulated at the Jozef Stefan Institute. The THAI experimental facility, located at Becker Technologies GmbH in Eschborn (Germany), is basically a single-volume cylindrical vessel, with a volume of 60 m3, an internal height of 9.2 m, and an internal diameter of the main part of 3.2 m. In the considered experiments, the following characteristics of the experimental conditions were varied:
- atmosphere composition (air-hydrogen or air-steam-hydrogen),
- ignition location (upward or downward flame propagation),
- atmosphere structure (homogeneous or stratified atmosphere – both in the sense of composition and in the sense of temperature).
A multi-volume input model of the THAI facility was developed for the ASTEC code, and the experiments were successfully simulated. An overview of the calculated results, and a general assessment of the agreement between experimental and simulation results (pressure and temperature), are presented.






12.09.2017 10:10 Poster session - RED

Severe accidents - 415

FLEX strategy implementation for LOCA sequences in PWR-Westinghouse

Marta Ruiz-Zapatero, Rafael Bocanegra, César Queral

Universidad Politécnica de Madrid (UPM), Nuclear Engineering Department, José Gutiérrez Abascal 2, 28006 Madrid, Spain

marta.rzapatero@upm.es

 

MELCOR is a fast and versatile tool which allows carrying out several applications mainly related to severe accident management. As a result of the Fukushima Dai-ichi nuclear accident, the nuclear industry began an introspective task of questioning its safety via the performance of stress tests. The resulting consequence from these evaluations was the implantation of portable equipment (available from the plant itself and from outside locations) and the development of FLEX strategies in which the already mentioned equipment will be employed; as well as the modification of severe accident management guidelines and procedures. Besides, the reactor cavity has been modified in several Spanish NPPs in order to allow direct injection with portable pumps.
The main purpose of this work is the assessment, by means of MELCOR 2.1 code, of these new and updated severe accident management strategies in the context of a LOCA sequence. There are two different ways of management to be analyzed: in-vessel and ex-vessel actuations. This analysis allows for the evaluation of their impact over the accident progression. The study takes into consideration portable equipment of various characteristics, and a range of actuation times for which the accident sequence is evaluated.
Results enable a review of these newly updated severe accident management actions and reactor cavity modifications in Spanish NPPs, and allow for the analyst to notice and highlight every advantage and possible drawback derived from them.






12.09.2017 10:10 Poster session - RED

Severe accidents - 416

Investigation of hydro accumulators influence on core degradation progression during SBO scenario with ASTECv2.1.1.0

Pavlin Petkov Groudev, Antoaneta Stefanova, Rositsa Gencheva

Institute for Nuclear Research and Nuclear Energy, 72 Tzarigradsko chaussee, Blvd., BG-1784 Sofia, Bulgaria

antoanet@inrne.bas.bg

 

The objective of this paper is to present and discuss the results obtained from performing the calculations with ASTEC computer code for the influence of Hydro Accumulators on core degradation for specific severe accident transient. The calculations have been performed with ASTECv2.1.1.0 computer code.
The purpose of this analysis is to assess the evaluation of ASTEC computer code with modelling of main phenomena arising during hypothetical severe accidents.
The performed analyses cover Station Blackout (SBO) scenario with and without injection of passive safety injection systems (hydro-accumulators). The main target of this study is to assess the influence of hydro-accumulators work on core degradation progression in case of simulation of severe accident scenarios (with and without activation of passive safety system) . The investigation is focused on investigation of in-vessel phenomena arising during the selected scenario such as dryout of reactor core, hydrogen generation, core material degradation, melting and relocation. It was simulate melt pool formation in the core and on reactor vessel bottom. The analyses have been performed until failure of reactor vessel bottom head in both investigated cases of SBO scenarios.
This investigation has been performed in the framework of CESAM project (under the Euratom 7-th framework program) by Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Science (INRNE-BAS).






12.09.2017 10:10 Poster session - RED

Severe accidents - 417

Preliminary Study of the MCCI Phenomena on the APR1400 using MAAP 5.04

Gilbeom Kang, Jaehwan Park, Mi-Ro Seo

POWER, Izpolni naslov!, USA

gilbeom.kang@khnp.co.kr

 

A severe accident at nuclear power plants is an accident that exceeds the design basis accident. It is included accident sequences since a core melt occurs. It is more important to figure out the severe accident phenomena at the nuclear power plant after the Fukushima accident. There are some codes that can simulate the severe accident phenomena; Modular Accident Analysis Program (MAAP), Method for Estimation of Leakages and Consequences of Releases (MELCOR), Accident Source Term Evaluation Code (ASTEC) and etc. Among them, MAAP is known for the fast-running and well-estimation. For this reason, many organizations have utilized the MAAP for their purposes. In this study, the MCCI phenomena are simulated to verify the reaction between a molten corium and concrete materials on Advanced Power Reactor 1400 (APR1400) using the MAAP.






12.09.2017 10:10 Poster session - RED

Research reactors - 506

Experimental Determination and Calculation of the Temperature Reactivity Coefficients for TRIGA MARK II reactor

Romain Henry1, Anže Jazbec2

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

romain.henry@ijs.si

 

Measurement of in core water temperature at JSI TRIGA has allowed the development of methods to evaluate reactivity temperature coefficients. Those quantities are the integral parameter to describe thermal-hydraulic/neutronic coupling. The sign and magnitude of these coefficients are important to know, as they suggest the consequences of sudden changes in the operating parameters of a nuclear reactor. First, experimental protocols to evaluate water and fuel temperature reactivity coefficients are specified. Then the methodology to calculate those coefficients with the neutron transport code TRIPOLI is presented. Finally, experimental and calculated results are properly compared and analysed.






12.09.2017 10:10 Poster session - RED

Research reactors - 507

Linear Stability Analysis of a Full TRIGA Reactor Plant in a Closed Loop

Sara Boarin1, Antonio Cammi2

1Politecnico di Milano - department of energy, Via La Masa 34, 20156 Milano, Italy

2Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

sara.boarin@polimi.it

 

This work investigates the dynamics of the TRIGA-Mark II located on the premises of Laboratorio Energia Nucleare Applicata (LENA) - Universita degli Studi di Pavia; in particular, it focuses on the system stability in a closed loop, with thermal-hydraulics feedback and poison effects on reactivity. The stability analysis is based on the linearized equation system that describes the plant physics, encompassing neutronics, thermal-hydraulics and reactivity feedback. The investigation of the linear approximation may provide relevant conclusions about the system stability; in particular, stability against small perturbations may often be deduced (Ogata, 2002). Laplace Transform is applied to the linearized system to derive the system transfer function at zero power. The transfer function completely represents the system differential equations and its poles and zeros effectively define the system response. The system dynamics is studied with particular attention to the power (normalized neutron density in the reactor core) transient with respect to the control input (control rods position), highlighting the role of poison accumulation.
TRIGA Mark II is a pool-type research reactor, with the core immersed in a demineralised water tank. The water inventory gives a thermal inertia that is a significant contribution to the system stability. Typically, a buoyancy force induces a natural circulation mass flow rate across the core, due to the different water density in the pool: a column of heated water in the core is pushed upwards replacing the cold water at the top of the pool. The nominal power in steady state condition is 250 kW while thermal neutron flux is of the order of 10^13 #n/cm2. This type of reactor has unique features in terms of safety: the specific composition of fuel (Uranium dispersed in a Zirconium-Hydride matrix), gives a significant moderating effect due to the presence of Hydrogen in the ZrH lattice. The result is a strong negative reactivity coefficient that contributes to the intrinsic safety of the plant. On the opposite, moderator has a net positive reactivity coefficient due to non-linear behaviour of the incoherent elastic scattering cross-section of water molecules, over the relevant spectrum. The latter prevails over the water density negative coefficient.
Three control rods (boron carbide and boron graphite) perform active control of the reactivity. During the reactor life, burn-up of the fuel produces fission products that reduce the neutron population due to a high absorption cross section. Control rods can compensate the effects of neutron poisons until they are completely extracted. (_54^135)Xe and (_62^149)Sm have the highest neutron absorption cross section and the highest fission yield; their dynamics is therefore included in this analysis.
The perimeter of this study includes the core and the natural circulation mass flow through it, the reactor pool and the thermal power exchange with the primary cooling loop.






12.09.2017 10:10 Poster session - RED

Research reactors - 508

Comparison of several RANS modeling for the Pavia TRIGA Mark II Research Reactor

Carolina Introini1, Antonio Cammi2, Stefano Lorenzi2, Davide Baroli3, Bernhard Peters3, Davide Chiesa4, Massimiliano Nastasi4, Ezio Previtali5

1Politecnico di Milano - department of energy, Via La Masa 34, 20156 Milano, Italy

2Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

3University of Luxembourg, 2, avenue de l'Université, 4365 Esch-sur-Alzette, Luxembourg

4Universita degli Studi di Milano-Bicocca, Piazza dell'Ateneo Nuovo, 1, 20126 Milano, Italy

5INFN, Largo Enrico Fermi, 2, I-50125 Firenze, Italy

carolina.introini@mail.polimi.it

 

The study of fluid flows is a complex phenomenon but it has a practical importance in making predictions about fluid-dynamics and heat transfer in nuclear reactors. This complexity is even higher in the case of turbulent flows. In terms of fluid-dynamics conditions, the TRIGA Mark II research reactor at University of Pavia, the system of interest in the present work, has not been fully characterized yet: its flow regime is to this day still not fully known.
The scope of this work is to compare different turbulent models based on the Reynolds-Averaged Navier-Stokes (RANS) equations and to find out which is the most suitable for the study of the channel thermal-hydraulics of the TRIGA Mark II reactor. The RANS models available in the open source CFD software OpenFOAM have been applied to the most internal channel of the TRIGA and assessed against a Large Eddy Simulation (LES) model. In general, LES models require a finer grid and are more expensive, however, they are able to capture the smaller features of the flow and thus better approximate a direct numerical simulation.
On the other hand, for most engineering applications, the time average quantities are appropriate for the requested accuracy. To this aim, a fine grid LES simulation of this channel is performed, using the dynamic Smagorinsky model and the k-? SST-DES model. The results of those two models, expressed in terms of axial velocity, turbulent viscosity, turbulent kinetic energy, have been compared with the results obtained by the RANS models available in OpenFOAM, evaluated using a coarser grid. Heat transfer is taken into account as well by means of the turbulent energy diffusivity parameter. Those RANS models can be grouped into three categories: k-? models, which describe turbulence by means of two transport equations, using as variables the turbulent kinetic energy and turbulent dissipation, and which focus on the mechanisms that affect the turbulent kinetic energy; k-? models, which again use two transport equations to close the Reynolds-averaged Navier-Stokes equations, but are specifically tailored for near-wall simulations; and Reynolds Stress Transport models (RMS), in which the components of the Reynolds stress tensor are directly computed, and which are considered the most complete classical turbulence model.
The simulation results demonstrate how, amongst the RANS models, the k-? SST-SAS one is the one whose results are closer to the LES simulation. This model seems to be the best one for the treatment of turbulent flow within the TRIGA subchannel, offering a good compromise between accuracy and computational requirements. Since it is much less expensive than an LES model, it can be applied even to full core calculation, in order to obtain accurate results with less computational effort.






12.09.2017 10:10 Poster session - RED

Reactor physics - 607

Results of EURADOS exercise on neutron spectrum unfolding in Bonner sphere spectrometry using GRUPINT

Bor Kos, Vladimir Radulović, Ivan Aleksander Kodeli

Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

bor.kos@ijs.si

 

Bonner sphere spectrometry (BSS) is a well-established neutron spectrometry method developed in the 1960s. The main advantage of the method is its nearly isotropic response and wide energy range, from thermal neutron energies up to several hundred MeV.
Bonner sphere spectrometers consist of neutron detectors, in our case an idealized 3He detector in a 1 mm thick steel case, located in the middle of moderating spheres of different diameters. The 13 spheres used in in our case are made of polyethylene of variable thickness. In one case the sphere is covered in Cadmium to act as a thermal neutron absorber. Combining thermal neutron sensors with varying thicknesses of polyethylene moderators results in a method which is sensitive to a broad energy range of neutrons. Neutron spectra in measurement locations can be deduced from these measurements using so-called unfolding techniques.
An inter-comparison was proposed by EURADOS (The European Radiation Dosimetry Group) of different unfolding techniques. A comparison of the results is crucial since the user plays a major role in the end quality of the results.
In this paper, unfolding techniques are tested on 4 different real-life neutronics scenarios. These are:
• an irradiation room with an iron moderated radionuclide source
• a medical accelerator within a treatment room, with two measurement positions
• a Sky-Shine scenario
• an irradiation room with a water moderated radionuclide source
The primary software used in the present work is the GRUPINT program developed by A. Trkov in 2000 for calculating constants for neutron activation analysis from energy-dependent cross sections. Further development for the field of BSS was envisioned by the author but not realized. Several other programs, such as MCNP, ANGELO, ST2ENDF, DICTIN and ZOTT99 were used to determine the final 100-energy group neutron spectrum and uncertainty determination for the four different real-life workplace neutronics scenarios.






12.09.2017 10:10 Poster session - RED

Reactor physics - 608

SCALE 6.1.3 evaluation of the erbia effect on the control rod worth of an Er-SHB PWR assembly

Antonio Guglielmelli, Roberto Pergreffi, Federico Rocchi

Italian National Agency for New Technology, Energy and Substainable Economic Development, Via Martiri di Monte Sole, 4 - Bologna , 40129, Italy

antonio.guglielmelli@gmail.com

 

In the last fifty years, one of the research efforts in the field of nuclear fuel design has been to search burnable neutron absorbers to be mixed directly and in low amount in the UO2 fuel powder matrix. The use of the burnable absorbers have opened the possibility to improve the safety (excess of reactivity compensation), the cost (long-life core with higher burnup cycle), the waste (reducing on-site fuel management) and the proliferation (decrease of the final plutonium inventory) in LWRs. In the 1980s Erbia (Er2O3) was recognized as a viable alternative absorber to gadolinia (Gd2O3) because it introduces a series of further safety improvements. Erbium, with respect to Gadolinia, shows a better effectiveness ? thanks to relatively low thermal absorption cross section ? at minimizing radial power peaking; better controlling ? thanks to the higher resonance integral – of the transients, and a nearly linear efficiency – thanks to a rather short evolution chain ? as a function of the content and number of poisoned rods. In the last years, the aforementioned Erbia specific physical features have suggested the introduction of the erbia-credit super high burnup (Er-SHB) fuel concept that is a possible pathway to realize very high-burnups (>70 GWd/MTU) fuels with higher 235U (>5 wt.%) enrichments. The Er-SHB concept consists of mixing low amounts (>0.2 wt.%) of erbia (Er2O3) in a highly enriched (>5 wt%) UO2 powder immediately after the reconversion process in order to respect the criticality safety requirements related to the design of fabrication plants and without the need of substantial adaptation and relicensing of fuel cycle facilities. The Er-SHB is an innovative concept because erbia ? opposite to fuels in which burnable absorbers are loaded only in certain fuel pins ? is added homogeneously in all highly-enriched pins. In this way, the initial excess reactivity can become equivalent to that expressed by a fuel with the current enrichment limit (5 wt%) for commercial-type LWR fuel. The present work is the extension of a previous neutronics study [1] on a 17x17 PWR assembly (10.27 wt.% in 235U and 1.0 at.% in erbia content) and aims to evaluate the effect of an Er-SHB fuel on the control rods worth (CRW) at the beginning of life (BOL). The CRW estimation was achieved using the deterministic neutronic code NEWT of the SCALE 6.1.3 package that is widely accepted and used worldwide for safety analysis and criticality of LWR systems. The control rods worth results for the erbia fully poisoned PWR assembly were further compared with those obtained in a standard 17x17 PWR assembly with a BOL equivalent reactivity (5.0 wt.% in 235U). A comparison between APOLLO2 and NEWT main neutronic parameters (multiplication factor, fast and thermal flux, spectral index, neutron spectrum) results was also realized.

[1] R. Pergreffi, D. Mattioli, F. Rocchi, “Neutronic characterization of an erbia fully poisoned PWR assembly by means of the APOLLO2 code”, EPJ Nuclear Sci. Technol. 3, 8 (2017)






12.09.2017 10:10 Poster session - RED

Reactor physics - 609

The comparison of the calculated and measured reactivity efficiencies of control rods during the safety analysis of VVER reactors

Nurbol Zhylmaganbetov

Scientific and Engineering Centre for Nuclear and Radiation Safety Gosatomnadzor of Russia, Malaja Krasnoselskaya, P.O.Box 2-8, Building 5, 107140 MOSCOW, Russian Federation

zhylmaganbetov@secnrs.ru

 

The term reactivity is one of the most fundamental in the physics, calculation methods and safety analysis of reactors of nuclear power plants (NPP). Requirements for reactivity as well as to parameters of reactor system defined utilising reactivity are stated in regulatory documents of various levels. These parameters include the efficiency of emergency protection (EP), the efficiency of EP without the most effective single control rod and others. Following Nuclear Safety Regulations of Nuclear Power Plants (NP-082-07) the results of calculations of these parameters must be confirmed by results of measurements at the minimum controlled power level during the physical start-up of the unit and before the operation start of each fuel load.
The measurement of the efficiency of EP and the efficiency of EP without the most effective single control rod at the operating units of NPP with VVER reactors is performed utilising so-called “rod drop method”. In this case, reactivity is recorded by reactivity meter that processes the signal of ion chamber (IC) located in the concrete of bio-shield of the reactor system.
The calculated reactivity in this paper means the quantity gained from the equation ? = 1 – 1/Keff. For example, when the efficiency of EP has calculated the values of Keff is substituted in the equation. These values of Keff are obtained using two steady-state calculations of reactor system – initial state (before the EP dropping) and the final state (after the EP dropping) while the transient process is not taken into account. In practice, when the reactivity has measured the behaviour of a transient process has a significant effect on the measured value of reactivity. Therefore, the results of steady-state calculations using the above equation are not equal with the results of performed measurements. For example, the difference between the values of measured and calculated efficiencies of EP of VVER reactor systems could reach up to 1.5-2 times. Consequently, for the correct comparison of calculated and measured results of defining the EP efficiency it is rational to calculate reactivity considering the transient process between two steady-states of the reactor in other words to perform numerical simulation of measurement of EP efficiency using the model of reactivity meter (considering the position of IC). Such reactivity is commonly called as reactivity obtained from numerical measurement/experiment simulation. The application of this calculation technique is demonstrated in the present paper using the example of calculations of the physical start-up of the Rostov NPP unit 3.
As a result of the analysis of the comparison calculations with measurements, a systematic discrepancy between the results of calculations and measurements is noted which can be minimised by determining the reactivity at the moment when control rods reach the lowest position. It is shown in the paper that the readings of IC immediately after control rods reaching the lowest position are the most stable therefore the comparison is recommended to conduct exactly at this moment of the transient process. The selection of this state makes the result independent of ambiguous justification and use of the asymptotic behaviour of a system with delayed neutrons and the registration features of this state with IC. It also makes possible to reduce the systematic differences between the results of calculations and measurements in the safety analysis.






12.09.2017 10:10 Poster session - RED

Reactor physics - 611

Investigation of the Performance of the GFR2400 Reactivity Control System in the Equilibrium Cycle

Stefan Cerba, Branislav Vrban, Jakub Lüley, Ján Haščík, Vladimir Nečas

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia

stefan.cerba@stuba.sk

 

In the age of modern calculation tools and high-tech computer systems Monte Carlo codes are getting into a strong dominance against deterministic methods. They are used for various application, but mostly for reactor core analyses. One of the most effective tools used for reactor core calculations is the SCALE 6 system. In addition to general applications, like criticality or burnup calculations, sometimes it is important to understand the behavior of the reactor core from the local point of view. One of the new unconventional features of the SCALE 6 system is the calculation of Local Multiplication Values (LMV). The LMV parameters provide an alternative method of calculating the keff of the system. They are based on the fission production matrix calculations. This method has already been used for several applications and reactor types, however no verification against deterministic approaches is available. Common, daily use of this feature requires at least partial validation or code to code verification of the LMV parameters. This paper is dealing with the issue of verifying the SCALE 6 LMV calculations using the nodal diffusion DIF3D code. The verification methodology is based on the calculation of region wise energy integrals of parameters, such as Buckling, gain balance or fission rate. The methodology is is described in the paper. The deterministic calculations are carried out using problem oriented SBJ_E71 multi-group XS libraries and the results are compared using the in-house DIFRES graphical utility program.






12.09.2017 10:10 Poster session - RED

Reactor physics - 612

Monte Carlo simulation of neutron interaction with materials of neutron detection interest

Djelloul Benzaid1, Bentridi Salaheddine1, Seghour Abdesslam2

1University Djilali BOUNAAMA de Khemis-Miliana, Route de Theniet El-Hed, 44225 Khemis-Miliana, Algeria

2Centre de Recherches Nucléaires, 23 Rue du Loess, F-67037 Strasbourg, France

d.benzaid@univ-dbkm.dz

 

It is well established that neutrons are not directly detected because they have no electric charge. And to do so one have first to convert these particles to charged detectable particles.
The behavior of charged particle products of neutron reaction with the convertor materials reveals all possible proprieties of incident neutrons.
The present work consider the simulation of mono-energetic neutron beam with adequate convertor materials, especially boron-10 and lithium-6, to find the optimal characteristics of the material layers to be used in the process of neutron detection, thickness in particular. The layer is supposed to be a cylindrical form.
The response function calculated is used as a strong tool to control the quality of the simulation method and the different parameters used as well.
To calculate the response function of the detector a program is developed. It consists mainly of modeling the interaction of both incident neutrons and particle products with the target nucleus of the boron and lithium converter layers separately, using the rejection sampling method of VON NEUMANN. Results are discussed and compared to existing data in the literature.






12.09.2017 10:10 Poster session - RED

Reactor physics - 614

MULTIGROUP NEUTRON TRANSPORT THEORY IN MOLTEN SALT REACTORS: THEORY AND APPLICATIONS

Ayhan Yilmazer

Hacettepe University, Nuclear Engineering Department, 06800 Beytepe, Ankara, Turkey

yilmazer@hacettepe.edu.tr

 

Neutron transport equation coupled with delayed neutron precursor equations is formulated against a moving molten salt background in a hydrodynamic representation using Lagrangian reference frame. Conservation of the neutrons is formulated in terms of relative velocities with respect to the medium. Moving molten salt introduce four generic correction factors to the transport equation expressed for static background material: a density modification to the time derivative of the angular flux, a material acceleration term, a cross section renormalization, and a (moving) source transformation.
Then, the transport equation formulated in relative velocity space is expressed in energy-direction configuration space with explicit source (i.e.inscattering plus fission plus delayed neutrons plus external source). This form allows us to obtain standard multigroup representation in which group parameters are defined over group energy intervals. The molten salt acceleration term is treated for one-dimensional slab and spherical geometry and two-dimensional cylindrical geometry.
As an application, one-speed and two-group criticality calculations are done for slab geometry using discrete ordinates method. Effect of the magnitudes and spatial distribution of the relative velocity and acceleration, and any directional anisotropy, between transport particle and material background is investigated.






12.09.2017 10:10 Poster session - RED

Nuclear fusion - 707

Investigation of laser polarization effect on Collisional Absorption in Inertial Fusion

Leila Gholamzadeh

Yazd University, Physic Department, P.O.Box 89195-741, Yazd, Iran

gholamzadeh@yazd.ac.ir

 

Collisional absorption is analytically considered in an unmagnetized and uniform plasma using kinetic theory. The laser and plasma parameters such as elliptical- polarized electric field, the tamper atomic number and electron –ion collision frequency have been investigated on the laser energy absorption. Maxwellian distribution function is used for the electrons distribution. Our results show that higher absorption is obtained with elliptical-polarized laser and high electron –ion collision frequency.






12.09.2017 10:10 Poster session - RED

Nuclear fusion - 708

PIC KINETIC MODELLING FOR ELM TRANSPORT IN THE SCRAPE-OFF LAYER

Ivona Vasileska1, Leon Kos1, David Tskhakaya2, Richard Pitts3, Tomaž Gyergyek4

1University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta. 6, 1000 Ljubljana, Slovenia

2Institute of Applied Physics, TU Wien, Fusion@ÖAW, Austria, Wiedner Hauptstr. 8-10/134, 1040 Wien, Austria

3ITER Organization, Cadarache Centre, Building 519, 13108 St. Paul lez Durance, France

4University of Ljubljana, Faculty of Electrical Engineering, Tržaška 25, 1000 Ljubljana, Slovenia

ivona.vasileska@lecad.fs.uni-lj.si

 

In this work we describe a kinetic model of JET SOL (scrape-off layer). It includes the dynamics of three particles: the main ions (D+), neutrals (D) and the electrons. This modelling aims at improving and understanding the edge localised mode (ELM) transport in the SOL at the JET tokamak. The simulations were done by particle-in cell (PIC) code BIT1, which is one of the most powerful tool for electrostatic parallel simulation of edge plasma. It usually includes Monte Carlo (PIC/MC) code for simulating particle collisions and plasma surface interaction. In our case we develop 1D3V (1D in usual and 3V velocity space) kinetic PIC model, where the charged and neutral particle dynamics and interaction between them is included in a fully self-consistent way. The simulations geometry corresponds to a SOL bounded between the divertor plates, separatrix and outer wall. The particles were injected by ambipolar source (electrons and ions) and the divertor plates have absorbing nature. The plasma parameters chosen are relevant for JET tokamak parameters: B = 2.2 T, the angle between the magnetic field and the normal from the plates is 84$^{\circ}$, the length of the simulated system is 0.8\,m. The simulated parameters were: the number of cells was equal to 120000 and the physical particles per computer particles was 1.0e8. As a result we obtained the profiles of the plasma and neutral parameters (density and temperature) of the fully independent run starting from empty system and reaching to the stationary state.






12.09.2017 10:10 Poster session - RED

Nuclear fusion - 709

Microstructural and mechanical characterization of W-based composites for DEMO divertor

Petra Jenuš1, Matej Kocen1, Andreja Šestan2, Janez Zavašnik3, Sabina Markelj4, Mitja Kelemen5, Saša Novak6

1Jožef Stefan Institute, Department for nanostructured materials, Jamova 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Centre for electron microscopy and microanalysis, Jožef Stefan International Postgraduate School, Jamova 39, 1000 Ljubljana, Slovenia

3Jožef Stefan Institute, Centre for electron microscopy and microanalysis, Jamova 39, 1000 Ljubljana, Slovenia

4Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

5Institut Jožef Stefan, Jamova cesta 39, 1000 Ljubljana, Slovenia

6Jožef Stefan Institute, Department for nanostructured materials, Jožef Stefan International Postgraduate School, Jamova 39, 1000 Ljubljana, Slovenia

petra.jenus@ijs.si

 

Lately tungsten-based composites have gained considerable attention of the scientific community due to their excellent performance at high temperatures, especially high melting point, good thermal conductivity and a low thermal expansion coefficient. On the other hand, tungsten is affected with a serious reduction of strength at elevated temperatures, latter being one of the main drawbacks of its usefulness as a plasma facing material in fusion reactors.1 Therefore, the main aim of this work has been to improve the material properties to sustain plasma-facing conditions when installed into divertor, especially to be able to resist high thermal loadings during operation. Among the available options we selected reinforcement of tungsten by incorporation of carbide nanoparticles, wherein the reinforcement should not chemically react with the matrix. In this respect, W2C nanoparticles offer an interesting option.
WC nanoparticles and graphene were used as a carbon source. Both used precursors resulted in formation of W-W2C composites. The results of their analysis suggested WC nanoparticles as the most convenient precursor for W2C. Using field assisted sintering technique (FAST, 1900°C, 60 MPa, 5 min), only two phases were detected in the sintered composite: cubic W and hexagonal W2C, which implies complete reaction of the WC precursor.
The results confirm that presence of small W2C grains enhance densification of tungsten, inhibit the tungsten grain growth at high temperature up to at least 1300 °C and significantly improve mechanical properties of the material. In addition to thorough microstructural analysis, XRD and analysis of mechanical properties at room and elevated temperature, we also verified the potential beneficial role of reinforcing W2C particles into the W matrix on helium accumulation, which might cause deterioration of mechanical properties. Helium implantation was achieved by irradiation of polished samples at 1 MeV at room temperature with average He ion fluence of 8 x 1016 He/cm2. After implantation, the composite samples were examined by focused ion beam-scanning electron microscopy (FIB-SEM) and compared with the W without carbide inclusions.
Here, we will present the main characteristics of W-based composites prepared by FAST, i.e. relevant microstructural properties of the composites in comparison with tungsten, their mechanical properties as well as the formation of He bubbles after thermal treatment at 1600 °C, which are the potential source of material degradation during operation.






12.09.2017 10:10 Poster session - RED

Nuclear fusion - 710

Angular dependence of Fe erosion yield studied with keV Ar ions

Mitja Kelemen1, Anže Založnik2, Primož Pelicon2, Sabina Markelj2, Rodrigo Arredondo Parra3, Thomas Schwarz-Selinger3

1Institut Jožef Stefan, Jamova cesta 39, 1000 Ljubljana, Slovenia

2Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

3Max-Planck-Institut für Plasmaphysik (IPP), Boltzmannstr. 2, D-85748 Garching b. München, Germany

mitja.kelemen@ijs.si

 

Erosion of the inner walls is one of the issues of a thermonuclear reactor. Material used for inner walls will be irradiated by fast ions which will erode the material and possibly contaminate the core plasma. Also re-deposition of eroded material contributes to additional loss of fusion fuel and retention and consequently reduce the operational life time of the inner walls. For perfectly smooth surfaces a distinct angular dependence of the sputter yield is expected [1]. For rough surfaces this dependence is smeared out. Within the Eurofusion work package PFC a dedicated task was launched to quantitatively determine the influence of roughness on the sputter yield.
To address this issue the erosion of magnetron sputtered iron (Fe) thin films on silicon by argon (Ar) ions was performed. To this end we constructed an experimental set up inside of an vacuum experimental chamber. Samples are exposed to Ar ions generated with Electron cyclotron resonance (ECR) ion gun. Ions are extracted with 1keV, focused with an Einzel lens and decelerated in front of the sample to 900eV.
The sample with 260 nm thick Fe film deposited on silicon (Si) was mounted on a rotating table. Samples were exposed to 1020 ions/m2 Ar ions at different impact angle to the surface. The thickness of the eroded layer was measured by Rutherford back scattering spectroscopy (RBS). Due to the small size of the erosion crater we perform the RBS measurements in our microbeam chamber with 1.5 MeV protons focused down to 3 µm. In the presented work results from experiment will be discussed and plans for future work will be presented.

[1] M. Küstner et al. JNM 265 (1999)






12.09.2017 10:10 Poster session - RED

Nuclear fusion - 711

Activation Material Selection for Multiple Foil Activation Detectors in JET TT campaign

Igor Lengar1, Aljaz Čufar2, Vladimir Radulovič2, Paola Batistoni3, Sergey Popovichev4, Luka Snoj5

1Institut "Jožef Stefan", Odsek za reaktorsko fiziko, Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

3ENEA Fusion Division, Via E. Fermi 45, 1-00044 Frascati, Rome, Italy

4Culham Centre for Fusion Energy, Abingdon, Oxon, OX14 3DB, United Kingdom

5Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

igor.lengar@ijs.si

 

In the preparation for the Deuterium-Tritium campaign, JET will operate with a tritium plasma and would allow for the study of isotope effects in various plasma scenarios. The T+T reaction consists of two notable channels: (1) T+T › 4He + 2n, (2) T+T › 5He(GS) + n › 4He + 2n, GS meaning the ground state of 5He. The reaction channel (1) is the reaction with the highest branching ratio and a continuum of neutron energies being produced. Reaction channel 2 does not produce a continuum of energies but a spectrum with a peak. A particular problem when dealing with the TT reaction is the ratio between the individual reaction channels, which is highly dependent on the energy of the reacting tritium ions. In contrast to accelerator data obtained at higher energies, the TT neutron spectrum does not show the n+5He peak at 8.7 MeV for CM energies lower than about 110 keV. There are, however, very few measurements and the study of the TT spectrum at JET would be interesting.
The work is focused on the determination of the spectral characteristics in the TT plasma discharges, especially on the presence of the 8.7 MeV peak, a consequence of the interaction between the particles in the final state. The standard JET spectroscopic capabilities are used for spectra determination. In addition the possibility to use an optimized set of activation materials in order to target the measurement of the 8.7 MeV peak is studied. The measurement could be very difficult in presence of traces of deuterium as the neutrons produced in the DT reaction could completely cover the relevant part of the neutron spectrum. The effort will be made to calculate the maximum concentration of deuterium allowed.






12.09.2017 10:10 Poster session - RED

Radiation and environment protection - 803

Draft Version of the Computational Module for Simplified Radiological Analyses

Paulina Dučkić, Krešimir Trontl, Mario Matijević, Radomir Ječmenica

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia

kresimir.trontl@fer.hr

 

In the framework of the DOCPAGANSA (“Development of Code Package for Advanced Gamma and Neutron Shielding Analyses”) research project the computational module for simplified shielding analyses has been developed. The computational module is based on the point kernel method which is used for gamma, as well as neutron dose calculations. The gamma and neutron buildup factors are determined by a novel method based on Support Vector Regression (SVR) which is one of the machine learning techniques. In this paper we report on the main characteristics of the developed preliminary SVR models for gamma and neutron buildup factors. We also present the preliminary results of the testing of the module draft version on the geometrically simple problem setup with gamma and neutron source.






12.09.2017 10:10 Poster session - RED

Radiation and environment protection - 804

Overview of Radiation Accidents at Industrial Accelerator Facilities

Helena Janžekovič

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

helena.janzekovic@gov.si

 

Industrial irradiators have been used for decades but in the last decade sterilisation of products became an important part of many fast growing industries. The required dose rates in order to make an efficient sterilisation might be up to thousands of Gy/s. Due to such extremely intense radiation fields, defence in depth shall be implemented in design of irradiation facilities and high safety culture shall be in place. The facilities are based either on a use of radioactive materials, i.e. Co-60 or Cs-137 radioisotopes, or on a use of accelerators. Due to security reasons irradiators with Cs-137 became obsolete in the last decade. In addition, mentioned radioactive sources of so-called Category 1 sources from the IAEA RS-G-1.9 are largely replaced by accelerators in order to avoid not only security issues but complex safety requirements related to management of disused radioactive sources. Sources of Category 1 are related to high risks as given in the document mentioned: “The sources if not safely managed or securely protected, would be likely to cause permanent injury to a person who handled it or who was otherwise in contact with it for more than a few minutes. It would probably be fatal to be close to this amount of unshielded radioactive material for a period in the range of a few minutes to an hour.” The use of sources in industrial sterilisation facilities is linked to high risks to human health in case of an accident.
In available literature accidents related to industrial irradiators using radioactive sources are well described including detailed analyses of accidents with fatalities, e.g. in Kjeller 1982 and Nasvizg in 1991. Much less information is available when industrial accelerator facilities are involved in accidents although the first accident related to industrial sterilisation using accelerators happened in 1965 in USA resulting in amputation of worker’s leg and arm. There is less accidents with accelerators than with radionuclides as decay of radionuclides used in Category 1 sources poses constant risk. In addition, it can be noted that safety features installed at accelerator facilities enable easier implementation of safety in such facilities than in facilities using radioactive sources, e.g. number of external events jeopardising safety is much lower. In addition, collimated fields used at accelerator facilities contribute to partial exposure to bodies in case of an accidents, e.g. in general, amputation of limbs could be required in worse-case scenario. The analysis of around ten reported events related to industrial accelerators used for sterilisation described in open literature, e.g. IAEA, OTHEA-RELIR database and IRPA proceedings, is presented systematically identifying:
• initial event,
• contribution factors,
• responsibilities,
• lessons learned and applicability of contemporary standards and guides.
The analysis uses not only international standards and guides, such as IAEA SS SSG-8, but also national standards and guides. The analysis focuses on physical phenomena such as dark currents, on safety systems installed and on safety procedures. It might help designers, users and regulatory bodies when preparing or studying safety assessment of industrial accelerator facilities used for sterilisation. It can contribute to better understanding risks associated with the use of these somehow very challenging facilities.







