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09.09.2013 19:00 Invited lecture 1

Invited lectures - 101

Building a New Safety Construct for Energy: Bringing the Socio-Political Factors to Bear

Marc G. Goldsmith1,2

American Society of Mechanical Engineers, Two Park Avenue, New York, NY 10016-5990, USA1

Marc Goldsmith & Associates LLC2

marc@mgallc.net

 

In 2011, following the Fukushima Dai Ichi Earthquake and Tsunami, ASME formed a Presidential Nuclear Safety Task Force to investigate and develop lessons learned from the accident. Chaired by former U.S. Nuclear Regulatory Commission Chairman Nils Diaz and Westinghouse Executive Regis Matzie, the Task Force developed a new Nuclear Safety Construct (1) that takes into account the socio-economic and political impacts of failure in a complex technical system. The report focuses on nuclear safety, although there are implications for other energy sources and for electric power in a broader sense. This paper brings some of the safety construct into a broader discussion of electric power, the role of nuclear power and the importance of balancing supply resources.






10.09.2013 09:00 Invited lecture 2

Invited lectures - 108

Severe accident research at Paul Scherrer Institute in Switzerland

Horst-Michael Prasser1,2

Paul Scherrer Institut, Villigen, CH-5232 Villigen PSI, Switzerland1

ETH Zurich, Institute of Energy Technology, ML K 13, Sonneggstrasse 3, 8092 Zürich, Switzerland2

hprasser@ethz.ch

 

Severe accident research has a long tradition and a preponderant place in the portfolio of the Laboratory of Thermal Hydraulics at the Paul Scherrer Institute. For many years, the emphasis was put on on aerosol retention in various components of the reactor (suppression pools (POSEIDON), passive containment coolers (AIDA, CONGA)) and recently on aerosol retention within the flow path during bypass scenarios caused by spontaneous and induced steam generator tube rupture scenarios. The most prominent activity, which found a wide international recognition, was the eight years long ARTIST (Aerosol Trapping In STeam generator tube rupture) project with its eight different phases. Its purpose was the assessment of aerosol retention in case of a core damage combined with a spontaneous or induced steam generator tube rupture. The decontamination factors were determined in experiments with scaled models of all components found in the flow path of the steam generator, some of them in original scale (e.g. the cyclone separators of the steam generators). The effect of both dry and flooded steam generator bundle was studied. In an ongoing activity (PASSAM) which investigates active and passive systems for source term mitigation, high-resolution measurements of the interfacial structure formed by vapor injection into a water pool and into a flooded bundle are performed at a new adiabatic non-pressurized test facility (TRISTAN) with wire-mesh sensors. The goal is the quantification of the interfacial area density as one of the determining parameters influencing the scrubbing efficiency.

Aerosol research at LTH is not limited to empirical quantification of transport and retention by way of experimentation in different scales, but there are considerable efforts put on the fundamental modeling by a multi-scale approach. Main working horses are CFD models ranging from DNS and LES coupled to a Langrangian particle transport model, up to the implementation in RANS and URANS of turbulence aerosol transport models based on stochastic random walk. The implementation of a model for turbophoresis in RANS (FLUENT) allowed reproduction of experimental results on aerosol deposition in many configurations, e.g. turbulent pipe or elbow flows, steam generator tube with active thermophoresis, etc. Here, the focus is put on particle dispersion, clustering and deposition. For model validation, a differentially heated cavity is used as test case for both simulations and experiments. Another important fundamental research is directed towards the experimental and theoretical modeling of the breakup of aerosol agglomerates by turbulence and in case of a wall impaction. Aerosol research has also found spin-offs in fields other than severe accidents, e.g. dust deposition in clean rooms, or in human airways, and it may be used for modeling graphite dust transport in HTRs.PSI has furthermore developed the infrastructure to test filtered venting devices. Aerosol scrubbing can be studied in a wet filter system represented by a single injection nozzle (VEFITA). Iodine retention efficiency can be measured, as well. PSI has developed a method for retention of organic iodine based on the enhancement of the efficiency of Thiosulphate by a ternary amino compound used as a catalyst.

Another important field with growing international attention, in particular after Fukushima, is cladding oxidation in spent fuel pools and in case of air ingress into the primary system. PSI is participating in the SANDIA Spent Fuel Project and has developed an own model for the departure from the protective parabolic oxidation model based on a transition region of the oxide layer thickness in which kinetics change from a diffusion regime controlled by an effective oxide layer thickness to a linear regime. It was implemented in SCDAP/RELAP and MELCOR. KIT and IRSN single effect tests are used for the validation. Both, the influence of oxygen and nitrogen are taken into account.

PSI has the ability to perform integral severe accident plant analyses using MELCOR and SCDAP/RELAP5. An example is the investigation of Severe Accident Management measures for the case of long-lasting Station Blackout, applied to both PWRs and BWRs. PSI furthermore takes part in the international Fukushima Benchmark. In connection with the Fukushima accident, at the early stages of the accidents PSI carried out investigation to determine the extent of damage in the spent fuel pool based on the activity ratio measurements of the different radioactive isotopes.

Experimental and theoretical work is carried out to address the safety of the steam generators, both in regimes relevant for DBA and BDBA conditions. CFD Calculations of the superheated gas flow in the hot leg and the steam generator primary side in case of a dry core predict circulation patterns that may lead to an induced failure of steam generator HX tubes. This leads to a containment by-pass whereas a rupture of the hot leg or of the pressurizer surge line will lead to the release into the containment. The small predicted time interval between the different failures underlines the big influence of uncertainties on the probability of an induced core bypass and implies the need of a careful experimental validation. PSI therefore proposed to construct a dedicated test facility for studying the mixing in a steam generator mockup including a core simulator, a hot leg, a tube bundle and inlet and outlet plena, with different geometries. Another issue is reflux-condensation in the steam generator tubes in case of DBA and BDBA conditions. The availability of experimental data obtained at prototypical conditions with regard to pressures, temperatures, the presence of non-condensable gases and finally aerosols, is still not satisfying. A test facility is being build that allows to study the condensation under the mentioned conditions in a shortened segment of a single steam generator tube. A test with full-scale steam generator tubes was proposed for an international project. The small test facility under construction is equipped with innovative heat flux sensors and needle probes that combine the measurement of the temperature with a phase detection in the condensate and the boundary layer from the side of the gaseous phase.

Last but not least, the laboratory of Thermal Hydraulics at PSI operates the large-scale containment test facility PANDA. Currently it is used to study the transport, accumulation and dispersion of hydrogen in a complex geometry under conditions of the containment atmosphere during different accident scenarios. Experiments in the extensively instrumented test facility are performed with helium as model fluid. Highlights are detailed studies on the erosion of a hydrogen-rich layer by different jets and plumes, the effect of gas plumes exiting from recombiners, the effect of catalytic recombiners and of containment spraying. A PANDA test was selected for an international OECD code benchmark on hydrogen layer erosion within the CFX4NRS network.






10.09.2013 09:40 Severe accidents

Severe accidents - 402

A Comparison of Core Degradation Phenomena in the CORA, QUENCH, Phébus SFD and Phébus FP Experiments

Tim Haste1, Martin Steinbrück2, Marc Barrachin1, Olivier De Luze3, Mirco Grosse2, Juri Stuckert2

Institut de Radioprotection et de Sureté Nucléaire, Bât. 702 Centre de Cadarache, BP 3-13115 Saint Paul lez Durance, France1

Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany2

Institut de Radioprotection et de Sureté Nucléaire, BP 3 , 13115 Saint Paul Lez Durance Cedex, France3

tim.haste@irsn.fr

 

Over the past twenty years, integral fuel bundle experiments performed at IRSN Cadarache, France (Phébus-SFD and Phébus FP – fission heated) and at Karlsruhe Institute of Technology, Germany (CORA and QUENCH – electrically heated), accompanied by separate-effect tests, have provided a wealth of detailed information on core degradation phenomena that occur under severe accident conditions, relevant to such safety issues as in-vessel retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. These data form an important basis for development and validation of severe accident analysis codes such as ASTEC (IRSN/GRS, EC) and MELCOR (USNRC/SNL, USA) that are used to assess the safety of current and future reactor designs, so helping to reduce the uncertainty associated with such code predictions.

Following the recent end of the Phébus FP project, it is appropriate now to compare and contrast the core degradation phenomena observed in these four major experimental series, indicating the main conclusions that have been drawn. This covers subjects such as early phase degradation up to loss of rod-like geometry (all the series), late phase degradation and the link between fission product release and core degradation (Phébus FP), oxidation phenomena (all the series), reflood behaviour (CORA and QUENCH), as well as particular topics such as the effects of control rod material and fuel burn-up on core degradation. It also outlines the separate-effects experiments performed to elucidate specific phenomena such as chemical reactions involving boron carbide absorber material. Finally, it indicates the remaining topics for which further investigation is still required and/or is under way.






10.09.2013 10:00 Severe accidents

Severe accidents - 415

Simulation of hydrogen deflagration experiment – Benchmark exercise with lumped-parameter codes

Ivo Kljenak1, Mikhail Kuznetsov2, Giovanni Manzini3, Pál Kostka4, Lubica Kubišova5, Mantas Povilaitis6

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany2

Ricerca sul Sistema Energetico – RSE S.p.A., via R. Rubattino 54, 20134 Milano, Italy3

NUBIKI Nuclear Safety Research Institute, Konkoly-Thege Miklós út 29-33. building 6, 1121 Budapest, Hungary4

Úrad jadrového dozoru SR, P.O.Box 24, 82007 Bratislava, Slovakia5

Lithuanian Energy Institute, Breslaujos str. 3, LT-44403 Kaunas, Lithuania6

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.