12.09.2017 10:10 Poster session - RED

Radioactive waste management - 901

Effect of pH on cation release from sodalite-based matrices for immobilization of spent chloride salt waste

Mirko Da Ros1, Giorgio De Angelis2, Francesca Giacobbo3, Marco Giola3, Elena Macerata3, Mario Mariani3

1Politecnico di Milano - department of energy, Via La Masa 34, 20156 Milano, Italy

2Retired from ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy

3Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

francesca.giacobbo@polimi.it

 

In the frame of Generation IV reactors a renewed interest in pyro-processing of spent nuclear fuel is underway. Molten chloride salt waste arising from the recovering of uranium and plutonium through pyro-processing is one of the problematic wastes for direct application of vitrification or ceramization. Sodalite, a naturally occurring aluminosilicate mineral containing chlorine, is among those mineral phases under consideration as potential matrix for confinement of spent chloride salt waste coming from pyro-processing. In this work, the effect of pH on metal release from sodalite-based matrices blended with borosilicate glass has been evaluated since there is a scarcity of such data in literature. To this aim, sodalite samples loaded with simulated chloride salts waste, with and without the addition of borosilicate glass, were leached for contact times up to 15 days at 90°C buffered at three different values of pH. Leaching results were compared in terms of normalized releases as a function of pH. SEM analyses were also performed in order to compare the matrix surface before and after leaching. According to this study it is apparent that the retention performances of sodalite-based matrices can be significantly worsened under acidic condition, even of some orders of magnitude for some elements of interest. Therefore it is fundamental to adopt suitable solutions in order to avoid acidic conditions and preserve the retention capabilities of the sodalite-based confining matrices.






12.09.2017 10:10 Poster session - RED

Nuclear power plant operation and new reactor technologies - 1007

Visual Examination of Primary Loop Components in VVER Nuclear Power Plants

Petar Mateljak

INETEC-Institute for Nuclear Technology, Dolenica 28, 10250 Zagreb, Croatia

petar.mateljak@inetec.hr

 

Besides nuclear power plant reactor, pressurizer, cooling tanks and steam generators are main parts of nuclear power plant primary loop. Generally, primary loop components are exposed to radioactive contamination and high mechanical and thermal stresses. In order to obtain safe operation during regular fuel cycle, all primary components have to be examined in accordance with strict nuclear regulations and relevant standards. In this paper, visual testing (VT) technique and equipment for visual examination of primary loop components in VVER nuclear power plants are presented.
INETEC has recently developed a system for visual inspection of VVER pressurizer, cooling tanks and steam generators. Taking into account increasing requirements to the safety enhancement during plant operation, shortening of the inspection time, radiation exposure to examination personnel and cutting the total inspection costs, system is designed as modular robotic system with integrated VT module. In this paper, system’s capabilities and features are described, and visual testing technique and performance are presented.

Keywords: VVER, pressurizer, cooling tank, steam generator, non-destructive testing, VT, automated inspection






12.09.2017 10:10 Poster session - RED

Regulatory issues, legislation, sustainability and education - 1108

Public Opinion about Nuclear Energy – Year 2017 Poll

Radko Istenič, Igor Jenčič

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

radko.istenic@ijs.si

 

The Information Centre established within the Nuclear Training Centre at the Jožef Stefan Institute 24 years ago informs the visitors about nuclear power and nuclear technology, about radioactivity and about Krško Nuclear Power Plant.
Information activities target mainly the schoolchildren from the 8th and 9th grade of elementary school with their teachers (in total close to 8000 per year). The visitors can choose between live lectures on nuclear technologies (fission and fusion), lecture on stable isotopes and energy workshops. The visit includes a demonstration of radioactivity and a guided tour of a permanent exhibition.
The opinion trends are monitored since 1993 by polling about 1000 youngsters every year. The poll is conducted before the youngsters listen to the lecture or visit the exhibition in order to obtain their opinion based on the knowledge from everyday life. Trends over the last 24 years will be presented, summarized and commented.






12.09.2017 10:10 Poster session - RED

Regulatory issues, legislation, sustainability and education - 1109

Nuclear heat utilization for campus heat supply at the research reactor of Budapest

Török Szabina, Börcsök Endre, Talamon Attila

Centre for Energy Research Hungarian Academy of Sciences, P.O.Box 49, H-1525 Budapest, Hungary

torok.szabina@energia.mta.hu

 

Significant part of global GHG emission is related to heat generation The development of nuclear cogeneration and recover heat utilization offers a convenient possibility for emission reduction and for replacement of fossil fuels. This study is focuses on feasible heat recovery for space heating system in case of small, nuclear reactor with 10MW thermal power (Budapest Research Reactor). The presentation shows the methodology to assign the economic and environmental role of nuclear cogeneration in the heat generation portfolio of a well defined region, taking into consideration the seasonal heat demand profile and the existing infrastructure (Demonstrate with Nuclear Power Plant Paks).
Cost benefit analysis is presented for nuclear congelation compared to the status quo fossil fuel (natural gas) consumption.
Vision for heat utilization of existing/future research facilities will be shown.






12.09.2017 10:10 Poster session - RED

Regulatory issues, legislation, sustainability and education - 1110

Media Content Analysis on India’s Nuclear Power Policy under Prime Minister Narendra Modi

Hijam Liza Dallo Rihmo

Jawaharlal Nehru University, New Mehrauli Road, Munirka, New Delhi, Delhi 110067, India

lizahij@gmail.com

 

The role of media is laudable for the success of Modi coming to power in 2014. It is a strong reification of media’s outreach and its potential to influence public opinion. Modi coming to power heralded with fervour transformative policies, development programmes, smart cities, “Make in India” Initiative and Digital India. His government policy highly resonated among the general public with reciprocity which were intuitively captured by the media. His policy on India’s energy efficiency was heard loudly as well. But in examining the mediatisation of India’s nuclear energy policy there should be an understanding of the nuances of the nuclear energy policy. This will highlight the different trajectories of India’s nuclear policy framing from the need for nuclear power generation to status and prestige of the country in international politics. The research paper process through the media frames in India for the identification of media construction nuclear energy policy under Modi’s leadership. The presentation of India’s nuclear energy policies in major newspapers is examined to understand how people integrate media frames with their pre-existing beliefs and values. It is a critical analysis of media as an institution where public consent is engineered by interested parties.






12.09.2017 10:10 Poster session - RED

Probabilistic safety assessment - 1201

Human Factor Importance in Probabilistic Safety Assessment - Shutdown Versus Full Power

Mitja Antončič1, Marko Čepin2

1University of Ljubljana, Faculty of Electrical Engineering, Tržaška 25, 1000 Ljubljana, Slovenia

2Fakulteta za elektrotehniko, Tržaška cesta 25, 1000 Ljubljana, Slovenia

mitja.antoncic@fe.uni-lj.si

 

Probabilistic safety assessment is a standard method for assessing safety of the nuclear power plants and other complex systems. It has been discussed in a number of documents and studies since its introduction in 1957. First probabilistic safety assessments were mainly focused on the safety of a nuclear power plants during a full power operation. Later, the methods and models have been developed also for other plant states including plant shutdown. The objective of this paper is to present the differences in the models and the results of probabilistic safety assessment for the plant shutdown versus the plant full power operation. The shutdown probabilistic safety assessment is developed in a similar manner as the full power probabilistic safety assessment, but considering specifics of each particular plant operating state under investigation. The differences include different system models, different connections between systems and different probabilities or frequencies of events. Consequently, the results are different for each particular plant operating state, which may include qualitative differences in terms of different minimal cut sets and quantitative differences in terms of different importance factors and different frequencies of accident scenarios. Importance of human errors in the process of initiating event mitigation (Post-accident) and also during initiating event development (Pre-accident) differs considerably between power operation and shutdown states. Consideration of different plant operating states within the probabilistic safety assessment causes notable increase of complexity of the models.






12.09.2017 12:00 Reactor physics - plenary

Reactor physics - 601

Utilization of thorium in PWR reactors- First step toward a Th-U fuel cycles

Jose Rubens Maiorino1, Giovanni Stefani2, Francesco Saverio D'Auria3

1Universidade Federal do ABC/ Universita di Pisa, Av. dos Estados, 5001, Santo André, Sao Paulo, Brasil, Brazil

2Universidade Federal do ABC - PROGRAMA DE PÓS-GRADUAÇAO EM ENERGIA, Av. dos Estados, 5001. Bairro Bangu. Santo André - SP , 09210-580, Brazil

3University of Pisa, Via Montebello di Mezzo 17, 19020 Bolano, Italy

joserubens.maiorino@ufabc.edu.br

 

Since the beginning of Nuclear Energy Development, thorium was considered as a potential fuel, mainly due to the potential to produce fissile 233U. Several Th/U fuel cycles, using thermal and fast reactors were proposed and are still under investigation. However, the technical feasibilty to use thorium was made in PWR; the USA PWR Indian Point Reactor was the first to utilize a core load with (Th0-0.9./U1-0.1)O2, with highly enriched U (93w/0), achieving a maximum burn up of 32 MWD/kg HM. Also the last core of the Shipping port PWR (shutdown in 1982) was ThO2 and (Th/U)O2, operating as a Light Water Breeder Reactor (Seed-Blanket Concept) during 1200 effective full power days of operation (60 MWD/kg HM). More recently, many researchers turned their attention to Th fuel cycles in PWRs aiming at reducing the generation of minor actinide waste, at improving the nuclear power sustainability and at better fuel utilization. These studies were interested in assessing the feasibility of using 233U-Th fuels in PWR without worrying about how to obtain the initial 233U fuel load or the transition from an uranium to a thorium core in the current nuclear power plants. In this paper a review is provided of recent initiatives, with emphasis in a study, demonstrating the feasilibility to convert an existing advanced PWR (AP 1000) from UO2 to a mixed U/ThO2 core. The study takes as criterion that the transition from the current UO2 AP 1000 core to one with mixed U/Th fuels should be such that minimum changes occur on its current core design and operational parameters. Thus one could consider the following requirements in this study: produce important amounts of 233U (maximization) for future 233U/Th cores; keep the current fuel assembly geometry, i.e., fuel rod diameter and pitch and meet the current thermal-hydraulic limits such as maximum center line fuel rod temperature and maximum linear power density; keep the current fuel cycle length of 18 months. The results obtained showed that the homogeneous concept with three different mass proportion zones, the first containing (32% UO2-68%ThO2); the second with (24% UO2-76% ThO2), and the third with (20% UO2-80% ThO2), using 235U LEU (20 w/o), and corresponding with the three enrichment zones of the AP 1000 (4.45 w/o; 3.40 w/o; 2.35 w/o), satisfies the optimization criteria as well as attending all thermal constraints. The concept showed advantages compared with the original UO2 core, such a lower power density and, keeping the same 18-months-cycle, a reduction of B-10 concentration as soluble poison, as well as eliminating the integral boron poison coated (IFBA).






12.09.2017 12:20 Reactor physics - plenary

Reactor physics - 602

Study of Uncertainty in Kinetic Parameters in the Scope of the UAM-SFR Project of OECD

Ivan Aleksander Kodeli1, G. Rimpault2, P. Dufay2, J. Tommasi2

1Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

2CEA-Cadarache DEN/DER/SPEX/LDCI, Bâtiment 238 - Piece 10, F13108 Saint-Paul-lez-Durance, France

ivan.kodeli@ijs.si

 

Responding to an increasing demand from the nuclear community for best estimate predictions to be provided with their confidence bounds, OECD/NEA started in 2007 an Expert Group on "Uncertainty Analysis in Modelling (UAM)" focused on benchmark activities for Light Water Reactors (LWR). In 2015, the scope of UAM was extended to the Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors of the new generation (SFR-UAM) with the objective to study the uncertainties at different stages of Sodium Fast Reactors. Such an effort has been undertaken within the framework of a program of international co-operation that benefits from the coordination of the NEA Nuclear Science Committee (NSC), and from interfacing with the Committee of safety of Nuclear Installations (CSNI).
Among the basic fission reactor parameters the kinetic parameters such as the effective delayed neutron fraction (beta-eff) and the neutron generation lifetime (lambda) play a major role in the reactor safety and control analysis. Its accuracy should be therefore well understood and evaluated. The values of beta-eff vary from one isotope to the other (from ~200 pcm for Pu239 to ~650 pcm for U235 and ~1500 pcm for U238), therefore the reactor systems containing actinide isotopes in their fuel have to face the problem of lower values of beta-eff due to the presence of Pu isotopes, making the reactor control of MOX fueled cores more challenging. However, U238 contributes to the beta-eff in proportion to its fission and increases the overall beta-eff value. This requires a specific validation.
Precise knowledge of the effective delayed neutron fraction (beta-eff) and the corresponding uncertainty is important for nuclear reactor safety analysis. New approach for calculating beta-eff and the corresponding sensitivity and uncertainty analysis developed at JSI and CEA.
The paper presents the overview of the computer code and method for the calculation of the kinetic parameters and the corresponding uncertainties. The intercomparison covers both Monte Carlo (TRIPOLI, MCNP) and deterministic (SUSD3D/PARTISN, ERANOS) codes applied to a vast list of fast reactor benchmarks.






12.09.2017 12:40 Reactor physics - plenary

Reactor physics - 603

A Serpent/OpenFOAM coupling approach for the study of the fuel burnup analysis

Paolo Bianchini1, Antonio Cammi1, Christian Castagna1, Eric Cervi1, Francesca Celsa Giacobbo1, Stefano Lorenzi1, Davide Chiesa2, Massimiliano Nastasi2, Monica Sisti2, Ezio Previtali3, Manuele Aufiero4, Massimiliano Fratoni4

1Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

2Universita degli Studi di Milano-Bicocca, Piazza dell'Ateneo Nuovo, 1, 20126 Milano, Italy

3INFN, Largo Enrico Fermi, 2, I-50125 Firenze, Italy

4University of California, Berkeley, Department of Nuclear Engineering, 4155 Etcheverry Hall, MC 1730, University of California, Berkeley, Berkeley, CA 94720-1730, USA-California

christian.castagna@polimi.it

 

In nuclear reactors, fuel burnup calculations assess the time evolution of both the fuel composition and the reactivity during the in-core reactor irradiation. Furthermore, this analysis is also relevant for the fuel management (i.e., the study of the fuel cycle) and for determining the amount of long-lived radionuclides in spent nuclear fuel. In general, burnup analyses are carried out by means of Monte Carlo or deterministic codes, solving the Bateman equation. In this equation, an uniform distribution of both the temperature and density is usually employed for the coolant and the fuel regions. On the other hand, the influence of the correct temperature distribution on fuel consumption can have a remarkable effect. In this regard, the development of accurate modelling approaches is recommended to investigate the coupling between the thermal-hydraulics and the fuel burnup.
In this paper, a simulation approach that solves the neutron transport problem and the time evolution of the nuclide concentration, and concurrently describes the heat transfer between the fuel and the coolant is presented. The goal is achieved by coupling the Monte Carlo code Serpent for neutronics/fuel burnup modelling with the CFD code OpenFOAM for thermal-hydraulics description thanks to an iterative procedure. In particular, Serpent provides the power distribution that is taken as input by OpenFOAM, which calculates the temperature distribution within fuel and coolant. These results are used by Serpent for the next iteration and the process continues until the solution converges. The adopted approach was tested on a case study of a simplified fuel cell, composed by an UO2 pin surrounded by water. The results show that the proposed procedure is able to properly consider the coupling between neutronics and thermal-hydraulics and it can be worthy for assessing the influence of the thermal-hydraulics on the evolution of the fuel composition.






12.09.2017 14:20 Invited Michel Maschi

Invited lectures - 105

The EDF Research and Development Strategy for (Nuclear) Electricity Generation

Michel Maschi

Electricite de France, Research and Development Division, Avenue les Renardieres, Ecuelles, 77818 Moret sur Loing Cedex, France

michel.maschi@edf.fr

 

The EDF group is an integrated energy company with a presence in a wide range of electricity-related businesses: nuclear, renewable and fossil-fuel fired energy production, transmission, distribution, marketing as well as energy management and efficiency services, along with energy trading. EDF is France’s leading electricity operator and has a strong position in Europe (United Kingdom, Italy, countries in Central and Eastern Europe), making it one of the world’s leading electrical providers as well as a recognized player in the gas industry.
The main goals of the EDF group’s Research and Development (R&D) Division are to contribute to the improvement in the performance of the operational units, to identify and prepare mid and long-term growth vectors and to anticipate the major challenges facing the Group in the global energy context. EDF’s R&D works for all the businesses of the Group. It provides technological solutions or innovative and economic business models that improve the performance of the businesses and prepares the long-term future of the Group through medium and long-term anticipatory programs. It contributes to making EDF a worldwide industrial group of carbon-free electricity networks.
In this context, R&D has a crucial role in finding solutions to all of these challenges. Its research areas focus on three major priorities:
- to consolidate and develop competitive low-carbon production mixes. One of the key issues is to ensure the effective coexistence of conventional means of production, in particular by further improving the safety, performance and operating lifespan of existing nuclear power plants, alongside the development and renewable energy;
- to develop and experiment new energy services for customers enabling a flexible and low carbon demand-side management: energy efficiency, promotion of new effective uses of electricity, often in combination with renewable energies (heat pumps, electric mobility, etc.), development of technical and economic modelling to engineer buildings, industry and sustainable cities, development of uses and consumption being integrated into the electricity system itself through the use of smart grids and appropriate pricing;
-to adapt the electricity systems by improving network asset management, optimization models and economic scenarios for new transmission infrastructure projects, integrating intermittent energy and developing smart grids.
As concern Nuclear Technology, EDF R&D has developed innovative initiatives to improve the performance of existing plants, to support the commissioning of EPR fleet with Flamanville , to extend the operation of nuclear plant towards 60 years , and to prepare the future of new nuclear fleet.
EDF R&D supports also international project in Europe such as UK, Germany , but also in RP China , USA , South America , and Middle East countries . EDF R&D is of course sharing its research with countries, universities, research institutes and main suppliers. Innovation is also coming for the economic network using startup and SME solutions.
The innovation initiative aims at facing very tough economic environment, to improve competitiveness of Nuclear.
The Nuclear for the Future Initiative covers many aspects out of which the main are
- Safety , reducing severe accident impact and improving operation margins
- Use of Digital and and data analysis in Nuclear processes
- New material and fabrication modes for maintenance and new built
- New reactors design and construction modes
EDF R&D has major challenges to achieve to answer to the needs of existing and new built plants and make Nuclear a sustainable solution for low carbon energy.






12.09.2017 15:00 Current and future reactors - plenary

Nuclear power plant operation and new reactor technologies - 1002

Plans for deployment of High Temperature Reactors in Poland

Grzegorz Wrochna

Narodowe Centrum Badań Jądrowych, Soltana 7, 05-400 Otwock, Poland

g.wrochna@ncbj.gov.pl

 

High Temperature Gas Cooled Reactors (HTGR) are considered as the most mature technology among advanced nuclear reactors. Their deployment, however, was blocked for years by a dead loop: no vendor was ready to invest into a design without having a contract with an end-user, and no end-user was ready to invest in a first of a kind reactor.
This may be changed with the ‘Strategy for Responsible Development” published recently by the government of Poland. Deployment of HTGR is listed there among projects intended to boost Polish economy. Polish industry uses over 6500 MW of industrial heat in form of 550°C steam. Currently it is produced by coal- and gas-fired boiler. Coordinated replacement of aged boilers can create a market large enough to break the investment dead-loop mentioned above.
Polish HTGR deployment programme is well rooted in the achievements of Sustainable Nuclear technology Platform (SNETP), especially its Nuclear Cogeneration Industrial Initiative (NC2I) pillar. Recently finished NC2I-R project and just started Gemini+ project, financed by H2020/Euratom, collects world-wide knowledge and experience in order to facilitate the HTGR deployment programme.






12.09.2017 15:20 Current and future reactors - plenary

Nuclear power plant operation and new reactor technologies - 1003

SMR - Is the Future of the Nuclear Power?

Andras Cserhati

Hungarian Nuclear Society, Fenyesa U. 4, 1036 Budapest, Hungary

cserhati@npp.hu

 

"The Small is Beautiful" stated in study of economics in '70s.
In nuclear energy there was always demand for smaller models, units. Nowadays it seems to increase in addition to the main stream. Evermore many professionals look them as savers of the nuclear future.
The paper and presentation cover the following items:
- The size distribution of NPPs.
- The SMR definitions.
- Opportunities and challenges.
- The integrated composition.
- Best sources for information about SMR.
- Overview, the most important SMR projects.
- SMR groups: the simple, the hot, the fast and the useful.
- Economic approach, prices, economy of size and series.
- IAEA, US DoE and UK support.
- Licensing problems.
- Under construction: the CAREM of Argentina, the floating NPP in Russia and the HTR-PM in China.
- Projects of near future.
- SMR in Central and Eastern Europe?






12.09.2017 15:40 Current and future reactors - plenary

Nuclear power plant operation and new reactor technologies - 1009

Innovative Technologies for Nuclear Applications: a collaborative effort

Abderrahim Al Mazouzi

Electricite de France, Research and Development Division, Avenue les Renardieres, Ecuelles, 77818 Moret sur Loing Cedex, France

abderrahim.al-mazouzi@edf.fr

 

Since its creation in 2012, NUGENIA association has been focusing on its mission to be an integrated framework for safe, reliable and competitive Gen II & III fission technologies.
After 5 years of effort, and thanks to the extensive collaboration between industry, SMEs, RTOs, academia and technical safety organizations, the association, with its more than 110 members, was able to launch important collaborative projects that are aligned with its roadmap. These projects allowed building new knowledge and expertise and also generating results with added value to the whole nuclear community.
This presentation will give a global overview on the recent progress of the association, its strategic planning to strengthen the role of nuclear energy in the mix portfolio. The focus will be on the on-going and or just launched projects to highlight their objectives and how they will help the community increasing the competitiveness of the nuclear energy sources.






12.09.2017 16:20 Invited Said Abousahl

Invited lectures - 107

The European Commission’s Joint Research Centre (JRC) nuclear research and training programme

Said Abousahl

European commission, Directorate-General for Research, Rue Montoyer 75, 3/62, B-1049 Brussles, Belgium

said.abousahl@ec.europa.eu

 

The Commission’s Joint Research Centre (JRC) was created by the Euratom Treaty in 1957and in its early years all its efforts were directed towards nuclear or nuclear-related research. Since the late 1960s the JRC has diversified its activities to include a wide range of strategic research areas that are of importance to EU policy initiatives. The Euratom Treaty established the ability to pursue Framework Programmes (FPs) for nuclear research and training activities. Euratom FP is organised in two parts (or ‘Specific Programmes’): one corresponding to ‘indirect actions’ on fusion energy research and nuclear fission and radiation protection; and a second investing in ‘direct’ research activities that is carried out by the Joint Research Centre in the field of nuclear safety and security. The general objective of the current Euratom Programme 2014-2018 is to pursue nuclear research and training activities with an emphasis on continuous improvement of nuclear safety, security and radiation protection, notably to potentially contribute to the long-term decarbonisation of the energy system in a safe, efficient and secure way. The activities of the JRC are developed in full alignment and complementary with the research programmes implemented by various EU MS and are based on the following specific objectives:
(a) improving nuclear safety including: nuclear reactor and fuel safety, waste management, including final geological disposal as well as partitioning and transmutation; decommissioning, and emergency preparedness;
(b) improving nuclear security including: nuclear safeguards, non-proliferation, combating illicit trafficking, and nuclear forensics;
(c) increasing excellence in the nuclear science base for standardisation;
(d) fostering knowledge management, education and training;
(e) supporting the policy of the Union on nuclear safety and security.

These objectives are aiming essentially to develop and assemble knowledge in order to provide input to the debate on nuclear energy production, its safety and reliability, its sustainability and control, its threats and challenges – including the assessment of innovative future reactor systems.

JRC contribute to maintaining competences and expertise in the EU by ensuring access to its infrastructures to other researchers, training young scientists and fostering their mobility to sustain nuclear know-how in Europe. In this context, a close cooperation with ENEN is developed and several agreements with national research organisations and universities are established. At international level, co-operation with partner courtiers (US, Japan, China,..) and organisations (IAEA, OECD/NEA) has been reinforced and grows in major international initiatives such as Generation IV international Forum.






12.09.2017 17:00 Regulatory issues and legislation - plenary

Regulatory issues, legislation, sustainability and education - 1111

STATE LEVEL CONCEPT: QUANTIFICATION OF STATE SPECIFIC FACTORS

Daria Andreevna Koneva, Dmitry Sednev

Tomsk Polytechnic University, 30, Lenin Avenue, 634050 Tomsk, Russian Federation

d.a.koneva@gmail.com

 

In accordance with Development and Implementation Support Programme for Nuclear Verification 2016-2017 (STR-382), there is a necessity for the improvement of the existing Safeguards Instrumentation and for the development of new systems and approaches for efficiency and effectiveness in nuclear nonproliferation purposes.
One of the defined milestones is implementation of the State-Level Concept (SLC) that allows a comprehensive evaluation of the activities of a state. The SLC is applicable to all States with safeguards agreements in force. The SLC includes State Specific Factors (SSF), Acquisition Path Analysis (APA) and the development of State-Level Approaches (SLAs). Under the SLC, safeguards will be focused on understanding the entirety of the nuclear program in the State and developing a customized SLAs for the States.
SSF is the first step of SLC, that dedicated to provide the fundamental information for the further analysis. These factors are based on technical considerations and will be used objectively and consistently for all States. SSF list consists of six factors, that should be taken into account. However, most of mentioned parameters could not be measured quantitively, but should be described qualitatively by group of experts. Main criticism of SLC was focused on suspicion of IAEA in possibility to apply double standards to different states in safeguards implementation. In order to have an independent view, an expert evaluation approach should be minimized as much as possible, but unfortunately, it could not be changed so much. Despite of this fact, several steps in direction to decrease human factor could be done.
In the paper an overview of SSF for APA is represented. The possibility of SSF quantification and its influence on the APA process is studied.






12.09.2017 17:20 Regulatory issues and legislation - plenary

Regulatory issues, legislation, sustainability and education - 1102

Hinkley Point C and EU State Aid Rules

Ana Stanic

E&A Law, 42 Brook Street, London W1K 5DB, United Kingdom

office@ealaw.eu

 

This paper will discuss the highly controversial Hinkley Point C nuclear power station project from the stand point of EU state aid rules.
Part 1 of the paper will examine the proposed financing structure for the project. The project is the first example of the UK's readiness to act as a single buyer of nuclear energy with the government now responsible for contracts supporting all new generation investments. Part 2 will discuss the EU state aid rules as they apply to nuclear industry as well as review the Commission’s decision of October 2014 approving UK’s proposed contract for differences state aid package. Part 3 will discuss the current status of the project including the legal challenge pending before the Court of Justice of the EU and the UN’s Economic Commission for Europe (UNECE) call for the project to be suspended for failure to comply with the Espoo Convention. Finally, part 4 will discuss the implications for Brexit for the project.






12.09.2017 17:40 Regulatory issues and legislation - plenary

Regulatory issues, legislation, sustainability and education - 1103

Transposition of the EU Basic Safety Standard Directive into the Slovenian Legal System

Andrej Stritar1, Igor Sirc1, Igor Osojnik1, Aleš Škraban1, Nina Jug2, Damijan Škrk2

1Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

2Ministrstvo za zdravje Uprava RS za varstvo pred sevanji, Ajdovščina 4 (p.p. 557), 1001 LJUBLJANA, Slovenia

andrej.stritar@gov.si

 

After many years of preparation on 5 December 2013 the European Union Council directive 2013/59/EURATOM laying down basic safety standards for protection against the dangers arising from exposure to ionising radiation (BSS Directive) was adopted. It is requesting from all the EU Member Countries to transpose its provisions into national legislations by 6 February 2018. Like all the other EU member countries we have initiated the process of transposition very early. The new BSS Directive is based on the similar directive from 1996 (Directive 96/29/Euratom) and is also integrating several other directives related to the use of sources of ionizing radiation (89/618/Euratom, 90/641/Euratom, 97/43/Euratom and 2003/122/Euratom).
As all above mentioned directives were already transposed into the existing Slovenian legal system, there was no need for major reshuffling. However, the new BSS Directive is introducing several new approaches, therefore some important adjustments in Slovenia had to be done. This article describes how the process of transposition was designed and implemented, what were the major challenges and what were the solutions.
The first step of the transposition was done already long before December 2013. Several years before that when the first draft of the future BSS Directive was still discussed in Brussels, we have prepared the first Table of Concordance (TOC), where for each article of the draft directive the corresponding article in the national legislation was identified. In cases where there were not yet appropriate legal requirements available in existing legal acts the provisional solution for their implementation was proposed.
After the adoption of the BSS Directive the TOC was reviewed again this time setting more concrete goals how and into which legal act to transpose individual articles and/or chapters of the BSS Directive. The transposition team was created from the members of Slovenian Nuclear Safety Administration and Slovenian Radiation Protection Administration. It was concluded that majority of requirements from the BSS Directive will be transposed into the revised Act on protection against ionizing radiation and nuclear safety, while some parts will go into subordinate legally binding acts.
Towards the end of 2016 the draft Act was open for public comment. All the comments received before the end of that year were considered and most of them also incorporated into the revised text. In the spring 2017 the Act is going through the inter-ministerial discussion, it will be adopted by the Government and is expected to be passed by the Parliament at the end of the summer. Subordinate legal acts are being prepared in parallel.
The major challenges we are facing because of the new BSS Directive are similar as in other member states: improvement of protection against radon and changed requirements for building materials and consumer products.






13.09.2017 08:30 Invited Thomas Walter Tromm

Invited lectures - 102

Reactor Safety Research at the Karlsruhe Institute of Technology (KIT)- investigations from design basis accidents up to source term modelling and rehabilitation strategies

Walter Tromm

Forschungszentrum Karlsruhe, Institute for nuclear and energy technologies (IKET) Hermann-von-Helmholtz-Platz-, Hermann-von-Heimholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany

walter.tromm@kit.edu

 

After the reactor accident in Fukushima (Japan) in 2011, the German parliament decided, based on a broad societal consensus, to terminate nuclear electricity production with the last nuclear power plant in Germany to be shut down in 2022. Nevertheless, the research programme NUSAFE, Nuclear Waste Management, Safety and Radiation Research performed by the three Helmholtz-Centers Forschungszentrum Jülich, Helmholtz-Center Dresden-Rossendorf and KIT, represents an integral part of national provident research and addresses the new challenges arising from these new constraints. Therefore, core competences on the internationally highest level of science and technology regarding nuclear safety research, together with an appropriate participation in international programmes and expert groups (e.g. IAEA, OECD NEA) is ensured.
Reactor Safety Research at KIT together with the two other Helmholtz-Centers includes research for the safe operation of nuclear power plants in Germany and abroad as well as design-basis and beyond-design basis accidents in nuclear power plants or nuclear facilities. This expertise in nuclear technology is maintained and further developed. It enables Germany to maintain its influence in international safety organisations. In terms of safety standards and design issues, German interests can be represented effectively. The further development and validation of simulation programmes required for this purpose include extensive experimental studies of accident phenomena in the primary circuit and the containment up to severe accidents with release of fission products to the environment. With the further development of severe accident integral codes such as ASTEC, risks are quantified and the scientific results are transferred to plant application and incorporated into models that define the basis of a uniform and harmonised emergency management.
Additionally, the scientific and technological basis for the safety assessment of innovative concepts, such as SMRs or facilities for the partitioning and transmutation strategy will be further explored mostly together with European partners, thus providing sufficient scientific know-how for a comprehensive evaluation of its feasibilities, advantages and drawbacks in the European context and world-wide.






13.09.2017 09:10 Severe accidents - plenary

Severe accidents - 401

CoreSOAR Update of the Core Degradation State-of-the-Art Report: Status of Code Reviews

Tim Haste1, Olivia Coindreau2, Florian Fichot2, Mohamed Torkhani3, Thorsten Hollands4, Henrique Austregesilo5, Gabor Horvath6, Leticia Fernandez-Moguel7

1Institut de Radioprotection et de Sureté Nucléaire, Bât. 702 Centre de Cadarache, BP 3-13115 Saint Paul lez Durance, France

2Multiple organizations possible, Unknown, Unknown, Slovenia

3EDF - R&D, 1, avenue du General de Gaulle, 92141 Clamart Cedex, France

4Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Forschungszentrum, Boltzmannstr. 14 , 85748 Garching, Germany

5Gesellschaft für Anlagen- und Reaktorsicherheit Forschungsgelände, Postfach 12-28, 85748 Garching b. München, Germany

6NUBIKI Nuclear Safety Research Institute, Konkoly-Thege Miklós út 29-33. building 6, 1121 Budapest, Hungary

7Paul Scherrer Institut Nuclear Energy and Safety Research Department, Reaktorstrasse, CH-5232 Villigen, Switzerland

tim.haste@irsn.fr

 

In 1991 the CSNI published the first State-of-the-Art Report on In-Vessel Core Degradation, which was updated to 1995 under the EU 3rd Framework programme. These covered phenomena, experimental programmes, material data, main modelling codes, code assessments, identification of modelling needs, and conclusions including the needs for further research. This knowledge is relevant to such safety issues as in-vessel melt retention of the core, recovery of the core by water reflood, H2 generation and fission product release.
In the following 20 years, there has been substantial progress in understanding, with major experimental programmes finished, e.g. the integral Phébus FP tests and the QUENCH series on reflooding degraded rod bundles, while others have many tests completed, e.g. LIVE on melt pool behaviour in the lower head. These is a similar situation regarding integral modelling codes such as MELCOR and MAAP (USA) and ASTEC (Europe), which encapsulate current knowledge in a quantitative way. After the two EC-funded projects on the SARNET network of excellence, now continuing in the NUGENIA association, it is timely to take stock of the knowledge gained both in the experimental and modelling areas.
The CoreSOAR project, in NUGENIA/SARNET, draws together the experience of 11 European partners, to update the state of the art in core degradation, over the next two years. The review of available data being well advanced, attention is now focussed on summarising the status of the main modelling codes covering in-vessel core degradation, which enable the knowledge gained to be applied to reactor accident situations in light water reactors. The codes considered here include MAAP, MELCOR and ASTEC as well as other independent codes such as ATHLET-CD and SCDAPSIM. The final CoreSOAR report will serve as a reference point for ongoing research programmes in NUGENIA, in other EU research projects such as in Horizon2020, and in CSNI, e.g. the Fukushima benchmark BSAF.