In 2012, the Upward Flame Propagation Experiment (UFPE) on hydrogen combustion, which was proposed by the Jozef Stefan Institute (JSI), was carried out at the Karlsruhe Institute of Technology (KIT) in Germany. The experiment was performed in the HYKA A2 facility, which is basically a cylindrical vessel with a volume of 200 m3. The initial conditions were the following: pressure 1.5 bar, temperature 90 °C, hydrogen concentration 12 vol.% and steam concentration 20 vol.%. The hydrogen-steam-air mixture was ignited near the bottom of the vessel, and the ensuing flame propagation was observed.For safety analyses of nuclear power plants, hydrogen deflagration in a containment should be adequately simulated. Although models of hydrogen deflagration, implemented in lumped-parameter codes (which model the reactor containment as a network of control volumes, in which the conditions are in principle modelled as homogeneous) do not reflect all the complexity of the combustion process, such codes are still likely to be used. Namely, the containment of actual NPPs are much too large for combustion to be simulated using so-called Computational Fluid Dynamics (CFD) codes, which solve the basic transport equations of fluid mechanics and additional constitutive equations on the local instantaneous scale.A benchmark exercise was organised by JSI, with the purpose to compare simulation results of the UFPE experiment, obtained either with different lumped-parameter codes or with same lumped-parameter codes but different input models. Apart from the organizer who participated with the ASTEC code, the following organisations also took part in the benchmark:- RSE (Italy) with the ECART code,- NUBIKI (Hungary), with the ASTEC code,- UJD SR (Slovakia), with the COCOSYS code,- LEI (Lithuania), with the COCOSYS code.In the proposed paper, the results obtained with different codes (pressure and temperature increase and flame propagation) are compared to the experimental results, and the merits of individual codes are discussed.






10.09.2013 10:20 Severe accidents

Severe accidents - 404

MELCOR Modeling of Fukushima Unit 2 Accident

Tuomo Sevon

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

tuomo.sevon@vtt.fi

 

A MELCOR model of the Fukushima Daiichi unit 2 accident was created. Plant data for the model was collected from various public sources available in March 2013. Missing pieces of plant data were taken mainly from the Peach Bottom plant in the United States, using approximate scaling factors due to the smaller size of the Fukushima reactor.

The calculated reactor water level and reactor and containment pressures were compared with the measurements. The RCIC (Reactor Core Isolation Cooling) system and the containment leakage model were adjusted so that the difference between the calculations and the measurement data is as small as possible. A significant uncertainty is related to the seawater injection rate by the fire engine. The calculation results are quite near to the measurements for most of the time. However, it is difficult to explain why the measured containment pressure did not increase at 75 hours after the earthquake, when the operators made the reactor depressurization. Calculated radionuclide release to the environment will be included in the paper.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 202

Analysis of LBLOCA using Best Estimate Plus Uncertainties for Framatome 3-Loop Plant Power Uprate

Dong Gu Kang, Byung-Gil Huh, Seung Hun Yoo, Chae-Yong Yang, Kwang-Won Seul

Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

littlewing@kins.re.kr

 

The best estimate (BE) calculation with uncertainty evaluation of large break loss-of-coolant-accident (LBLOCA) has been increasingly applied to the licensing applications such as fuel design change, power uprate, and licensing renewal of nuclear power plants (NPPs). The KINS-Realistic Evaluation Model (KINS-REM) was developed for the independent audit calculation based on Best Estimate Plus Uncertainty (BEPU) method pioneered by the Code Scaling, Applicability, and Uncertainty (CSAU) methodology, and the code accuracy and statistical method have been improved. To support the licensing review and to confirm the validity of licensee’s calculation, regulatory auditing calculations have been also conducted. Currently, the modification of Ulchin 1&2 (Flamatome 3-loop plant) operating license for 4.5% power uprate is under review. In this study, the LBLOCA analysis of Ulchin Unit 1&2 with 4.5% power uprate was performed by applying KINS-REM. The MARS-KS 1.3 code was used as a frozen BE code, and 18 uncertainty parameters were considered in the analysis. Several code models including the gap conductance, heat transfer in the core were selected as the major model uncertainty parameters affecting the important phenomena during LBLOCA. For the plant parameters, core power, accumulator pressure and temperature, etc. were considered. By assuming a conservative bounding value for some parameters such as peaking factor and containment pressure, they were excluded from the set of uncertainty parameters treated statistically. The 95 percentile peak cladding temperature (PCT) was determined by 124 calculations based on Wilk’s formula of the non-parametric statistics, and additional biases for ECC bypass and steam binding were added to account for untreatable phenomena and models. It was confirmed that the analysis results of LBLOCA for Ulchin 1&2 power uprate met the PCT acceptance criteria.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 204

Thermal-hydraulics modeling and analyses of the supercritical water loop

Jelena Zmitkova

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

zmj@cvrez.cz

 

Research Centre Rez (CVR) is a research company focusing a significant part of its activities to realization and implementation of the Sustainable Energy Project (SUStainable ENergy, SUSEN). This project is aiming at developing a robust infrastructure for sustainable R&D activities to support the Czech participation on European effort for safe and efficient energy generation in Europe. The objective of this research program is to build up large-scale experimental facilities allowing research and development in the area of generation IV nuclear reactor concepts and in the field of fusion reaction. Experimental data acquired from such facilities will extend the existing knowledge of material properties and behavior in specific conditions, and will be used during the development of the given type of reactor.

One of the R&D directions is dedicated to the supercritical water-cooled reactor concept (SCWR). Supercritical water-cooled reactor SCWR is high temperature, high pressure reactor cooled by supercritical light water (operational parameters in primary circuit are: temperature 600°C, pressure 25 MPa). Due to this parameters and simplification of the facility (in comparison to current power plants) it is possible to achieve the thermal efficiency up to 45%. The concept combines the advantages of pressurized water reactor, boiling water reactor and fossil-fueled supercritical water-cooled power plant to ensure good utilization of fuel.Among the developed experimental facilities and infrastructure in the framework of the SUSEN project is construction and experimental operation of the supercritical water loop SCWR. At the first phase, this SCWR loop is assembled and operated out-of-pile in the dedicated loop facilities hall. At this out-of-pile operation various operational conditions are tested and verified. After that, in the second phase, the SCWR loop will be situated in-pile, in the core of the research reactor LVR-15, operated in CVR.This paper describes the design of the SCWR loop and presents the results of its thermal-hydraulics analyses using the RELAP5/MOD3.3 computer code. The thermal-hydraulics modeling and the performed analyses are focused on the SCWR loop model validation through a comparison of the calculation results with the experimental results obtained at various operational conditions. The results obtained by the validated SCWR loop model will be used in the safety documentation to be presented to the Safety Authority when the SCWR loop is placed into the research reactor LVR-15. In addition, the performed calculations and validation exercise make it possible to improve the calculation model and to propose an update/modification of the code itself.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 205

Evaluation for Effect of Upper Head Nodalization and Temperature in OPR1000 Plant

Byung-Gil Huh, Seung Hun Yoo, Dong Gu Kang, Kwang-Won Seul

Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

k686hbg@kins.re.kr

 

Nowadays, the best estimate (BE) method with the uncertainty evaluation has been broadly used in Korea for licensing of NPP. In the BE method, the nodalization for base calculation was defined from the comparative study of the experimental data and it could influence the accuracy of code for specific phenomena such as ECC bypass and blowdown quenching. The nodalization of upper head for OPR1000 and APR1400 was generally composed of axial several single volumes and the flow recirculation between the upper plenum and the upper head/dome was not considered significantly since the recirculation rate was much smaller than the primary coolant flow rate. However, there are actually the upward and downward flows in the upper head and this recirculation flow can transfer the heat of the upper plenum to the upper dome. Up to now, it was assumed that the temperature of the upper dome was close to that of the cold leg in most LOCA analyses to consider the effect of bypass flow from the downcomer to the upper dome. However, it was recently found that the temperature of the upper dome might be larger than that of the cold leg according to the heat exchange due to the recirculation flow and the detailed design data. Therefore, the high upper head temperature can influence the modification of nodalization for upper head and the blowdown quenching behavior. In this study, the LBLOCA in OPR1000 was evaluated to identify the effects of the upper head nodalization and the temperature.

Actually, the upper head temperature was determined by a key-ways bypass and CEA guide flow rate. In order to consider of the flow exchange between the inner- and the outer-region of the CEA shroud assembly, two nodalization methods were considered instead of the conventional upper head nodalization (Case-0). One is that the guide structure was modeled as 5 single volumes and the single junction was connected from the upper guide structure to the upper head (Case-1). The other is that the upper head was separated into 2 axial volumes to simulate the actual recirculation flow (Case-2). Two axial volumes were connected each other with the cross flow junctions. From the calculation results, as the downcomer bypass flow increases, the temperature of upper head decreases due to the increment of cold water which flows into the upper dome. The highest upper head temperature was shown in Case-2. Compared to Case-0, the cladding temperature behaviors for Case-1 and Case-2 have the following characteristics, 1) Reduction of the depth for blowdown quenching, 2) Increase of the reflood temperature and 3) Delay of the fuel final quenching. Also, the effect of nodalization change was also evaluated in the KINS-Realistic Evaluation Methodology (KINS-REM). The 9 cases, which have a high reflood temperature, were selected as the evaluation targets. For Case-1 and Case-2, blowdown quenching depth was reduced dramatically and the reflood PCT was increased by ~ 100 K in comparison with the results for Case-0.In conclusion, if the nodalization is changed to exchange the flow between the upper plenum and the upper head/dome, the temperature of the upper dome increases and the blowdown quenching happens at the higher temperature. Also, the modification of nodalization influences the BE method with the uncertainty and the PCT behavior can be changed. The more detailed analysis for the effect of nodalization would be needed to consider the temperature distribution in the core appropriately.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 206

Powder’s Conductivity Measurements in the TxP Facility

Davide Rozzia1, Giuseppe Fasano2, Mariano Tarantino2, Pietro Agostini2, Alessandro Alemberti3, Oriolo Francesco1

Universita degli Studi di Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione, Largo Lucio Lazzerino 1, 56100 Pisa, Italy1

ENEA CR Brasimone, Localita Brasimone, 40032 Camugnano (BO), Italy2

Ansaldo Nuclear S.p.a., C.so F.M. Perrone 25, 16152 Genova, Italy3

daviderozzia@libero.it

 

One of the most promising GEN-IV concepts is the Lead cooled Fast Reactor (LFR). The actual conceptual design of LFR deals with the compact pool type reactor in which the steam generators (SG) are located inside the reactor tank. The Lead cooled European Advanced Demonstrator Reactor (LEADER) is an European project that belongs to the seventh Framework Programme (FP-7) and is aimed to develop a 300Mth Advanced Lead cooled Fast Reactor European Demonstrator (ALFRED). In this framework, a new configuration of SG has been proposed for ALFRED: the super-heated steam double wall once through bayonet type. An example of facility that operates with this concept is CIRCE (ENEA Brasimone), nevertheless the application is limited to the heat exchange function.

The study of this configuration is motivated by safety improvement. In fact, it allows the double physical separation between lead and water sides by mean of an intermediate gap. Furthermore, gap pressurization (with Helium), allows to check leakages from the system in order to minimize the probability of incident scenarios (i.e. steam generator tube rupture with lead - water interaction).Each SG includes about 500 bayonet tubes, operates at 180 bar (water side) and should generate superheated steam at 450 °C. Therefore, R&D is necessary to check and improve (if necessary) its TH performance. In particular, due to the introduction of a double wall tube with annular gap, the thermal performance of the Steam Generator (SG) could be decreased.The present work aims to determine a candidate material as gap filler with the purpose to increase the SG thermal efficiency. In order to fulfil the objective, a specific facility for conductivity measurements has been designed and constructed: the Tubes for Powders (TxP) Facility. The facility consists of three concentric tubes instrumented with 48 thermocouples and relies on two annular gaps of different size for conductivity measurements. The paper will present the main results of the calibration tests and the first tests carried out at ENEA CR Brasimone.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 209

Computational Fluid Dynamic model of TRIGA Mark II reactor

Romain Henry, Iztok Tiselj

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

romain.henry@ijs.si

 

The objective of this work is to build a thermo-hydraulics model of the TRIGA Mark II research reactor at the Jozef Stefan Institute (JSI) by using the CFX 13.0 computational fluid dynamic (CFD) code. Indeed, CFD is a very accurate tool describing fluid flow mechanic by solving numerical algorithms.

This work is a step of a project aiming to build a computational model of TRIGA Mark II research reactor coupling thermal-hydraulics and neutron physics in order to see how each one influence each other and what are the main parameters involved in this process.Whereas TRIGA at JSI was studied in great detail from neutron physics point of view, models describing flow pattern inside the reactor tank do not exist. Indeed, such model is not needed from a safety point of view; however, the benefit of such research is to establish an accurate benchmark test for coupled thermal-hydraulics and neutronic models. Starting from a given power distribution, which was already been established as a main future input from neutron physics model, output such as temperature distribution and velocity field will be computed. In the paper, the model will be presented (geometry, mesh, boundary conditions…), verification and validation will be made from data collected during the many experiments performed at JSI. Finally results will be analysed.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 211

Evaluation of Different Heat Conduction Correlations of the Supercritical Water by the ATHLET code

György Hegyi, Andras Kereszturi, István Trosztel, Hunor György

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

gyorgy.hegyi@energia.mta.hu

 

The High Performance Light Water Reactor (HPLWR) cooled by supercritical water is one of the Gen4 concepts. The main advantage of this reactor type is the high thermal efficiency exceeding even 43 %. The supercritical water properties are essentially different from those at normal pressures and temperatures and – due to the complicated behavior in this region – sufficiently accurate analytical formulae could not be set up yet. Concerning the heat conduction between a heated surface and the coolant, several empirical Nusselt number correlations were elaborated in the past 70 years. The qualification of the different correlations was performed by comparing them to a large number of experiments.

In the frame of HPLWR project an appropriate design was chosen. When selecting the experiments, the parameter range of this reactor type was taken into account. The pressure values were in the range of 200- 230 bar, the heat flux 60 – 240 W/cm2, the inlet temperature 200 – 320 oC, the coolant enthalpy 850 – 1500 kJ/kg, the tube diameter 5 – 10 mm. Special attention was paid to the geometry and the measurement positions being also very important for the correct comparison of the measurement and the calculation.Ten supercritical heat conduction correlations were selected from the literature and programmed into the ATHLET system thermo-hydraulic code. After modifying the corresponding heat transfer correlations, a set of input decks specifying the different inlet mass flux, enthalpy values were prepared. The nodalizations were also different for each case, taking additional care for the measurement positions. Three hundred measurements were chosen and simulated by the ATHLET system thermo-hydraulic code. The comparison was based on the measured and calculated values of the coolant temperature and enthalpy increment, moreover, the wall temperatures and heat transfer coefficients were also compared.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 213

How to reduce Main Steam Line vibration

Tobias Zieger

IMI Nuclear, CCI AG, Itaslenstrasse 9, 8362 Balterswil, Switzerland

tzieger@iminuclear.com

 

Main Steam Isolation Valves (MSIV’s) are used during plant emergencies to isolate the containment from the environment. These valves are open during operation of the plant. If a Main Steam Line Breaks (MSLB) these valves have to close quickly and reliably within the specified closing time. This is usually in the range of 2 to 5 seconds. Especially in Nuclear Power Plants of the type Boiling Water Reactor (BWR) this isolation function is very important to guarantee sufficient cooling of the reactor and to avoid contamination of the environment by fission products, released with the steam.

MSIV’s are typically in a size range between 300mm (12in) and 600mm(24in). In the past they were usually of the globe style, Y-pattern design, actuated either by an electrohydraulic or a pneumatic system or by system medium. Since the valves are open during operation with a high steam flow thru the valves they cause a relatively high pressure loss. In the last years lots of plants went thru a Plant Life Extension (PLEX) project. Very often this PLEX projects comes along with a project to increase the power output of the plant. To increase the power the steam flow needs to be increased too. Unfortunately the pressure drop over the valves (and any other components too, by the way) increases with the square of the steam velocity.To increase the power from 100% to 120% the steam flow must be increased by about 20%. But this will increase the pressure drop over the valves and with this the loss of energy to from 100% to 144%.Therefore very often the globe style MSIV’s are replaced by gate valves. Gates valves cause typically only about 10% of the pressure loss of Y-pattern globe valves, provided the same size and the same flow are compared.In one plant the Y-pattern, globe style MSIV’s were replaced by gate valves. Before replacement a relatively high vibration level was measured at the MSIV’s. This is caused by the special flow pattern thru the valve in combination with the steam line isometry and fluid-born pressure pulsations in the steam line.After replacement of the Y-pattern globe valves by gate valves the vibration level at the valves went down significantly.But the Main Steam Line started to vibrate at a very discrete frequency about 15m behind the valve. It was a resonance problem. One frequency of the pressure pulsation in the steam did match one of the natural frequencies of the main steam pipe at this location. These effects are almost un-foreseeable at this kind of fundamental modifications of the main steam system.The vibration was unacceptable high. The plant could not be operated at the rated capacity of 129%.The problem was solved by the installation of a special structured seat at the upstream side of the MSIV. This modified seat caused a shift of the pressure pulsation frequencies in the steam. The steam line vibration disappeared. The plant is now operating at 129% power as planned.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 214

Transient Analysis of Integrated Pressurized Water Reactor(IPWR)

Salah Ud-Din Khan1, Minjun Peng2, Shahab Ud-Din Khan3

King Saud University, Sustainable Energy Technologies Center, P.O.Box 800, Riyadh 11421, Saudi Arabia1

Harbin Engineering University College of Nuclear Science and Technology Nuclear Power Simulation and Research Center, NO.145-1, Nantong Street, Harbin, China2

Chinese Academy of Sciences, Institute of Plasma Physics , P.O.Box 1126, 230031 Anhui, Hefei, China3

drskhan@ksu.edu.sa

 

In this paper, research has been conducted on the accidental conditions of Small nuclear reactor which is named as Uranium Zirconium Hydride Nuclear Reactor INSURE-100 designed by Nuclear Power Institute of China, Chengdu. This reactor is an integral type pressurized water reactor (IPWR) in which all the major components lies in the same reactor pressure vessel. This reactor is designed in such a way that pressurizer and once through steam generator (OTSG) lies in the reactor pressure vessel whereas the Passive residual heat removal system (PRHRS) is situated outside the reactor pressure vessel. PRHRS consists of heat exchanger dipped in a water tank and has a function of absorbing core residual heat in case of transient. The current research is focused on the safety aspects depicting into the preliminary safety assessment of reactor. The reactor was tested at 100% full power by using thermal hydraulic system code RELAP5/MOD3.4 before analysing the transients. Two types of accidents have been simulated i.e., loss of coolant accident (LOCA) and loss of feed water accident (LOFW). These accidents can be caused by many unpredictable reasons but in this paper some assumptions were made for the preliminary safety assessment of the reactor. Finally the graphs with explanations are given along with time sequence of events.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 216