13.09.2017 09:30 Severe accidents - plenary

Severe accidents - 402

Progression Of Postulated Severe Accidents In The Wet Storage Pool Of NPP Gösgen-Däniken

Bernd Jaeckel1, Pascal Steiner2, Jens-Uwe Klügel2

1Paul Scherrer Institut, Villigen, CH-5232 Villigen PSI, Switzerland

2Kernkraftwerk Gösgen-Däniken, P.O.Box, CH-4658 DÄNIKEN, Switzerland

bernd.jaeckel@psi.ch

 

The wet storage building of the nuclear power plant Gösgen-Däniken is the first building for the long term wet storage of spent nuclear fuel in Switzerland. Due to the Swiss moratorium against the export of spent nuclear fuel in 2006 it was necessary to store the spent fuel locally at the nuclear power plant site. The prolongation of this moratorium in 2016 does not allow reprocessing of spent fuel in the next years. As a direct conclusion from this moratorium it was decided to build a wet storage pool at Gösgen-Däniken which entered operation in 2008 with a capacity of 504 fuel assemblies. In a second stage of completion another 504 fuel assemblies can be stored in the pool. Two cooling towers will be available for the cooling of the wet storage pool with a maximum cooling capacity of 1.5 MW after the second stage of completion.
The present work describes postulated severe accidents with total loss of cooling in the wet storage pool (WSP) of the Gösgen-Däniken nuclear power plant. The accident progression in a spent fuel pool and even more in a WSP is very slow due to the low nuclear decay heat. Therefore the investigation of such kind of accident was not assumed to be of high importance for a long time. The overall accepted assumption of an energy recovery after at least 24 hours would not lead to boiling of a spent fuel pool due to the low decay heat.
The severe accident in Fukushima Dai-Ichi (11.03.2011) following the Tohoku earthquake and the related tsunami has shown that the power recovery can be delayed much longer than 24 hours, even for more than a week. This long time of loss of cooling could endanger a spent fuel pool especially if the whole core is unloaded into the pool like in Fukushima Dai-Ichi unit 4 at the time of the accident.
Main goal of the work is to achieve information about the timing of postulated severe accidents in the WSP and to define time frames for accident management measures to prevent fuel damage and fission product release.






13.09.2017 09:50 Severe accidents - plenary

Severe accidents - 403

Penetration Tube Failure Experiments for APR1400 and Fukushima Daiichi Nuclear Power Plant

Sang Mo An1, Hwan Yeol Kim1, Jin Ho Song1, Masanori Naitoh2

1KAERI (Korea Atomic Energy Research Institute), 989-111 Daedeok-daero, Yuseong-gu, 305-353 Daejeon, South Korea

2IAE (The Institute of Applied Energy), Shimbashi SY Building, 14-2 Nishi-Shimbashi 1-Chome, Minato-ku, Tokyo, 105-0003, Japan

sangmoan@kaeri.re.kr

 

For the LWRs (light water reactors), a few dozen to several hundred penetration tubes are located at the RPV (reactor pressure vessel) lower head. From the PRV failure point of view during a severe accident, these tubes have inherent structural weakness in that they are attached to the inside of the reactor lower head by a partial weld with a small gap between the tube and reactor vessel. Thus the penetration tube can be ejected out of the RPV if the tube weld failure occurs and RCS (reactor coolant system) pressure is high enough to overcome the binding shear force at the gap. In addition, molten corium can be discharged out of the RPV through several holes inside the tubes when they are damaged by molten corium and the melt flows into the holes. That is, the penetration tube failure determines success or failure in the IVR (in-vessel corium retention) strategy and provides the initial boundary condition regarding the melt jet discharge in the EVR (ex-vessel corium retention). Therefore, the penetration tube failure is one of the most important research subjects in the severe accident mitigation strategies.
This paper introduces the experimental works on the penetration tube failure for APR1400 and Fukushima Daiichi NPPs (nuclear power plants). 61 ICI (in-core instrumentation) tubes are installed at the APR1400 reactor lower head, in which several sensors are inserted to monitor the in-core status such as core exit temperature and neutron flux. For the Fukushima Daiichi NPPs, more than one hundred penetration tubes are installed such as ICM-GT (in-core monitoring guide tube), CRGT (control rod guide tube) and drain tube. In general, BWR (boiling water reactors) has more penetration tubes than the PWR (pressurized water reactor) at the reactor lower head due to the structural complexities at the reactor vessel upper head. Thus, the research on the penetration tube failure is of higher importance in a BWR.
Several penetration tube failure experiments were performed for the ICI tubes of APR1400, and ICM-GT and CRGT of Fukushima Daiichi NPP. The objectives are to investigate the penetration tube failure mechanisms experimentally, develop a penetration failure model, and validate the SAMPSON code. All the penetration tube specimens have similar features that one penetration tube is located at the center of a unit reactor vessel wall, and they were manufactured according to the real manufacturing process with the same materials and dimensions. About 2500°C corium simulant (ZrO2) or prototypic melt (ZrO2 and UO2 mixture) were generated in a cold crucible by induction heating and interacted with the penetration tube specimens. The vertical temperature distributions of the reactor vessel and tube were measured during the melt-specimen interaction. The penetration tube weld failure was observed in the APR1400 ICI tube; however, the tube ejection did not occur in the present experimental conditions. A melt discharge phenomenon was observed for the ICM-GT tube, while several holes inside the CRGT tube was found to be blocked with solidified melt ingot partially or completely.






13.09.2017 10:10 Poster session - GREEN

Thermal-hydraulics - 215

Comparison of pool scrubbing simulations with POSEIDON-II experiments

Matic Kunšek1, Ivo Kljenak2, Leon Cizelj1

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 Ljubljana, Slovenia

matic.kunsek@ijs.si

 

During a hypothetical severe accident in a nuclear power plant, the reactor fuel could melt and there is a possibility, that some of the radioactive material could be released as particles to the surrounding area. The highest risk of contamination comes from particles with diameter below 1 µm. The releases of the radioactive material can be reduced with the application of pool scrubbing, where the release of contaminated gases is filtered thought a pool of liquid water. The interfacial particle transfer from gaseous bubbles to liquid water removes most of the particles from gases and greatly reduces the amount released to the environment.
Some challenges and specifics of pool scrubbing modelling in nuclear power plants are their sheer size and the complexity of interactions. Because of the complexity of interactions between particles and fluids and their behavior in different flow regimes, knowledge in the field, both phenomenological and predictive, is still lacking despite decades of research.
The POSEIDON II experiment is located at Paul Scherrer Institute in Switzerland. The experiment facility consists of DRAGON system (Diverse purpose aerosol generation facility) and POSEIDON tank. With the DRAGON system, particles of different compositions can be generated and mixed with gas composed of steam and nitrogen. The POSEIDON tank is a cylindrical vessel which is 5 m high and has 1 m diameter. The design of the tank allows the changes of the injector and its alignment prior of tests. The goal of the experimental facility is to examine the dependence of the decontamination factor on water height, carrier gas steam mass fraction, particle diameter and temperature dependencies.
In the proposed paper, numerical simulations of pool scrubbing in OpenFoam solver reactingMultiphaseEulerFoam are performed, using Euler-Euler approach. First, the pool scrubbing and its four main features are presented. Then, the POSEIDON-II experiments are described. In the end, the results of numerical simulation are analyzed and compared with the experimental results.






13.09.2017 10:10 Poster session - GREEN

Thermal-hydraulics - 216

Application of new turbulence modeling in stratified flow of PTS

Mohsen Ghafari1, Mohammad Bagher Ghofrani1, Francesco Saverio D'Auria2

1Sharif University of Technology, Azadi Ave, Tehran, Iran

2University of Pisa, Via Montebello di Mezzo 17, 19020 Bolano, Italy

ghaffari@energy.sharif.ir

 

High neutron flux in core of a nuclear reactor can affect the material of Reactor Pressure Vessel (RPV). The neutron radiation has a detrimental impact on the mechanical properties of the RPV material such as hardening (or embrittlement) while neutrons are absorbed by the material. A major concern in embrittled RPVs is propagation of critical flaw which may cause through-wall cracks. Some transients leading to overcooling of RPV intensify the propagation of the cracks and known as Pressurized Thermal Shock (PTS). Such situation could be created in case of Emergency Core Cooling System (ECCS) actuation which leads to injection of cold water into the cold leg of the primary loop in some accidents, e.g. Loss Of Coolant Accident (LOCA). The investigation of PTS is performed throughout three steps including Probabilistic Safety Assessment (PSA); thermal-hydraulics and structural mechanics i.e. fracture mechanics. The final goal of the thermal-hydraulics analysis step is the prediction of the imposed temperature gradient on the downcomer as a result of single-phase and two-phase phenomena. Depending on the leak size, its location and the operation condition of the plant after LOCA, water in the cold leg can be in single-phase or two-phase condition. Two-phase PTS will occur when steam and water are in the cold leg during the injection of ECCS water. The ECCS water enters the hot steam flow environment in the cold leg and two phase stratified flow propagates in the cold leg by means of density difference between water and steam. Condensation of steam and mass transfer between two phases is the only heat source in this zone and the accurate modeling of DCC plays a significant role to predict temperature profile in downcomer. The interfacial heat transfer coefficient is defined according to eddy contact time with both steam and water at interface which is a function of turbulence characteristics. The high gradient of velocity generates too high turbulence when differential eddy viscosity models are used and some modifications should be considered in turbulence models at the interface. Implementation of turbulence damping function in turbulence eddy frequency transport equation is one of this modification. Although the new source function improves velocity profile of smooth stratified flow, but significant deviations reveal when the vertical motion of the interface is considerable. Also, the value of turbulence kinetic energy decreases substantially by employment of damping function without the other modification. The reduction of turbulence kinetic energy at the interface changes the value of heat transfer coefficient. In this paper, the other source function of turbulence is proposed by consideration of different boundary condition at the interface and the effect of turbulence characteristics on condensation rate is demonstrated. The results show that implementation of damping function, without any special treatment of turbulence kinetic energy, leads to a considerable overestimation of condensation rate which would be improved with employment of proposed turbulence source function at the interface. then, the new turbulence model is used for stratified flow zone in PTS scenario in VVER-1000 RPV to find the effect of interfacial heat transfer coefficient on temperature profile in the cold leg. For this purpose, the plant response to LOCA is simulated by RELAP system code until the injection of ECCS water. The results of RELAP simulation before injection point of ECCS is considered as input of CFD part. CFX code is employed for mixing and stratification zone of cold leg needing 3D nodalization for prediction of temperature profile.






13.09.2017 10:10 Poster session - GREEN

Thermal-hydraulics - 217

Modeling of NEK Main Steam Line Break in Intermediate Building with Computer code Apros 6

Jure Jazbinšek

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia

jure.jazbinsek@zel-en.si

 

A Model of Krško Nuclear Power Plant (NEK) Intermediate Building (IB) was developed in computer code Apros 6. In order to validate the model, simulations performed in Apros 6 environment were compared to similar calculation results made with the coupled RELAP5 and GOTHIC codes.
A double-ended Main Steam Line Break (MSLB) in IB model was simulated in Apros 6, representing best estimate analysis of transient. The break was assumed between Steam Generator (SG) outlet and Main Steam Isolation Valve (MSIV), so the blowdown of affected SG could not be prevented. In first part, Apros stand-alone IB model results was compared to selected reference mass energy release vectors used in GOTHIC calculations. After stand-alone IB model was validated to GOTHIC results, Stand-alone IB model was integrated into large Apros model, for simulation of primary and secondary systems response of NEK model to MSLB transient. MSLB energy release in coupled Apros model of NEK was adjusted on the point of break to achieve similar mass release parameters than previous GOTHIC reference vectors, previously used to validate stand-alone model of MSLB in IB.
Simulated results of Apros MSLB in IB transient present pressures, temperatures and break flows in IB rooms. Large coupled model presents additional comparison of multiple NEK primary and secondary system response parameters to MSLB transient for first 5000 seconds.






13.09.2017 10:10 Poster session - GREEN

Thermal-hydraulics - 218

Development of CFD models and pre-test calculations for thermal-hydraulics and freezing experiments on Lead coolant

Matteo Iannone1, Walter Borreani2, Ivan Dofek1, Guglielmo Lomonaco2, Vincent Moreau3

1Research centre Rez, Hlavni 130, 250 68 Husinec-Řež, Czech Republic

2INFN, Largo Enrico Fermi, 2, I-50125 Firenze, Italy

3CRS4, Polaris Edificio, CP25, 09010 Pula (CA), Italy

matteo.iannone@cvrez.cz

 

Heavy Liquid Metals (HLM) are objects of interest in the nuclear research sector because of their optimal thermal and neutronic properties; the development and the validation of models allow to predict their behaviour and are fundamental for the future development of the Generation IV energy systems.
An experimental facility named Sesame-stand, is planned to be operated in the second half of 2017 at Research Centre Rez (RC-Rez) under the framework of Sesame project. The aim of the facility is to study the solidification of Lead under GEN-IV Heavy Liquid Metal pool type nuclear reactors relevant conditions and to provide a database for the benchmarking and validation of numerical models. Corresponding CFD models are developed using commercial software and are used for the pre-test assessment and to support the experimental work. The aim of this paper is to describe the CFD models, explain how they are tested and used in order to define a valuable experimental matrix that will be needed in order to run the facility itself.
First of all the facility is introduced together with the range of foreseen investigations. The numerical models are then presented. Emphasis is given to the geometrical and physical assumptions. Different approaches of modelling are compared and discussed. Results from the pre-test simulations are illustrated. Encountered challenges and their relevance with regard to the experimental matrix and setup are commented.






13.09.2017 10:10 Poster session - GREEN

Thermal-hydraulics - 219

Semiscale Natural Circulation Tests Calculations

Andrej Prošek1, Catur Febriyanto Sutopo2, Azwidovhiwi Emmanuel Nengudza3

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Nuclear Energy Regulatory Agency (BAPETEN) Directorate for Licensing of Nuclear Installation and Nuclear Materials, JI. Gajah Mada No.8, Jakarta 10120, Indonesia

3Assessment Group, National Nuclear Regulator, P.O.Box 7106, 0046 Centurion, South Africa

andrej.prosek@ijs.si

 

Tests S-NC-2 and S-NC-3 were used to assess the RELAP5/MOD3.3 Patch05 computer code. The code developers reported that the calculated mass flow rate showed a slower increase with decreasing mass inventory and that major factor appeared to be related to the interphase drag for the S-NC-2 test cases. Therefore the purpose of the present study was to perform sensitivity study for RELAP5/MOD3.3 Patch 05 best-estimate system thermal-hydraulic code by varying interphase drag in the primary system.
The natural circulation experiments were performed in the Semiscale Mod-2A test facility, which is a small-scale model of the primary system of a four-loop pressurized water reactor (PWR). The facility incorporates the major components of a PWR including steam generators, vessel, downcomer, pumps, pressurizer, and loop piping. The tests selected were Semiscale natural circulation tests S-NC-02 and S-NC-03. The S-NC-02 tests simulated were performed at 60 kW (6% of full Semiscale core power). The objective of the steady-state separate effects S-NC-02 natural circulation test was to study thermal hydraulic response during the three modes of natural circulation: single-phase, two-phase, and reflux. The secondary side conditions were constant, while on primary side the mass inventory was varied. The S-NC-3 tests were performed at a core power of 62 kW varying steam generator secondary side mass inventory. The objective of the test was to study the effect of different steam generator secondary condition on two-phase natural circulation.
For sensitivity calculations the latest RELAP5/MOD3.3 Patch computer code has been used. The ASCII input deck was obtained in the frame of RELAP5 code distribution for the auto validation purposes. The nodalization view of input deck was prepared using Symbolic Nuclear Analysis Package (SNAP). In addition, SNAP graphical user interface animation masks have been created to better understand the influence of varying interphase drag on natural calculated physical phenomena and processes. The results of the sensitivity study will be presented showing the influence on the result when comparing to experimental data.






13.09.2017 10:10 Poster session - GREEN

Thermal-hydraulics - 222

Building an Experimental Apparatus for Advanced Heat Transfer and Fluid Flow Studies during Convective Single-Phase and Two-Phase Flows

Marko Matkovič1, Martin Draksler2

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

marko.matkovic@ijs.si

 

New experimental test facility for heat transfer and fluid flow studies during single-phase and two-phase flows in different systems and on different scales has been designed and is currently under construction at Reactor Engineering Division of Jozef Stefan Institute. There, the scientists strived for reliable experimental measurements on CFD scale, which would help them understand and model local heat transfer and fluid flow governing mechanisms. It is, however, no easy task to obtain these kind of data nor build such a device. In this context, building process of the apparatus with emphasizes laid on key decision criteria during the designing phase are explained and discussed in the present paper. With the goal to operate in wide range of test conditions and to acquire experimental results at the highest accuracy levels, the main design parameters were set. The first implies, but is not limited to, the use of various working fluids at broad range of mass flow rates, control over the diabatic wall heat flux, wall temperatures and variable gravity to flow direction angles. It enables instantaneous supply of the test section with the subcooled liquid and superheated vapor at arbitrary thermodynamic states. Besides, the design offers an insight into the velocity field within the boundary layer at the diabatic wall where local heat transfer coefficient is measured. The second, though, infers the use of highly accurate instruments, adoption of appropriate state-of-the-art measuring techniques and thoughtful run of experiments. As the construction of the apparatus is approaching the completion, the forthcoming preliminary experiments are about to judge the success of the designed apparatus described herein.






13.09.2017 10:10 Poster session - GREEN

Materials, integrity and life management - 307

Effect of Power Spectral Density Profile on Fatigue Predictions of Pipes under Turbulent Fluid Mixing

Oriol Costa Garrido, Samir El Shawish, Leon Cizelj

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

oriol.costa@ijs.si

 

Thermal fatigue has been the cause of primary water leakages in the safety related piping of nuclear power plants. In several cases, the structural damage due to thermal fatigue evolved in T-junction piping where fluids at different temperatures turbulently mixed. Under these circumstances, the frequency content of the temperature fluctuations near the pipes is an important factor driving the fatigue damage. The power spectral density (PSD), computed with temperature readings from experiments and simulations of the mixing fluids, is used to characterize the temperature fluctuations in the frequency domain. It is known that the turbulent temperature fluctuations have a multi-frequency and variable amplitude content and that higher amplitudes are located in the low frequency range. The measured PSD is typically roughly described by the theoretical PSD derived from the theory of turbulent flows. The main differences between the two, such as peaks in the frequency range between 1 and 10 Hz and the lower PSD levels in the frequencies above 10 Hz, have been measured but are not represented in the theoretical PSD.
The aim of this paper is to analyze the influence of different PSD profiles on the thermal fatigue predictions of pipes. The fatigue assessment employs the improved spectral method, for the generation of synthetic temperature histories, and the simplified one-dimensional pipe model, for the heat transfer and mechanical analyses. The fatigue analyses follow the codified ASME rules for fatigue design together with the rainflow counting method. The life predictions are moreover compared with those with the sinusoidal (SIN) method which assumes single frequency temperature fluctuations. The results of the analyses show that shorter fatigue lives are obtained for peaks in the PSD located at frequencies below 10 Hz and that the most damaging peaks are located at about 1 Hz. Consistent with the SIN-method results, the analyses in this paper also verify the ability of rainflow method to identify damaging events in complex stress histories.






13.09.2017 10:10 Poster session - GREEN

Materials, integrity and life management - 308

Zirconium Material Science Studies in Hungary

Emese Slonszki1, Zoltan Hozer1, I. Groma2, Gy. Gémes3, Gabor Lajtha4

1MTA Centre for Energy Research, Konkoly Thege ut 29-33, Budapest-1121, Hungary

2Eötvös Loránd University, Egyetem tér 1-3, H-1053 Budapest, Hungary

3TÜV Energy and Systems Technology, Baden - Wüerttemberg, Gottlieb-daimler-str. 7, D-70794 FILDERSTADT, Germany

4NUBIKI Nuclear Safety Research Institute, Konkoly-Thege Miklós út 29-33. building 6, 1121 Budapest, Hungary

emese.slonszki@energia.mta.hu

 

Nowadays more than 400 nuclear power reactors operate and most of them use zirconium cladding for fuel. The cladding is a barrier which prevents the release of radionuclides from the pellet to the environment. Operation experiences of reactors demonstrate that these zirconium alloys reliably perform the protection. At the same time the neutron radiation, the high temperature and pressure result structural changes in the cladding material. Using new methods and equipment more detailed knowledge can be obtained about the structural changes of the material in this research program.
A consortium established by the Hungarian Academy of Sciences Centre for Energy Research, NUBIKI Nuclear Safety Research Institute Ltd., TÜV Rheinland InterCert Ltd. and Eötvös Loránd University received a grant from the National Research, Development and Innovation Office of Hungary. The contract supports a three-year research project started in January 2017. The primary purpose of the project is the evaluation of the effects of structural changes in zirconium alloys used in nuclear power plants and investigation of their consequences on the integrity of fuel elements and environmental load.
The research program consists of four main parts:
• Experiments will be performed with zirconium claddings in order to simulate the loads that may take place during normal reactor operation and accident conditions. The measurement program includes oxidation, hydriding of different Zr specimens as well as zirconium tube irradiation in the Budapest Research Reactor.
• Mechanical and material structural tests will be performed to identify the mechanisms that can lead to the loss of cladding integrity.
• Using the measured data numerical models will be developed in order to predict changes in the cladding under operating conditions, accidents and storage.
• New instrument will be developed which can be used for X-ray diffraction testing of irradiated zirconium cladding for checking the integrity of the zirconium alloy.
Results of Zirconium Material Science Studies (CAK) contribute to the safety of using and handling of nuclear fuels under operating condition, during spent fuel storage and deposing in final repository.
The present work is supported by the National Research, Development and Innovation Fund of Hungary (contract number: NVKP_16-1-2016-0014).






13.09.2017 10:10 Poster session - GREEN

Materials, integrity and life management - 309

Polycrystalline simulations of an irradiated stainless steel: a comparison between Finite Element and Fast Fourier Transforms based simulations

Pierre-Guy Vincent1, Hervé Moulinec2, Samir El Shawish3, Leon Cizelj3

1IRSN - Institut de radioprotection et de sureté nucléaire, Nuclear Safety Division , BP17, 92262 Fontenay-aux-Roses Cedex?, FRANCE, France

2Laboratoire de Mécanique et d’Acoustique CNRS, 4 impasse Nikola Tesla, 13453 Marseille, France

3Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

samir.elshawish@ijs.si

 

This work concerns a comparison between Finite Element Method (FEM) simulations and Fast Fourier Transforms (FFT) based simulations on an irradiated austenitic stainless steel (SA304L) currently used as internals in nuclear reactors. The main objective was to model the elasto-viscoplastic behaviour of this material with 3-dimensional numerical full-field simulations of polycrystals. The constitutive law for the grains of the polycrystals of [X. Han, PhD. Thesis, Ecole Nationale Supérieure des Mines de Paris (2012)] has been used together with the parameters of [J. Hure et al., Journal of Nuclear Materials 476 (2016) 231-242] for the selected steel at 330°C and 13 dpa. JSI have performed FEM simulations with Abaqus software and IRSN-LMA have performed FFT simulations with CraFT software. Polycrystalline aggregate models have been generated upon Voronoi tessellations with periodicity in all 3 directions of the space and random grain orientations. Simple tension with fixed strain rate has been imposed. A good agreement has been observed between FEM and FFT results on macroscopic tensile curves and also on grain averaged stresses. This work is performed as a specific contribution in the frame of the general SOTERIA H2020 research project*.

*SOTERIA – Safe long-term operation of light water reactors based on improved understanding of radiation effects in nuclear structural materials – is a research project in the field of nuclear safety. This project received funding under the Euratom research and training programme 2014-2018 under grant agreement N° 661913.






13.09.2017 10:10 Poster session - GREEN

Severe accidents - 420

Simulation of a Low-Momentum Steam Jet Interaction with a Light Gas Layer in a Containment Facility

Rok Krpan1, Ivo Kljenak2

1Jožef Stefan Institute, Reactor Engineering Division, Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 Ljubljana, Slovenia

rok.krpan@ijs.si

 

During a severe accident, a hydrogen explosion could threaten the integrity of the nuclear power plant containment, which could lead into release of radioactive material into the environment. Various experiments are performed to simulate physical phenomena occurring in containment during severe accidents and results are used to validate Computational Fluid Dynamics (CFD) codes in order to simulate phenomena in actual power plants.
Within the OECD project SETH-2, which lasted from 2007 to 2010, experiments of erosion of a helium gas layer with a vertical steam jet were performed in the PANDA containment facility at Paul Scherrer Institute (PSI) in Villigen (Switzerland). The PANDA facility, used in this experimental campaign, consists of two cylindrical vessels and an interconnecting pipe, with a total volume of 183 m3. The interaction of axisymmetric low momentum steam jet on a previously established horizontal layer of helium-steam mixture in the upper part in one of the vessels was observed. Six experiments performed in the ST1 experimental campaign with different jet velocities and helium concentrations were simulated using the open-source CFD code OpenFOAM v1606+.
A two-dimensional mesh using a wedge boundary condition and a three-dimensional axisymmetric numerical model of a single cylindrical vessel of the PANDA facility were developed. Simulation results (essentially local helium concentrations and atmosphere temperatures) from both models are compared with experimental results. Results obtained with the k-Epsilon turbulence model with additional buoyancy term implemented in OpenFOAM with default model constants are not satisfactory enough, so the model constants were adjusted to achieve better agreement between simulation and experimental results.






13.09.2017 10:10 Poster session - GREEN

Severe accidents - 421

Influence of Metal Corium Oxidation on Ex-vessel Steam Explosion

Tomaž Skobe1, Matjaž Leskovar2

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

tomaz.skobe@ijs.si

 

A steam explosion may occur, when during a severe reactor accident the corium melt comes into contact with the coolant water. Steam explosions are an important nuclear safety issue because they can potentially jeopardize the primary system and the containment integrity of the nuclear power plant.
In the paper the influence of the metal corium oxidation on the ex-vessel steam explosion is investigated. A PWR ex-vessel steam explosion study was carried out with the MC3D computer code. Premixing and explosion simulations were performed, varying the metal corium oxidation rate. The explosion was triggered at the time of melt bottom contact. The premixing simulations were performed with the global jet breakup model. Based on experimental findings, the hydrogen film hypothesis was applied. The hypothesis proposes that only a limiting amount of zirconium may be oxidized in sub-cooled conditions during the premixing phase, while the rest of non-oxidized zirconium is available for the oxidation during the explosion phase. With the comparison of the simulation results, the influence of the oxidation of the melt on the strength of the steam explosion was analysed.






13.09.2017 10:10 Poster session - GREEN

Severe accidents - 422

Summary of the Nordic collaboration: Impact of air radiolysis, fission product and control rod species on the transport of ruthenium in the primary circuit of NPP in a severe accident

Teemu Kärkelä1, Ivan Kajan2

1VTT Technical Research Centre of Finland, Tietotie 3, Espoo, 02044 VTT, Finland

2Chalmers University of Technology, Kemirägen 4, SE-41296 Goeteborg, Sweden

teemu.karkela@vtt.fi

 

When ruthenium is released from the fuel to the environment in a severe NPP accident, ruthenium isotopes 103Ru and 106Ru with half-lives of 39.35 days and 373.5 days, respectively, cause a radiotoxic risk to the population both in a short and long term. The transport of ruthenium through a reactor coolant system, after being released from the fuel, has been investigated in several experimental programmes recently. The VTT Ru transport programme has shown that the release rate of Ru from RuO2 powder was dependent on the oxygen partial pressure, as well as temperature, in air-steam atmospheres at 827, 1027, 1227 and 1427 °C. The highest fraction of gaseous RuO4 at the outlet of the model primary circuit was observed at 1027 °C oxidation temperature. At higher temperatures, ruthenium was transported mainly as RuO2 aerosol. In the experiments the partial pressure of RuO4 reaching the outlet of model primary circuit was in the range of 10-7 to 10-6 bar, which is significantly higher than what is expected based on thermodynamic equilibrium calculations.
As the previous studies have mainly been conducted in pure air-steam atmospheres, the current Nordic study was dedicated to air ingress conditions with representative airborne air radiolysis (N2O, NO2, HNO3), control rod (Ag) and fission product (CsI) species which were mixed with vaporized Ru oxides. The aim was to study the impact of these additives on the transport of ruthenium as gas and aerosols through the primary circuit of nuclear power plant in a severe accident (SA). As a main outcome, the transport of gaseous ruthenium compound through the facility (heated up to 1027, 1227 and 1427 °C; outlet at 30 °C) increased significantly when the oxidizing NO2 gas was fed into the atmosphere when compared to the pure air-steam atmosphere. A notable increase in the transport of gaseous compound was also observed with the HNO3 feed even at the highest temperature. Introduction of N2O into the atmosphere led to a decrease of gaseous ruthenium transport through the facility as well as to an increased fraction of ruthenium transported in the form of aerosols at 1027 °C and 1227 °C. Similarly, the feed of pure silver particles into the gas flow showed an immediate decrease in gaseous RuO4 reaching the outlet of the facility. Simultaneously, an intense increase of ruthenium in form of RuO2 trapped on the filter was observed. The feed of CsI into the flow of ruthenium oxides had a strong effect on the thermodynamic equilibrium of Ru species. The transport of gaseous ruthenium compound (10-5 bar) was the highest ever observed with this facility, whereas the aerosol transport of ruthenium decreased significantly.
Based on experiments it was concluded that the composition of atmosphere in the primary circuit will have a notable effect on the speciation of ruthenium transported into the containment building during a severe accident. These experimental observations should be considered when developing the SA analysis codes.






13.09.2017 10:10 Poster session - GREEN

Severe accidents - 423

Study on the Severe Accident Management Strategy against Over-Pressurization at PWRs in Korea

Yu Jung Choi1, Jaehwan Park2

1Korea Hydro & Nuclear Power Co. Ltd., 70gil, Yuseong-Daero 1312, Yuseong-Gu, 34101, Daejeon,, South Korea

2POWER, Izpolni naslov!, USA

yina.choi@gmail.com

 

The containment is the last barrier against releasing radioactive materials to the environment and the public during a severe accident. Integrity of a containment is likely to be threaten due to the containment over- pressurization, came from decay heat of fission products and massive steam, when a severe accident is progressed without the appropriate accident managements or any countermeasures. Following the Fukushima accidents, Korean government carried out safety inspection for all nuclear power plants in Korea. As a result of the safety inspection, nearly 50 items for safety enhancement (so called, ‘Post-Fukushima enhancement’) of operation Nuclear Power Plants(NPPs) were out. Among them, Filtered Containment Venting System(FCVS) was chosen as a countermeasure system for over-pressurization of containments at NPPs in Korea. Installations of FCVS for Pressurized Heavy Water Reactors (PHWRs) have been underway and for Pressurized Water Reactors (PWRs) are also in preparation. Also, the external coolant core injection systems have been installed at all NPPs in Korea as a new safety enhancement feature against severe accidents. In this study, preliminary evaluations of the FCVS operation strategy under severe accident conditions with the external coolant core injection were performed for establishment of severe accident management strategy appropriately for PWRs in Korea.






13.09.2017 10:10 Poster session - GREEN

Severe accidents - 424

Development and Validation of the Passive Autocatalytic Recombiner Analysis Module

Jaehwan Park1, Gilbeom Kang1, Yu Jung Choi2

1POWER, Izpolni naslov!, USA

2Korea Hydro & Nuclear Power Co. Ltd., 70gil, Yuseong-Daero 1312, Yuseong-Gu, 34101, Daejeon,, South Korea

park.jaehwan@khnp.co.kr

 

It is critical to maintain the integrity of the containment building at a nuclear power plant (NPP) during a severe accident. It is the last barrier to prevent from the releasing radioactive materials into the environment and the publics. Especially, after hydrogen explosion of Fukushima accident, it has been concerned that as one of the reasons to damage the containment wall. Passive Autocatalytic Recombiner (PAR) is one of the methods to reduce the hydrogen which is converted into water vapor composing with oxygen. Its characteristic, passive mechanism, does not need the active power to operate PAR under the situation without the electricity like station black out. PAR system is being installed or will be installed in the near future at all NPPs in Korea. With being installed the PAR system, PAR analysis module to evaluate the performance and results of PAR under severe accident conditions has also been developed. This module was initiated by Korea Hydro and Nuclear Power (KHNP) and is undergoing verification and validation, as well as further development.






13.09.2017 10:10 Poster session - GREEN

Research reactors - 510

Neutronic Analysis of Control Rod Effect on Safety Parameters in Tehran Research Reactor

Mina Torabi1, Ahmad Lashkari2

1K. N. Toosi University of Technology, 470 Mirhamad Ave. West, 19697, Tehran, Iran

2Atomic Energy Organization of Iran, KH Africa, K. Tandis, No. 7 P.O.Box 14155-1339, 19156 Tehran, Iran

minatorabi90@gmail.com

 

The measurement and calculation of neutronic parameters in nuclear research reactors has an important influence on control and safety of nuclear reactor. The power peaking factors, reactivity coefficients and kinetic parameters are the most important neutronic parameter for determining the state of reactor. The position of the control shim safety rods in the core configuration effects on these parameters depend. The main purpose of this work is using the MTR_PC code to evaluate the effect of the partially inserted of the control rod on the neutronic parameters at core No: 61 of the Tehran Research Reactor. The simulation results show that with increasing the amount of inserted control rods (bank) in the core, the absolute values of PPF, reactivity coefficients and effective delayed neutron fraction increased and only prompt neutron life time increased. The changes of Moderator Temperature Coefficients value verse the control rods positions was very significant. The average values of MTC increase about 98% in the range of zero to 70% inserted of control rods.






13.09.2017 10:10 Poster session - GREEN

Research reactors - 511

Characterization of neutron field in the TRIGA irradition facilities inside and outside the biological shield

Klemen Ambrožič1, Bor Kos2, Anže Jazbec3, Vladimir Radulovič2, Luka Snoj1

1Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

3Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

klemen.ambrozic@ijs.si

 

The "Jožef Stefan" Institute (JSI) TRIGA reactor is equipped with several larger, ex-core irradiation facilities located in the reactor biological shield, which are used for irradiation of larger electronic assemblies. Until now the irradiation positions close to the reactor core have been computationally characterized by analog (no variance reduction is used) Monte Carlo (MC) simulations. However calculations of neutron spectra and flux inside shielded irradiation facilities by analog MC simulations are computationally very intensive. In this paper we present our effort on decreasing the computational time for neutron characterization of irradiation facilities outside the reactor pool tank i.e. inside biological shield and elsewhere, by using the ADVANTG code for variance reduction of Monte Carlo neutron transport simulations using MCNP on a complete 3D model of the TRIGA reactor. The results presented will be compared and validated against neutron activation foil measurements in the respective irradiation positions.
In the past year, a complete 3D CAD model of JSI TRIGA Mark II was developed. The model consists of the complete reactor core including irradiation channels and nuclear instrumentation, complete reactor pool, biological shield and irradiation facilities inside it, reactor hall interior along with reactor basement and spent fuel pool, on basis of which a complete MCNP model with all the necessary data on materials was made. The model can be used to calculate various neutron parameters inside the reactor hall or reactor basement and even outside the reactor walls. In this particular case, the MCNP 3D model was used to characterize neutron field inside the dry chamber irradiation facility.
The Automated Variance Reduction Generator (ADVANTG) code is a Monte Carlo/Deterministic Hybrid transport code. The fundamental concept is to generate an approximate importance function from a fast-running deterministic adjoint calculation and use the importance map to construct variance reduction parameters that can accelerate tally convergence in the Monte Carlo simulation. The code has proven to be very ffcient in producing variance reduction parameters for various fission and fusion neutronic applications, but is currently limited to fixed source problems. In this paper it will be used to produce statistically relevant neutron fluences outside of the TRIGA biological shield where analog Monte Carlo simulations are not viable.
Due to the nature of ADVANTG code, being only applicable to fixed source computational models, attempts on redefining the neutron multiplication eigenvalue (k-code) problems into a fixed-source problems with different degrees of accuracy have been made. In this paper, efforts on redefining the JSI TRIGA k-code problem to fixed source, making use of most of the available MCNP fixed source parameters for higher accuracy, are presented.
The denition of a fixed source was performed per slice of an active part of fuel element (fuel meat), comprised of 100 slices. Fresh fuel is considered in the analog MCNP model, therefore a single fission source neutron spectrum has been applied. In order to construct a neutron source for deterministic calculations by ADVANTG code, fission rates of each fuel meat slice have to be calculated.