An analytical study for LBLOCA with loss of emergency core cooling system using MARS-KS-CANDU

Jin-Hyuck Kim, Byung-Gil Huh, Chae-Yong Yang

Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

jhkim@kins.re.kr

 

If a large break loss of coolant accident(LBLOCA) of the primary coolant system with loss of the emergency core cooling system (LOECC) once occurs in a heavy water type nuclear power plant, the prolonged cooling by the vapor flow was generated following loss of the coolant. Heat removal will not be almost done due to the vaporization of the fuel channels, and there is a sharp rise in temperature of the pressure tube and the fuel sheath. The deformation in the direction of the diameter of the pressure tube with the high-pressure and high-temperature can be done, and it is in direct contact with the calandria tube if deformation persisted. Once the contact occurs, the heat flux spike phenomenon occurs through the moderator of the low-temperature, which heat is rapidly removed. For the safety analysis of the accident without the emergency core cooling system, the development of modeling for a pressure tube and calandria tube contact and of heat removal load evaluation methodology of the moderator must be required. In this study, the purpose is to evaluate the heat removal load of the moderator with the pressure tube deformation model of MARS-KS-CANDU, which the emergency core cooling system damaged with a large break of the primary coolant system occurs.

Pressure tube deformation model is considered as a part of the heat structure of MARS-KS-CANDU, in which there are three different layers. If a pressure tube deformation occurs, the wall thickness of a pressure tube keep thinning and expanding while maintaining the volume of pressure tube. When the deformation continues to happen, there is direct contact between the pressure tube and the calandria tube. On the other hand, the heat transfer between the pressure tube and the calandria tube is composed by the radiation heat transfer between pressure tube and calandria tube and the convection by the CO2 gas layer. Convection and radiation which is the actual mechanism of heat transfer, cannot be simulated because the pressure tube, CO2 gas layer, and the calandria tubes consist of one heat structure here. Therefore, the heat transfer by convection is to be calculated to the heat transfer by conduction by using the thermal conductivity of CO2 gas here.It was used for modeling of the primary coolant system in Wolsong unit 1, in order to simulate an accident with a LBLOCA in the primary coolant system with the loss of Emergency core coolant injection. In this case, the condition of the calculation was 35% in size of the break in the fourth core path.Modeling of the fuel channels was modified for prohibiting the coolant injection even if the emergency core coolant injection requirements was needed by changing open condition some valves which the emergency core coolant to connect with each header. The analysis calculations were performed by using the previously described pressure tube deformation model. For this, the pressure tube, CO2 gas layer and the calandria tube were modeled by using one heat structure. Also, in the case that the pressure tube deformation model was not be used, it was set to separate heat structure between pressure tube and the calandria tube, and in the case of the CO2 gas layer by modeling as a separate volume.To analyze the LBLOCA with LOECC in CANDU type reactor, it is imperative to adequately predict the pressure tube deformation and subsequent heat removal load to the moderator. Therefore, in this study, we evaluate whether the pressure deformation model of MARS-KS-CANDU code is properly work through calculation of Wolsong Unit 1. It is estimated that pressure tube deformation model work properly and reliably predict the moderator`s heat load in LOECC accident. However, these assessments should be limited for the qualitative part due to the difficulty of acquiring the experimental data for the CANDU type reactors.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 218

Critical Power Prediction by CATHARE2 of the OECD/NRC BFBT Benchmark

Sergii Lutsanych1, Luben Sabotinov2, Francesco D'Auria1

University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Livornese 1291, 56122 Pisa, Italy1

Institut de Radioprotection et de Sureté Nucléaire, 31, avenue de la Divison Leclerc, 92260 Fontenay Aux Roses, France2

sergii.lutsanych@gmail.com

 

The main purpose of this study is to evaluate the performance of the French best estimate thermal-hydraulic code CATHARE 2 regarding critical power prediction, comparing the analytical results with the experiments, available in the framework of the International OECD/NRC Benchmark, based on NUPEC (Nuclear Power Engineering Corporation) BWR Full-size Fine-mesh Bundle Tests (BFBT).

Two-phase flow calculations and prediction of the void fraction distribution and the critical power were carried out both in steady state and transient cases, using 1D and 3D modeling. Comparison with the test results shows the ability of CATHARE 2 code to predict reasonably the critical power, using appropriate modeling.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 219

Analysis of gas-liquid churn flow in a vertical pipe

Matej Tekavčič, Boštjan Končar, Ivo Kljenak

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

matej.tekavcic@ijs.si

 

Understanding of multiphase flow phenomena is of great importance in reactor engineering. In a loss-of-coolant accident, so called "flooding" in vertical pipes is one of such phenomena that are of particular interest for safety analyses.

After reflux condenser mode of cooling is established during such an accident, the counter-current flow of steam in the core region of a vertical pipe can limit the downward flow of the liquid water film on the pipe wall. Flooding occurs when the liquid film flow reverses and cannot penetrate further downwards into the reactor primary system. A thorough understanding of this initiating mechanism is required to be able to predict the onset of flooding conditions.Even for the simplest cases, the prediction of the main mechanisms is still very uncertain. One of such flow examples is an isothermal counter-current air-water flow in the churn flow regime of a vertical pipe. The churn flow regime can be viewed as a transitional regime between slug flow and annular flow. It is believed that this transitional regime is related to the onset of flooding mechanism. Large waves of liquid travelling upwards can typically be observed in the churn flow regime.The visualization and modelling study of the churn flow regime of counter-current air-water flow in a vertical pipe performed at Imperial College, London [1], will be used as a benchmark reference. In this experiment, flooding type wave formation and motion was observed, modelled and analysed. The study provides us with visualization data of wave shape evolution as well as wave frequency, wave position and velocity data which can be used to validate and evaluate our modelling results.Based on the force balance over such flooding type waves, an analytical model can be established to describe its growth and movement. An appropriate physical model for huge wave movement in gas-liquid churn flow was found in the open literature [2]. The model assumes a sinusoidal shape of the disturbance wave. The boundary layer in the liquid film is proposed as turbulent in contrast to other similar analytical models that use laminar liquid film assumption [1]. In this model, the most important force on the wave is due to the gas pressure difference between leeward and windward side.This chosen model is implemented and applied to the considered experimental case data. The calculated results are compared to experimental data and CFD (Computational Fluid Dynamics) simulation that used a two-fluid modelling approach with interface tracking to determine the shape of liquid waves.References:[1] Barbosa Jr., J. R., Govan, A. H., and Hewitt, G. F., 2001. "Visualisation and modelling studies of churn flow in a vertical pipe". International Journal of Multiphase Flow, 27(12), pp. 2105 – 2127.[2] Da Riva, E., and Del Col, D., 2009. "Numerical simulation of churn flow in a vertical pipe". Chemical Engineering Science, 64(17), pp. 3753 – 3765.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 220

RELAP5 Simulation of Thermal Hydraulic Test on VEFITA facility

Jun Yang1, Detlef Suckow1, Terttaliisa Lind1, Horst Michael Prasser2

Paul Scherrer Institut, Villigen, CH-5232 Villigen PSI, Switzerland1

ETH Zurich, Institute of Energy Technology, ML K 13, Sonneggstrasse 3, 8092 Zürich, Switzerland2

toyangjun@gmail.com

 

This paper summarizes the RELAP5/Mod3.3p3 simulation on the experiment performed in the VEFITA (Venting Filter Assessment) test section of Paul Scherrer Institut (PSI) operated by the severe accident research group of the laboratory for thermal-hydraulics (LTH). The experiment is a hydraulic pre-test for the venting filter assessment program. The test section is a vessel with about 0.6m diameter and 5m height, filled with water to a level 2.25 m above the gas injection nozzle. The gas space above the water is at a gauge pressure of 2.8 bar and the water is brought to nearly saturation temperature of 140 °C. Non-condensable nitrogen at a mass flow rate of 450 kg/h and a gas temperature of 145 °C is fed through the injection nozzle into the test section for a time duration of 6000s. The test section wall is trace heated to 140°C. The thermal-hydraulic behaviours are characterized with respect to the water temperature, the water swell level, pressure and flows. A one dimensional input model is developed for the VEFITA facility and the experiment transient is simulated. The RELAP5 code result has a good agreement with the experiment data.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 221

1D Two-Fluid Model Simulations of Flashing Flows in TOPFLOW Facility

Blaž Mikuž1, Iztok Tiselj1, Dirk Lucas2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstrasse 400, D-01328 Dresden, Germany2

blaz.mikuz@ijs.si

 

Two phase flows occur in pressurized water reactor (PWR) under normal conditions as well as under accident scenarios such as Loss of Coolant Accident (LOCA). The typical LOCA involves rupture of major primary piping, steam generator tube rupture, inadvertent opening of a relief or safety valve and similar events where pressure decreases down to the saturation pressure resulting in the generation of steam from liquid water. In the present study evaporation of liquid water to steam caused by depressurization was simulated in an 8 m long vertical tube with an inner diameter of 195.3 mm. Simulations were prepared using an in-house code WAHA, which was designed for one-dimensional simulations of two phase flows for water hammer transients. Results of simulations are compared with the experiment measurements of TOPFLOW (acronym for Transient twO Phase FLOW) test facility, where pressure relief were done using two different procedures: during circulating of water in tubes and from stagnant water in tube. In the first case the superficial velocity of circulating water was about 1 m/s and in both cases the water was saturated at initial pressure values of 1, 2, 4 and 6.5 MPa.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 222

A Numerical Analysis of Direct Spring Loaded Type - Steam Safety Valve Using CFD Simulation

Sangho Sohn1, Jin Hyun Nam2, Kyungha Ryu1, Woo Tae Lim3

Korea Institute of Machinery and Materials, 171 Jang-Dong, Yusung-Gu, 305-343 Daejeon, South Korea1

Daegu National University Division of Automotive, Industrial and Mechanical Engineering, 15 Naeri-ri, Jinryang-eup, Gyungsan 712-714, South Korea2

Choongnam National University, , South Korea3

sangho@kimm.re.kr

 

A safety valve is one of pressure relief valves well-known to protect pressurized equipments such as a boiler or pressure vessel from exceeding the maximum allowable working pressure. When pressure inside a vessel reaches or excesses a set pressure of a safety guideline, the safety valve assures its safety to release its fluid by a quick opening called as popping. The performance of a safety valve is evaluated by set pressure, full open, blow-down, leakage and flow capacity. The relevant technical requirements are described in the international ASME code. The object of this paper is to investigate the fluid dynamics of a spring-loaded type safety valve operated with steam fluid by computational fluid dynamics (CFD). The opening characteristic is evaluated by 10 quasi-static simulations according to opening lift levels from 0 to 100%. The results show fluid flow, pressure, forces on the disc and mass flow at each simulation step. The CFD model of safety valve is beneficial to investigate the detailed flow dynamics. In addition, the effect of back pressure at the outlet is investigated by analyzing lifting force and mass flow. The constant superimposed back-pressure is usually determined from the outlet source tank. It is expected that the back pressure has an effect on opening pressure, flow capacity and instability. Finally, the effect of huddle chamber or control chamber is shown by dynamic analysis based on CFD simulation results such as lifting force.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 223

Thermal Hydraulic Evaluation of Liquid Metal Flow Simulation System

Kyungha Ryu, Byoungmin Ban, Jaehyoung Kim, Sangho Sohn

Korea Institute of Machinery and Materials, 171 Jang-Dong, Yusung-Gu, 305-343 Daejeon, South Korea

khryu@kimm.re.kr

 

To maintain sustainability of nuclear energy as an important energy source, it should be solved that both safety problem and Spent Nuclear Fuels (SNFs) problem.

In case of Gen-IV reactors such as fast reactor, SNFs can be used as fuels by using fast neutrons. It can be suitable treatment method of high-level waste in near future. Liquid metals such as Sodium or Lead-Bismuth Eutectic (LBE) are possible coolant to use fast neutrons.The liquid metals as coolant also has the advantage of design of full passive system. The full passive design uses natural circulation without active components such as valves and pumps. With the design, it is considered that core damage frequency (CDF) can be reduced about 1000 times lower than commercial Nuclear Power Plants (NPPs) [1].To verify and validate the natural circulation ability of the liquid metal, the Korea Institute of Machinery and Materials - Liquid Metal Flow Simulation system (KIMM–LiMSi system) was developed. It is composed with 2.5 inches (outer diameter: 0.0635 m) STS 316L pipe. The fluid in the system circulates with gravity and possible difference of density caused by heat.Lots of experiential are performed. The limitations of test temperature were 400? that is limit operation temperature in STS316L in oxygen condition [2]. The ability of natural circulation was sufficient within suitable temperature.Fluid analyses are performed by using both finite element analysis and thermal hydraulic code analysis. The analytical results were matched with the experimental results in good agreement.With these experimental and analytical achievements, the study on flow characteristics of two phase fluid, mixture of LBE and noble gas, would be scheduled. This study can be contribute to the securing of properties of two phase flow as well as the ensuring of nuclear safety.1. S.H. Jeong, “Development of an integral test loop, HELIOS and investigation of natural circulation ability for PEACER” Ph.D. thesis, Seoul National University, 2006.2. Fazio, C., G. Benamati, C. Martini, G. Palombarini, “Compatibility Tests on Steels in Molten Lead and Lead-bismuth”, Journal of Nuclear Materials, 296, p. 243, 2001






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 224

Investigation of primary circuit loop transfer properties in VVER-440 reactors

Sándor Kiss, Sándor Lipcsei

Centre for Energy Research, Hungarian Academy of Sciences , Konkoly Thege M. út 29-33, H-1121, Hungary

lipcsei.sandor@energia.mta.hu

 

Enhancing the evaluation of reactor noise diagnostics measurements requires feedback originating from the effect of the circulation of the coolant in the primary circuit loops to be taken into account. Therefore circulation period of the coolant and transfer properties of the steam generator and its effect on the perturbations traveling with the coolant have to be known. Circulation period was determined from the correlation of the noise signals of thermocouples installed in the loops as a part of the standard instrumentation. Circulation time of the perturbations traveling with the coolant was found to be smaller than expected by the raw estimation from total mass and mass flow rate of the coolant. The reason of this is partly coming from stagnating volumes not exactly known in the raw estimation and from the effect of the steam generator. The latter is well demonstrated by the transfer function of the steam generator which was produced by taking its inner structure into account.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 227

Spectral Benchmark for Natural Convection Flow in a Tall Differentially Heated Cavity

Jure Oder, Iztok Tiselj

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

jure.oder@ijs.si

 

Natural convection is an important phenomenon in nuclear engineering. It represents a passive safety mechanism that is at the core of many reactor designs. The tools used to analyse the safety of particular design of nuclear reactor as well as of surrounding structures depend on accurate prediction of this and accompanying phenomena.

In the year 2002 Christon et. al. published a summary of a workshop that took place at the First MIT Conference on Computational Fluid and Solid Dynamics held in June 2001. The object of the workshop was to better understand the fluid dynamics of the two dimensional square cavity. Benchmark cases on such simple geometries are used for verification and validation of different codes used to simulate fluid flows. This case is particularly interesting because of plethora of different solutions that can be attained.In this paper we in part recreated the spectral method by Xin and Le Quéré that was chosen to be the most accurate within the workshop. Their method is second order in time and uses Chebyshev collocation method to solve for spatial dimensions. The solution at which we arrived using their method agreed with their results to within the order of E-4.Additionally, we developed a spectral method to solve the spatial part of the equations that is based on trigonometric functions. The results with this method differ from reference for about few per cent.Using these two methods we explored different solutions that can be reached and are meaningful in two dimensions. The result is a tree of different solutions for Rayleigh numbers between 3E5 and 4E5.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 229

Numerical and experimental design of multi-stage orifice FWRO-004

Franci Vehar1, Andrej Lipej1, Rok Pavlin1, Aljaž Škerlavaj1, Bogdan Jančar1, Matjaž Černec2

Turboinstitut d.d., Rovsnikova 7, 1210 Ljubljana, Slovenia1

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia2

aljaz.sk@gmail.com

 

The orifice plate was built into the feedwater (FW) long recirculation as the pressure reducing device. During the operation very intensive cavitation was developed, generating vibrations and noise of high level. Also fractures of certain welds in the piping system were noticed as the final result.

The aim of the research is an analysis of existing orifice FWRO-004 and hydraulic analysis and design of a new multi-stage orifice. The main goal in the design process is minimization of cavitation and reduction of pressure pulsations and vibrations.In the literature the information about the pressure losses in the orifice in dependence onpipe diameter and opening of the orifice are given. In the single-stage orifice the pressure drop is known. In the multi-stage, due to avoiding the occurrence of cavitation it is important that the pressure drop on the last two stages is lower than on the first stages. According to the recommendations in the literature the pressure drop on the orifice for existing geometric conditions is around half of the absolute pressure before the orifice. This criterion necessarily brings an increase in the number of stages. We limited the number of orifice stages to four and designed orifice plates with up to four eccentric holes.In the design process Computational Fluid Dynamics (CFD) software ANSYS CFX has been used. Computations have been done for steady state analysis using k-omega SST turbulence model and for unsteady analysis using SAS SST turbulence model. To obtain final optimal solution more than ten different geometries of the orifice have been numerically analysed.The numerical results have been verified with the measurements on the model in Turboinstitute's laboratory and finally confirmed on-site in Nuclear Power Plant Krško.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 230

GRNSPG/UNIPI Activities on KALININ-3 Coolant Transient Benchmark: Best Estimate Coupled Plant Transient Modeling

Raul Gonzalez Gonzalez, Francesco D'Auria

University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Livornese 1291, 56122 Pisa, Italy

r.gonzalez@dimnp.unipi.it

 