13.09.2017 10:10 Poster session - GREEN

Research reactors - 512

TRIGA Pulse Experimental Benchmark Database

Anže Pungerčič1, Luka Snoj2

1Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

anze.pungercic@student.fmf.uni-lj.si

 

In 1991, the research reactor TRIGA Mark II at the Jožef Stefan Institute was reconstructed and upgraded for pulse mode operation. With addition of a 4th, transient control rod, pulse operation was made possible due to a hydraulics system of fast rod ejection. In this kind of operation large positive reactivity is inserted and the reactor core becomes prompt-super-critical. This process results in a quick and intense reactor power rise from sub critical system (P ? 10 mW) to high power (P ? 500 MW). TRIGA reactors are especially convenient for pulse experiments, because of the homogeneous fuel mixture of uranium and hydrogen; therefore featuring prompt negative reactivity coefficient. One pulse only lasts a couple of milliseconds, depending on the inserted reactivity. During each pulse operation, information regarding reactor power and fuel temperature is recorded with a sample rate of 20 kHz. From 1991 more than 300 pulse experiments have been made, which gave us the motivation to construct a database containing all pulse parameters and recorded quantities that could be used for validation purposes.
The main goal of this paper is to present the pulse experimental benchmark database containing all information regarding each pulse for further analysis and studies. The parameters which were included in the database are the inserted reactivity, transient control rod calibration curve and initial position, energy released, fuel temperature, maximum power and FWHM of a pulse. Furthermore, from the analysis of TRIGA’s normal operational history, the burnup and in-core position of each element for every pulse experiment are also present in the database.
This gathered information enables us to study the adiabatic Fuchs-Hansen model and to validate improved models with additional parameters, where fuel temperature and fuel element position could be taken into consideration. Due to a high number of parameters extracted from the operational analysis, the database would be also useful for validation of Monte Carlo calculations of transients, a feature currently under development in the SERPENT Monte Carlo neutron transport code.






13.09.2017 10:10 Poster session - GREEN

Reactor physics - 619

TRANSURANUS Modelling of MOX Fuel for Fast Reactors: Oxygen Redistribution and its Effect on Fuel Thermal Conductivity

Rolando Calabrese1, Arndt Schubert2, Dragos Staicu3, Paul Van Uffelen2

1ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy

2European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Hermoltz-Platz 1, 76344 Eggenstein-Leopolshafen, Germany

3EC, Directorate-General Joint Research Centre Institute for Energy Safety of Present Nuclear Reactors Unit (SPNR) Plant Operation Safety, PO Box 2, 1755 ZG Petten, Netherlands

rolando.calabrese@enea.it

 

The design of FBR fuel pin is strongly influenced by the centerline fuel temperature and the margin to fuel melting. The behaviour of MOX fuel is affected among others by the oxygen content. The local O/M (oxygen to metal) ratio changes under irradiation because of two main factors: fuel burn-up, radial redistribution under temperature gradients. The O/M ratio is one of the parameters influencing the thermal conductivity and melting temperature of fuel. Moreover, it affects the chemical interaction between fuel and cladding. The role of the O/M ratio in determining the thermal conductivity of MOX fuel is more pronounced at the beginning of irradiation while other factors are more relevant at higher burn-up such as the diffusion of fission products towards the cladding.
The TRANSURANUS code is a well-known tool for the analysis of the fuel pin thermal-mechanical behaviour. In view of the potential development of innovative fast reactor systems, significant efforts have been devoted by the developers to refine and improve several aspects relevant for the description of MOX fuel such as the release of helium, and the redistribution of actinides under irradiation. A coupling between TRANSURANUS and MFPR to describe the migration of fission products is ongoing in the frame of a collaboration JRC-IRSN. The modelling of oxygen redistribution was also considered and an improved diffusion coefficient of oxygen has been proposed.
The TRANSURANUS modelling of oxygen redistribution and its relationship with the evaluation of MOX thermal conductivity at the first stages of irradiation are discussed in the paper. A detailed analysis of the experiments L5 and L6 performed in the SILOE reactor is presented. These experiments are focused on the analysis of the oxygen redistribution through a proper selection of the irradiation conditions that permitted to reduce the role of interfering phenomena. After irradiation, the radial profile of the O/M ratio together with other PIEs was measured. The uncertainties affecting the O/M ratio, gap width, and thermal conductivity are considered in the TRANSURANUS analysis. Finally, the results are discussed in the light of the studies on the thermal conductivity of unirradiated MOX fuel recently published in the literature.






13.09.2017 10:10 Poster session - GREEN

Reactor physics - 620

Uncertainty Analyses of Unprotected Transients in Fast Reactors from Reactor Physical Point of View

Bálint Batki1, Andras Kereszturi2, István Panka1

1Hungarian Academy of Sciences Centre for Energy Research, Budapest 114, P.O. Box 49, Hungary, H-1525, Hungary

2Hungarian Academy of Sciences, Centre for Energy Research , P.O. Box 49, 1525 Budapest 114, Hungary

batki.balint@energia.mta.hu

 

Fast spectrum reactors highly differ from thermal systems considering the reactivity coefficients. Reactivity feedbacks play important role during unprotected transients, where the reactor is expected to stabilize itself. Several neutron physical parameters are to be taken into account with uncertainties, to accurately determine the evolving temperatures during these transients. Uncertainty analyses show the relevant parameters and highlight the main sources of the target parameter uncertainty. The object of this study was to identify the important reactor physical parameters and the sources of the uncertainty on different target parameters during unprotected transients for two types of fast reactors.
The most important reactor physical parameters were calculated with the Serpent Monte Carlo code for the ALLEGRO demonstrational GFR core and for an SFR core with 3600 MWth power. Using the obtained coefficients of reactivity, point-kinetic parameters and power peaking factor, unprotected transient simulations were performed with the ATHLET 3.1A code in point-kinetic approach on simplified reactor core models. Uncertainty calculations were based on a statistical methodology, where the considered reactor physical uncertainties were taken from international benchmark results, but thermal-hydraulic uncertainties were ignored.
It was found that during unprotected overpower transients the most important uncertain parameters are the Doppler and the fuel thermal expansion coefficient of reactivity, the power, and the power peaking factor. In addition, the large uncertainty of the positive coolant and cladding temperature coefficient of reactivity have significant influence on the uncertainty of the cladding peak temperature in case of unprotected loss of flow transients. The results show that the uncertainty of some reactor physical parameters should be reduced.






13.09.2017 10:10 Poster session - GREEN

Reactor physics - 621

Xenon concentration determination in VVER type reactor

Áron Vécsi

Hungarian Academy of Sciences Centre for Energy Research, Budapest 114, P.O. Box 49, Hungary, H-1525, Hungary

vecsi.aron@energia.mta.hu

 

We have a good experience to operate VVER-440 type nuclear power plant with our own on-line core monitoring system. The Centre for Energy Research and the Paks Nuclear Power Plant made the system named VERONA in cooperation at the year 80’s. This system work well nowadays after many modernizations. Our plan is to build two new VVER-1200 type reactors. This reactor type will be bigger than the current ones and we may want to operate this in load following mode. Because of these we need to track the distribution of xenon concentration on-line.
I will show the main reasons that result deviation of xenon concentration in space and time. I made a program to calculate the xenon concentration with point kinetics equations with groups of the delayed neutrons are used. The model assumes feedback from lumped fuel and coolant temperatures. I will present the comparison of these simulations and the measurements of Paks Nuclear Power Plant (what are VVER-440 type).






13.09.2017 10:10 Poster session - GREEN

Reactor physics - 622

ASSEMBLY LEVEL HOMOGENIZATION USING MONTE CARLO METHOD

Dušan Ćalić

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia

dusan.calic@zel-en.si

 

The Serpent-GNOMER code sequence was validated for Krsko NPP using first fuel cycle in 2015. This is a three step sequence where we applied the Monte Carlo method to obtain the cell homogenized cross sections and to use them in fuel assembly calculations. The results showed good agreement between the Serpent-GNOMER simulation and the reference WIMSD-GNOMER calculation. This study also showed that the difference in computational time is about 3 orders of magnitude so for routine calculations the computational time must be shorter.
Shorter calculational time can be achieved using new approach that will be developed in the following years. In this new approach we will obtain the cell homogenized cross sections in fuel assembly calculations and to use them in full core nodal diffusion calculation. In recent years we have learned that it is very important to have a really good reference results. We will compare the results with MIT BEAVERS benchmark, which provides a detailed description of a commercial PWR with a lot of experimental measurements.
In this paper the results of 3D heterogeneous model of full core using Monte Carlo code Serpent 2 will be presented using zero power conditions and fresh core configuration.






13.09.2017 10:10 Poster session - GREEN

Reactor physics - 623

Comparison of Simulated and Actual Reactivity Parameters during a Reactor Power Transient at NEK

Barbara Grobelnik, Matjaž Božič, Bojan Kurinčič

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

barbara.grobelnik@nek.si

 

The controlled transient from nominal reactor power to HZP followed by a return to 100% reactor power was performed due to a maintenance activity. The reactor power transient was simulated and predicted with advanced nuclear code programs and compared to the actual measured reactivity parameters of the transient. The advanced nuclear code programs use on-line data from plant instrumentation to develop and provide information on the actual core conditions. The evaluation of simulated and measured data for the main reactivity control parameters as of boron concentration, control rods position, xenon concentration and axial offset is presented.






13.09.2017 10:10 Poster session - GREEN

Reactor physics - 624

Mathematical aspects of nuclear data covariance matrix preparation - an example of delayed neutron data

Solene Tarride1, Ivan Aleksander Kodeli2, Karl-Heinz Schmidt3, Pierre Dossantos-Uzarralde1

1École Nationale Supérieure d’Informatique pour l’Industrie et l’entreprise, 1, Square de la Résistance, F-91025 Évry, France

2Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

3Centre Etudes Nucléaires de Bordeaux Gradignan, 19 Rue du Solarium, 33170 Gradignan, France

ivan.kodeli@ijs.si

 

The Bayesian approach was used to derive the variance-covariance of the delayed fission yields of actinides based on theoretical model. The GEF code was used for theoretical calculations. The code calculates the distribution of delayed fission yield data as a function of input model parameters which are randomly varied according to the assigned standard deviation levels. The correlations between different isotopes and energy ranges were deduced from GEF runs using the same input parameters. Example of results in the form of standard deviations and correlation coefficients for U-238 and Pu-239 at few different energies will be presented.






13.09.2017 10:10 Poster session - GREEN

Nuclear fusion - 717

Combined Steady-State and Transient Plasma Impact on Tungsten Surface Behaviour

Oleksii Girka1, Ivan Bizyukov1, Maksym Myroshnyk1, Igor Garkusha2, Stanislav Herashchenko2, Vadim Makhlaj2, Sergii Surovitskiy3, Sergey Malykhin3

1V. N. Karazin Kharkiv National University, Kharkiv, Ukraine

2Nuclear Safety Consultacy, Akenwerf 35, 2317 DK Leiden, Netherlands

3Lviv's National Polytechnical University, Computer Science Departmet, 21, Ivan Trush Str, ap.2, Lviv 79057, Ukraine

oleksiigirka@karazin.ua

 

Lifetime of plasma-facing components (PFCs) defines work time of fusion reactors such as ITER and DEMO. It has been experimentally shown that the combination of transient heating and hydrogen plasma exposure lead to severe surface damage and modifications, such as crack formation, enhanced erosion/ejection, roughening, melt layers formation and motion [1]. The damage behaviour strongly depends on the loading conditions and the sequence of the particle and heat flux exposure. The stress-free surface demonstrates a high resistant to cracking under transient heat loads. For small number of plasma pulses combined with steady-state irradiation a faster relaxation of residual stresses occurs [3]. The damage of exposed surface was caused by physical sputtering and cracks appearing. Nevertheless, contribution of combined steady-state and transient heat and particles loads to damage of tungsten needs for further research in the course of large number of powerful plasma pulses.
Experimental studies on combined exposure of texturized tungsten samples were carried out. The damage features of the tungsten surfaces has been studied under combined exposure of steady-state hydrogen ion fluxes (2 keV, 1.7 MW•m-2, 1022 m-2s-1, 2×1026 m-2) generated by FALCON ion source [2] and the pulsed plasma loads generated by QSPA Kh-50 facility [1]. The energy of pulsed plasma loads (0.45 MJ/m2 and duration each pulse of 0.25 ms) was chosen to remain the surface temperature below the tungsten melting point.
Three Plansee tungsten samples with residual stresses about -350 MPa and texture (100), (100) and (110) were chosen for irradiation. Samples have been irradiated with combined steady-state and pulsed plasma exposes after preliminary SEM and XRD analysis. Steady-state irradiation results in full annealing of initial compression residual stresses. Pulsed QSPA plasma loads lead to creation of symmetrical thermal tensile residual stresses of +650 MPa in two samples with (100) texture. The third sample with texture (110) demonstrates tensile residual stresses of +125 MPa after irradiation by QSPA plasma. Such stress relaxation is caused by cracks and pores formation. The number of pores rises with increasing of steady-state fluence in to 3 times in the target with cracks. That is agreed with increase of number of vacancies complexes in irradiated surface.
Stationary heat loads cause slower relaxation of tensile residual stresses developed after numerous pulsed plasma loads in comparison to the relaxation after a small number of pulsed plasma loads (5 pulses of 0.45 MJ/m2) [3]. The combined plasma loads result in development of tungsten surfaces roughness, which is caused by cracks appearing and growth of grain edges on exposed surfaces. It has been shown that the tungsten structure could be improved by steady-state plasma irradiation with slight differences in structure evolution of samples with different textures.

[1] I.E. Garkusha, V.A. Makhlai, N.N. Aksenov, B. Bazylev, I. Landman, M. Sadowski, E. Skladnik-Sadowska. Tungsten Melt Losses under QSPA Kh-50 Plasma Exposures Simulating ITER ELMs and Disruptions. Fusion Science and Technology, 2014, Vol. 65, P. 186-193.
[2] O. Girka, I. Bizyukov, K. Sereda, A. Bizyukov, M. Gutkin, V. Sleptsov. Compact steady-state and high-flux FALCON ion source for tests of plasma-facing materials. Review of Scientific Instruments, 83, 083501 (2012)
[3] S.S. Herashchenko, V.A. Makhlaj, O.I. Girka, N.N. Aksenov, I.A. Bizyukov, S.V. Malykhin, S.V. Surovitskiy, K.N. Sereda, A.A. Bizyukov. Erosion Features of Tungsten Surfaces under Combined Steady-State and Transient Plasma Loads. Problems of Atomic Science and Technology. 2016, ? 6. Series: Plasma Physics (22), p. 69-72.






13.09.2017 10:10 Poster session - GREEN

Nuclear fusion - 718

Effect of DEMO Load Cases on Rectangular Bellows Design

Oriol Costa Garrido1, Boštjan Končar1, Richard Brown2, Christian Bachmann2

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2PPPT, PMU, EUROfusion Consortium, Boltzmannstrasse 2, 85748 Garching, Germany

oriol.costa@ijs.si

 

It is foreseen that the connection between the vacuum vessel ports and the cryostat in the DEMO fusion reactor will be performed with bellows expansion joints, named as cryostat bellows. One of the main functions of the cryostat bellows is to accommodate the relative displacements anticipated to occur between the two connected components due to thermal expansions or seismic motion, i.e. load cases. This is achieved by the deformation of the convolutions that form the corrugated walls of the expansion joints.
The paper presents the work performed in 2016 at Jožef Stefan Institute within “Initial definition of Cryostat Bellows” project, in the “Project Management and Integration” (PMI) work package of the “EUROfusion” framework program. It analyzes the dimensioning of the upper port cryostat bellows with the available standardized analytical procedures for rectangular bellows. To this end, a sensitivity analysis of possible load cases, materials and dimensions of the bellows is performed. The aim of the analyses is thus to find out, with the available analytical procedures, if there exist sets of bellows convolutions parameters for which the design constraints are fulfilled. At the same time, the final dimensions of the bellows should be accommodated within the available space in the current DEMO design. The results of the analyses show that the available sets of bellows parameters are indeed very dependent on the inputs, given the challenging dimensions and loadings of the DEMO bellows. With the design assumptions taken at this stage, it has been found that the high mass of some to the available bellows may actually represent an important design constraint for the cryostat bellows.






13.09.2017 10:10 Poster session - GREEN

Nuclear fusion - 719

Development of preliminary test acceptance procedures for components of the TIR area in DONES Facility

Petar Mateljak

INETEC-Institute for Nuclear Technology, Dolenica 28, 10250 Zagreb, Croatia

petar.mateljak@inetec.hr

 

In this paper, preliminary test acceptance procedures for components installed in TIR area of DONES Facility are presented. Procedures that will be used for inspection of vacuum pumps, valves and beam diagnostics are written in manner to cover all relevant tests that should be performed in order to obtain proper installation and operation of critical components in TIR area. As relevant nuclear standard, ASME Boiler and Pressure Vessel Code Section XI has been used, since it is commonly used in great majority of western nuclear power plants. According to ASME Code, plant components are divided in several examination categories and each category has specific testing requirements according to specification, function and risk assessment. Apart from non-destructive testing and ASME Code acceptance criteria, functional test of each component and TIR vacuum system as a whole, is proposed to ensure safe operation when Facility is started.
All components that will be installed in TIR area should comply the technical and safety standards of DONES Facility. For that reason, all components should be delivered with relevant documentation and certificates of factory acceptance tests in accordance with requirements given for each component. Moreover, site acceptance test should be performed by DONES Facility responsible personnel. Examination personnel should have level of expertise as requested in relevant international standards. Also, all activities should be recorded in accordance with requirements given in this document.
According to dose rate classification, all activities in TIR area, even during maintenance period, should be performed by remote handling system. For this reason, all components installed in TIR area should be designed in such manner to allow component handling, examination, repair and replacement by automated robotic system.

Keywords: DONES, TIR, non-destructive testing, vacuum pump, valve, bellow, diagnostic systems






13.09.2017 10:10 Poster session - GREEN

Nuclear fusion - 720

Development of DEMO thermal shield concept: design requirements and expected thermal loads

Boštjan Končar1, Oriol Costa Garrido1, Martin Draksler1, Richard Brown2, Christian Bachmann2

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2PPPT, PMU, EUROfusion Consortium, Boltzmannstrasse 2, 85748 Garching, Germany

bostjan.koncar@ijs.si

 

The main function of thermal shields is to minimize the thermal radiation load from the warm components, like vacuum vessel and cryostat to superconducting magnets operating at cryogenic temperature of about 4 K. Thermal shields (TS) system shall be robust and reliable in terms of long-term operation. Initial concept of DEMO thermal shields, based on the recent DEMO baseline design with 18 sectors, is presented in this study. Main TS functions, its design features and requirements are specified. These include also space requirements for TS installation that are based on thermal expansion study of relevant DEMO tokamak systems. The expected static heat loads on thermal shields and magnets at normal DEMO operating conditions have been evaluated using the validated theoretical model. The model considers thermal radiation and heat conduction loads through several physical supports. The results of thermal analysis show that thermal radiation from the vacuum vessel presents by far the largest contribution to the overall heat load on TS and magnets. The majority of thermal radiation is intercepted by the actively cooled TS system, hence the magnets are also strongly affected by the heat conduction loads, primarily through the gravity supports carrying the heavy toroidal field coils.






13.09.2017 10:10 Poster session - GREEN

Radiation and environment protection - 806

Evaluation of the radiological impact on the Italian territory of a severe nuclear accident at Krsko NPP by means of a statistical methodology

Antonio Guglielmelli, Federico Rocchi

Italian National Agency for New Technology, Energy and Substainable Economic Development, Via Martiri di Monte Sole, 4 - Bologna , 40129, Italy

antonio.guglielmelli@gmail.com

 

Italy’s four nuclear power plants (NPPs) are presently under decommissioning, but on its territory, four research reactors are still in operation; moreover, Italy is surrounded, at less than 200 km from the borders, by several foreign NPPs. This particular situation implies the need of having some level of anticipated preparedness in order to have a preliminary idea of the degree of the radiological impact, and its geographical distribution over the Italian regions, of a severe accident. ENEA, in order to implement an Emergency Preparedness and Response (EP&R) methodology, has recently adopted the latest approach developed in this field which consists of a series of statistical studies of the radiological impact by simulating several hundreds or thousands of identical accidents happening at different times ? making recourse to real meteorological data from the past years ? and then getting an average outcome of their consequences. This approach cannot give a precise answer to a real-time accident, nonetheless it is, at the present, the most powerful one in giving indications on the type of fixed and permanent countermeasures (i.e. number and location of radiation measurement stations, location and number of infrastructures for response) and for preparing the long-term recovery phase (i.e. preparation of sampling strategies, food ban prescriptions, decontamination procedures). The aforementioned statistical approach has been therefore adopted to perform a statistical analysis of the consequences of a severe accident at the Krsko NPP which is one of the closest NPPs to the Italian borders. The code used for the statistical study is ldx, a long-range, 3D, Eulerian atmospheric dispersion code specifically developed by the French IRSN, for which the ENEA-FSN-SICNUC Division has signed a specific bilateral cooperation agreement. The meteorological dataset, defined over a geographical domain with a 50 km spatial resolution, covers a period of ten years, namely 2002-2011. In order to simplify the calculations and to allow comparisons, the transported Source Term (ST) consists of only one isotope, namely Cs-137, of relevant radiological impact, evaluated with the fast-running code RASCAL 4.3. The ST values chosen for the simulation were 5E+15 and 5E+16 Bq with a release dynamics of both 1 hour (Puff type) and 72 hours (Unit type) of emission. The physical quantity involved is the ground deposition cumulated at the end of the simulation which has been compared with the Italian legislation equivalent threshold limits. It was also performed a preliminary RASCAL 4.3 standard calculation of I-131 integrated air concentration whose values were used to evaluate the tyroid dose equivalent, than compared to that obtained by the Italian Civil Protection in the context of the national plan of protective measures in case of the radiological emergencies.






13.09.2017 10:10 Poster session - GREEN

Radiation and environment protection - 807

Draft Version of the DOCPAGANSA GUI and Control Module

Krešimir Trontl, Mario Matijević, Dubravko Pevec, Ivan Mihaljević

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia

kresimir.trontl@fer.hr

 

The DOCPAGANSA (“Development of Code Package for Advanced Gamma and Neutron Shielding Analyses”) research project aims to develop an integrated program environment that would enable the shielding designer to employ simplified, as well as advanced computational procedures and codes. The simplified code is based on the point kernel method, which is a fast numerical integration technique utilizing concept of buildup factors for neutrons and photons. The advanced procedure is based on hybrid shielding methodology, designed to optimize the final Monte Carlo (MC) answer using deterministic adjoint transport theory method of discrete ordinates (Sn). Such computational environment will allow the user to make preliminary radiation shielding calculations, resulting in only the approximate answers, and/or to proceed to the final calculations involving detailed calculations. An appropriate graphical user interface and control module are required to enable efficient shielding project management and interaction between simplified and complex calculations. In this paper we report on the preliminary version of these two program modules.






13.09.2017 10:10 Poster session - GREEN

Radioactive waste management - 905

Interactive Assessment of the Radioactive Waste Management Options Using the SITEX-II Pathway Evaluation Process

Nadja Železnik

Regionalni center za okolje za srednjo in vzhodno Evropo , Slovenska cesta 5, 1000 Ljubljana, Slovenia

nzeleznik@rec.org

 

The "Pathways Evaluation Process" or PEP is a tool that emerged from the reflections of the SITEX-II European research project ("Sustainable network of Independent Technical Expertise for radioactive waste disposal - Interactions & Implementation") related to the Radioactive Waste Management (RWM). The PEP objective is to identify, structure and discuss issues that would really matter for different types of stakeholders (all institutions involved and especially civil society) and that concerns all the steps of the different possible RWM “Pathways” that may be considered over a timescale of several generations.
The proposed methodology is based on the concept of “Pathways”, which describe strategies, or future visions, retracing the steps of a possible evolution from the current situation of radioactive waste management as a whole (including waste already produced and potentially waste to be produced) to a final state called “Safe Terminus” (ST). The ST is defined as a situation where the safety of all considered categories of waste do not anymore entail an active human contribution, at least after a period that does not exceed an order of several generations. Several options or "pathways" constituting the basis for discussion and made from a combination of technical options for RWM: a) straightforward Geological Disposal (GD), b) interim storage and search of ST and c) GD with possibility of reversibility of decisions and retrievability of RW to look for ST. The pathways are assessed for the robustness by using “Testing conditions” (disruptive events, unplanned changes, decision making challenges) and “Evaluation Criteria" on governance quality, management of risk and values & ethics.
The PEP approach was exercised for several times, including the participants coming from technical support organisations, nuclear regulators, research entities, radioactive waste management organisations and representatives of civil society. The results very promising and emphasize the importance of transversal elements (to have in mind in all the pathway), notably institutional structure and background, meaningful public participation, pluralistic expertise, availability of financial resources, monitoring and memory in long-term horizons. The paper will present the PEP methodology and the achieved results.






13.09.2017 10:10 Poster session - GREEN

Nuclear power plant operation and new reactor technologies - 1008

Development and Its Application for Prevention the Reactor Trip when single 12-finger CEA is dropped

Chang Ho Kim

KEPCO-E&C , 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353, South Korea

kimch@kepco-enc.com

 

I. Reactor trip due to single 12-finger CEA drop
The CPCS for APR1400 in Korea is designed to initiate the reactor trip when single 12-finger CEA is dropped using the Penalty Factor (PF) calculated by CEA processor to ensure the fuel integrity. The Optimized Power Reactor 1000 (OPR1000) in Korea has experienced the reactor trips more than fourteen times due to the low Departure Nucleate Boiling Ratio (DNBR) and high Local Power Density (LPD) trips by CPCS when single 12-finger CEA drop event or its spurious signal during normal plant operating from 1995 to 2008 since the Hanbit NPP units 3 and 4 commercial operating as follows;
- Reactor trips based on the real CEA drop : Seven (7) times
- Reactor trips based on the spurious signals of RSPT : Seven (7) times

II. Prevention the Reactor Trip when single 12-finger CEA is Dropped
To prevent the reactor trip when single 12-finger CEA is dropped or its spurious signal exists, a solution is presented in this paper. Whenever single 12-finger CEA is dropped or its spurious signal exists during normal plant operating, a turbine runback signal from CPCS through Reactor Power Cut-back System (RPCS) to Turbine Control System (TCS) is initiated to reduce the reactor power automatically, and the Penalty Factor (PF) in CPCS is deleted using thermal margin in Core Operating Limit Supervisory System (COLSS). Without the PFs in CPCS, the DNBR and LPD trip signals is not initiated when single 12-finger CEA is dropped or its spurious signal exists.

However, in the case of single 12-finger CEA drop, the proposed method in this paper is to ensure the safety of the core by setting a sufficient Required Overpower Margin (ROPM) of the COLSS to compensate for the radial peaking factor increase, instead of applying the PF into the CPCS. In order to do this enhancement, more ROPM is needed than the existing one. This can be achieved by introducing new PLUS7 fuel with improved thermal margin.

Nevertheless, there is one more thing to consider. The trip due to the CPCS required by the safety analysis has been removed, but another trip from the behavior of the reactor due to the 12 finger CEA drop is additionally generated. When CPCS PF for 12-finger CEA drop is removed, if no action is taken in the system, the reactor core coolant temperature decreases due to the output imbalance of the primary and secondary sides. In this case, according to the performance evaluation, the reactor shutdown will occur due to the CPC Variable Overpower Trip (VOPT).

III. Application in Man-machine Interface Systems
To prevent the VOPT reactor trip, the secondary side output (i.e., turbine output) must be reduced quickly, which can be achieved through turbine runback/setback systems. Proposed a new set of systems is configured to generate turbine runback to prevent reactor shutdown due to CPCS VOPT after dropping the CEA. The changes of MMIS will be presented in the full paper.

IV. Conclusion
Through this solution, the EPRI URD 7.2.1.6 requirement to prevent the reactor trip when single 12-finger CEA is dropped is fully satisfied. This will enhance the plant safety and reduce the plant trip possibility in APR1400 NPPs.






13.09.2017 10:10 Poster session - GREEN

Nuclear power plant operation and new reactor technologies - 1010

An ultrasonic non-destructive testing system for detection and quantification of early stage subsurface creep damage in the thermal power generation industry

Petar Vejić1, Marko Budimir1, Channa Nageswaran2, Liudas Mažeika3, Aristeidis Arvanitis4, Antonatos Alexandros5, Pierre-Adrien Itty1

1INETEC-Institute for Nuclear Technology, Dolenica 28, 10250 Zagreb, Croatia

2TWI Ltd, , United Kingdom

3Kaunas University of Technology, K. Donelaicio 73, LT-44239 Kaunas, Lithuania

4SA, Tivolska 50, 1000 LJUBLJANA, Slovenia

5Public Power Corporation S.A. - Testing Research and Standards Center, Leontariou 9, Leontario 153 51, Greece

marko.budimir@inetec.hr

 

Creep damage detection in pressurised steam line components is a major concern in the power generation industry. Currently, replica metallography is used to inspect these components. This method can only detect surface defects; however, evidence indicates that creep damage develops first inside the pipe wall and does not appear at the wall surface until the pipe is almost ready to fail. This results in catastrophic component failures which cost the industry more than €500,000 in lost revenue per day out of operation.
To combat this, we are currently commercialising the CreepUT system, which employs a proprietary Ultrasonic (UT) technique that enables the early detection of sub-surface creep damage. We are an industrially driven consortium with significant expertise in field inspection services and NDT product development. The upgrade towards the TRL-9 is being done by refinement of the hardware electronics and software, and by making the CreepUT system more ergonomic, to be used by technicians. A major part of the project is focused on validating the performance of the system in an industrial power plant. Activities as well target towards certifying the CreepUT product and setting-up a customer services department.
This paper presents current technical achievements of the project - ongoing improvements in the system mechanical design, electronics and software.






13.09.2017 10:10 Poster session - GREEN

Nuclear power plant operation and new reactor technologies - 1011

Topical study on NPP Design Deficiency

Miodrag Stručić

Joint Research Centre of the European Commission, Westerduinweg 3, 1755 ZG Petten, Netherlands

miodrag.strucic@ec.europa.eu

 

The EU Nuclear Safety Clearinghouse for Operating Experience Feedback (Clearinghouse) has been established in 2008 to enhance nuclear safety through dissemination of lessons learned from Nuclear Power Plant (NPP) events, and to provide help in Operational Experience Feedback (OEF) process primarily to nuclear safety Regulatory Authorities and to their Technical Support Organizations within the EU. One of the main Clearinghouse deliverables is Topical Study and so far, there are twenty studies already published. The most recent one is Topical study on NPP Design Deficiency.
Topical study on NPP Design Deficiency has been conducted to review the worldwide OE from events where design deficiencies are addressed. The sources of analysed events are the IAEA International Reporting System (IRS), the US NRC Licensee Event Reports (LER), the French Institut de Radioprotection et de Sureté Nucléaire (IRSN), and the German Gesellschaft für Anlagen und Reaktorsicherheit mbH (GRS) Operational Experience (OE) databases. The main objective of this study was to extract the generic and case-specific lessons from the events contained in the databases, and to provide recommendations to the members of the European Clearinghouse. More specifically, study is aimed to reveal latent weaknesses of nuclear power plants which dates from the design phases i.e. before start of NPP operation. Thousands of the event reports have been screened and 774 were selected for further study.
For the purpose of the study all applicable events were categorised, in the first place, to provide events phenomenology type's distribution. The objective was to group the events around a common topic which presents a general issue. All selected event reports were classified into families according to different criteria. Finally, the six Common Major Issues have been defined to properly summarise most of the concerns related to design. Accordingly, the 29 recommendations are grouped under six topics: Unanalysed condition, Robustness of design, Ageing, Internal and External Hazard, and Quality of Documentation.
Additional finding in this study came from the evaluation of the detection mode, i.e. how design deficiencies have been detected by licensees. It has been identified that, for IRS database, the design issues have been most often revealed by an event with actual consequences, rather than, as it would be desirable, by engineering reviews, inspections or surveillance activities. The number of detection by actual consequences is increasing over time. Therefore, design reviews and other methods to detect design deficiencies need to be more encouraged.
Licensees have been probably already informed about all important events described in this study through different OE communication channels and implemented appropriate actions to prevent reoccurrence on their site. Therefore, recommendations from described reports should be used to check if applicable ones have been adequately implemented in affected NPPs, or if further assessment should be performed to prioritize necessary changes.
However, it is also important to note that design process management flaws are evident in all events on which main recommendations are based. Therefore, for an effective management system with a strong commitment to safety and strong culture of safety, it is necessary to establish effective safety strategy, i.e. integrate good design and engineering features providing adequate safety margins, diversity and redundancy, necessary for preventing accidents and mitigating the consequences.






13.09.2017 10:10 Poster session - GREEN

Regulatory issues, legislation, sustainability and education - 1113

New Thinking Approach in the Back-End Strengthening for Small Nuclear Programmes

Andrejs Dreimanis

Radiation Safety Centre of the State Environmental Service, Rupniecibas Str.23, Riga LV-1045, Latvia

andrejs.dreimanis@rdc.vvd.gov.lv

 

An actual challenge of European nuclear safety and security consists in finding of safe mode of final disposal of nuclear and high-level radioactive waste for small countries having minor nuclear programmes. We consider an approach how to promote development of shared international/regional deep repositories - the back end for nuclear energy production providing disposal of radioactive waste (RW) and spent fuel, at the same time contributing to physical protection of nuclear material.
Recognizing the choice of the host country as a prior problem in siting such facility we propose an interdisciplinary approach to multilevel stakeholder involvement and consensus building for siting of shared deep repositories. The approach is based on societal optimization of RW management in an extended environment - a multitude of physical, economic, social and psychological factors - by using the self-organization (SO), chaos and fuzziness concepts as well as the principle of requisite variety.
In the shared repository siting problem there appears an essential novel component of stakeholder consensus building: to reach consent – political, social, economic, ecological – among international partners, in addition to the intra-national consensus building tasks. An entire partnering country is considered as a national stakeholder, represented by the national government, being faced to simultaneous seeking an upward (international) and a downward (intra-national) consensus in a stressed environment, having possibly diverse interests.
Following theses about building of multilevel consensus are developed:
a) owing to stakeholder informational SO via their cooperation and competition there are formed knowledge-creating stakeholder communities,
b) building of international stakeholder consensus could be promoted by activating and diversifying multilateral interactions between intra- and international stakeholders, including various networks for international cooperation,
c) development of partnership between inter-national and intra-national stakeholders - a key towards democratic dialogue, with the aim to observe distinguishing interests and to reach a shared understanding of disputable issues,
d) arisen controversies are resolvable using synergetic non-rigid step-by-step approach to the choice of the host country, thereby reducing mutual misunderstanding in decision-making.