During the last several years a considerable effort was devoted and progress has been made in various countries and organizations in incorporating full three-dimensional (3D) reactor core models into system transient codes, it allows performing of a “best-estimate” calculation of interactions between the core behavior and plant dynamics. Several benchmarks have been developed to verify and validate the capability of the coupled codes in order to analyze complex transients with coupled core-plant interactions for different types of reactors. In December 2008 the NEA/OECD Nuclear Science Committee (NSC) Bureau has expressed support for the coupled Kalinin-3 benchmark problem in general to become an international standard problem for validation of the best-estimate safety codes. This benchmark defines a coupled code problem for further validation of thermal-hydraulics system codes for application to Russian-designed VVER-1000 reactors based on actual plant data from the Russian NPP Kalinin Unit #3 (Kalinin-3). The selected transient “Switching-off of one Main Circulation Pump (MCP)” is performed at a nominal power and leads to asymmetric core conditions with broad ranges of the parameter changes. The available real plant experimental data is very well documented and made these benchmark problems very valuable. Measurements were carried out with a quite high frequency and their uncertainties are known for almost all measured parameters. This fact allows applying the studied transient not only for validation purposes but also for uncertainty analysis as a part of the NEA/OECD LWR Uncertainty Analysis in Modeling (UAM) Benchmark. The purpose of this benchmark is four-fold: to verify the capability of system codes to analyze complex transients with coupled core-plant interactions and complicated fluid mixing phenomena, to fully test the 3D neutronics/thermal-hydraulic coupling, to evaluate discrepancies between predictions of the coupled codes in best-estimate transient simulations with measured data, to perform uncertainty analysis having at disposal not only the measured values but also their accuracy. This report provides the calculations for first three points.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 231

TRACE Code Analysis of a Two-Loop PWR Coolant Mixing in the Reactor Pressure Vessel

Ovidiu-Adrian Berar, Andrej Prošek, Borut Mavko

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

adrian.berar@ijs.si

 

The coolant mixing phenomena in the reactor pressure vessel plays an essential role for Pressurized Water Reactor (PWR) safety analyses using computer codes. Especially in the case of Main Steam Line Break accident, the accuracy of the predicted coolant mixing is of key importance as the incomplete mixing of the reactor coolant occurring in the downcomer and the lower plenum of the reactor pressure vessel creates an asymmetric temperature distribution at the core inlet. The mixing phenomena are particularly difficult to model and require specialized computer codes able to predict thermal-hydraulic conditions in three dimensional space. The TRAC/RELAP Advanced Computational Engine (TRACE) is an advanced, best-estimate reactor systems code developed by the U.S. Nuclear Regulatory Commission for analyzing light water reactors, and it has the capability of solving the fluid-dynamics equations in three dimensional space. The TRACE code prediction of the coolant mixing phenomena for a two loop PWR is analyzed using a three dimensional TRACE input model of the reactor pressure vessel. The computational domain of the vessel model is discretized in cylindrical geometry by using axial levels, radial rings and azimuthal sectors. Different nodalization schemes for the downcomer, lower plenum and core region are presented. The TRACE predictions of the temperature field in the reactor pressure vessel are analyzed for asymmetric reactor pressure vessel inlet conditions. A sensitivity study assessing the influence of the reactor pressure vessel nodalization on the coolant mixing phenomena is presented and discussed.






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 232

Thermal-Hydraulic Analysis of Fast Breeder Reactor: Protected Loss Of Flow (PLOF) Transient

Chirayu Batra1, Marco Cherubini1, Francesco Saverio D`Auria1, Alessandro Petruzzi1, Tomasz Kozlowski2

University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Livornese 1291, 56122 Pisa, Italy1

University of Illinois at Urbana-Champaign, 104 South Wright Street, Urbana IL 61801, USA2

chirayu.arya@gmail.com

 

Experimental Breeder Reactor II (EBR-II) is a Liquid Metal Reactor(LMR) at Argonne National Laboratory (ANL). It is a pool type Sodium-cooled Fast Reactor (SFR). In order to demonstrate the inherent safety of this type of reactor, several loss of flow tests were done, as a part of Shutdown Heat Removal Test (SHRT) series. One of the tests is SHRT-17, which is a Protected Loss of Flow transient. At the beginning of this test the primary pumps were tripped and at the same time full control rod insertion was done. The test successfully demonstrated the effectiveness of natural circulation cooling capability of the reactor, which makes them inherently safe under this accident conditions.

The design and geometry of the EBR II is quite complex and different form normal PWR reactors. The aim of this paper is to demonstrate the modeling of the reactor using RELAP5-3D and to perform its Thermal-Hydraulic Analysis. The data was released as a part of an IAEA Coordinated Research Project which aims at improving the simulation capabilities in the fields of research and design of SFR through data and codes validation and qualification. Key words: EBR-II, RELAP5-3D modeling, IAEA benchmarking exercise for EBR II






10.09.2013 10:40 Poster session 1

Thermal-hydraulics - 233

Simulation of a SBLOCA in a Hot Leg. Scaling Considerations and Application to a Nuclear Power Plant

Andrea Querol, Sergio Gallardo, Gumersindo Verdú

Universidad Politecnica de Valencia, Departamento de Ingeniería Química y Nuclear, Camino de Vera s/n, 46022 Valencia, Spain

sergalbe@iqn.upv.es

 

During a Small Break Loss-of-Coolant Accident (SBLOCA) transient, depressurization can be slow enough to delay the Accumulators (ACC) entry for a long time. Actuation of High Pressure Injection (HPI) system is then necessary in order to maintain the core temperature low enough to avoid core boil off, and consequently avoiding the core level to fall below fuel rods level, thus producing a temperature excursion in the fuel cladding. In this frame, and with the aim of understanding the thermal hydraulic phenomena associated with this kind of scenarios, the Test 1.2 of OECD/NEA ROSA Project, performs a 1% hot leg SBLOCA in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA).

LSTF is a Full Height Full Pressure (FHFP) facility simulating a PWR reactor Westinghouse type, of 4 loops and 3423 MW of thermal power. The facility is electrically heated, scaled 1:1 in height and 1:48 in flow areas and volumes.In this work, LSTF and Test 1.2 transient has been simulated using the thermal-hydraulic code TRACE5 patch 2.In order to know if physical phenomena observed in LSTF during a SBLOCA transient can be reliably extrapolated for an actual 4-loop and 3400 MW nuclear power plant, a TRACE5 model has been developed, conserving the power-to-volume scaling ratios of LSTF components, initial and boundary conditions. In this 4-loop NPP model, horizontal length and diameters has been scaled from LSTF model in order to conserve Courant number. Power-to-volume scaling is frequently used to preserve the time, power and mass inventory for FHFP facilities and for prototype plant, because the fluid exhibits the same properties at the full pressure. Transient 1.2 has been simulated by TRACE5 using this 4-loop model.Finally, due to the fact that actual Spanish NPP are 3-loop Westinghouse design, a 3-loop PWR TRACE5 model has been adapted to meet also the Test 1.2 conditions.Transient is analyzed by means of the main thermal hydraulic variables: primary and secondary pressures, mass flow rates, discharged coolant inventory through the break, pressurized vessel collapsed liquid levels, fluid temperatures in hot and cold legs and Core Exit Temperature and Peak Cladding Temperature. These thermal hydraulic variables are obtained by simulation for all the three TRACE5 models (LSTF, 3-loop and 4-loop NPP). Results show that the main thermal hydraulic phenomena of a hot leg SBLOCA (core boil-off, etc) are well predicted by TRACE5 in the LSTF model and in the power-to-volume scaled models.






10.09.2013 10:40 Poster session 1

Severe accidents - 401

Analysis of Ex-Vessel Steam Explosion in 3-D

Matjaž Leskovar

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

matjaz.leskovar@ijs.si

 

The importance of severe accident research has been again seen after the March 2011 Fukushima nuclear reactor accident in Japan. As the accident analysis showed, it seems that in the Daichi Unit 3 a steam explosion occurred. A steam explosion is an energetic fuel coolant interaction (FCI) process, which may occur when the hot reactor core melt comes into contact with the coolant water. During the steam explosion the energy of the molten corium is transferred to the coolant water in a timescale smaller than the timescale for system pressure relief and so induces dynamic loading of surrounding structures. 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. Direct or by-passed loss of the containment integrity can lead to radioactive material release into the environment, threatening the safety of the general public.

To resolve the remaining open issues on the FCI processes and their effect on ex-vessel steam explosion energetics, the OECD project SERENA (Steam Explosion Resolution for Nuclear Applications) was launched in the year 2007, consisting of an experimental and an analytical part. To verify the progress made in the understanding and modelling of FCI key phenomena for reactor applications a reactor exercise has been performed. The exercise comprises three cases, a BWR axial melt release, a PWR axial release and a PWR side release.In the paper the PWR ex-vessel steam explosion study, which was carried out with the MC3D code in conditions of the SERENA reactor exercise for the PWR side release case, will be presented. The analysed domain was modelled in the cylindrical coordinate system in 3-D. In reactor calculations the largest uncertainties in the prediction of the steam explosion strength may be expected due to the large uncertainties related to the jet breakup. They propagate through different premixing processes and result in uncertainties in the generation rate and size of the melt droplets, the distribution of the melt droplets in the premixture, the droplets solidification and the void fraction, which all influence the steam explosion strength. To get some insight in these uncertainties, the premixing simulations were performed with both available jet breakup models – the global model and the Kelvin Helmholtz model. Various premixing phase and explosion phase simulation results will be presented and discussed, among them also the pressure histories and the pressure impulses at different locations in the cavity.






10.09.2013 10:40 Poster session 1

Severe accidents - 405

Modeling of ESBWR Passive Containment Cooling System with MELCOR

Tuomo Sevon

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

tuomo.sevon@vtt.fi

 

GE Hitachi ESBWR reactor design has a passive containment cooling system (PCCS) for removing heat from the containment during an accident. The system has vertical tubes in a water pool. Steam from the containment flows through the tubes and condenses on their inner surfaces. The condensed water and non-condensable gases are returned to the containment.

Capability of the MELCOR code to model the vertical PCCS was investigated in a three-step process, progressing from a simple case to a more complicated. First, Purdue University experiments of steady-state condensation in a single vertical tube were calculated. Second, VTT experiments of steady-state condensation and aerosol deposition in a vertical tube were calculated. They involved a wider range of non-condensable gas concentrations than the Purdue University experiments. Third, PANDA T1.1 experiment, performed by PSI in Switzerland, was calculated. It was the most complicated of the calculated experiments, involving transient operation of a scaled model of three ESBWR PCCS units with multiple tubes and helium injection for simulating hydrogen release in a severe accident.When calculating the Purdue University experiments, it was noticed that with default parameters MELCOR underestimated the condensation rate by about 20 %. When Reynolds number limits for calculating heat transfer through the condensate film were modified to values found in a heat transfer textbook, very good results were obtained. In the VTT experiments the calculated condensation rates and aerosol deposition were near to the measured values. In the PANDA experiment the helium caused a reversal of flow direction in some of the condenser tubes, which increases uncertainties in the calculation. With proper modeling of drainage of condensed water out of the condenser, the model is sufficiently accurate for plant calculations.






10.09.2013 10:40 Poster session 1

Severe accidents - 406

Analysis of Passive Reactor Cavity Cooling System for a 600 MWTH HTGR Using MELCOR Code

Changwook Huh, Chang-Yong Jin, Jin-Hyuck Kim

Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

huhcw1@naver.com

 

High Temperature Gas-Cooled Reactor (HTGR) is one of Generation-IV reactor concept which is being developed to generate high temperature heat for other industrial processes and hydrogen production. Since one of the most important requirements for HTGR is passive safety, most HTGR designs typically use passive reactor cavity cooling system (RCCS) designed to remove all the core afterheat without the use of any active safety systems during all postulated accidents. On the researches of the HTGR licensing technologies in Korean Institute of Nuclear Safety (KINS), MELCOR code is under consideration as a safety evaluation tool for HTGR, which is used for thermal-fluid and accident analysis, including fission products transport release. The latest version of this code, MELCOR 2.1 has been modified for the Next Generation Nuclear Plant (NGNP) by the U.S. Nuclear Regulatory Commission (NRC).

In this study, the MELCOR 2.1 input model of HTGR with RCCS was developed for the design of 600 MWth HTGR which is based on General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) to assess the ability of MELCOR to predict the RCCS performance in accident condition. The characteristics of HTGR were modelled including conduction and radiation heat transfer in RCCS and between fuel and/or reflector blocks, oxidation phenomena of graphite, insulation effect of RCCS downcomer etc. The steady state and depressurization accident conditions were analysed using the developed input model to evaluate the applicability of MELCOR code in HTGR with RCCS.






10.09.2013 10:40 Poster session 1

Severe accidents - 407

Post-Test Calculation of the QUENCH-17 Bundle Experiment with Debris Formation and Bottom Water Reflood Using Thermal Hydraulic and Severe Fuel Damage Code SOCRAT/V3

Alexander D. Vasiliev1, Juri Stuckert2

Nuclear Safety Institute of Russian Academy of Sciences , 52, B. Tulskaya, 115191 Moscow, Russian Federation1

Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany2

vasil@ibrae.ac.ru

 

The thermal hydraulic and SFD (Severe Fuel Damage) best estimate computer modelling code SOCRAT/V3 was used for the calculation of QUENCH-17 experiment which was the first test in the QUENCH tests series simulating debris behaviour.

The QUENCH-17 test conditions simulated a representative scenario of nuclear power plant severe accident sequence with debris bed formation in which the overheated up to 1800K core would be reflooded from the bottom by ECCS (Emergency Core Cooling System). The QUENCH-17 test included the following phases:- Heat-up phase (heat-up rate up to 0.25 K/s);- Oxidation phase (the cladding temperature T»1800K in hottest region, steam mass flow rate 2 g/s);- Bottom flood phase (characteristic cooling time » 600 s, water mass flow rate 10 g/s).The test QUENCH-17 was successfully conducted at the KIT, Karlsruhe, Germany, on January 30-31, 2013. The objective of this test was to examine the formation of a debris bed inside the completely oxidised region of the bundle without melt formation and to investigate the coolability behaviour during the reflood.QUENCH facility is designed for studies of the PWR fuel assemblies behaviour under conditions simulating design basis, beyond design basis and severe accidents.The test bundle for QUENCH-17 test was intentionally changed in comparison to basic QUENCH tests with the aim to get more relevant and comprehensive results for debris behaviour phenomena.The test bundle was made up of 21 fuel rod simulators with a length of approximately 2.5 m. Only 12 periphery fuel rod simulators were heated over a length of 1024 mm, 9 unheated fuel rod simulators were located in the inner part of the test bundle.Due to such geometry, the porous debris formation in the inner part of the bundle was not influenced by the presence of tungsten heaters as it would be in old geometry.The rod cladding for 9 inner rods was identical to that used in LWRs: Zircaloy-4, 10.75 mm outside diameter, 0.725 mm wall thickness. The unheated rod simulators were filled with segmented ZrO2 pellets (without centre holes). The rod claddings for 12 outer heated rods were made of hafnium. The shroud was also made of Hf. The high melting temperature of this material ensured that the claddings did not melt during high temperature phase of the test.SOCRAT/V3 computer modelling code was used for calculation of basic thermal hydraulic, oxidation and thermal mechanical behaviour during all phases of the experiment.The calculated results are in a good agreement with experimental data which justifies the adequacy of modeling capabilities of SOCRAT code system.






10.09.2013 10:40 Poster session 1

Severe accidents - 408

Internal Vessel Retention for VVER-440/V213 reactors: relevant experiments on the RESCUE facility and modelling with NEPTUNE_CFD

David Guenadou, Verloo Eric, Casado Remi

CEA Cadarache, DTN/STRI/LHC, Bar. 728, FR-13108, Saint Paul lez Durance, France

david.guenadou@cea.fr

 

During a severe accident, the decay power can lead to the core fusion producing a magma called corium. This one, composed of the high temperature melted materials of the core, is very aggressive regarding the Reactor Pressure Vessel (RPV). It can induce its failure and the dissemination of nuclear products into the environment. Various systems have been designed to avoid this scenario. One is based on the cooling of the external vessel wall by water. In case of accident, a pool surrounding the RPV is flooded to dissipate the decay power. This is a passive system, fed with a 4 m high reservoir ,a natural circulation being initiated due to the steam generated on the warm vessel wall.

The RESCUE facility, located at CEA Cadarache in France, was designed and built to check the efficiency of such a system and to understand the main phenomena that take place. This last point aims at improving the modelling of this safety system. The test section is a 1/10th portion in azimuth (that is to say 36° in angle) of a 4 m high vessel, the bottom head of which is elliptical. The inner wall of the test section is segmented into 12 heated zones, independently controlled, allowing heat fluxes up to 450 kW/m2: this allows simulating different accident scenarios and locations of the corium.NEPTUNE_CFD is a thermohydraulic code devoted to two-phase flow developed by EDF.This mock-up was used in the framework of SARNET 2 (7th EU Framework program) for studying the EVRC of VVER440/V213 reactors. VVER440/V213 heat flux profiles calculated using the ASTEC code has been adapted to the RESCUE geometry. A series of relevant experiments were carried out and modelled. This article is devoted to present both the experimental and numerical results.






10.09.2013 10:40 Poster session 1

Severe accidents - 409

Severe Accident Analysis in the NPP Krško with the ASTEC Code

Siniša Šadek, Milan Amižić, Davor Grgić

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

sinisa.sadek@fer.hr

 

The ASTEC/V2.0 computer code was used to simulate a severe accident sequence in the nuclear power plant Krško which is a two-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Sureté Nucléaire (IRSN, France) and Gesellschaft für Anlagen- und Reaktorsicherheit (GRS, Germany) to assess the nuclear power plant behaviour during a severe accident. The main focus was on the in-vessel phase of the accident. The model of the plant included the detailed nodalization of primary and secondary circuits without the containment. The reactor coolant system, steam generators, steam lines, feedwater and auxiliary feedwater pipes were modelled as a set of thermal hydraulic volumes connected by junctions, to which heat structures were attached to simulate heat losses to the environment. The ICARE module was used to model the reactor core and the CESAR module to model all the other plant systems: primary and secondary circuit piping, the pressurizer and the steam generators. In the reactor pressure vessel (RPV), the analysis of all relevant core damage phenomena (Zircaloy oxidation, chemical interactions, liquefaction of the fuel and core structures) and the failure of the RPV wall in the lower head after accumulation of the molten material was performed.

The analysis was conducted in two steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout (SBO) accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. The analyzed SBO scenario included the loss of both off-site and on-site AC power. The only systems available were passive systems: the accumulators and the turbine driven auxiliary feedwater (AFW) system.Two scenarios were analyzed: one with and one without the AFW. The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with the core melt and the reactor pressure vessel failure with significant release of hydrogen.At the end, the results of the ASTEC calculation were compared with the results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results showed a good agreement between predictions of those two codes.






10.09.2013 10:40 Poster session 1

Severe accidents - 410

Investigation of oxygen depletion during hydrogen recombination for VVER1000 of Kozloduy NPP

Pavlin Petkov Groudev, Antoaneta Stefanova, Petya Vryashkova

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

pivryashkova@inrne.bas.bg

 

In the paper is presented an analysis of hydrogen generation during Station Blackout (SBO) scenario for both phases (in-vessel and ex-vessel) as well as removing of hydrogen by using already installed Passive Autocatalytic Recombines (PARs) and comparison with additionally modeled new type PARs. The new 15 PARs will be installed in every one of containments of unit 5 and 6 in connection with reducing the risk of detonation during severe accidents. It has been investigate the behavior of main gases in the containment rooms with estimation the conditions for flammability and detonation in different zone of containment using Shapiro diagram. For this purpose the MELCOR 1.8.5 computer code is used for performing calculations of SBO accident for VVER-1000 with modeling of 8 old type of PARs and 23 PARs (including 15 new PARs to the 8 old PARs) in the containment.