13.09.2017 10:10 Poster session - GREEN

Regulatory issues, legislation, sustainability and education - 1114

Sensitivity Study of Soil Poisson's Ratio Used in Soil Structure Interaction (SSI) Analyses

Seung Ju Han

Korea Hydro & Nuclear Power Company, Ulsan, 45014, South Korea

hansj914@khnp.co.kr

 

Korea Electric Power Corporation (KEPCO) and Korea Hydro and Nuclear Power (KHNP) began the process of seeking US Nuclear Regulatory Commission (NRC) approval for the design of Advanced Power Reactor 1400 (APR1400). The USNRC issued that use of Poisson’s ratio approaching values may result in numerical instability of the SSI analysis results. It is generally known that a dynamic Poisson’s ratio value approaching 0.5 will cause numerical sensitivity problems in SSI analyses using the SASSI program. In the SSI analysis of the APR1400, the dynamic Poisson’s ratio of the soil is limited to not greater than 0.48 in order to avoid numerical sensitivity problems. To demonstrate that the Poisson’s ratio values used in the SSI analyses soil profile cases do not produce numerical instabilities in the SSI analysis results, a sensitivity study is performed using the ACS SASSI Nuclear Island (NI) model. Sensitivity study is performed using the ACS SASSI NI model of APR1400 with soil profiles to demonstrate that the Poisson’s ratio values used in the SSI analyses of S1 and S2 soil profile cases do not produce numerical instabilities in the SSI analysis results.






13.09.2017 10:10 Poster session - GREEN

Regulatory issues, legislation, sustainability and education - 1115

Differences between IAEA and EU Basic Safety Standards

Helena Janžekovič

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

helena.janzekovic@gov.si

 

The European Union (EU) published in 2013 the new basic safety standards “for protection against the dangerous arising from exposure to ionising radiation” as given in the text of the Council Directive 2013/59/Euratom. This directive was published after more than 15 years of thorough analyses of radiation protection issues and shall be transposed in all 28 EU Member States (MS). The Directive is based on experiences with implementation of five Euratom Directives already transposed, on ICRP 103 publication published in 2007 and on new scientific data. In 2014 the IAEA published International Basic Safety Standards regarding radiation protection and safety of radiation sources, i.e. IAEA SS GSR Part 3, which is also based on the publication ICRP 103 and experiences of more than 160 IAEA Member States.
Despite the fact that both texts regarding basic safety standards have been drafted in the same period of time and are based on the ICRP 103, i.e. introducing three exposure situations, there are some significant differences in both documents, e.g. in managing exposure situations related to industries using natural occurring radioactive materials. The differences span from variations in basic principles of radiation protection to distinctness in technical standards, such as the standard related to the exposure due to radon in workplaces. The presented systematic analysis of differences shall facilitate communication among EU MS and other countries which is particularly important taking into account world market of equipment producing ionising radiation, world market of materials and control of exposure of itinerant workers in all exposure situations.






13.09.2017 11:00 Severe accidents

Severe accidents - 404

The SCONE software for corium-sodium interaction

Magali Zabiégo, Christophe Fochesato

CEA, DEN, Cadarache, DTN, SMTA, Batiment 224, 13108 Saint-Paul-Lez-Durance, France

magali.zabiego@cea.fr

 

The ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration) is a technological demonstrator, currently designed by CEA and its partners, with very high levels of requirements. In the frame of the conception of this reactor, the possible occurrence of a severe accident has to be considered.
The present paper focuses on a particular phenomenon that could occur during a severe accident sequence: the Fuel Coolant Interaction (FCI). A significant FCI would take place if corium (mixture of hot (~3000°C) molten core materials) came into contact with the more volatile liquid sodium, after flowing down into the reactor lower plenum (where the sodium temperature is around 400°C) via dedicated transfer tubes under conception.
Heavy corium fragmentation would then occur and, depending on the physical conditions of both corium and sodium, the sudden vapor production would lead to the formation of a pressure wave that could threaten the reactor structures. Apart from the potential explosive nature of the vapor generation, the features and the relocation of the corium debris generated by the interaction have to be precisely assessed in order to ensure post-accident debris bed cooling and avoid re-criticality.
In this paper we intend to present the SCONE software under development at CEA for the assessment of corium-sodium interaction. SCONE aims at being a mechanistic tool to precisely represent the phenomena involved when hot molten corium interacts with liquid sodium. After recalling the physics of the interaction that SCONE will have to simulate, we will describe the object-oriented architecture that was developed as well as our first numerical choices. Indeed, the versatile architecture allowed us to implement a first numerical engine that will be described in terms of representation of the different phases composing the so-called interaction zone (liquid sodium, sodium vapor, coherent and dispersed corium) and in terms of numerical scheme for solving the mass, momentum and energy balance equations. This engine was developed to handle a compressibility-evolving flow (liquid sodium vaporization), the pressure build-up observed during a corium-sodium interaction as well as the pressure wave expansion into the bounded system.
We intend to present the first results that demonstrate the operationality of the SCONE architecture and of the numerical engine.






13.09.2017 11:20 Severe accidents

Severe accidents - 425

Experimental Investigation of Iodine Decontamination Performance of a Filtered Containment Venting System in ARIEL Facility

Jaehoon Jung1, Jae Bong Lee1, Hwan Yeol Kim2

1Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, 305-353 Daejeon, South Korea

2KAERI (Korea Atomic Energy Research Institute), 989-111 Daedeok-daero, Yuseong-gu, 305-353 Daejeon, South Korea

jhjung@kaeri.re.kr

 

A filtered containment venting systems (FCVS) is one of the strategies maintaining the containment pressure by releasing high temperature and pressurized gas from the containment to the environment. During the releasing process, fission products are filtered simultaneously by FCVS to reduce the leakage of radioactive materials to the environment. Korea plan to install the FCVS and is developing a new Korean FCVS for the light water reactors as a depressurization system of the containments under a severe accident. The Korea Atomic Energy Research Institute (KAERI) constructed a large-sized test facility, called Aerosol Removal & Iodine Elimination (ARIEL) test facility, to verify the performance of the FCVS. The ARIEL test facility consists of three parts; a performance test facility which is a scaled-down, full height, and reduced diameter of Korean FCVS, thermal-hydraulic facility and aerosol/iodine generation and measurement facility. For the iodine retention test , the elemental/methyl iodine measurement systems were developed to evaluate the performance of Korean FCVS. Experiments on iodine retention in FCVS have been conducted in ARIEL test facility . The main parameters of the test were: i) pressure 5, 7 bar (a), ii) total mass flow rate 0.14 kg/s(5bar), 0.21kg/s(7 bar), iii) the initial pH=13. More experiments are planned to contribute to the development iodine analysis model.






13.09.2017 11:40 Severe accidents

Severe accidents - 405

Raising Nuclear Reactor Safety to a Higher Level - The Supercritical CO2 Heat Removal System - "sCO2-HeRo"

Joerg Starflinger1, Dieter Brillert2, Otakar Frybort3, Petr Hajek4, Martin Rohde5, Thomas Freutel6

1Institute of Nuclear Technology and Energy Systems (IKE) University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart, Germany

2UNIVERSITÄT DUISBURG-ESSEN, Forsthausweg 2, 47057 Duisburg, Germany

3Research centre Rez, Hlavni 130, 250 68 Husinec-Řež, Czech Republic

4ÚJV Řež, a.s., Hlavní 130, Řež, 250 68 Husinec, Czech Republic

5Delft University of Technology, TRI, Afd. RF, Mekelweg 15, 2629 JB Delft, Netherlands

6Das Simulatorzentrum KSG - GfS, Deilbachtal 173, 45257 Essen, Germany

joerg.starflinger@ike.uni-stuttgart.de

 

The “supercritical CO2 heat removal system”, sCO2-HeRo, safely, reliably and efficiently removes residual heat from nuclear fuel without the requirement of external power sources. This system therefore can be considered as an excellent backup cooling system for the reactor core in case of a Fukushima-like scenario, combined station blackout, loss of ultimate heat sink, and loss of emergency cooling.
sCO2-HeRo is a very innovative reactor safety concept as it improves the safety of both currently operating and future BWRs and PWRs through a self-propellant, self-sustaining and self-launching, highly compact cooling system powered by an integrated Brayton-cycle using supercritical carbon dioxide as working fluid. Since this system is powered by the decay heat itself, it provides new ways to deal with accidents that are beyond design. The turbine of a Brayton-cycle provides more energy than necessary to drive the compressor, which means that the sCO2-HeRo system provides electricity, which could be beneficial e.g. in case of a station blackout.
The sCO2-HeRo provides breakthrough options with scientific and practical maturity, which will be proven by means of numerical tools, like advanced CFD, and small-scale experiments to determine the performance of the components like a compact heat exchanger and a turbo-machine set. A demonstration unit of the sCO2-HeRo system will be installed in a unique glass model in order to demonstrate the maturity of the system. Finally, the potential of this system to deal with a range of different accident scenarios and beyond-design accidents will be shown with the help of the German nuclear code ATHLET.
The paper contains a summary of the scientific achievements. The project sCO2-HeRo leading to this application has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 662116.






13.09.2017 11:00 Reactor physics

Reactor physics - 604

The 117Sn(n,n’)117mSn reaction: a suitable candidate to investigate the epithermal neutron spectrum by reactor dosimetry techniques

Christophe Destouches1, Gilles Gregoire2, Domergue Christophe2, Thiollay Nicolas3, Casoli Pierre3, Radulovic Vladimir4, Snoj Luka4, Trkov Andrej5, S. Bourganel6

1Commissariat a l'Energie Aromique - Centre d'Etudes de Cadarache DRN/DER, Izpolni naslov!, 13108 St Paul Lez Durance Cédex, France

2CEA-Cadarache DEN/DER/SPEX/LDCI, Bâtiment 238 - Piece 10, F13108 Saint-Paul-lez-Durance, France

3CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France

4Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

5International Atomic Energy Agency, Wagramerstr. 5, P.O.Box 100, A-1400 Vienna, Austria

6CEA/DEN/DANS/DM2S/SERMA/LPEC, CEA/Saclay, Izpolni naslov!, F91191 Gif /yvette, France

christophe.destouches@cea.fr

 

Neutron dosimetry technique relies either on capture and fission reactions for thermal - epithermal neutron flux (< 1keV) characterization or on inelastic, fission, proton or alpha production reactions which have an energy threshold at roughly around 1 MeV for the fast neutron spectrum. The intermediate part of the neutron spectrum, although of great interest for the study of material damage under irradiation in Material Testing or Reactors or Nuclear Power Plants, is poorly covered by dosimetry techniques due to the lack of effective nuclear reactions in this energy range.
The CEA has launched studies several years ago to try to identify new reactions suitable for use in research and irradiation reactors. These studies have resulted in the identification of specific reactions on isotopes of Tin and Zirconium [a] to cover the epithermal energy range of the neutron spectrum.
This paper focuses on the 117Sn(n,n’) 117mSn reaction which has been identified as a good candidate for the characterization the upper energy part of the epithermal spectrum. Indeed, this reaction presents a relatively high level cross section (1.25 barns) and an energy threshold around 300 keV in a typical MTR reactor neutron spectrum. The 117mSn radio-isotope is also suitable for a spectrometry measurement with a 14 days half-life and a gamma ray emitted at 158.56 keV. Based on highly enriched 117Sn metallic samples developed for medical applications, experimental irradiation campaigns have been performed from 2011 to 2016 in several French CEA reactors and at the TRIGA reactor of the JSI (Slovenia). The purposes of these irradiations were to establish the feasibility of both the irradiation process and the activity measurement, to experimentally validate the 117mSn half-life value and other decay data and finally to assess the possibility to use this dosimeter with a satisfying accuracy.
After reviewing the main tin dosimeter nuclear data features, the paper briefly describes the irradiation campaigns and their objectives. Efforts made for improving the spectrometry measurement techniques are then presented. The analysis of the experimental activity results is exposed in terms of comparison with results of other experimental dosimeter types as well as calculated activities derived from neutron modeling codes. Then, through an uncertainty analysis process of experimental and calculated data, feedback on the need of improvement of the tabulated nuclear data is given. Finally, the evaluation of the maturity of the use of the enriched tin dosimeter is stated and upcoming works and data needs are suggested.
[a] V. Sergeyeva and al. «Determination of Neutron Spectra Within the Energy of 1 keV to 1 MeV by Means of Reactor Dosimetry» in IEEE Transactions on Nuclear Science, October 2015 - DOI: 10.1109/TNS.2015.2480889






13.09.2017 11:20 Reactor physics

Reactor physics - 605

Qualification of the WWER Benchmarks for Criticality Safety Validation

Jakub Lüley, Stefan Cerba, Branislav Vrban, Filip Osuský, Ján Haščík

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia

jakub.luley@stuba.sk

 

The computational systems, including software, hardware and nuclear data, can be characterized by a computational bias which demonstrates how accurately the computational system is able to simulate a reality. The criticality safety criteria required that the effective multiplication of the investigated system is less than defined limits. Therefore, the computational bias of criticality safety calculations, as an additional margin, must be established through the validation of the applied methods to critical experiments. The similarity assessment method, implemented in SCALE system, can be very effectively utilized in the process of identification of the relevant experiments and extrapolation of the obtained results to the target system. This method is based on sensitivity and uncertainty analysis of the several hundred critical benchmark experiments (DICE database) processed though the VALID procedure by the Oak Ridge National Laboratory. Previous analysis revealed that the DICE database does not contain sufficient number of benchmarks with required level of similarity for WWER reactor technology. Therefore, this paper tries to extend the database by experiments dedicated to WWER reactors presented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. The first part of the paper is devoted to the description of selected experiments and their transformation to SCALE system environment format. Results from the comparison of the criticality calculation with the experimental data are also presented. The second part of the paper is aimed to the sensitivity and uncertainty calculation where all sensitivity coefficients are validated by the Direct Perturbation method. Similarity assessment method is applied to the selected set of benchmark experiments and new obtained correlation coefficients are presented in the last part of the paper. Within the conclusion, a relevancy of selected experiment is assessed and recommendations for the next validation are given.






13.09.2017 11:40 Reactor physics

Reactor physics - 617

CIELO Evaluated Nuclear Data Files for Reactor Calculations

Andrej Trkov, Roberto Capote

International Atomic Energy Agency, Wagramerstr. 5, P.O.Box 100, A-1400 Vienna, Austria

a.trkov@iaea.org

 

The starting point for nuclear reactor neutronics (deterministic and Monte Carlo transport) calculations are evaluated nuclear data files, which are processed into application libraries. However, there is a long way from experimentally measured data and the data in evaluated data files. The task of an evaluator is to assess the most probable value of a parameter (e.g. a cross section) at any energy by suitably averaging experimental data, identifying and eliminating discrepant data and providing values in energy ranges or for nuclides for which certain experimental data types are missing. Considering the complexity of the data, this is not a trivial task.
The Working Party on Evaluation Cooperation of the OECD set up a subgroup WPEC-SG40 (alias CIELO) to focus on the evaluated nuclear data of the major nuclides in reactor technology, namely 1H, 16O, 56Fe, 235U, 238U and 239Pu. Different research groups in various parts of the world were working on improved evaluated nuclear data and their uncertainties for these nuclides; the ultimate test of improvement is the performance of the data in simulating integral experiments.
The Pilot project CIELO is practically completed. New evaluations for 235U and 238U that were co-ordinated through the IAEA as one of the participants in the project were produced and are available from the IAEA-CIELO web page https://www-nds.iaea.org/CIELO/. Together with the contributions from other partners they were extensively tested against criticality benchmarks from the Handbook of International Criticality Safety Benchmark Experiments Project (ICSBEP). They show significant improvement in performance compared to the previous evaluated data libraries. The key principle followed in the evaluation process was to respect the available differential data and use integral benchmarks as guidance in assigning preferences in the case of discrepant differential data, or to make adjustments to quantities with large uncertainties. The new evaluations for Standards_2017 were also fully respected.
The IAEA-CIELO evaluations were adopted for the new ENDF/B-VIII library, which is scheduled for release in 2018. We are looking forward to receiving feedback from general users on the overall performance of the new library.






13.09.2017 12:00 Thermal-hydraulics

Thermal-hydraulics - 203

Simulation of ROSA/LSTF Test SB-HL-02 using RELAP5 and TRACE

Andrej Prošek

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

andrej.prosek@ijs.si

 

Confidence in the computational tools, and establishment of their validity for a given application depends on assessment. The purpose of this study is therefore to independently assess the TRACE computer code for hot leg break test. A pressurized water reactor (PWR) hot leg break loss-of-coolant accident (LOCA) simulation experiment SB-HL-02 was performed on the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) program with a break size equivalent to 10% cold leg cross sectional area. For calculations the latest RELAP5/MOD3.3 and TRACE V5.0 computer codes were used.
The RELAP5/MOD2 input model was obtained within the framework of International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on Evaluation of Uncertainties in Best Estimate Accident Analysis (2006-2010). The obtained input model was first adapted to RELAP5/MOD3.3 and then converted to TRACE using Symbolic Nuclear Analysis Package (SNAP). Manual corrections were also needed (break model, tees for accumulator connection, steady-state calculation). TRACE input model consists of 171 Hydraulic Components and 44 Heat Structures.
The LSTF simulates a Westinghouse-type four-loop 3423 MW (thermal) PWR by a full-height and 1/48 volumetrically-scaled two-loop system. The break was located at the side of horizontal hot leg pipe in the loop without pressurizer. Total failure of high pressure injection system and auxiliary feedwater as well as loss of off-site power concurrent with the scram were assumed as the experimental conditions. The accident started with break valve opening. Scram signal and safety injection signal are generated. Only passive accumulators and low pressure safety injection were available for injection. Core heatup was experienced before first injection. The experimental data were available for the first 1000 s. In the paper the comparison between calculated and experimental data will be shown. For results presentation the SNAP animation of the ROSA/LSTF facility will be used too. The focus will be on graphical presentation of the physical phenomena and processes in an easy and understandable way.






13.09.2017 12:20 Thermal-hydraulics

Thermal-hydraulics - 204

Experimental investigation on long closed two-phase thermosyphons to be applied in spent fuel pools for passive heat removal

Claudia Grass, Rudi Kulenovic, Joerg Starflinger

Institute of Nuclear Technology and Energy Systems (IKE) University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart, Germany

claudia.grass@ike.uni-stuttgart.de

 

The removal of decay heat from spent fuel pools is presently dependent on an active cooling system. In case of a station blackout with loss of active cooling modes, a passive heat removal system based on closed two-phase thermosiphons can maintain an adequate spent fuel pool cooling and hence can significantly contribute to nuclear reactor safety. Because of their high heat transport capability by latent heat transfer and their simple design, closed two-phase thermosyphons have a great potential to satisfy as basic components the demands of a reliable and independent passive heat removal device.
In this paper, an experimental laboratory setup for the investigation of the heat transport behaviour of long closed two-phase thermosiphons is introduced and first measuring results of systematic test campaigns are presented. Due to the particular requirements of the application in spent fuel pools, closed two-phase thermosiphons of long lengths (?10 m) with extended adiabatic sections are investigated in a low operating temperature range between 45 °C and 80 °C. Two thin-walled stainless steel (wall thickness approx. 1.5 mm) pipes with a length of 10 m and inner diameters of 32 mm respectively 45 mm are examined. The heat source is realized by high performance tubular cartridge heaters on the evaporation section of the closed two-phase thermosiphons, which provide a heating power of approximately 6 kW. The heat sink is given by a double-pipe heat exchanger, which is connected to a secondary water cooling circuit. The temperature distribution along the outer pipe wall of the adiabatic section as well as pressures and temperatures within the evaporator and condenser section are measured. The heat transport performances of the closed two-phase thermosyphons are determined by means of heat input and heat removal balances. All experiments are performed with degassed, deionized water as working fluid. Various filling ratios from 50 % to 100 % are investigated. Depending on the imposed thermal operating conditions, e. g. the heat input and the inlet temperature of the cooling circuit, steady-state and pulsating operations of the closed two-phase thermosiphons are observed , whereas especially at low operating temperatures and high filling ratios a periodic geyser boiling in the closed two-phase thermosyphons can be detected. Based on these two basic operating conditions the heat transfer characteristics of both investigated pipe configurations, e. g. temperature profile, thermal resistance, pressure drop and heat transport performance, are discussed and compared each other.






13.09.2017 12:40 Thermal-hydraulics

Thermal-hydraulics - 205

Investigation of Turbulence Characteristics of the Flow in a 1x3 Array Fuel Rod Bundle with Spacer Grids by Using Computational Fluid Dynamics Method

Umut Karaman1, Cemil Kocar1, Adam Rau2, Seungjin Kim3

1Hacettepe University, Nuclear Engineering Department, 06800 Beytepe, Ankara, Turkey

2Pennsylvania State University, 231 Sachett, University, PA 16802, USA

3Department of Mechanical, Aeronautical and Nuclear Engineering School of Engineering and Applied Science University of California, Los Angeles, Izpolni naslov!, Los Angeles, California 90024, USA

ckocar@hacettepe.edu.tr

 

In nuclear reactor analysis, understanding the effects of spacer grids on flow have an important place in determining the reactor thermal-hydraulic performance and ensuring that reactor works under reliable operating conditions. Computational Fluid Dynamics (CFD) methods which could reflect the physics of the real flow are capable of presenting precise analysis results. In this study effects of simple support grids, independently of the mixing vanes, on flow through a 1 x 3 array rod bundle has been investigated with CFD methodology and the most appropriate turbulence model reflecting physics of the flow has been determined. Experimental studies performed in the flow laboratories of Penn State University has been referenced in comparison of the simulation results (Wheeler, 2014) where, single phase flow in the subject rod bundle was examined.
In the first part, mesh convergence study has been carried out on tetra, hybrid and poly type mesh in order to determine the most appropriate type and density. In this section, only k-e Standard and RSM LPS turbulence models have been utilized. In the second part of the study, using the mesh determined in the first section, comparison has been made covering all turbulence models examined in the study. Velocity distribution in the center of the rods, velocity and turbulence intensity contour plots on the upstream and downstream of the spacer grid at –3dh, +3dh and +40dh locations have been examined and compared with the experimental data.
The results of the study reveal the effects of the mesh type on calculations where, hybrid mesh having the most structured elements exhibit better performance over the others. Comparison between numerical and experimental results of rod central velocities shows an overall agreement with all turbulence models while, complex models and exceptionally k-e Realizable present better results than other two equation models. As a result of this study, flow through a simple support grid has been examined and the most appropriate turbulence model reflecting the physics of the flow has been determined.






13.09.2017 12:00 Waste management and emergency response

Severe accidents - 406

SAM strategy&modifications and SA simulator at Paks NPP

Éva Tóth

MVM Paks Nuclear Power Plant Ltd., P.O. Box 71, H-7031 Paks, Hungary

tothnel@npp.hu

 

This presentation would provide short description of Hungarian Severe Accident Management hardware modifications developed before/after Fukushima accident and based on Level 2 PSA results. It would be also presented sequences choosen for SAMG verification process which are used in so called “SA Simulator” at Paks NPP.






13.09.2017 12:20 Waste management and emergency response

Regulatory issues, legislation, sustainability and education - 1104

Achieving Coordinated Emergency Response with Non-harmonized Planning Zones - Fiction or Real Option?

Davor Šinka1, Saša Medaković2, Davor Rašeta2

1ENCONET d.o.o., Miramarska 20, 10000 Zagreb, Croatia

2State Office for Radiological and Nuclear Safety, Frankopanska 11, 10000 ZAGREB, Croatia

davor.sinka@enconet.hr

 

Many European countries have developed their nuclear emergency preparedness and response systems without giving much attention to what their neighbors are doing, which has lead to more or less significant differences. Should a nuclear emergency occur, these differences can lead to uncoordinated response. As a result, populations would feel unequally protected depending on where they live.
Croatia and Slovenia are the examples of neighboring countries where several components of the nuclear emergency preparedness and response system have jet to be harmonized. As Krsko NPP is located only some 10 km from the border, the efforts are being made by both sides to improve the status in this area.
The paper focuses on emergency planning zones. The first part provides the summary of current international recommendations and the overview of how planning zones are defined in European countries. In the second part the situation concerning Krsko NPP emergency planning zones is described. As the efforts to harmonize the zones didn't produce any results so far, two ways to proceed are examined. The first option would be to continue with the harmonization efforts until the solution acceptable for both sides is reached. The second option, which seems more likely at the moment, would be to try to ensure coordinated response with non-harmonized planning zones. Advantages and disadvantages of both options are presented and discussed.






13.09.2017 12:40 Waste management and emergency response

Regulatory issues, legislation, sustainability and education - 1105

National Survey on Nuclear Energy and Radioactive Waste in Croatia

Dubravko Pevec1, Mile Baće1, Krešimir Trontl1, Mario Matijević1, Radomir Ječmenica1, Paulina Dučkić1, Ana Holjak1, Irena Jakić2

1University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia

2Hrvatska elektroprivreda, Ulica grada Vukovara 37, 10000 Zagreb, Croatia

kresimir.trontl@fer.hr

 

During the year 2016 the national public opinion survey on nuclear energy and radioactive waste has been carried out in Croatia on the representative sample of 2002 participants. The stratified sampling method based on county distribution, gender, age, and the level of education has been employed on the Croatian population older than 14. Out of five defined research hypotheses, only the one claiming that the public feels inadequately informed on the subjects of interest was fully confirmed. The remaining four claiming that the public is not inclined to nuclear power, that the nuclear power plants in neighbouring countries do not effect public opinion, that the level of knowledge on the radioactive waste is low, and that there is a correlation between individual’s attitude on nuclear power and the level of knowledge, were only partially confirmed or disputed. Where possible, the results of the survey were compared to the relevant Eurobarometer surveys.






14.09.2017 08:30 Invited Tony Donné

Invited lectures - 104

The European Roadmap towards Fusion Electricity

Tony Donné

EUROfusion Consortium, JET, Culham Science Centre, OX14 3DB, Abingdon, United Kingdom

tony.donne@euro-fusion.org

 

The European Roadmap to the realisation of fusion energy breaks the quest for fusion energy into eight missions. For each mission, it reviews the current status of research, identifies open issues, proposes a research and development programme and estimates the required resources. It points out the needs to intensify industrial involvement and to seek all opportunities for collaboration outside Europe.
A long-term perspective on fusion is mandatory since Europe has a leading position in this field and major expectations have grown in other ITER parties on fusion as a sustainable and secure energy source. China, for example, is launching an aggressive programme aimed at fusion electricity production well before 2050. Europe can keep the pace only if it focuses its effort and pursues a pragmatic approach to fusion energy. With this objective the present roadmap has been elaborated. The roadmap covers three periods: The short term which runs parallel to the European Research Framework Programme Horizon 2020, the medium term and the long term.
ITER is the key facility of the roadmap as it is expected to achieve most of the important milestones on the path to fusion power. Thus, the vast majority of resources proposed for Horizon 2020 are dedicated to ITER and its accompanying experiments. The medium term is focussed on taking ITER into operation and bringing it to full power, as well as on preparing the construction of a demonstration power plant DEMO, which will for the first time supply fusion electricity to the grid. Building and operating DEMO is the subject of the last roadmap phase: the long term. It might be clear that the Fusion Roadmap is tightly connected to the ITER schedule. A number of key milestones are the first operation of ITER (presently foreseen in 2025), the start of the DT operation foreseen in 2035 and reaching the full performance at which the thermal fusion power is 10 times the power put in to the plasma.
DEMO will provide first electricity to the grid. The Engineering Design Activity will start a few years after the first ITER plasma, while the start of the construction phase will be a few years after ITER reaches full performance. In this way ITER can give viable input to the design and development of DEMO. Because the neutron fluence in DEMO will be much higher than in ITER (atoms in the plasma facing components of DEMO will undergo 50-100 displacements during the full operation life time, compared to only 1 displacement in ITER), it is important to develop and validate materials that can handle these very high neutron loads. For the testing of the materials a dedicated 14 MeV neutron source is needed. This DEMO Oriented Neutron Source (DONES) is therefore an important facility to support the fusion roadmap
The presentation will focus on the strategy behind the fusion roadmap and will describe the major challenges that need to be tackled on the road towards fusion electricity. Encouraging recent results will be given to demonstrate the outcome of the focused approach in European fusion research.






14.09.2017 09:10 Nuclear fusion - plenary

Nuclear fusion - 701

FUSION MATERIALS UNDER NEUTRON IRRADIATION AND DEUTERIUM PLASMA EXPOSURE

Olga Ogorodnikova1, Mitja Majerle2, Vladimir Gann3, Sergey Stepanov4, Leonid Dubov4

1Moscow Engineering Physics Institute National Research Nuclear University, "MEPhI", Kashirskoye shosse 31, 115409 Moscow, Russian Federation

2Nuclear Physics Institute of the CAS, v. v. i., Řež 130, 250 68 Řež, Czech Republic

3National Centre for Scientific Research “Demokritos”, Institute of Nuclear & Radiological Sciences & Technology, Energy & Safety, Nuclear Research Reactor Laboratory, Agia Paraskevi Attikis, P.O.Box 60037, 153 10 Athens, Greece

4Institute of Theoretical and Experimental Physics, B. Cheremushkinskaya ul., 25, 117259 Moscow, Russian Federation

olga@plasma.mephi.ru

 

Materials used in fusion reactors suffer from neutron and hydrogen isotope irradiation at high temperatures. Additionally, hydrogen and helium isotopes will be generated in fusion materials. A fundamental understanding of radiation damage and deuterium (D) retention in neutron-irradiated W-based and Fe-based materials is critical for the design of materials for fusion reactors. However, working on irradiated materials is costly. To simulate neutron-induced damage in nuclear materials ion beams are widely used. However, it is not always clear if the mechanisms under the ion irradiation are relevant to lower dose rate and the primary knock-on atom (PKA) spectrum under neutron irradiation. In order to simulate the real experimental conditions of irradiation of reactor materials for reliable predictions of radiation damage in fusion reactors, it is necessary to establish the adequacy of the resulting damage produced by different types of irradiation. For this reason, we compare radiation-induced defects created by self-ion irradiation, protons (from 20 to 35 MeV), and neutrons (with different spectrum up to 35 MeV) and their influence on the D accumulation in W- and Fe-based materials. Study of different distributions of radiation-induced vacancies and vacancy clusters of different sizes created by different types of irradiation allows us an experimental validation of the value of “displacement per atom” (dpa) when comparing different types of irradiation and improve the evaluation of dpa. In the present work, intrinsic and radiation-induced defects (vacancies, vacancy clusters, vacancy-impurity complexes, disordered regions) in polycrystalline W, pure Fe and reduced-activation ferritic/martensitic (RAFM) steels have been studied by well-established method of positron-annihilation lifetime-spectroscopy (PALS). The positron lifetime of about 130 ps for ‘as-received’ W indicates the presence of such defects as dislocations or vacancy decorated impurities. We have found single lifetime component of about 105±5 ps for both W recrystallized at 2000 K for 20 minutes and for Fe well-annealed at 1173 K for 1 hour. It may indicate that density of defects is very low and positrons annihilate from a non-localized state in the lattice. To minimize the initial defect concentration, all specimens were well-annealed prior to irradiation. Size distribution of defects obtained as a result of irradiation by different species at the same dpa was different. Distribution of defects after self-ion irradiation better reproduces the neutron damage compared to proton irradiation. It is in an agreement with theoretical predictions because both fast neutrons and heavy ions create dense collision cascades. However, majority of defects are vacancies and small vacancy clusters in W and Fe-based materials for all types of irradiation. In RAFM steels, lower density of vacancy clusters was found. The difference in the defect size distribution in W and Fe-based materials is discussed. An increase of the positron lifetime was observed upon annealing for all irradiated specimens indicating a growth of vacancy clusters. After investigation of radiation-induced defects by PALS, samples were exposed to D plasma to decorate those defects with D followed thermal desorption spectroscopy (TDS) measurements. By combination of PALS and TDS data, we correlate the D binding energy with each specific type of defects and predict the tritium retention in plasma-facing materials in ITER and DEMO conditions.






14.09.2017 09:30 Nuclear fusion - plenary

Nuclear fusion - 702

Comparison of spark plasma and conventional sintering for consolidation of W-based composites for DEMO divertor

Matej Kocen1, Petra Jenuš2, Andreja Šestan3, Saša Novak4

1Jožef Stefan Institute, Department of Nanostructured Materials K-7, Jamova 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Department for nanostructured materials, Jamova 39, 1000 Ljubljana, Slovenia

3Jožef Stefan Institute, Centre for electron microscopy and microanalysis, Jožef Stefan International Postgraduate School, Jamova 39, 1000 Ljubljana, Slovenia

4Jožef Stefan Institute, Department for nanostructured materials, Jožef Stefan International Postgraduate School, Jamova 39, 1000 Ljubljana, Slovenia

matej.kocen@ijs.si

 

Currently, one of the fields of research and key problems within EUROfusion project is a selection of materials for plasma facing materials. These components in DEMO will be subjected to harsh conditions, like high-heat flows, high thermal loads, and erosion by hot plasma.
A potential candidate for this material is tungsten (W). It has already been used in fusion research due to its good chemical and mechanical properties. Tungsten is a metal which is brittle at room temperature, but its good mechanical properties deteriorate at temperatures above 1000°C due to the recrystallization process and exaggerated grain growth. Therefore, current efforts are focused on research of W-composites. The main aim is to limit the grain growth of W at high temperatures while other chemical and mechanical properties (such as thermal conductivity, toughness, low activation) should not impair.
One of the possibilities is to reinforce W by the incorporation of W2C particles. The W2C particles can be synthesized by reaction of W with WC, graphene or another source of carbon at high temperatures during sintering.
Conventional sintering in high-temperature vacuum furnace is one of the possible ways to densify W and assure complete reaction between matrix and C-precursor. More novel approach for consolidation of powders is spark plasma sintering (SPS), where uniaxial pressure and a pulsed (on-off) direct electrical current (DC) is used to perform high-speed consolidation of powders. In a short period of time powders can be fully sintered, exhibiting great mechanical properties of an end product.
After sintering, phase analysis of samples was made by XRD, and results show, that only two phases were detected in composites; cubic W and hexagonal W2C. Although the sintering process in SPS is very short (30 min), compared to conventional sintering (1 day), W and C-precursor react completely into W-W2C composite in all samples with no detection of free carbon.
Analyses also showed that W sintered by conventional technique reached 90% of theoretical density (TD) while SPS prepared samples achieved up to 96% of TD. With small amounts of WC, the density of samples increases and the highest density using conventional sintering was 97 % of TD in a sample W + 5 vol.% WC, while samples W with 5 and 10 vol.% WC sintered with SPS all exhibit 99 % of TD. The strength of W + 1 or 5 vol.% WC are 10 and 30 % higher in samples sintered with SPS compared to samples sintered with a conventional technique, respectively. W with 10 vol.% WC had the highest tensile strength (1431 MPa). Samples exhibit around 30 % higher hardness if sintered with SPS compared to the conventional sintering method. The hardest sample was W + 25 vol.% WC with Vickers hardness HV/50 7.0 GPa. This sample contained 29 wt.% of W2C, which influenced on the increase in hardness.
The results of mechanical testing show that small amounts of W2C in W matrix significantly improve mechanical properties of the material as well as inhibit W grain growth. W and W-W2C composites tend to sinter better using SPS, resulting in a denser product with better mechanical properties. Moreover, this novel sintering technology will enable an important energy consumption reduction and a fine control of material micro- and nanostructures due to short processing times compared to traditional sintering method.






14.09.2017 09:50 Nuclear fusion - plenary

Nuclear fusion - 703

Comprehensive Neutronics Simulation Program SuperMC for Nuclear Energy Applications

Jing Song1, Lijuan Hao2, Shengpeng Yu1, Mengyun Cheng1, Guangyao Sun1, Bin Wu1, Huaqing Zheng1, Qi Yang1, Peng He1, Jun Zou1, Quan Gan1, Pengcheng Long1, Liqin Hu1, Yican Wu1

1Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031, China

2Vinča Institute of Nuclear Sciences, P. O. Box 522, 11001 Beograd, Serbia

shengpeng.yu@fds.org.cn

 

Super Monte Carlo Program for Nuclear and Radiation Simulation (SuperMC) has been widely used for the nuclear design and safety evaluation of fusion system. It supports the comprehensive neutronics simulation. It has been developed since 1998. The latest version is 3.1. Advanced and in-novative methodologies have been developed in the latest version to meet the challenges of Monte Carlo method in latest nuclear energy application.
To describe intricate structures of new nuclear energy systems, hybrid solid-surface modeling method which describes solids with arbitrary irregular boundary using free surfaces, facets and un-structured mesh was developed. With this method, the calculation model can be constructed from CAD models with twisted free surfaces directly without time-consuming pre-processing.
The burnup calculation based on Chebyshev Rational Approximation Method (CRAM) was in-ternal coupled with Monte Carlo transport calculation to improve calculation efficiency and precision. Data decomposition based lock-free thread-level parallel calculation method was developed to tackle the large memory consumption problem in high fidelity full-core transport-burnup calculation.
SuperMC implemented point activation calculation and coupled transport activation calculation, which consider the activation physical processes of 2231 nuclides and 66256 reactions. Similar as burnup calculation, CRAM was adopted to carry out activation calculation, taking into account the effects of light nuclide yield, short-lived nuclides and excited-state nuclides on the activation calcu-lation results. Various material activation properties can be calculated, including activity, decay heat, biological hazard, dose rate, and clearance index.
To accelerate the whole core simulation, the Global Weight Window Generator (GWWG) and GWWG coupled Uniform Fission Site (UFS) methods were proposed in SuperMC for the fusion and fission reactor respectively. Lately, considering the influence of different particle energy, the weight window parameters with energy spectrum were generated by GWWG, improved the acceleration ef-fect of GWWG. Furthermore, to avoid difficulties of applying GWWG to the simulations of fission reactors, coupled GWWG-UFS method is proposed.
SuperMC has been verified and validated by more than 2000 benchmark models and experi-ments, such as ICSBEP, SINBAD, ITER C-Lite, and etc. For the newly developed methods, it has been verified on the PWR burnup assembly model and the production of ITER activation handbook for burnup and activation calculation respectively. It has accelerated the transport calculation by 634 times on ITER reference model based on energy refined GWWG, and accelerated neutron transport simulation on HM model by more than 30 times based on coupled GWWG-UFS methods.






14.09.2017 10:10 Poster session - BLUE

Thermal-hydraulics - 207

Modeling of Loss of Coolant Accident on Bethsy Facility with Apros Software

Klemen Debelak, Luka Štrubelj

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia

klemen.debelak@gen-energija.si

 

In this paper, the analysis of international standardized test ISP 27 using APROS process simulation software is presented. ISP 27 also known as Bethsy 9.1b was performed on Bethsy test facility, which is a scaled down model of a three loop, 900 MWe Framatome PWR. The test simulates small LOCA, with 2-inch cold leg break combined with High pressure Injection System (HPIS) failure and state oriented approach, which requires operators to start an Ultimate Procedure. Model was built in APROS using elementary (branch, nodes…) and complex (advanced steam generator, pipes with heat structure…) modules in order to describe the volumes, heat structures and regulation of the test facility.
The verification of APROS model was made on heights, volumes and mass of heat structures. The events and actions were triggered with modelled regulation. The results presenting core cladding temperature, time integrated break mass flow, core liquid level and pressurized pressure…, are in good agreement with experimental data. The results clearly show all the processes such ass loop seal clearing, core uncover and rise of cladding temperature and other processes taking place in the experiment.






14.09.2017 10:10 Poster session - BLUE

Thermal-hydraulics - 208

Application of Several Numerical Schemes Solving Debris Bed Formation

Wael Hilali, Michael Buck, Joerg Starflinger

Institute of Nuclear Technology and Energy Systems (IKE) University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart, Germany

wael.hilali@ike.uni-stuttgart.de

 

One of the crucial questions in the management and mitigation of the consequences of a severe accident in light water reactors (LWR) is how to cool and stabilize the molten corium. For several designs of LWR, a deep pool of water is foreseen in the lower drywell of the containment. In the case of the failure of the reactor pressure vessel, the core melt materials will be discharged into the pool. By contact with water, it will fragment, solidify and settle on the bottom forming a porous debris bed. A two-dimensional continuum model of the deposition and relocation of particles is described in this paper. The mathematical model is based on a hyperbolic system of partial differential equations determining the distribution of the flowing layer depth and the depth-averaged velocity component tangential to the sliding bed. Because of the hyperbolicity of the system, a successful implementation of a solver is challenging, particularly when large gradients of the physical variables occur, e.g. for a moving front in the flowing layer or possibly formed shock waves during the deposition. In this paper, several numerical methods are applied to solve the system and compared, including traditional difference schemes, e.g. central and upstream difference schemes, as well as the Roe scheme and high-resolution NOC (Non-Oscillatory Central Differencing) schemes, in which several TVD (Total Variation Diminishing) limiters and reconstruction methods are applied. The implemented solver has provided promising results, which will be validated with other experimental investigations. The performed simulations with this modeling approach give some useful insights for the study of the most important parameters influencing granular bed formation process. It will contribute to the enhancement of the capabilities of the system code COCOMO simulating real reactor applications and providing more realistic data.






14.09.2017 10:10 Poster session - BLUE

Thermal-hydraulics - 214

Calculation of NEK CILR Test Using GOTHIC Code

Davor Grgić, Siniša Šadek, Tomislav Fancev, Vesna Benčik

University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia

davor.grgic@fer.hr

 

Containment is last barrier for release of radioactive materials in case on accident in NPP. Its overall integrity is tested during Containment Integrated Leak Rate Test (CILRT) at design pressure, in regular intervals. Thanks to applied risk based licensing the test intervals are increased up to once in 10 years and beyond. Taking that into account it is important to prepare test properly and to use obtained results to assess real status of containment. The test can be used to verify existing containment calculation models with potential benefit to use that verified models for explanation of some test results.
NPP Krsko has performed CILRT during plant outage in 2016. The paper presents comparison between measured and results calculated using multivolume GOTHIC model. The test scenario was reproduced using limited available data up to the end of pressurization phase. The depressurization phase is calculated by the code and measured leakage rate is implemented in the model. Taking into account necessary adjustments in the model overall prediction of the measured results (in terms of pressure, temperature and humidity) is very good. In the last phase of the test some non-physical behavior is noticed (without influence on overall test results), probably caused by combination of air redistribution within the containment and influence of heat transfer to plant systems that were in the operation during the test. Gothic model was used to check sensitivity of the predicted pressure (leak rate) to different heat inputs and to investigate the influence that operation of only one RCFC train during pressurization can have on the mixing of air within the containment. In addition, the influence of currently used weighting factors (weighting of measured temperature, RH and pressure data) on the used test methodology is investigated. The possible non-conservative direction of the influence (currently used weighting factors are giving lower leakage rate) was demonstrated and new set of weighting factors is proposed too.






14.09.2017 10:10 Poster session - BLUE

Thermal-hydraulics - 221

MOTIVATION AND DESIGN FOR AN IDEAL TEST FACILITY

Sujith Kumar Velumula, Nikhil Sai Ponna, Francesco Saverio D'Auria

University of Pisa, Via Montebello di Mezzo 17, 19020 Bolano, Italy

s.velumula@studenti.unipi.it

 

Sophisticate computational tools exist in nuclear thermal-hydraulics; however, approximations and hypotheses are needed to derive solutions or results applicable for the design and safety analysis of Nuclear Power Plants (NPP). A vast amount of experimental data has been gathered from so-called basic experiments, separate effect and integral effect test facilities. The measured data has been extensively used to prove the validity of computational tools.
A number of issues arises when comparing experimental data with the code results (validation process) as well as when attempting to apply the codes to the actual configuration of nuclear reactors. Examples are: a) precision targets are not established; b) in one assigned scenario some measured parameters maybe very well predicted and other at the same time may be mispredicted c) full demonstration of scaling capability cannot be achieved; d) some phenomena, whose importance is recognized within accident analysis in NPP, like two phase critical flow are predicted with large errors also due to the lack of modeling of parameters as density of the nucleation sites (either in the fluid and on the solid wall) and possible void formation following sharp edge cavitation upstream the break location.
We propose the design of a virtual (so far) test facility aiming at prioritization of the research in nuclear reactor thermal-hydraulics with respect to the importance on safety and design of NPP. The test facility is entitled as µ?-I4TF (µ=Modular, ?=Large, Advanced, Multi-Basics & discipline apparatus, I4TF=Ideal (four times) Test facility) and has the following key features:
• The scaling of the reference configuration is based upon the findings and recommendation of a recently issued NEA/CSNI State-of-Art-Report on Scaling.
• The size is the maximum reasonable (even though it remains four times ideal).
• Occurring phenomena are expected to cover an entire spectrum of accidents in Light Water Reactors (LWR), including primary system and containment.
• Calculated trends are (to be demonstrated) fully consistent with the experimental data base available today.
The global strategy behind this idea can be summarized as following: 1) To fix the minimum list of accident scenarios which covers all the phenomena mentioned above with the reasonable ranges of variation of dominating parameters, including their combination; 2) To perform calculation and analysis of each selected scenario (one example is provided in the present paper); 3) To show consistency between the calculated phenomena and parameter ranges with the experimental data; 4) To change features of the µ?-I4TF (e.g. by introducing CFD portions, by changing scaling laws, etc.) and performing new calculation; 5) The comparison between the calculation results at previous step with the reference calculation at step 2) may involve large differences in Safety Margins or in important design features: those differences (following a specific qualification analysis) will support the prioritization of new research in nuclear thermal-hydraulics.
The purposes of the present paper are:
- to describe the overall pattern of prioritization-driven research using µ?-I4TF facility,
- to present the main features of µ?-I4TF,
- to perform one reference calculation of the selected accident scenario.






14.09.2017 10:10 Poster session - BLUE

Materials, integrity and life management - 305

Temperature Dependence of Molybdenum Isotope Fractionation in Basic Aqueous Solution Using Anion Exchange Chromatography

Yu Tachibana1, Tatsuya Suzuki1, Toshitaka Kaneshiki2, Shin Okumura2, Masao Nomura2, Masanobu Nogami3

1Nagaoka University of Technology, 1603-1, Kamitomioka, Nagaoka, 940-2188, Niigata, Japan

2Tokyo Institute of Technology, Research Laboratory for Nuclear Reactors, 2-1 Hirosawa, Wako-shi,, Saitama, 351-0198, Japan

3Kindai University, 3-4-1 Kowakae, Higashiosaka City, Osaka 577-8502, Japan

yu_tachibana@vos.nagaokaut.ac.jp

 

Diagnostic nuclear medicine is making remarkable progress. In the world, the number of nuclear medical diagnostics has reached more than 30 million per year. Especially, 99mTc, a metastable isomer of 99Tc, is of great interest from the viewpoint of the medical use of nuclear diagnostics due to the half-life of T1/2 = 6.015 h and 143 keV. The diagnostics using 99mTc accounts for ca. 80 % among them. In case of Japan, 99Mo, a raw material of 99mTc, has been imported from foreign countries. Nowadays, most 99Mo is produced by using nuclear research reactors with highly enriched 235U, which has intrinsically some serious worries for nuclear proliferation. These reactors have been getting decrepit and the realistic costs for specialized facilities for chemical treatments, storages, and the disposal of large amounts of highly radioactive wastes are not reasonable. To improve these problems, the low-enriched 235U has been used to produce 99Mo in Australia’s Open Pool Australian Lightwater Reactor. Meanwhile, some researchers have suggested that 99Mo can be produced using the respective reactions of 98Mo(n, ?)99Mo, 100Mo(n, 2n)99Mo, and 100Mo(p, x)99Mo reactions. Before their nuclear reactions, it also has been required to enrich 98Mo or 100Mo for preparation of the enriched 99Mo because of comparatively lower natural abundance of 98Mo and 100Mo. Some researchers have studied the chemical enrichment of various nuclides by using chromatography. However, the data on isotope fractionation of medium-heavy elements such as Mo are not well-known. Hence, in analogy of them, we have also performed some chromatographic isotope separation experiments of Mo(VI) using benzimidazole-type anion-exchange resin embedded in high-porous silica beads (AR-01(Cl form)) in HCl solutions. As a result, the degree of the isotope fractionation, the isotope separation coefficient (?) is proportional to the reciprocal square of the atomic weight, which is called mass shift effect. However, the degree of the ? values of Mo(VI) was found to be insufficient practically. In other words, the systematic understanding of adsorption and desorption behavior of Mo is inevitable in the Mo isotope separation reactions, compared with other elements. We have examined it profoundly, that is, the effect of different valence states on chromatographic fractionation of Mo isotopes in HCl solutions has been examined. The ? values of Mo(VI) were obtained by using the isotope fractionation curve of Mo(VI) with AR-01(Cl form) and weakly basic porous-type WA20(Cl form) while we could not observe the Mo(VI) isotope fractionation in case of PA316(Cl form), which is one of porous-type strongly basic anion-exchange resin and the Mo(V) isotope fractionation using these resins. By using WA20, more than 15 % of the ? values of Mo(VI) can be improved but the degree still has not been sufficient. It has been known that Mo has various chemical forms in aqueous solutions and seven stable isotopes in nature. We have prepared MoO42- in basic aqueous solutions. Because the volume of MoO42- is smaller than that of Mo7O21(OH)33-. For the preparation of the single Mo isotope such as 98Mo and 100Mo, MoO42- may be desirable and little information on the effect of reaction temperature on chromatographic fractionation of Mo isotopes in basic aqueous solutions is available. In addition, the Mo isotope fractionation in chemical reactions is also particularly interesting for geochemists. Our works which have examined the mechanisms of Mo isotope fractionation in solutions, may contribute to the understanding the nature of the isotope fraction of Mo in the natural world. From these backgrounds, we have examined the effect of reaction temperature on Mo(VI) isotope fractionation behavior in basic aqueous solutions using PA316 resin with the high adsorption ability for MoO42- at a wide temperature range.






14.09.2017 10:10 Poster session - BLUE

Materials, integrity and life management - 306

Experimental measurement of the neutron emission rate

Filip Osuský, Branislav Vrban, Jakub Lüley, Stefan Cerba, Ján Haščík

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia

filip.osusky@stuba.sk

 

The experimental measurement of the neutron emission rate with manganese sulfate bath technique has been constructed at the Institute of Nuclear and Physical Engineering. The measurement system consists of two main parts: the spherical shaped Plexiglas bath vessel, where the neutron source is placed during measurement, and the special aluminum beaker, where the gamma detector is placed. The continuous flow of manganese aqueous solution increases the measurement precision thanks to the small circulation pump which interconnects the spherical bath vessel with the aluminum breaker. To ensure homogenous manganese concentration in the solution, the mixing equipment is placed inside the spherical vessel. NaITl and HPG detectors are used to measure the activity of 56Mn, the isotope which is formed by neutron absorption on 55Mn. In order to prevent the personnel from adverse effects caused by neutrons, a safety transportation container has been constructed. To thermalize and subsequently absorb the neutrons, the transportation container consists of a suitable thickness of polyethylene moderator as well as a thin layer of a cadmium absorber. To minimize the gamma radiation, the cadmium absorber is surrounded by a sufficient volume of lead. In order to perform the radiation shielding calculation and to create a map of neutron and gamma radiation, the MONACO code was used. Due to the finite dimensions of the spherical bath, a significant portion of neutron leaks from the system, therefore this effect needs to be assessed by the measurement or calculation of correction factors. The paper proposes correction factors for spherical bath vessel and aluminum beaker. Furthermore, both detector responses (NaITl and HPG) are compared in the paper and the first experimental results of activity of the manganese aqueous solution are presented.






14.09.2017 10:10 Poster session - BLUE

Severe accidents - 408

Modelling of premixed layer formation in stratified melt-coolant configuration

Janez Kokalj, Matjaž Leskovar, Mitja Uršič

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

janez.kokalj@ijs.si

 

A hypothetical severe accident in a nuclear power plant has the potential for causing severe core damage, including meltdown. The hot melt coming in touch with the coolant rapidly transfers the internal energy, which can cause a vapour explosion. This can jeopardize the integrity of the containment and can cause damage to the systems inside. Consequently, the possibility of a radioactive leakage presents danger for the surrounding. Similar explosion phenomena can be a concern in some industrial processes such as foundries and liquefied natural gas operations or in certain volcanic activity where water is present.
In severe accidents analysis in nuclear power plants, the melt jet pouring into a coolant pool is the mostly addressed geometry. In a stratified configuration, a continuous layer of melt lies beneath the water and both layers are separated by a steam film. In the past, this configuration was believed to be incapable of creating a significant premixed layer and producing strong explosions. However, the results from the experiments performed in the PULiMS and SES facility (KTH, Sweden) with corium simulants materials contradict this hypothesis. A clearly visible premixing layer and strong spontaneous vapour explosions were observed in some of the tests.
In the paper, the premixed layer formation in the stratified melt-coolant configuration will be presented. As of now, no validated models for the description of the premixed layer formation exist yet. Our developed modelling approach of the premixed layer formation and the implementation of the model in the MC3D computer code (IRSN, France) will be discussed. Numerical tests, applying the developed model, will be performed and the results will be analysed and discussed.






14.09.2017 10:10 Poster session - BLUE

Severe accidents - 409

Simulation of loss-of-coolant accident in spent fuel pool with ASTEC code

Ivo Kljenak1, Marko Matkovič2

1Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 Ljubljana, Slovenia

2Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

ivo.kljenak@ijs.si

 

Spent fuel pools (SFPs) are large structures equipped with storage racks designed to temporarily store irradiated nuclear fuel removed from the reactor of a nuclear power plant (NPP). SFP severe accidents have long been considered as highly improbable since the accident progression is slow (in comparison with reactor core accidents) and let time to corrective operator action. However, the accident at the Fukushima Dai-ichi NPP (Japan) in 2011 has highlighted the vulnerability of nuclear fuels that are stored in spent fuel pools in case of prolonged loss of cooling accidents. The Fukushima Dai-ichi nuclear accident has consequently renewed interest in the safety of SFPs.
In the frame of the NUGENIA project Air-SFP within the NUGENIA association, a benchmark was organised, in which participants simulated two accident scenarios (loss-of-coolant and loss-of-cooling) in the Fukushima SFP, using severe accident codes that were basically designed to model the degradation of the reactor core during a severe accident. The purpose of the benchmark was to assess the applicability of these codes to SFP accidents. The Jozef Stefan Institute participated in the benchmark with the ASTEC severe accident code.
In the paper, the developed input model of the Fukushima SFP for the ASTEC code is presented first. Then, the modelling of the main physical phenomena is described. Finally, the results of the simulation of the loss-of-coolant accident are presented and analysed.






14.09.2017 10:10 Poster session - BLUE

Severe accidents - 410

Simulation of a severe accident in a generic German PWR with the code system AC2

Livia Tiborcz1, Peter Pandazis2, Sebastian Weber3

1Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Forschungszentrum, Boltzmannstr. 14 , 85748 Garching, Germany

2Gesellschaft für Anlagen- und Reaktorsicherheit Forschungsgelände, Postfach 12-28, 85748 Garching b. München, Germany

3Gesellschaft für Reaktorsicherheit (GRS), Schwertnergasse 1, 50667 Köln, Germany

livia.tiborcz@grs.de

 

Safety analysis is a crucial part of nuclear safety, both for the licensing and the design assessment of new as well as for existing nuclear power plants. Particular interest has been drawn to severe accident analyses after the accident in Fukushima. Key points of the analyses are the time required to reach core degradation, the extent of the degradation and the release and transport of radioactive material first into the containment later possibly to the environment. Another main interest of the analyses is to assess the possible counter measures preventing core degradation and eventually the radioactive release.
The simulation environment AC2 contains the thermal hydraulic system codes ATHLET, ATHLET-CD (severe accident module) alongside with the containment code COCOSYS and the interactive analyses simulator ATLAS. It is being developed by GRS (Gesellschaft für Anlagen- und Reaktorsicherheit gGmbH) for the analysis of the whole spectrum of operational transients, design-basis accidents and beyond design-basis accidents anticipated for nuclear energy facilities. The main advantage of this code system assembly is its capability to simulate severe accidents as a whole starting from the normal operation throughout the core degradation phase and finally the assessment of the containment behavior, enabling the user to determine the source term into the environment.
Present study focuses on a severe accident scenario in a German generic PWR. The simulated scenario is a medium break cold leg LOCA during a station blackout. The aim of the study has been to assess the extent of core degradation and radioactive release into the containment. The calculations have been performed with the code system AC2.
Analyses demonstrated the capability of AC2 to simulate the whole transient scenario starting from normal operational condition till the release into the containment. During the accident the main points such as the relocation of the melt and its behavior in the lower plenum as well as the transport and release of fission products into the containment have been successfully simulated. In particular, the deposition of fission products and aerosols in the cooling circuit has been investigated. Especially the transport and retention of FP and aerosols in the steam generator are of great interest due to the bends and great temperature gradient found within.






14.09.2017 10:10 Poster session - BLUE

Severe accidents - 411

Numerical and Experimental Investigation of Debris Bed Formation in Degraded Cores of Light Water Reactors with Meltdown

Wael Hilali, Michael Buck, Joerg Starflinger

Institute of Nuclear Technology and Energy Systems (IKE) University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart, Germany

wael.hilali@ike.uni-stuttgart.de

 

The coolability of particulate debris bed is a major issue in the severe accident (SA) research. As an accident mitigation strategy for several designs of light water reactors (LWR), a deep pool of water is foreseen in the cavity below the reactor pressure vessel (RPV), with the aim of cooling the core melt materials discharged from the RPV after its failure. The molten corium jet into water will fragment, solidify and settle at the bottom as a porous particles bed, which generates residual heat due to the radioactive decay of the fission products. The preeminent goal becomes how to prevent the re-melting of the debris in consequence of insufficient cooling. One of the main factors determining the ability of decay heat removal is the geometrical configuration of the bed. For this purpose, the present work is part of ongoing numerical and experimental investigations into the formation process of debris beds by particles deposition and relocation. The numerical model is based on depth-integrated conservation equations describing a rapid flow of granular material over two-dimensional topography. The heap-formation model takes into account also the effect of the coolant boiling within the porous bed, which may lead to its leveling. A series of experiments were conducted to validate the numerical results using single- and multi-size mixtures of particles and pressurized air to simulate steam generation by decay heat. The comparison between the model prediction and the experimental observations shows a good agreement.






14.09.2017 10:10 Poster session - BLUE

Severe accidents - 418

CFD and LP Simulations of ENACCEF2 Hydrogen Fast Deflagration Experiment

Tadej Holler1, Ed Komen2, Ivo Kljenak3

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2NRG-Nuclear Research and Consultancy Group Dept. Fuels, Actinides and Isotopes, P.O.Box 25, 1755 ZG Petten, Netherlands

3Jožef Stefan Institute, Reactor Engineering Division , Jamova cesta 39, 1000 Ljubljana, Slovenia

tadej.holler@ijs.si

 

During a severe accident in a Nuclear Power Plant, large amounts of hydrogen can be generated. When hydrogen is released into the containment, it is mixed with air and may form a highly flammable gas mixture. Considerable attention has been given to the risk of hydrogen deflagration after the Three Mile Island accident in USA, 1979, and after the Fukushima Daiichi accident in Japan, 2011, where the destructive power of hydrogen was displayed. Hydrogen mitigation systems, e.g. passive auto-catalytic recombiners, can be installed in order to reduce the risk of hydrogen combustion as far as possible. However, as the possibility of hydrogen combustion is still not completely eliminated, both experiments and computer modelling are necessary to assess the hypothetical consequences of hydrogen combustion in NPPs.
Within ETSON (European Technical Safety Organisation Network), a benchmark exercise on hydrogen combustion was organised. The benchmark is based on experiments in the ENACCEF2 test facility, located at the Institut de Combustion, Aérothermique, Réactivité et Environnement (ICARE) within the CNRS (Centre National de la Recherche Scientifique) research centre in Orléans (France). The ENACCEF2 facility is a vertical acceleration tube of 7.650 m height and 25 cm inner diameter. The hydrogen-air mixture is ignited at the bottom of the tube, and the ensuing hydrogen combustion (fast deflagration) is observed.
Within the benchmark, one of the experiments was proposed as “open”, which means, that experimental results were provided to the participants prior to the simulations. The Jožef Stefan Institute participated in the benchmark both with the CFD code ANSYS Fluent and with the lumped-parameter code ASTEC. In the proposed paper, simulation results (maximum pressure, pressure increase rate and flame propagation) of the open test are compared to the experimental results and discussed.






14.09.2017 10:10 Poster session - BLUE

Severe accidents - 419

Pressurization process modelling in sodium during fuel-coolant interaction

Mitja Uršič, Matjaž Leskovar

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

mitja.ursic@ijs.si

 

In the frame of safety studies for sodium cooled fast reactors, it is important to estimate the risk for the environment in case of energetic fuel-sodium interaction. An energetic event in a nuclear power plant could jeopardize its integrity. The comprehensive fuel-coolant interaction computer codes are considered to be the most appropriate tool for the safety studies related to the issues of fuel-coolant interaction.
The modelling capabilities of fuel-coolant interaction codes to cover the fuel-water interaction in the reactor cases were already demonstrated. Because of large differences in thermo-dynamical and physical properties of sodium compared to water, the applicability of fuel-coolant interaction codes for the fuel-sodium interaction is currently under examination.
One of the main challenges in the energetic fuel-sodium interaction modelling is related to the precise nature of the pressurization mechanisms. The pressurisation during the energetic fuel-sodium interaction is mainly governed by the significant increase of the melt-coolant surface area during the fine fragmentation process. Understanding the heat transfer processes between the fine fragments and the coolant seems to be crucial for the pressurization modelling.
Our purpose is to assess and discuss the level of understanding of the pressurization process during the energetic fuel-sodium interaction The first objective is to review the direct boiling concept and the micro-interaction concept that are currently used in the fuel-coolant interaction codes. The second objective is to assess the applicability of different concepts to cover the fuel-sodium interaction.






14.09.2017 10:10 Poster session - BLUE

Severe accidents - 426

INVESTIGATION OF DESIGN CHANGE FOR THE SAFETY IMPROVEMENT OF APR1400 FAMILY OF REACTORS

Ihn Namgung

KEPCO International Nuclear Graduate School, 45014 Haemaji-ro Seosaeng-myeon, Ulju-gun, 689-882 Ulsan, South Korea

inamgung@kings.ac.kr

 

Since the Fukushima accident in 2011, the safety of Nuclear Power Plant becomes very important issue. Current APR1400 family of reactors consisting of APR400 and OPR1000 adapted multiple safety design features for defence in-depth. Korean nuclear fleet has been shown excellent safe operation recode. However, for public safety as well public acceptance of NPP, it would be better if we modify or improve current design so that it may be even safer in case of severe accident conditions. This paper attempts to look into the improvement of APR1400 reactor. Specifically, two design change items were investigated.
One is elimination of RV (Reactor Vessel) bottom nozzles that are used as conduit for ICI (In-Core Instrumentation) probe and cable. The RV bottom nozzles poises danger in case of core melt condition. It has yet to show these penetrations will cause early breach of pressure boundary. This study will suggest how to relocate the ICI penetration to RV Closure Head.
The other design change considered in this study is removal of long RV support column. There are four RV support columns in APR1400 family of reactors. Long support column tend to amplify seismic load in the core due to lowing the eigenvalue of reactor system. This long column type support was replaced a plate type support that is attached to the RV ledge location. For this design change, RV ledge need to be redesigned to accommodate support and sustain all load. The modal value and the mode shape of RV support system was computed and compared with that of the plate type RV support system. The plate type RV support produce much higher modal value, hence the RV support system will transfer less load in the low frequency range. This will alleviate seismic load to the core where nuclear fuel will undergo seismic excitation. These two design improvements can reduce the risk of severe accident damage among other benefits.






14.09.2017 10:10 Poster session - BLUE

Research reactors - 504

Preliminary calculations in support of the experimental campaign to evaluate the neutron energy spectrum inside the JSI TRIGA Mark II research reactor

Tanja Kaiba1, Vladimir Radulović2, Damien Fourmentel3, Loic Barbot4, Christophe Destouches5, Gašper Žerovnik1, Luka Snoj6

1Institut "Jožef Stefan", Odsek za reaktorsko fiziko, Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

3CEA France, CEN Saclay ORE/SRO, France

4CEA-Cadarache DEN/DER/SPEX/LDCI, Bâtiment 238 - Piece 10, F13108 Saint-Paul-lez-Durance, France

5Commissariat a l'Energie Aromique - Centre d'Etudes de Cadarache DRN/DER, Izpolni naslov!, 13108 St Paul Lez Durance Cédex, France

6Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

tanja.kaiba@ijs.si

 

Within the collaboration between the CEA Cadarache (Commissariat a l’Energie Atomique et aux Energies Alternatives) and the Jožef Stefan Institute (JSI) an experimental campaign is planned in autumn 2017 to evaluate the neutron energy spectrum inside the JSI TRIGA Mark II research reactor. The goal of the bilateral project is to establish an on-line neutron spectrum adjustment method using fission chambers containing different fissile deposits and covered with different thermal neutron shields. With the aim to establish optimal experimental conditions and experiment configuration, preliminary calculations have been performed and are presented in this paper. Using different fissile materials within the fission chambers will enable sensitivity to different neutron energy regions in the spectrum due to the different energy dependence of their fission cross sections. It was decided to use 235U, 242Pu and 237Np fission chambers in combination with Cd (1 mm thick) and Gd (250 µm thick) thermal neutron shields. The shields absorb thermal neutrons before they reach the fission chamber and therefore enable measurements of the epithermal and fast part of neutron spectrum. Neutron energy spectra in different measurement positions inside the reactor core were calculated and compared. The impact of different shield configurations and materials on the neutron energy spectrum was evaluated. Effects of the different shields on the detector response were investigated by calculations of fission rates for bare and shielded fission chambers. It can be concluded that the effect of the screen (opening for insertion of cables, etc.) has high impact on the detector signal due to the streaming of neutrons and can in case of 235U fission chamber result in even more than 15 % difference in the detector response. However, in case of the 237Np and 242Pu fission chambers this effect is much smaller and is evaluated to be below 1 %. It was found that a shield length of 11 cm would sufficiently minimize the contribution to the total signal of neutrons coming from the opening. Calculations were performed using the Monte Carlo particle transport code MCNP 6.0 in combination with the ENDF/B-VII.1 nuclear data library. They were performed using the JSI TRIGA MCNP model, which was developed at the JSI Reactor Physics Department (F8) and was previously validated by numerous different experiments. In future work the final experimental configuration will be added to the computational model and final calculations will be done. Measurements will serve as additional validation of the JSI TRIGA MCNP model and will contribute to improve the characterization of the JSI TRIGA Mark II research reactor.






14.09.2017 10:10 Poster session - BLUE

Research reactors - 505

Using the CEA In-Core Miniature Fission Chambers for Control Rod Calibration by the Rod Insertion Method

Vid Merljak1, Loic Barbot2, Damien Fourmentel3, Gašper Žerovnik4, Christophe Destouches5, Jean-François Villard3, Luka Snoj6

1Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

2CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France

3CEA France, CEN Saclay ORE/SRO, France

4Institut "Jožef Stefan", Odsek za reaktorsko fiziko, Jamova cesta 39, 1000 Ljubljana, Slovenia

5Commissariat a l'Energie Aromique - Centre d'Etudes de Cadarache DRN/DER, Izpolni naslov!, 13108 St Paul Lez Durance Cédex, France

6Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

vid.merljak@ijs.si

 

Calibration of control rods at the Jožef Stefan Institute TRIGA reactor is routinely performed by using the rod insertion method. The signal corresponding to the neutron population in the reactor is taken from a compensated ionization chamber (IC) located outside the reactor core. A correction factor has to be applied to eliminate flux redistribution effects, especially when the signal is used for on-line monitoring of the reactor power. In order to average-out these spatial effects a series of tests was performed where four miniature fission chambers (FC) were used simultaneously while placed inside the reactor core in the so-called measuring positions in-between fuel elements, at core middle height.
This paper presents a comparison of the total control rod reactivity worth as well as the integral and differential reactivity worth curves which were calculated from the signals of the IC and four FC during execution of the rod insertion method. A few ways to combine the FC signals are also examined. Results indicate a reasonable agreement between measurements with IC and FC. In general, reactivity worth deduced from FC signals tends to overestimate the one deduced from IC measurement, which could partially be attributed to the fact that flux redistribution correction factors were deduced from static calculations. To quantify this effect, a direct 3D dynamic numerical simulation would be needed. Also, a clear motivation to reduce the signal noise of the FC is raised, as the uncertainty in the background noise value affected the estimated total reactivity worth up to ±5 %.






14.09.2017 10:10 Poster session - BLUE

Research reactors - 509

Comparison of Stochastic and Deterministic Burnup Codes on TRIGA Research Reactor

Anže Pungerčič1, Dušan Calić2, Luka Snoj3

1Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

2Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

3Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

anze.pungercic@student.fmf.uni-lj.si

 

The criticality benchmark experiment at the Jozef Stefan TRIGA reactor was performed in 1991, after the reconstruction of the reactor. The evaluated criticality benchmark experiment was later published in the International Handbook of Evaluated Criticality Safety Experiments (ICSBEP).
After some years of operation the criticality benchmark was repeated with burned fuel. This benchmark provides useful information for testing of different cross section libraries of burnup calculation codes as well as the reactivity effect of burnup. In addition, the fuel burnup of individual fuel elements was measured by reactivity experiments. Recently we initiated activities to thoroughly record the operational history of the reactor together with excess reactivity and control rod worth measurements, which could be used for burnup validation of different core management codes, such as TRIGLAV [1] or Monte Carlo codes such as SERPENT [2].
Calculation parameters, such as energy released and fuel element positions, will be obtained from the complete operational history, where the information regarding every single TRIGA operation and experiment is recorded. For more than 200 elements used in the early years of operation, reactor calculation is the only viable method for determining accumulated fuel element burnup, as they were shipped back to the USA in 1999. For the elements still in use, gamma-ray spectrometry and measurements of the elements reactivity worth could be performed in the future and later compared with calculations.
In this paper we have performed reactor calculations on specifically chosen reactor cores, using deterministic (TRIGLAV) and stochastic (SERPENT) neutron transport and burnup codes. We determined burnup of each fuel element and its isotopic composition. Physical parameters, such as excess reactivity and fuel element burnup will be compared with the measurements for the purpose of studying reactivity effect of burnup and influence of different cross section libraries.

[1] Andreja Peršič, Tomaž Žagar, Matjaž Ravnik, et all. TRIGLAV: A program package for TRIGA reactor calculations, Nuclear Engineering and Design 318, 2017, pp. 24-34.
[2] Jaakko Leppanen, et all.,The Serpent Monte Carlo code: Status, development and applications in 2013, Annals of Nuclear Energy, 82, 2015, pp. 142-150.






14.09.2017 10:10 Poster session - BLUE

Reactor physics - 606

Generic model for natural nuclear reactors: from Oklo to a possible prior Georeactor

Bentridi Salaheddine1, Gall Benoit2, Gauthier-Lafaye Francois3, Naima Maiza-Amrani4, Djelloul Benzaid1, Mustapha Guerrache4

1University Djilali BOUNAAMA de Khemis-Miliana, Route de Theniet El-Hed, 44225 Khemis-Miliana, Algeria

2Institut Pluridisciplinaire Hubert CURIEN , 23 rue du loess - BP28, 67037 Strasbourg, France

3Laboratoire d’HYdrologie et de GEochimie de Strasbourg, 1, rue Blessig, 67084 Strasbourg, France

4Université de Sétif, El Bez Campus, Sétif 19000, Algeria

s.bentridi@univ-dbkm.dz

 

The Oklo natural nuclear reactors (located in Franceville basin, Gabon) present a real case of long-life operating nuclear system with thermal neutrons. Without any possible human intervention and considering their geological history, Oklo reactors were always considered as the natural analogue of geological storage of nuclear waste. Under thermal effects, altered surroundings of reactor cores evolved into a clay envelope of this high U-rich ore. The key to understand the operating of Oklo reactors undergo the understanding of thermal effects besides the neutron physics of such a system. The modern and recent processing of historical drills and outcrops of Oklo situation shed light on some unrevealed feature on the way how nature acted to ignite and maintain a sustained chain of fission reactions. Initially, MCNPX simulations permit us to explore and investigate related neutron physics of Oklo situation with real geological constraints and limits. Later, with developed shell scripts dedicated to natural U-rich configurations it became possible to interact with Monte-Carlo simulations and optimize time and calculation power to define all possible critical configurations, even for different geological ages. We extrapolate then, from Oklo case to a generic model covering any possible natural nuclear fission reactors occurrence. Very small dimensions cylinder (about few centimeters radius by few tens centimeters length) could be obtained as possible critical configurations for an older age before 2.0 b.y. Here, nature made it in easier way with long life systems including waste management with fuel confinement in space and time. Probably, prior natural nuclear reactors to Oklo have been occurred in an ancient past but their confirmation needs more investigation and new way to observe their signatures, unlike the Oklo case which presents a physical presence with several observations, measurements and analyses.






14.09.2017 10:10 Poster session - BLUE

Reactor physics - 610

Parametric Study of Burnup Modelling Aspects of VVER-440/V213 Fuel

Branislav Vrban, Jakub Lüley, Stefan Cerba, Dana Baratova, Vladimir Nečas, Ján Haščík

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia

branislav.vrban@stuba.sk

 

The properties of nuclear fuel depend on the actual composition of the fuel. The characteristics of the reactor core therefore undergo changes during burnup. In practice, precise prediction of core lifetime and reactor behaviour during burnup is essential part of reactor core analysis. The prediction accuracy of burnup calculations is a critical factor in the reactor analysis sequence. In the recent years there has been growing interest in further increasing the burn-up of all LWRs UO2 fuel including VVERs as much as possible in order to decrease power production cost. This paper investigates various calculation modelling issues associated with VVER-440 fuel depletion relevant to burnup credit. The well-known SCALE system and the TRITON sequence are used for the calculations. The effects of variations in the depletion parameters, input data tolerances and used calculation methods on the isotopic vectors, neutron and gamma spectra are investigated. The analyses in this paper include determination and ranking of the most important actinides and fission products and the emphasis is put on the fuel temperature distribution and its influence to the final isotopic vector of burned fuel. Finally, the impact of different inventories of spent nuclear fuel assemblies on release rates (activity rates) in the radionuclide migration calculations is examined by GoldSim simulation software.






14.09.2017 10:10 Poster session - BLUE

Reactor physics - 613

Uranium Enrichment Determination without Reference Material using HpGe, CdZnTe and NaI Detectors by Monte Carlo Method

Onur Murat1, Faruk Logoglu2, Mehmet Tombakoglu1

1Hacettepe University, Nuclear Engineering Department, 06800 Beytepe, Ankara, Turkey

2Hacettepe University/Turkey, Beytepe Çankaya/Ankara, 6400 Ankara, Turkey

onurmurat@hacettepe.edu.tr

 

There is a need in uranium enrichment facilities in order to check the process and inspect the uranium material whether enrichment percentage of material is made by intended percentage or not. In addition, after the production of nuclear fuel and during transportation of it, uranium isotopic content needs to be checked. In order to perform that there is a non destructive method which can ensure inspections can be carried out quickly. Uranium isotopes have signature gamma ray energies and each emits gamma rays at different energies. Intensities of these signature gamma rays can relate the amount of indicated uranium isotope in the material. In traditional methods characterization of enriched uranium compound is performed by taking ratio of most likely gamma ray peaks of U235 and U238 isotopes. In this study, full spectrum of U235 and U238 isotopes have been used to determine the enrichment of uranium compound. Besides, this action could be performed without using any reference material. Predetermined gamma spectrometry responses of U235 and U238 isotopes could assure the information about isotropic masses of uranium material. Performing this inspection is non destructive and could be accomplished only by generating gamma ray spectrum. Based on the monte carlo method HpGe, CdZnTe and NaI detectors have been modelled and performance comparison for this non destructive method have been carried out.






14.09.2017 10:10 Poster session - BLUE

Reactor physics - 615

Development, Validation and Application of the TRANSURANUS Code for Westinghouse Fuel Designs

Csaba Gyori1, Paul Blair2, Martin Jonson2, Paul Van Uffelen3, Arndt Schubert3, Branislav Hatala4, Radim Meca5

1NucleoCon, , Slovakia

2Westinghouse Electric Sweden, Finnslatten, Fredholmsgatan 22, SE-72163 Västeras, Sweden

3European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Hermoltz-Platz 1, 76344 Eggenstein-Leopolshafen, Germany

4VÚJE Trnava Engineering, Design and Research Organization, Ltd., Okružna 5, 91864 Trnava, Slovakia

5Nuclear Research Institute UJV Rez a.s., Husinec-Rez 130, 250 68 Praha Vychod, Czech Republic

gyori@nucleocon.com

 

TRANSURANUS is a thermo-mechanical code of the European Commission, Joint Research Centre to simulate nuclear fuel rod performance under normal operation, transients and accidents. A specific version of the code has been extended with new best-estimate and conservative models to support deterministic and probabilistic safety analyses of Westinghouse's PWR and VVER fuel rod designs. The recent model refinements and code validation analyses have been carried in the EU-funded project ESSANUF aiming at the development of an alternative nuclear fuel, as well as licensing methods and methodologies for the Russian-designed pressurized water reactor VVER-440. The Westinghouse-specific developments of TRANSURANUS have covered fission gas release, fuel swelling, cladding corrosion and hydrogen uptake, rod growth, cladding creep, rod burst, inner cladding steam oxidation and oxide breakaway. The code validation has been based on two major data sources: (1) an extensive database of Westinghouse consisting of pool-side measurements and post-irradiation examinations (PIE) for a large number of commercial and experimental fuel rods and separate-effect tests representing a wide range of conditions and (2) data of on-line measurements in specific test fuel assemblies of the Halden Reactor Project (HRP). A statistical evaluation method has been applied to prove the correct simulation of the different physical phenomena and the adequate bounding of the experimental data. Safety analysis methodologies for design basis accidents (DBAs) are also developed in ESSANUF, proposing the application of the extended TRANSURANUS code coupled to the thermo-hydraulic system codes RELAP and ATHLET, and to the reactor dynamics code DYN3D that consists of a 3D neutron kinetics core model. The paper gives an overview of the incorporated models and the most relevant TRANSURANUS code validation results and statistics and discusses the application of coupled code systems in DBA analyses.






14.09.2017 10:10 Poster session - BLUE

Reactor physics - 616

Validation of the ADVANTG on the ICSBEP Skyshine Benchmark

Domen Kotnik1, Bor Kos1, Luka Snoj2

1Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

domen.kotnik@student.fmf.uni-lj.si

 

ADVANTG, Automated Variance Reduction Generator [1], is a code developed by the Oak Ridge National Laboratory that aims to automate the process of generating variance reduction parameters for fixed source MCNP calculations. In previous research, ADVANTG has been tested on a shielding benchmark (ALARM-CF-AIR-LAB-001, i.e. the neutron flux in a concrete labyrinth [2]) and has proven to be a powerful tool for acceleration of MC simulations in term of required CPU time. The purpose of this paper is to validate the use of ADVANTG on another computationally demanding benchmark, i.e. skyshine experiment dealing with neutron and photon radiation scattering in air from the open operating reactor.
We chose the Baikal-1 Skyshine experiment as the benchmark experiment. It is published in the ICSBEP (International Criticality Safety Benchmark Evaluation Project) handbook under identifier ALARM-REAC-AIR-SKY-001 [3].
Three experiments were done near Semipalatinsk (Russia) during the years 1996-1997 at the RA research reactor which is a part of the “Baikal-1” unique complex of research reactors belonging to the Kurchatov Institute of Atomic Energy in the Kazakhstan National Nuclear Centre (IAE NNC RK). The aim of the experiment was to obtain benchmark data for validation of the computer codes that are used to calculate the radiation and ecological safety of the population locating near nuclear power plants. In the experiments the source of radiation was the RA research reactor with nominal power of 300 kW, which was able to release a high-intensity flux of neutrons and photons into the atmosphere by removing upper shielding block. Basic parameters (flux, dose rates, energy and spatial distribution) of neutron and photon radiation released to the atmosphere were measured at various altitudes (18 cm - 260 cm) directly above the reactor cover and at various distances from the reactor axis (50 m – 1500 m) by a different set of spectrometric instruments, i.e. 1H spectrometer, 3He spectrometer, scintillation spectrometer with a stilbene crystal, multisphere spectrometer, sets of threshold and resonance detectors. The experiments were performed at a reduced power level of reactor and the results were then normalised to the rated reactor power of 300 kW.
In the paper we model the Baikal-1 Skyshine benchmark in the MCNP and the ADVANTG codes and compare the calculations against the benchmark values. In addition the performance of ADVANTG calculations versus analog Monte Carlo calculations is analysed.

References:
[1] S.W. Mosher, A.M. Bevill, S.R. Johnson, A.M. Ibrahim, C.R. Daily, T.M. Evans, J.C. Wagner, J.O. Johnson, R.E. Grove. 2013. ADVANTG-An Automated Variance Reduction Parameter Generator. Oak Ridge National Laboratory, Oak Ridge.

[2] E.A. Belogorlovet al. 1984. Neutron fields investigation in three sectional concrete labyrinth from Cf-252 source, ALARM-CF-AIR-LAB-001. Serpukhov, IHEP.

[3] O. F. Dikareva, I. A. Kartashev, M. E. Netecha and V. P. Zharkov, Baikal-1 Skyshine Experiment, NEA/NSC/DOC/(95)03/VIII, ALARM-REACAIR-SKY-1, Sep. 30, 2009, revision 1, NEAICSBEP Database






14.09.2017 10:10 Poster session - BLUE

Reactor physics - 618

Radioactive Inventory Data for Severe Accident and Consequence Calculation Codes

Davor Grgić1, Radomir Ječmenica1, Štefica Vlahović1, Ivica Bašić2

1University of Zagreb, Faculty of Electrical Engineering and Computing , Unska 3, 10000 Zagreb, Croatia

2APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia

davor.grgic@fer.hr

 

During severe accidents in the nuclear power plant it is possible to experience some fuel damage in the reactor core or within the Spent Fuel Pool (SFP). Radioactive elements released from fuel can reach the containment or Fuel Handling Building (FHB). In order to model transport of the radioactive material and to estimate dose to the people, the equipment and the environment, it is required to calculate activity and isotopic composition of the fuel (source term) and to group released isotopes according to their transport properties. The source term depends on initial amount of the fuel, its enrichment and acquired burnup. Taking into account number of isotopes and specific requirements imposed by the codes used for calculation of the radioactive materials transport that can be challenging task.
The paper presents procedure for automatic preparation of radioactive inventory data (mass, activity) for severe accident (MAAP, MELCOR) or consequence calculation codes (RADTRAD) based on ORIGEN2 depletion calculation. The depletion is performed on assembly-by-assembly basis using power/burnup 3D data calculated by PARCS code and prescribed time vector for radioactive decay calculation. The source term data can be prepared for reactor core or for whole or part of SFP inventory. In SFP case information obtained from SFPFA spent fuel assembly management system is used. The calculations are performed for NPP Krsko core cycle 29 and current content of NPP Krsko SFP.






14.09.2017 10:10 Poster session - BLUE

Nuclear fusion - 712

IMAS for SOLPS-ITER

Leon Kos1, Dejan Penko1, Xavier Bonnin2, Simon Pinches2

1University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta. 6, 1000 Ljubljana, Slovenia

2ITER Organization, Cadarache Centre, Building 519, 13108 St. Paul lez Durance, France

leon.kos@lecad.fs.uni-lj.si

 

SOLPS-ITER is an important element of the Integrated Modeling (IM) strategy for ITER. It contains an input file generator (DivGeo [or DG]/Uinp), a grid generator suite (CARRE/ Tria/ Triageom), a solver for plasma fluid equations (B2.5) and for neutral kinetic transport (Eirene), often (but not always) run together in coupled mode in what is called B2.5-Eirene, and post-processing tools in the form of analysis scripts and a plotting program (b2plot). B2.5-Eirene is to become an actor in the Integrated Modelling Analysis Suite (IMAS) being developed at ITER, and interfacing tools are designed with this transition in mind. The IMAS relies on standardized Interface Data Structures (IDSs) to transfer data from one code component to another within larger integrated modelling workflows.
In order to be able to fully integrate SOLPS-ITER within IMAS, therefore, it is necessary that it contain tools that allow it to read and write the relevant IDSs for its scope, namely edge_profiles, edge_sources, edge_transport, and transport_solver_numerics, including their underlying Generalized Grid Description (GGD) data.
It is also necessary to be able to import into the IMAS database older results currently stored in Consistent Physical Object (CPO) form that resulted from earlier work done within the framework of EUROFusion''s EU-IM Task Force and Work Programme on Code Development (WPCD).
The IMAS support for "edge" IDSs is an integral part of the SOLPS-GUI as well as fully integrated within the IMAS framework in place at the ITER organisation. In particular, users of the SOLPS-ITER code suite need to have a means by which they can save and archive their results to the IMAS database, and inversely be able to retrieve the results from older runs from the same database. Moreover, one wishes to be able to translate existing SOLPS output previously stored using the CPO formalism into IDSs for addition into the IMAS database. In this article we present the main aspects of SOLPS-ITER integration in the IMAS that performs these functions.






14.09.2017 10:10 Poster session - BLUE

Nuclear fusion - 713

SOLPS COUPLED FLUID-KINETIC MODELLING OF HYDROGEN LOW POWER PLASMA IN ITER DIVERTOR

Ivona Vasileska1, Bonnin Xavier2, Leon Kos1, Richard Pitts2

1University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta. 6, 1000 Ljubljana, Slovenia

2ITER Organization, Cadarache Centre, Building 519, 13108 St. Paul lez Durance, France

ivona.vasileska@lecad.fs.uni-lj.si

 

SOLPS-ITER is a code package consisting of the fluid plasma code, B2.5, and a kinetic Monte-Carlo neutral solver, Eirene, for the plasma edge of tokamaks including the outer core edge, scrape-off layer (SOL) and divertor regions. The B2.5- Eirene code package has a non linear model of neutral particle transports including neutral-neutral and molecule-ion collisions. It has been used to model edge plasmas and to show the physical mechanisms determining the behaviour of the particles in the tokamak edge. The paper describes the SOLPS-ITER kinetic modelling of a hydrogen plasma with low plasma density at the ITER divertor. The choice of the plasma density was done to reduce the frequency of the ELMs. We consider the plasma state in the quiescent period between ELMs, for a typical scenario of the Pre-Fusion Power Operation (PFPO) phase of ITER, with a power flux into the SOL of 40 MW. We present some initial results in the form of temperature and density profiles across the main SOL and at the divertor targets.






14.09.2017 10:10 Poster session - BLUE

Nuclear fusion - 714

CAD Model Preparation in SMITER 3D Field Line Tracing Code

Marijo Telenta1, Leon Kos1, Rob Akers2, Richard Pitts3

1University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta. 6, 1000 Ljubljana, Slovenia

2EUROfusion Consortium, JET, Culham Science Centre, OX14 3DB, Abingdon, United Kingdom

3ITER Organization, Cadarache Centre, Building 519, 13108 St. Paul lez Durance, France

marijo.telenta@lecad.fs.uni-lj.si

 

SMITER software library is a suite of object oriented FORTRAN modules designed to map profiles of SOL heat flux density flowing parallel to magnetic field lines onto plasma facing component (PFC) surfaces. This requires to follow the magnetic lines on flux surfaces within the magnetic equilibrium in 3D geometry until intersection with a solid surface. The high-resolution meshed surfaces are obtained from CAD model of the PFC structures.
In order to make SMITER suite more useful and easier to use, a CAD ``modeller'' is included in the framework that allows global CAD geometry to be displayed in a SALOME framework representing a general run environment. The tool offers selection and editing of the CAD elements constituting the model in which field lines are traced. The field lines must take into account the full neighbouring structures around the object of interest to ensure that the filed lines are not intersected locally by other surfaces. Also, meshing of the CAD model is performed if no mesh is provided. Work in this paper addresses the issue of PFC surface preparation from CAD model for meshing inside SMITER GUI framework. This is done using Open Cascade Technology (OCCT) CAD kernel that allows import of CAD files in STEP format and its visualization and manipulation.






14.09.2017 10:10 Poster session - BLUE

Nuclear fusion - 715

Kinetic simulation of parallel particle and momentum transport of blobs using a particle-in-cell code

Stefan Costea1, Jernej Kovačič2, David Tskhakaya3, Tsviatko K. Popov4, Miglena Dimitrova5, Bernd Schneider1, Codrina Ionita1, Roman Schrittwieser1

1University of Innsbruck Institut of Ion Physics Plasma Department, Techniker str. 25, A-6020 Innsbruck, Austria

2Institut "Jožef Stefan", Odsek za reaktorsko fiziko, Jamova cesta 39, 1000 Ljubljana, Slovenia

3Institute of Applied Physics, TU Wien, Fusion@ÖAW, Austria, Wiedner Hauptstr. 8-10/134, 1040 Wien, Austria

4Faculty of Physics, St. Kliment Ohridski University of Sofia, 5 James Boulcher Blvd., 1164 Sofia, Bulgaria

5Institute of Plasma Physics, Czech Academy of Science, Za Slovankou 3, 18200 Prague, Czech Republic

jernej.kovacic@ijs.si

 

The plasma heat flux to the solid walls of the fusion devices is one of the main limiting factors in the design of the future fusion reactors. As the plasma in the edge region of tokamaks is very complex, so is the modelling of the heat flux in the edge region. The large gradients of macroscopic plasma properties and the presence of many different charged and neutral particle species lead to many non-linear processes, e.g. inelastic collisions, which can be self-consistently described only by kinetic modelling [1].
Turbulent radial transport across the separatrix to the walls comes in the form of filaments (blobs) [2], which then disperse in the parallel and radial directions. Since the blobs bring into SOL particles with significantly higher energies than the bulk SOL plasma, this dynamic transport should be modelled with kinetic approach, as different particles can be subjected to vastly different collisional processes due to highly non-linear nature of the cross-sections describing them. This can drive edge plasma even further away from the thermodynamic equilibrium and non-Maxwellian velocity distribution should be assumed.
We propose to model the temporal evolution of the blobs in the SOL in the parallel direction by using a fully-kinetic massively-parallel particle-in-cell code BIT1 [3]. From the simulations we expect to study the behaviour of the particle and momentum transport of a passing blob in relation to the background conditions (collisionality, recycling etc.) and blob properties (size, temperature etc.). The obtained results should help in better understanding of the blob dynamics as well as make implications on the temporal dependence of the kinetic factors during blobs and consequently the heat flux [4].

[1] O. V. Batishchev et al, Phys. Plasmas 4 (1997), pp. 1672-1680
[2] S. I. Krasheninnikov, Phys. Letters A 283 (2001), pp. 368-370
[3] D. Tskhakaya et al, J. Nucl. Materials 438 (2013), pp. S522-S525
[4] D. Tskhakaya et al, Contrib. Plasma Phys. 48 (2008), pp. 89-93






14.09.2017 10:10 Poster session - BLUE

Nuclear fusion - 716

Modelling of neutron emission in tokamaks

Žiga Štancar1, Luka Snoj2

1Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

ziga.stancar@ijs.si

 

Neutrons represent one of the key mechanisms through which information about the plasma state in modern and future fusion devices is being conveyed. Using measurements of neutron emission it is therefore possible to perform diagnostics of some of the most important plasma parameters, such as its temperature, energy deposition of heating systems, fast ion distribution etc. Neutrons are crucial for fusion power measurements as well – namely the energy released in a DD or DT fusion reaction is proportional to the number of emitted neutrons. Two additional neutron based tokamak systems are considered to be fundamental for the operation of future fusion power plants. These are the production of tritium fuel, with the reaction between a fusion neutron and lithium taking place in the so-called breeding blankets in the walls of a tokamak, and the heating of the tokamak blankets and coolant through neutron slow-down.
The connection between plasma physics and neutron transport computations is thus crucial for understanding the behavior of plasmas in tokamak devices. Modern neutron transport computations in support of fusion experiments are performed with the Monte Carlo method. The initial step of a simulation is the creation of a source neutron whose characteristics are sampled on the basis of the system’s physical model. In order to study the relation between neutron emission, diagnostics response and its uncertainty it is necessary to analyze and improve the source models used in computations so far. In the paper the generic neutron plasma source models commonly used for neutron simulations are presented, together with a proposed methodology for creating a detailed plasma neutron source for Monte Carlo neutron transport calculations. This is based on the use of the state-of-the-art plasma transport codes, enabling the computation of neutron emission profiles as well as neutron spectra.






14.09.2017 10:10 Poster session - BLUE

Radiation and environment protection - 805

Extended Monitoring Programme on Environmental Radioactivity of the Krško NPP by reason of the construction of the Hydro Power Plant at Brežice

Milko Janez Križman1, Gregor Omahen2

1upokojenec, , Slovenia

2ZVD ZAVOD ZA VARSTVO PRI DELU d.o.o., Chengdujska cesta 25, 1260 Ljubljana Polje, Slovenia

milko.krizman@gmail.com

 

The chain of five hydropower plants (HPP) from Vrhovo to Brežice was built in the period from 1991 to 2017 on the lower course of the Sava river. The last one, the Brežice HPP, is located 7.2 km down-stream the Krško NPP and its construction considerably influenced the configuration of terrain. Thus, the previous riverbed is replaced by an accumulation basin downstream the NPP dam with maximum width of 680 m. The velocity of the river flow was slowed down to some cm/s, resulting in an increased sedimentation of suspended matter. The groundwater level was raised due to the elevated Sava level by 3 meters. An eutrophication of the Sava river basin due to the warmer water is expected and which might impact water biota. It is also presumed that liquid radioactive discharges from the Krško NPP will flow close to the left riverbank. In case of flooding, the contaminated water will therefore be spilled to retention areas on the left bank of the Sava river.
The above changes of the environment will have a substancial impact on the spread of radionuclides and their transfer to the water environment. Therefore, a conceptual framework (basis) for the extended radioactivity monitoring programme was elaborated under the co-ordination of the Slovenian Nuclear Safety Administration (SNSA).
The monitoring programme takes into account the following kinds of samples: surface water, sediment, water biota (fish) and ground-water. It determines the sampling locations, such as a new continuous pumping station for river water close and upstream the dam and introduces three additional boreholes for groundwater. The programme further determines the locations for grab sampling of water samples, sediments, and water biota (fish) within the accumulation basin, the natural water habitat situated near the left bank, and the widest part of the basin. Along with the standard analytics (for gamma emitters, strontium isotopes Sr-90/Sr-89 and H-3), the specific analysis of radionuclide C-14 in water environment is applied for the first time.
The extended water monitoring programme due to the Brežice HPP envisages (at the moment not yet precisely defined) an enlargement of 60-74 % of the existing annual scope of 266 analytic measurements. It requires modification of the Krško NPP documents (USAR, RETS) and the approval decree of the competent authority (SNSA).
The aim of the paper is to present and to discuss the monitoring programme extension with regard to the current state-of-the-art and to the applicable legislation.






14.09.2017 10:10 Poster session - BLUE

Radiation and environment protection - 808

Experimental stopping power data fitting for Hydrogen and Helium ions using MLP artificial neural network

Mohsen Kheradmand Saadi1, Javad Jamei2, Davood Alizadeh3

1Department of Nuclear Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran, 1477893855, Iran

2Islamic Azad University, Department of Nuclear Engineering, Science and Research Branch, Hesarek, 1477893855 Tehran, Iran

3University of Tabriz, 29 Bahman Blvd., Tabriz 5166616471, Iran

mohsen.kheradmand@gmail.com

 

A novel technique based on an intelligent Multi-Layer Perceptron (MLP) artificial neural network has been proposed for experimental stopping power data fitting. The unknown complex nonlinear stopping power functional form is fitted to hydrogen and helium experimental data by a set of linear combination of neurons. Experimental data from various simple and compound materials including Ag, Al, Au, B, C, Ca, Cu, Fe, He, H2, Kr, Mg, Pb, Ti, W, Air, CO2, Ethylene, Methane, and Polycarbonate are collected. The resilient back-propagation algorithm has been used for network training. The results showed that the use of artificial neural network is more efficient than conventional iterative methods such as least-square procedure. The proposed technique allows us to obtain stopping power values where no experimental data exists.






14.09.2017 10:10 Poster session - BLUE

Radioactive waste management - 902

Determination of Source Intensity for Neutron Flux Calculation Using MCNP5 in the Cold Neutron Guide for Activation Source Term Assessment

Gil Yong Cha1, Soonyoung Kim1, Kyung-Woo Choi2, Bo Kyun Seo3

1RADCORE, #503, 65 Technojoongang-ro, Yuseong , 34014 Daejeon, South Korea

2KINS(korea institute of nuclear safety), 62 Gwahak-ro, Yuseong-gu, 34142, DAEJEON, South Korea

3Korea Institute of Nuclear Safety , 62 Gwahak-ro, Yuseong-gu, Daejeon, 34142, South Korea

mars1222@radcore.co.kr

 

A method of determination of source intensity was studied when MCNP5 is used for neutron flux calculation at the guide in cold neutron research facility(CNRF) to be employed for the activation source term assessment. CNRF contributes to the development of relevant industries for researching nano- and bio-material structures. Cold neutron guide is a tube that can transfer neutrons from cold neutron source to the external experimental device minimizing a loss of cold neutron. The inner surface of the guide is coated with nickel or nickel and titanium alloy with a thickness of 5 ~ 10 ?, and it totally reflects cold neutrons incident within a critical angle. The inside of the guide is vacuumed so that cold neutrons are not leaked as much as possible. This specially treated guide minimizes the loss of particle, like optical fibers. Nevertheless, some cold neutrons leak into the guide, which causes the activation reaction in the guide. In general, two-step calculations were required to evaluate the activation inventory in the guide. First, the neutron flux in the guide is calculated by using the neutron transport code, and the result is used as input to calculate the activation inventory. The MCNP5 computer code is frequently used to calculate neutron flux, but it does not apply reflection physics for the cold neutron. In this study, we quantitatively assessed the effect of the reflection in the course of using MCNP5 to calculate cold neutron flux. At the time of calculation, the length of the guide is assumed to be 30 m, and cross section of the guide a square of 10 cm × 10 cm. First, we have calculated neutron flux in the guide near the guide entrance and the experimental device without the effect of the reflection. As a result, the neutron flux was calculated as 8.92E-06 #/cm2·s·source near the guide entrance and 8.04E-11 #/cm2·s·source near the experimental device. On the other hand, we also evaluated the case assuming that the cold neutron reflectivity of the inner surface of the guide is 90%. At this calculation, new value of neutron intensity was applied considering the reduction of neutron intensity at the collision point. As the result, the neutron flux in the guide was calculated to be 8.92E-07 #/cm2·s·source near the guide entrance and 2.64E-07 #/cm2·s·source near the experimental device. Thus, the neutron flux near the guide entrance will be overestimated, and underestimated the experimental device without considering the effect of the reflection on the surface. In particular, near the experimental device, there were several thousand times of neutron flux difference between both cases. The neutron induced activation inventory is expected to increase or decrease in proportion to neutron flux, because the cold neutron is a nearly mono-energetic beam. These results show that the effect of the reflection is an important point in calculating the activation inventory in the guide of CNRF when the MCNP5 is employed to calculate cold neutron flux.






14.09.2017 10:10 Poster session - BLUE

Radioactive waste management - 903

Reducing the activation of a reactor pressure vessel

Christoph Weindl1, Sam Karimzadeh2, Eileen Langegger3, Walter Binner3, Helmuth Böck4

1Atominstitut, Schüttelstr.115, A-1020 Wien, Austria

2Atominstitut der Österreichischen Universitäten, Stadionallee 2, 1020 Wien, Austria

3Osterreichische Kerntechnische Gesellschaft (Austrian Nuclear Society) Atominstitut, Stadionallee 2, A-1020 Vienna, Austria

4Technical University Vienna, Atominstitut, Stadionallee 2, 1020 Vienna, Austria

christoph.weindl@oektg.at

 

In PWRs decommissioning, the timeframe of Decontamination and Decommissioning (D&D) depends on the level of radiation exposure from long-lived activation products and safety issues. One of the major components that is highly irradiated and activated by neutron irradiation is the reactor pressure vessel (e.g. radiation from Co-60 and Nb-94). Under some assumptions it is possible to reduce the neutron activation in the outer layers of pressure vessel and decrease the radiation level accordingly. In this paper, a model based on planar geometry for neutron flux penetration calculation is presented that includes core reflector, core barrel, water annulus, pressure vessel and biological shield. For this purpose, Monte Carlo code is used to perform the calculation. In addition, methods of reducing or eliminating of thermal neutron flux in the outer layers of pressure vessel are discussed (e.g. use of neutron absorbers such as boron) in this paper.






14.09.2017 10:10 Poster session - BLUE

Radioactive waste management - 904

Safety and safeguards of spent nuclear fuel dry storage casks by means of ultrasonic tomography

Dmitry Sednev1, Yana Salchak2, Irina Bolotina1, Andrey Lider1

1Tomsk Polytechnic University, 30, Lenin Avenue, 634050 Tomsk, Russian Federation

2Unknown Organization, Unknown, Unknown, Slovenia

sednev@tpu.ru

 

Safety of spent nuclear fuel (SNF) is one of the most vital points in a back-end of any nuclear fuel cycle. Special meaning it has for states with closed fuel cycle concept, e.g. Russia, due to high amount of accumulated SNF and intention to use it in the foreseen future.
The project within Federal program performed by Tomsk polytechnic university was aimed to develop a robust automated system of nondestructive quality assurance and residual life assessment of spent nuclear fuel cask. The multichannel ultrasonic tomograph was developed on basis of Digital focusing array method (DFA). Implementation of DFA is required to get a quantitative evaluation of discontinuties in real-time. In contradiction to traditional X-ray testing ultrasonic tomography is much more sensitive to different crack types, especially kissig-bond type. Moreover, ultrasonic testing is faster, cheaper in operation and has no radiation hazard. All these features could significantly improve a safety level of dry storage procedure when ultrasound will be applied on manufacturing and operation stages.
The most important region of control during quality assurance procedure is welded joints of a cask body. Testing of welding gives us rather precise information about its structure. The preliminary research has shown that weld structure could be applied as a unique fingerprint of material with ultrasonic examination. The study was performed by means of single-channel instrument. However, the positioning was a weak point of this development. But proposed DFA-based system could form 3 dimensional fingerprints, that could be matched with a proper correlation algorithms. Volumetric signatures much less sensitive to a position error. Abovementioned fingerprints can serve as a unique intrinsic feature of cask with SNF and it can identify the cask in unique manner during safeguard verification inspections. Moreover, the structure of weld does not possible to reproduce on state-of-art level of technology.
In this paper the hardware and software for automated ultrasonic examination of dry storage casks is presented. Both safety and security approaches to data analysis are discussed.






14.09.2017 10:10 Poster session - BLUE

Nuclear power plant operation and new reactor technologies - 1006

LOSS OF ESSENTIAL POWER SYSTEM

Andrija Volkanovski1, Miguel Peinador2

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Joint Research Centre of the European Commission, Westerduinweg 3, 1755 ZG Petten, Netherlands

andrija.volkanovski@ijs.si

 

The non-interruptible power supply system is essential for safe operation and accident mitigation and recovery at the currently operating nuclear power plants. This system provides power to essential systems during the event of loss of offsite and onsite alternate current power sources referred to as station blackout (SBO).
This paper presents the results of statistical and engineering analysis of non-interruptible power supplies failure or deficiency events registered in two reviewed databases. The report includes events registered in the Nuclear Regulatory Commission (NRC) Licensee Event Reports (LERs) database and the IAEA International Reporting System (IRS) database. Both databases are screened for the relevant events registered in period 1.1.2000 to 1.1.2016.
The identified events will be classified into the multiple categories considering: plant state, circumstances, type of events, type of equipment failed or concerned, detection of the event, direct cause, root cause, consequences and event duration.
Each event will be classified into single best matching category with the exception for the characteristic related to the “failed or concerned equipment” and the “consequences” which can be multiple.
A trend analysis of the identified events will be performed with the application the approach for trend analysis of accident data developed in previous studies and utilising four measures of data trends.
Main results of the analysis and observations for prevention and mitigation of the loss of essential power system will be given.






14.09.2017 10:10 Poster session - BLUE

Regulatory issues, legislation, sustainability and education - 1101

Regulating New Nuclear Power Stations in Great Britain: An Overview of the Generic Design Assessment Process

Rebecca Thorington

Office for Nuclear Regulation, Building 4 Redgrave Court, Merton Road, Bootle, L20 7HS, United Kingdom

rebecca.thorington@nucleargrads.com

 

The Office for Nuclear Regulation (ONR) is responsible for independently regulating nuclear safety and security in Great Britain (G.B.). ONR also works closely with our sister regulatory body, the Environment Agency, who are responsible for regulating discharges to the environment from G.B. Nuclear Licensed Sites.
In 2006 the United Kingdom (U.K.) Government revised its policy towards nuclear energy, as part of the U.K.’s objective of moving towards a low carbon economy. In light of this revived interest in new nuclear power in the U.K., in partnership with the Environment Agency, ONR developed the Generic Design Assessment (GDA) process.
The GDA is a step-wise process consisting of four stages, where ONR undertakes an increasingly detailed assessment of the safety and security aspects for the generic design of a specific reactor type, in advance of an application for a Nuclear Site License from the prospective reactor operator. GDA is not a U.K. legal requirement, nor is it site-specific; it aims to give a prospective operator a clear indication of whether the design would in principle meet UK regulatory requirements. The process delivers a number of benefits, including: early engagement, which maximises ONR’s influence on nuclear safety from the outset, early identification of potential issues, and a reduction in regulatory and project risks.
There are several reactor designs which have either completed the GDA process or are currently progressing through it. EDF Areva’s UK EPR was the first to successfully complete GDA at the end of 2012. Westinghouse’s AP1000® completed GDA in March 2017. Two more designs are currently at different stages of the process; Hitachi-GE’s UK ABWR is scheduled to complete GDA in 2017 and China General Nuclear’s Hualong One design formally commenced the process in January 2017. Given the U.K. Government’s wider strategic objectives for new nuclear, there may also be the possibility of other reactor designs entering the process in the near future.
This paper provides an outline of the GDA process, along with some practical examples to articulate some of the lessons learnt and how these may be applied to future GDAs. The paper also covers ONR’s role as a regulator in the U.K.’s new nuclear build programme.






14.09.2017 10:10 Poster session - BLUE

Regulatory issues, legislation, sustainability and education - 1112

Early Warning Arrangements in Case of Potential Aircraft Attack to the NPP

Andrej Stritar, Blaz Vene, Andreja Peršič, Djordje Vojnovič

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia

blaz.vene@gov.si

 

After September 11th events, the Krško NPP determined the strategy and proposed some modifications related to preventive and mitigate measures in case of aircraft crash in accordance with the publication NEI 06-12, B.5.b Phase 2 & 3 Submittal Guidance. In parallel the Slovenian Nuclear Safety Administration has initiated activities for assurance of as early as possible notification on malicious aircraft potentially heading towards the nuclear power plant. The possibilities for effective communication between the organizations involved in the aircraft control in case of aircraft swerve were investigated.
After several years of discussion and administrative arrangements in 2016 the agreement was signed between civil and military air traffic controls, the Krško NPP and the Slovenian Nuclear Safety Administration. Based on the agreement the procedures were established to promptly, in the matter of seconds, alert operators of the NPP about potential danger from the incoming aircraft. By such alert operators would have several minutes available for eventual protection of crucial resources and personnel.
This measure comes as an addition to number of other measures at the nuclear power plant prepared for mitigation of large aircraft crash on site including severe accident management guidelines and special firefighting equipment.
Development and implementation of a multi-party agreement on notification of malicious aircraft to enable mitigation actions was recognised as a good performance for Slovenia at the 7th Review Meeting of the Convention on Nuclear Safety held in Vienna.






14.09.2017 11:00 Nuclear fusion

Nuclear fusion - 704

Irradiations of Mn, Au, Li2O foils and TLDs in the JSI TRIGA reactor for Potential Use as Tritium Production Monitors in Fusion

Ivan Aleksander Kodeli1, Vladimir Radulović1, Gregor Veniger2, Darko Kavsek2, Tadeusz Kuc3, M. Ciechanowski3, Władysław Pohorecki3

1Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

2Institut Jožef Stefan, Jamova cesta 39, 1000 Ljubljana, Slovenia

3AGH - University of Science and Technology, Izpolni naslov!, 30-059 Cracow, Poland

ivan.kodeli@ijs.si

 

Energy production in the future fusion reactors such as ITER or DEMO will be based on the fusion reaction of deuterium and tritium atoms (D-T reaction). Tritium is a radioactive isotope of hydrogen with a relatively short half-life, therefore it is available in nature only in negligible concentration and for fusion purpose it must be produced locally. Tritium can be produced by bombardment of lithium atoms with neutrons. Several types of special Tritium Breeder Modules (TBM) will be installed in the ITER reactor to demonstrate the self-sufficiency of tritium production.
Thermo-luminescent detectors (LiF - TLDs) can be used to measure tritium production. In the scope of the Fusion for Energy (F4E) project of the European Commission we proposed to investigate, as an alternative, the potential use of Manganese detectors for monitoring the tritium production in fusion machines. The similarities between the sensitivity profiles of the reaction of tritium production in 6Li and those of the 55Mn(n,?)56Mn reaction in the TBMs indicated that the latter reaction could be used as a tritium production monitor, at least for short-term monitoring, the half-life of 56Mn being 2.579 h. However, improvements and validation of the Mn cross-sections are needed in order to meet the required accuracy.
The first series of measurements was performed in 2014. This year we performed an additional set of experiments using aluminium-manganese foils irradiated at the JSI TRIGA reactor together with the aluminium-gold foils, TLDs and Li2O samples, with the principal objective to study the energy response of the 55Mn(n,?)56Mn reaction and correlations between the Mn and TLD / Li2O measurements. The samples were irradiated in different TRIGA irradiation channels , i.e. in the Central Channel, Pneumatic Tube in the core periphery and the IC40 irradiation channel in the graphite reflector. The analysis of the measurements is underway and includes the evaluation of uncertainties involved in the measurements and the calculations and the simulation using the Monte Carlo transport code MCNP.
The experiment was performed in the scope of the F4E supported project of the EC. The results of the measurements as well as the preliminary analysis and comparisons with the calculations will be presented in the paper.






14.09.2017 11:20 Nuclear fusion

Nuclear fusion - 705

The influence of helium on deuterium transport and retention in tungsten

Sabina Markelj1, Thomas Schwarz-Selinger2, Anže Založnik1, Stefan Elgeti2

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Max-Planck-Institut für Plasmaphysik (IPP), Boltzmannstr. 2, D-85748 Garching b. München, Germany

sabina.markelj@ijs.si

 

Recent laboratory experiments revealed that He admixture to deuterium (D) plasmas reduces D retention at elevated temperature for tungsten [1]. This change in retention is accompanied by reduced blistering and the growth of He nano bubbles below the surface. While there are several speculative attempts to explain these observations, the actual cause for the reduced retention remains unclear. One possibility is that He might act as diffusion barrier for D. Likewise nano-sized bubbles might open up additional pathways for D to reach the surface thereby decreasing its transport into the bulk. Contrary to these experimental findings density functional theory (DFT) calculations show strong attraction between He and hydrogen [2], indicating increased trapping and hence retention of D around He clusters should be expected.
In order to unravel this mystery we took a novel experimental approach. We separated the surface effect from the He effect by moving the He interaction zone into depth and decoupled the He implantation from D transport. The study was conducted on 20 MeV self-damaged W to minimize the possible influence of displacement damage created by He irradiation. Before implanting 0.5 MeV He in one half of the sample defects were decorated with D obtaining a homogenous D depth profile within the first 2 µm with a D concentration of about 2 at.-%. The range of He implantation fluences was chosen small enough that surface blistering would not yet set in [3] but large enough to reach relevant He concentrations [4]. He is located 1 µm deep, approximately in the middle of the W ion damage zone as visualized by focused ion beam cutting and scanning electron microscopy. D depth profiling was conducted by nuclear reaction analysis (NRA) in-situ during isochronal annealing between 300 and 800K. As expected from previous studies, in the no He half of the sample D content decreased with increasing temperature due to thermal de-trapping. However, accumulation of deuterium in the He implantation zone was observed, doubling the local D concentration. Exposing the sample to D atoms (0.2eV, flux 3.7×1018Dm-2s-1) at 600 K to study D uptake showed the same behaviour: The speed of D loading matched the no-He side and we again observed strong accumulation of D around the He implantation zone. Thermal desorption spectroscopy (TDS) of both sample halves shows no indication for a change in de-trapping energy but only a change in trap density. The effect of increased D trapping remains even after TDS to 1200K.
With this study we show for the first time unambiguously that the presence of He does locally increase D trapping. He does not act as a diffusion barrier. Rate equation modelling is presented which can explain the observed effects without the need for free fitting parameters. Results are discussed in the context of existing mixed He/D experiments.


[1] M. J. Baldwin et al. Nucl. Fusion 51 (2011) 103021
[2] H-B. Zhou, Nucl. Fusion 50 (2010) 115010
[3] S.B. Gilliam et al. J. Nucl. Mater. 347 (2005)
[4] R.P. Doerner, et. al, Nucl. Material Energy,in press (2016)






14.09.2017 11:40 Nuclear fusion

Nuclear fusion - 706

Manufacturing Challenges of fusion pebbles bed material

Rosa Lo Frano1, Daniele Del Serra2, Monica Puccini1, Donato Aquaro1, Eleonora Stefanelli1, Sandra Vitolo1, Riccardo Ciolini1, Aringhieri Marco3, Stefania De Santis3, Nicola Forgione4, Maurizia Seggiani1

1University of Pisa, Department of Civil and Industrial Engineering, Largo Lucio Lazzarino 2, 56124 Pisa, Italy

2Dipartimento d'Ingegneria Civile e Industriale - Universita di Pisa, Largo Lucio Lazzarino, 1, 56122 - Pisa (PI), Italy

3Industrieanlagen - Betriebsgesellschaft mbH, Einsteinstr. 20, D 85521 Ottobrunn, Germany

4Universita degli Studi di Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione, Largo Lucio Lazzerino 1, 56100 Pisa, Italy

rosa.lofrano@ing.unipi.it

 

The thermal stress on breeding blanket structure is one of the main design driver; consequently, an open issue for fusion power reactor is the choice of breeding blanket material.
The mostly promising and worldwide-investigated solution refers to the ceramic material, e.g. Li4SiO4, in form of pebble beds instead of a bulk form such as a block or a disk. This would introduce some advantages, especially for what concerns the heat transfer phenomena, low activation characteristics, low thermal expansion coefficient, high thermal conductivity, etc.
In this paper, particularly the reliability/availability of a methodology capable to produce stable and well-sized ceramic pebbles is a major challenge as well as the evaluation of the thermal conductivity, that is a necessary input data for the understanding of blanket behavior (support the thermal-structural and thermo-hydraulic analyses).
As for this latter, moreover, a sample holder considering the heat transfer mechanism through the pebble-bed was duly designed such to determine the pebble-bed thermal conductivity by means of a hot rig with guard resistance method.
This paper introduces (and analyses) preliminary results of the effective thermal conductivity on the pebble-bed as well as a description of the adopted methodology.
This research activity was developed in the framework of PRA2016 project.






14.09.2017 11:00 Radiation and environment

Radiation and environment protection - 801

Simulation of Consequences of Severe Accidents of TRIGA Mark II Reactor in Vienna with RODOS

Eileen Langegger

Osterreichische Kerntechnische Gesellschaft (Austrian Nuclear Society) Atominstitut, Stadionallee 2, A-1020 Vienna, Austria

eileen.langegger@euronuclear.org

 

The TRIGA Mark II Reactor of the TU Vienna – Institute for Atomic and Subatomic Physics (ATI) received new fuel elements in November 2012.
The responsible Austrian regulatory authorities have a big interest in the safety of the reactor and need to be prepared in case of major reactor accidents.
With the simulation tool RODOS several scenarios were calculated. RODOS is a simulation tool that can show the possible environmental impact in the closer and further proximity of a nuclear installation after a possible release of radioactive material. The paper will discuss those impacts for different scenarios, including the damage of 1 fuel element, up the the complete destruction of the reactor core.
The meteorological data, which are essential for the simulation, have been collected through many years at the site by the ATI weather station. Therefor realistic scenarios for typical spring, summer, autumn and winter days can be simulated, as well as unusual weather events (thunderstorm or snow days).
A holistic approach was taken to find out which meteorological scenario would create the biggest impact, and if this scenario changes regarding the damage to the reactor.
The outcome of the simulations could be used for other TRIGA Mark Reactors to show the safety of this type of research reactor.






14.09.2017 11:20 Radiation and environment

Materials, integrity and life management - 301

A RADIATION CAMPAIGN OF LUBRICANTS AT LENA NUCLEAR REACTOR FOR THE ESS AND THE SPES PROJECTS

Matteo Ferrari1, Antonietta Donzella1, Aldo Zenoni1, Diego Paderno1, Valerio Villa1, Monika Hartl2, YongJoong Lee2, Kristoffer Sjögreen2, Alberto Andrighetto3, Michele Ballan3, Francesca Borgna3, Stefano Corradetti3, Fabio D'Agostini3, Mattia Manzolaro3, Alberto Monetti3, Massimo Rossignoli3, Daniele Scarpa3, Andrea Salvini4, Fabio Zelaschi4

1Department of Mechanical and Industrial Engineering (DIMI) University of Brescia (UniBs), Italy, via Branze 38, 25123 Brescia, Italy

2European Spallation Source ERIC, Box 176, S-221 00 Lund, Sweden

3Legnaro National Laboratories of INFN, Viale dell'Universita, 2, 35020 Legnaro (PD), Italy

4Laboratorio Energia Nucleare Applicata Universita degli Studi di Pavia, Via Aselli 41, 27100 - Pavia, Italy

matteo.ferrari@unibs.it

 

Mechanical components used in the neighbourhood of high power targets for radioisotopes and/or spallation neutron production absorb high neutron doses during machine operations. The absorbed doses induce substantial modifications of physical and mechanical properties of the components, which can lead to a premature structural failure. In particular, non-metallic components are known to be sensitive to ionising radiation.
Greases and lubricant oils will be used for the target drive unit of the 5 MW class spallation target of the European Spallation Source (ESS) in Lund, Sweden and for the automatic handling unit of the SPES target at the National Institute of Nuclear Physics (INFN) in Legnaro, Italy. It is important to understand the neutron radiation-induced material degradation of different lubricant materials for a reliable operation of these target stations.
As the understanding of the lubricating material in a radiation environment is a common interest to the ESS and SPES projects, an irradiation campaign of these materials in the TRIGA Nuclear Reactor of the University of Pavia, Italy is planned in a collaboration frame. Grease and oil samples are irradiated in the mixed neutron and gamma field. Preliminary tests are made. The radiation-induced amount of evolved gas and its acidity level in chosen materials are assessed. The composition in terms of the elemental mass fraction is determined using different techniques. The main light elements are measured via CHN analysis and the metallic traces are measured via Neutron Activation Analysis. The composition of the material is needed to calculate the neutron dose and residual activation using MCNP, a Monte Carlo code simulating the interaction between radiation and matter.
We developed a test set-up to perform irradiation safely and efficiently. An extensive test campaign is ongoing. A wide selection of commercial off-the-shelf products is chosen for the tests, as a result of an extensive market research. The selected products have to satisfy the functional and constraint requirements for the reliable operations of target stations at ESS and SPES. Some of the lubricants are declared by the producer as radiation resistant, on the base of experimental tests performed usually with photon fields only. Others are general purpose ones.
The goal of the experimental procedure is to determine the evolution of the most relevant mechanical quantities as a function of the absorbed dose. Viscosity is measured for liquid oils using a rotational viscometer. Worked consistency is measured for semi-solid greases using a penetrometer. These parameters are considered as representative of the behaviour of the component in operations and they are the most relevant to the ESS and SPES applications.
The feasibility of other chemical-physical and rheological tests is currently under evaluation for a deeper understanding of the radiation-induced modifications at the microscopic level, which include Infrared spectroscopy, Cone-plate viscosity measurements, NMR analysis, etc.
The irradiation and testing procedures allow a comparison between the performance and lifetime of the analysed lubricants. The data from the post-irradiation examination will be used to determine the radiation resistance of the lubricants in target environments. The experimental results will represent a reference concerning the radiation resistance assessed in mixed neutron and gamma fields, largely lacking in the literature.






14.09.2017 11:40 Radiation and environment

Radiation and environment protection - 802

Gamma Dose Field due to Activated Cooling Water in a Typical PWR

Andrej Žohar, Luka Snoj

Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

andrej.zohar@ijs.si

 

In pressurized water reactors (PWR) the cooling water in primary loop is radioactive due to activation of corrosion products, migration of fission and activation products through the cladding or ruptures of the cladding and due to activation of water itself. The major contributors to the activity of the clean cooling water are oxygen and nitrogen nuclides produced by O-16(n,p)N-16, O-17(n,p)N-17 and O-19(n,gamma)O-19 reactions. N-16 and O-19 are gamma emitters (6 and 7 MeV for N-16 and 0.2 and 1.4 MeV for O-19) and N-17 decays by emitting neutrons (E ~ 1 MeV). The interest of the study is to develop methods for modelling the activated water radiation source and use it to determine the gamma dose field outside the steam generator during the operation of the reactor at full power.
A computational model of the steam generator is made in Monte Carlo neutron photon transport code MCNP. The constructed detailed model includes U-tubes, the secondary feeding ring, primary and secondary moisture separators, etc. The model is made according to specifications of the steam generator in a typical two loop PWR.
Special attention was devoted to modelling of the radiation source. Due to relatively short half-lives (7.13 s for N-16, 4.14 s for N-17 and 26.9 s for O-19) of activation products, the activity of water and consequently the intensity of the radiation source changes significantly along the pipes, hence modelling of the radiations source is relatively demanding. This was performed by discretisation of source and modelling disc shaped sources for selection of location for particle birth. In U-tubes the discs are centered to the centers of U-tubes and placed in planes with 50 cm distance between two planes. The probability for selection of a disc is determined from the half-life of the activated isotope and the flow rate of primary water. Due to different half-lives of activated isotopes, the sources are defined separately for each activated isotope.
The plan is to insert the model of the steam generator into the full scope model of the containment, which is relatively complex as well. In order to avoid potential issues with too complicated geometry, a simplified model of the steam generator was developed as well. The differences in dose fields around steam generator between the two models are investigated.
Our analysis of dose field around steam generator in PWR shows that the total gamma dose rate is order of several mSv/h at the bottom of the steam generator, while the gamma dose rate at the top of the steam generator is below mSv/h. The largest contribution to the gamma dose rates is N-16, which contributes more than 99.96 % to the total dose rate and is followed by O-19, which contributes 0.03 % to the total dose rate. The lowest contribution to the gamma dose rate is due to N-17, which contributes 0.01 % to the total dose rate.






14.09.2017 12:00 Current and future reactors

Nuclear power plant operation and new reactor technologies - 1004

Evaluation of Long Term Cooling Capability for APR1400 under Loss of Ultimate Heat Sink Accident

Nam-Seok Kim1, Yun-Il Kim2, Su-Hyun Hwang3, Soon-Il Jung3

1Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

2Multiple organizations possible, Unknown, Unknown, Slovenia

3Fakulteta za naravoslovje in tehnologijo, Izpolni naslov!, 1000 LJUBLJANA, Slovenia

nskim@kins.re.kr

 

A nuclear power plant should be able to maintain the heat removal capability of the reactor core even if accidents have occurred. Aircraft impacts and resulting consequence can cause damage to safety-related systems needed to maintain a core cooling such as ultimate heat sink or component cooling water systems. Especially, a loss of the ultimate heat sink (LOUHS) can lead to lose an operability of safe shutdown equipment, and the fuel can be seriously damaged as a consequence of loss of cooling to fuel in the reactor. In the present study, a long term cooling capability under LOUHS for Advanced Power Reactor 1400 (APR1400) is analysed by MARS-KS code. The realistic approach with reasonable assumptions included operator actions was applied to investigate a thermal behaviour in a full power operation. The plant reached the hot-shutdown condition with the safety grade equipment within the first 24 hours after LOUHS. The external cooling water injection systems of APR1400 were used to reach the cold-shutdown or maintain the hot-shutdown conditions for 60 consecutive days. The results showed that the reactor coolant system (RCS) can be reached the hot shutdown condition within 24 hours by the operation of the turbine driven auxiliary feed water pump (TDAFWP) and the steam generator atmospheric dump valve (SG-ADV). The RCS could be maintained the hot shutdown condition by only use of the steam generator, but could not be reached the cold shutdown condition. The once through cooling with the external cooling water injection in the primary system was required to reduce and maintain the RCS temperature below the cold shutdown condition.






14.09.2017 12:20 Current and future reactors

Nuclear power plant operation and new reactor technologies - 1005

AP1000 passive cooling containment analysis with computational fluid dynamics codes

Gonzalo Jimenez1, Zurine Goni1, Gonzalo Del Río1, Samanta Estévez1, César Queral2

1Universidad Politechnica de Madrid (UPM), Alenza 4, 28003 Madrid, Spain

2Universidad Politécnica de Madrid (UPM), Nuclear Engineering Department, José Gutiérrez Abascal 2, 28006 Madrid, Spain

gonzalo.jimenez@upm.es

 

The passive safety systems of the AP1000 nuclear reactor are based on natural phenomena to ensure containment integrity during an accident. One of the most important passive systems is the Passive Containment Cooling System (PCS), responsible for cooling the containment in operation and during an accident, guaranteeing the core residual heat evacuation.
The main objective of this research is to simulate the performance of this passive refrigeration system under operating conditions and in a LBLOCA type accident. Simulations were performed with the codes STAR-CCM+ 10.04 and GOTHIC 8.1 and reproduce the air flow established by natural circulation to ensure the reactor containment cooling.
The results of these simulations show how the cooling caused by the natural circulation is enough to evacuate the reactor residual heat. The temperature and flow stabilization during the transient evidences a balance in the heat exchange between the air flow and the containment surface, maintaining a suitable heat transfer in the containment.
In addition, the different simulations manifest how both the air flow and the flow pattern formed depend on the containment temperature, and the geometry of the components of the Passive Cooling System and the Shield Building. The Shield Building model is a part of the AP1000 full containment model with the GOTHIC code, and it has been connected thermally to the inner containment model to simulate dynamically the full PCS actuation.






14.09.2017 12:40 Current and future reactors

Nuclear power plant operation and new reactor technologies - 1001

Nuclear Safety Program of the Cernavoda Nuclear Power Plant

Dan Serbanescu

National Company "NUCLEAR ELECTRICA" s.a., 33 Blvd Gh. Magheru, P.O.Box 22-102, 70164 BUCHAREST 1, Romania

dserbanescu@nuclearelectrica.ro

 

Romanian National Nuclear Company (Societatea Nationala Nuclearelectrica S.A.) is the owner of two CANDU 6 operating units in Cernavoda. Both units are in mature operation phase and they are operating at high safety performance indicators, with very good production objectives achieved.
Maintaining and improving continuously the safety level of these units is a high priority for the company. The Cernavoda NPP units 1 and 2 are operating in full compliance with the national regulatory envelope aligned at the best practice level for CANDU reactors and with the latest safety requirements defined at international and national level as a post Fukushima set of feedback actions.
The items underline the actions to strengthen and maintain further compliance with the safety requirements of the regulatory envelope and the specifics of application of those actions for the two units. Each unit is in a different phase of the mature operation: unit 1 is starting the preparation for refurbishment, which is expected to be started in a medium term, while unit 2 is still in a period of full mature operation.
The paper presents the most important aspects of the safety aspects of the Cernavoda NPP and of the post Fukushima action plan






14.09.2017 12:00 Education and training

Regulatory issues, legislation, sustainability and education - 1106

An approach to attract, retain and develop new nuclear talents beyond academic curricula: the ENEN+ Project

Pedro Diequez Porras1, Csilla Pesznyak2, Behrooz Bazargan-Sabet3, Abdesselam Abdelouas4, Filip Tuomisto5, Leon Cizelj6, Michele Coeck7

1European Nuclear Education Network Association, Centre CEA de Saclay, INSTN, Bldg. 395, F-91191, Gif-sur-Yvette Cedex, France

2Budapest University of Technology and Economics, Műegyetem rkp 3-9, Budapest 1111, Hungary

3Laboratoire de Science et Génie des Matériaux Métalliques, CNRS, UMR 7584, Ecoles Mines, Nancy, France

4École des Mines de Nantes, 4, rue Alfred Kastler - La Chantrerie, CS 20722 44307 Nantes cedex 3, France

5Aalto University, School of Science, P.O. Box 11000, 00076 Aalto, Finland

6Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

7SCK CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 MOL, Belgium

sec.enen@cea.fr

 

The ENEN+ project proposes cost-effective actions to attract develop and retain new talents in the nuclear professions. This is seen as an essential part of the common strategic goal of all nuclear stakeholders, which is to preserve, maintain and further develop the valuable nuclear knowledge for todays and future generations. This is to be achieved by pursuing the following main objectives:
• Attract new talents to careers in nuclear.
• Develop the attracted talents beyond academic curricula.
• Increase the retention of attracted talents in nuclear careers.
• Involve the nuclear stakeholders within EU and beyond.
• Sustain the revived interest for nuclear careers.
The ENEN+ consortium will focus on the learners and careers in the following nuclear disciplines:
• Nuclear reactor engineering and safety,
• Waste management and geological disposal,
• Radiation protection and
• Medical applications.
The presentation will outline the most important expected contributions of the projects and the approaches taken to deliver those contributions.






14.09.2017 12:20 Education and training

Regulatory issues, legislation, sustainability and education - 1107

Nuclear Education in Slovakia

Ján Haščík

Slovak University of Technology, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava 1, Slovakia

jan.hascik@stuba.sk

 

Slovakia has limited and low quality fossil fuels resources. Moreover, almost eighty percent of its hydro potential is utilized. To face projected growth of electricity demand, Slovakia has to rely on nuclear power plants electricity production. At present more than 56% of the electricity consumed in the Slovak Republic comes from the production of domestic nuclear power plants. As predicted, this share will increase in the future. Thus, it is necessary to sustain the high quality nuclear education at secondary schools and especially at universities. The paper discusses the possibilities and adopted strategies of nuclear education at universities in the Slovak Republic. Three-level structure of university education provides opportunities to ensure adequate training of nuclear specialists for the needs of the nuclear industry in the Slovak Republic and engagement in the international network of nuclear education. Since 1962, hundreds of nuclear specialists have been trained in SVŠT in Bratislava (Slovak Technical University predecessor) and effectual postgraduate courses and international courses have been organized. The wide range of activities related to nuclear education are also presented in the paper.






14.09.2017 12:40 Education and training

Regulatory issues, legislation, sustainability and education - 1116

The first 15 years of the ENEN Association

Joerg Starflinger1, Filip Tuomisto2, Behrooz Bazargan-Sabet3, Veronique Decobert4, Pascal Anzieu5, Michele Coeck6, John Roberts7, Tzany Kokalova Wheldon8, Leon Cizelj9, Pedro Diequez Porras10

1Institute of Nuclear Technology and Energy Systems (IKE) University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart, Germany

2Aalto University, School of Science, P.O. Box 11000, 00076 Aalto, Finland

3Laboratoire de Science et Génie des Matériaux Métalliques, CNRS, UMR 7584, Ecoles Mines, Nancy, France

4Westinghouse Nuclear Services, 86, rue de Paris, 91401 Orsay, France

5CEA-INSTN Institut national del sciences & techniques nucleaires, Bât.121, 91191 Gif-sur-Yvette Cedex, France

6SCK CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 MOL, Belgium

7The University of Manchester, Materials Performance Centre, School of Materials, PO Box 88, M60 1QD Manchester, United Kingdom

8School of Physics & Astronomy, University of Birmingham, Edgbaston, Birmingham B15 2TT, United Kingdom

9Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

10European Nuclear Education Network Association, Centre CEA de Saclay, INSTN, Bldg. 395, F-91191, Gif-sur-Yvette Cedex, France

joerg.starflinger@ike.uni-stuttgart.de

 

The European Nuclear Education Network (ENEN) was established in 2003 through a EU Fifth Framework Programme (FP) project, as a legal nonprofit-making body. Its main objective is the preservation and further development of expertise in the nuclear fields by higher education and training. This objective is realized through the cooperation between EU universities involved in education and research in nuclear disciplines, nuclear research centers and the nuclear industry. As of March 2017, ENEN has 53 members in 18 EU countries and has concluded Memoranda of Understanding (MoU) with partners beyond Europe for further cooperation, including organisations in South Africa, Russian Federation, Ukraine, Canada and Japan. ENEN also has good collaboration with national networks and international organizations such as the Belgian Nuclear Education Network (BNEN) and the International Atomic Energy Agency (IAEA).
The main activities developed, and results achieved, within the first 15 years of the ENEN Association will be presented and discussed. These include, for example, the launch of the European Master of Science in Nuclear Engineering (EMSNE), the annual ENEN Ph.D. competition and the portfolio of more than 10 EURATOM projects dealing with nuclear education, training and knowledge management through development of teaching methods and materials, courses, and exchange of students and teachers within EU and beyond. Those projects were all supported by the European Commission with the ENEN Association acting as the coordinator or as a partner.






14.09.2017 14:20 Invited Benoit Tanguy

Invited lectures - 106

Irradiation embrittlement of austenitic stainless steels in PWR vessel’s internals

Benoit Tanguy

CEA, Member of SNETP Executive Committee, Gif sur Yvette 91191, France

benoit.tanguy@cea.fr

 

In PWR’s, internals made of austenitic stainless steels (Cold Worked (CW) 316 austenitic steels for bolts, 308 for welds and Solution Annealed (SA) 304 for baffle plates, former and core barrel) are subjected to irradiation embrittlement with doses that will reach up to 80 dpa locally after 40 years of operation at temperatures between 280°C and 380°C. The irradiation exposure alters the nanostructure and so consequently do the mechanical properties of stainless steels: it increases the yield strength, decreases ductility and promotes plastic instabilities. Moreover, neutron irradiation at low temperature generally causes a reduction in fracture toughness of stainless steels. In addition, these changes seem to be the basis of an increased sensitivity to stress corrosion cracking said to be assisted by irradiation (IASCC). The tensile properties evolutions are commonly ascribed to the formation of a high density of nano-sized irradiation defects clusters mainly fine dislocation loops in stainless steels. Others defects (e.g. precipitates, bubbles and voids) may develop at higher doses. Homogeneous deformation that is observed for virgin material or at low doses is shifted to heterogeneous deformation at higher doses. Ductility loss is mainly ascribed to increased plastic strain localization. So far, much effort has been made to identify radiation effects on material microstructure and to determine experimentally the evolution of mechanical properties. Security and reliability of nuclear power plants have become a main issue and a better understanding of the physical mechanisms leading to the change in mechanical properties is now required. In addition it is also necessary to develop micromechanical constitutive equations to be able to carry out numerical simulations of core internals structures to improve safety and better study the possibility of extending the lifetime of nuclear power reactors.
This talk will present some recent works developed to assess the irradiation embrittlement of austenitic stainless steels in PWR’s environment.
In the first part of the paper, recently proposed crystal plasticity physically-based constitutive equations for neutron-irradiated austenitic stainless steel that have been used in this study are described. Tensile curve predictions are presented and compare to experimental data.
Irradiation leads to voids formation in grains of austenitic stainless steel. When these austenitic stainless steels are subjected to loadings, ductile fracture happens as the results of void growth and coalescence.
The second part of this talk will be devoted to the description of a multiscale modelling of growth and coalescence of pre-existing voids leading to transgranular ductile fracture of irradiated austenitic stainless steels.
Last part of the paper will be devoted to IASCC.






14.09.2017 15:00 Research reactors - plenary

Research reactors - 501

STRIGA - A COMPUTER TOOL FOR TRIGA RESEARCH REACTOR

Dušan Calić1, Žiga Štancar2, Luka Snoj3

1Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia

2Jožef Stefan Institute, Reactor Physics Department , Jamova cesta 39, 1000 Ljubljana, Slovenia

3Jožef Stefan Institute, Reactor Physics Department, Jamova cesta 39, 1000 Ljubljana, Slovenia

dusan.calic@ijs.si

 

TRIGA Mark II research reactor at Jožef Stefan Institute is operating more than fifty years and has more than 200 core configurations changed. Determining the current fuel composition is very important to calculate various parameters. Current reactor burnup calculations are performed using TRIGLAV code. TRIGLAV code is based on diffusion model in two dimensions where the group constants are calculated using lattice cell code WIMSD.
New techniques, methods, software and increase of the processing capacity of the new computers motivates us to develop STRIGA package. Based on a loading scheme for TRIGLAV code we have developed a tool that can read old TRIGLAV inputs and make a 3D Monte Carlo Serpent input for TRIGA research reactor. This is a very detailed 3D model that has already been validated on core configurations 132 and 133.
With the use of STRIGA package we can perform the full burnup history of TRIGA research reactor without major effort since the code only requires simple loading scheme and some informations about previous burnup cycles. Also some utility codes which performers various functions: such as library generation and data management are part of the STRIGA package. In this paper the package will be presented since it is very useful for inexperienced and experienced Serpent users.






14.09.2017 15:20 Research reactors - plenary

Research reactors - 502

Installation of a Thermal White Neutron Beam Facility at the TRIGA reactor in Vienna

Wilfried Mach1, Erwin Jericha1, Michael Bacak1, Dieter Hainz1, Andreas Musilek2, Mario Villa2, Hartmut Abele1

1Atominstitut, Schüttelstr.115, A-1020 Wien, Austria

2Technical University Vienna, Atominstitut, Stadionallee 2, 1020 Vienna, Austria

wilfried.mach@ati.ac.at

 

The TRIGA research reactor of the Atominstitut in Vienna has celebrated its 55th birthday this year but the science around the neutron source stays up-to-date and vivid. As a commitment to the future new fuel elements were installed in 2012, the reactor instrumentation and control system was renewed in 2016 and the whole radiation protection system was renewed completely this year.
Furthermore, a new multi-purpose instrument was installed this year: the thermal white neutron beam. Its high thermal neutron flux of ~107 neutrons/cm2s will be available for several very different new experiments. The new reactor instrumentation offers the possibility to pulse the reactor for achieving a peak flux of ~1x1010 neutrons/cm2s with this instrument. The accessible experimental site has a size of ~3.7x3.7x2.2 m3, the cross section of the beam is 65x65 mm2.
High flux and a direct view to the core lead to high dose rates. Therefore, perfect single crystals made of sapphire and bismuth were installed inside the reactor beam tube to filter out fast neutrons and gamma radiation to decrease the total dose rate. Further, a special kind of radiation absorbing concrete was designed with boron carbide (B4C), serpentine and hematite as additives to increase neutron absorption, moderation and gamma absorption. The performance of this new radiation protection concrete was simulated with MCNP6 and also tested at the TRIGA reactor. Then the shielding of the instrument was simulated in MCNP6.
Several very different experiments can be performed with this new neutron instrument. Fundamental physics will be pushed by investigating new devices for neutron experiments as on neutron beta decay. New tools like neutron detectors with µm-resolution or new kind of neutron optics can be developed faster with this instrument. The radiation damage of semiconductors will be investigated. The new instrument also can be used as a station for neutron radiography.
The wide variety of experiments shows the high flexibility of the new instrument and the broad impact on neutron physics.
This presentation will introduce the new neutron beam instrument and some future experiments and will also discuss technical details about the work on MCNP6 and the development of the special radiation absorbing concrete.






14.09.2017 15:40 Research reactors - plenary

Research reactors - 503

A 3D CFD Model for the Study of Natural Circulation in the Pavia TRIGA Mark II Research Reactor

Carolina Introini1, Antonio Cammi2, Stefano Lorenzi2, Davide Baroli3, Bernhard Peters3, Davide Chiesa4, Massimiliano Nastasi4, Ezio Previtali5

1Politecnico di Milano - department of energy, Via La Masa 34, 20156 Milano, Italy

2Politecnico di Milano, Dipartimento di Energia, Via Ponzio 34/3 , 20133 Milano, Italy

3University of Luxembourg, 2, avenue de l'Université, 4365 Esch-sur-Alzette, Luxembourg

4Universita degli Studi di Milano-Bicocca, Piazza dell'Ateneo Nuovo, 1, 20126 Milano, Italy

5INFN, Largo Enrico Fermi, 2, I-50125 Firenze, Italy

carolina.introini@mail.polimi.it

 

The TRIGA (Training, Research, Isotopes, General Atomics) Mark II reactor of the University of Pavia is a pool-type research reactor. It was brought to its first criticality in 1965 and since then it has been used for several scientific and technical applications, including reactor physics studies, training and educational activities thanks to the possibility of performing experimental measurements for the validation of reactor modelling codes. One of the most challenging reactor feature to be considered in the modelling task is the natural circulation. Indeed, there is no primary pump in the pool and the heat removal from the core is achieved by free convection. The driving force behind natural circulation is buoyancy, which results from the difference in fluid density between the reactor core inlet and outlet. The modelling of this phenomenon requires the modelling of not just the core of the reactor but it should include the pool in order to consider the total pressure drop along the core that establishes the equilibrium flow rate. In this work, we present a 3D thermal-hydraulic model of the TRIGA Mark II aimed at assessing the natural circulation capabilities of the reactor. In this light, a CFD (Computational Fluid Dynamics) approach is used, considering also the complexity and the asymmetric geometry of the reactor. The contribution of some components of the reactor core (e.g., the lower grid) to the pressure drop is modelled with a porous media approach, the latter been tuned with a separate detailed model. As a major outcome, the model is able to provide the water mass flow rate induced by the natural circulation for each channel in the reactor, giving a fundamental input for simplified thermal hydraulic model of the reactor core.





Dates to Remember
January 31   Call for papers
April 30
May 15
  Abstracts submittal
June 30   Acceptance of abstracts
August 20   Young author papers
August 31   Full-length papers
September 11-14   Conference
November 30   Proceedings

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