The main aim of presented investigation is to assess the PAR efficiency and capacity during severe accident with hydrogen generation. It has been investigate the evaluation of hydrogen-oxygen-steam concentrations at the different zone of containment and especially the oxygen depletion resulting of PARs hydrogen recombining. The selected accident is Station Blackout with opening and stuck in open position of Pressurizer safety valve. In this way it was simulated small break loca, which allows maximum generation of hydrogen during the in-vessel phase. The calculations have been performed in the Institute for Nuclear Research and Nuclear Energy with the model for VVER-1000 developed for MELCOR computer code. The investigation has been a work performed in the frame of Kozloduy NPP project to provide data for main parameters (as pressure, temperature etc.) and gases behavior in the containment with 8 and 23 PARs. The group of 8 PARs was installed 10 years ago for DBA accident as well as for accidents until failure of reactor vessel. The capacity of 8 PARs was approved as enough and after changing reactor fuel with new one with increasing mass of Zr with 20% in reactor core. The performed comparison shows that in case with additionally installed 15 PARs in the containment it is observed oxygen depletion as a result of PARs recombining. The performed analyses shows that in both cases it is not observed violation of selected criteria for detonation and flammability in the containment.






10.09.2013 10:40 Poster session 1

Severe accidents - 411

A MUSIG model for melt fragmentation during Fuel-Coolant Interaction

Sebastian Castrillon Escobar1, Renaud Meignen1, Nicolas Rimbert2, Michel Gradeck2

Institut de Radioprotection et de Sureté Nucléaire, PSN-RES/SAG, BP-17, 92262 Fontenay-aux-Roses, France1

LEMTA Université de Lorraine CNRS , 2, av. de la Foret de Haye - TSA 60604, 54518 Vandoeuvre les Nancy, France2

sebastian.castrillonescobar@irsn.fr

 

This paper deals with melt fragmentation processes during a Fuel Coolant Interactions (FCI) that might happen during the course of a severe accident in a Nuclear Power Plant.

FCI might lead to a steam explosion that can damage the containment building. FCI will also contribute to the amount of hydrogen generated (by fuel oxidation) and affect the fuel debris coolability. At IRSN, for the purpose of modeling FCI, the Eulerian multiphase flow code MC3D is developed in partnership with CEA (France) and JSI (Slovenia). A crucial point for the accuracy of the evaluations is the melt fragmentation process occurring during melt mixing with coolant. The general modeling approach in MC3D describes separately the jet and the drops issued from melt jet fragmentation. Recently a homogeneous Multiple Size Group model (MUSIG) model was introduced in MC3D to improve the precision regarding the prediction of the drops population characteristics, namely surface (leading heat transfers) and volume (leading solidification characteristics). The drops production needs to be described and the objective of the paper is to show the first results of an original approach combining primary fragmentation (jet to drops) and secondary fragmentation (drops to drops) models. Primary jet fragmentation is modeled either thanks to a correlation either through a Kelvin-Helmholtz approach using local flow properties. In both cases, an initial drops size distribution can be prescribed, based on experimental findings. Then, secondary breakup is accounted directly in the jet fragmentation process. Alternatively, secondary breakup may be used for the purpose of drops size distribution prediction and the development of such method is the main intent of the present work. In such case, the jet fragmentation process is used mainly to provide a fragmentation rate and drops are generated with an arbitrarily large size. Different hydrodynamic fragmentation models have been introduced and are discussed. Daughter drops sizes are evaluated according to a constant characteristic Weber assumption and the fragmentation rate is based on the classical Ranger & Nicholls characteristic time. Classically, the characteristic Weber number is taken as a constant. Here, we also use a correlation proposed by Hsiang & Faeth, where the characteristic Weber number depends on the flow parameters. The correlation is normally applicable for high density ratios (liquid/gas). However, we use it here for smaller density ratios involved in liquid/liquid fragmentation; indeed, the range of initial mother drops Weber number, on which is based the correlation, varies within the range of validation given by the authors.The applicability of the model is firstly verified on single-drop fragmentation experiments and on isothermal jet fragmentation experiments. It is found that, when the mean final diameter is small, the homogeneous approach (all drops have the same velocity) is not fully satisfactory because the large drops are too rapidly entrained with the smallest ones and thus fragmentation is limited. The MUSIG model was then extended to a heterogeneous approach, using two different velocity fields to separate large and small drops, yielding more satisfactory results.Finally, we show the impact of various approaches on evaluations of real corium experiments (FARO). We highlight the strong impact of solidification effect in such cases. The method is promising and related phenomena like drops solidification will get more attention in future activities.






10.09.2013 10:40 Poster session 1

Severe accidents - 413

Significant Nuclear and Radiological Events in Europe in the Past

Helena Janžekovič1, Milko Janez Križman2

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia1

Independent qualified expert for RP, Ljubljana, Ljubljana, Slovenia2

helena.janzekovic@gov.si

 

Nowadays accidents in nuclear industry are categorised using so called the International Nuclear and Radiological Event Scale (INES scale). It was established by international organisations active in nuclear safety field after the Chernobyl accident. The scale can be applied not only to events associated with nuclear facilities, but also to the activities as transport, storage and use of radioactive material and radiation sources. The primary aim of the scale was to facilitate the communication about the safety significance of the nuclear accidents as well as other events. During and after the accident communication between different groups of people takes place, i.e. competent authority, technical community, media or public. Nuclear accidents are relatively rare and communicating facts about an accident can easily lead to misunderstanding or misinterpretation. On the other hand, nuclear accident can have a large effect on people everyday life not only in areas affected by an accident but much larger, e.g. decisions of a state to phase out nuclear power plants can influence everyday life.

Today incidents and accidents are reported to the database maintained by the IAEA. The safety significance of events using pre-defined parameters is determined. INES scale has seven levels, starting with “anomaly” as the lowest level of incidents. The most serious events rated at the scale by 7 which is the highest level of so-called “major accident” are accidents in Chernobyl and Fukushima. However, the accidents and events also happened before the establishment of the INES scale, e.g. before 1989 when OECD/NEA and IAEA set up the INES scale. The data of some of them are scarce.The aim of the paper is the overview and a systematic analysis of some events and accidents in Europe before establishment of the INES scale. The analysis is based on the available literature including also events which are not analysed using INES scale. The overview of the accidents with a special attention to handling or controlling contamination is given. A list of serious events includes among others the Pallomares accident in Spain in 1966 where nuclear bombs were involved in plan crash, the Lucens reactor accident in Switzerland in 1969 and both events in Jaslovske Bohunice in Czechoslovakia in 1976-77. The analysis is in line with the conclusions given at the IAEA International Conference on Effective Nuclear Regulatory Systems in Canada in 2013 giving that lessons to be learned from such events and accidents should be taken into account systematically. This is one of the main tasks of regulatory authority, operators and others involved in nuclear safety arena. A time distance from the events which happened before the establishment of the INES scale combined with new comprehensions could be a contribution to better understanding nuclear accidents.






10.09.2013 10:40 Poster session 1

Severe accidents - 414

Microstructure and Mechanical Properties of Zircaloy-4 Cladding Hydrogenated at Temperatures Typical for LOCA Conditions

Anton P. Pshenichnikov, Juri Stuckert, Mario Walter

Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe, Germany

anton.pshenichnikov@partner.kit.edu

 

The series of single rod tests was performed at KIT in framework of the new QUENCH-LOCA programme to investigate the properties of claddings hydrogenated to values between 600 and 4600 wppm. The samples were charged with hydrogen from gaseous phase under temperatures of 600-900 °C.

Tensile tests of these samples have shown the strong impact of hydrogenation on mechanical properties. Even the small periods of hydrogenation have led to significant decrease of failure strain and to a slight hardening in comparison to the original samples annealed in Ar. To investigate the different impact of annealing and hydriding on the material properties the microhardness measurements were performed. It was shown that microhardness changes in nonlinear way. The distinct transition from annealing softening to hydrogen hardening was obtained.Different treatment periods and different hydrogen content have led to a change in microstructure. The different stages of structure transformation were clearly obtained. The Scanning Electron Microscopy of the ruptured sample surface was performed to observe the fracture behaviour with respect to hydrogen content. The X-ray diffractometry (XRD) analysis was applied to observe the existing phases in the tested samples including possibly precipitated hydrides as well as change in the lattice parameters a and c. It has been shown that every treatment conditions used in our investigation have led to the formation of hydrides detectable by means of XRD. The presence of ?- and ?-hydrides was clearly shown. The evolution of peak intensities and peak shift was analysed to estimate the texture change and concentration of dissolved hydrogen correspondingly.






10.09.2013 10:40 Poster session 1

Severe accidents - 416

Validation of SOCRAT/HEFEST-EVA code capabilities of MCCI modeling on experimental data

Evgeny Moiseenko, Aleksandr Filippov

Nuclear Safety Institute of Russian Academy of Sciences , 52, B. Tulskaya, 115191 Moscow, Russian Federation

moi@ibrae.ac.ru

 

HEFEST-EVA code (Highly Efficient Finite Element Solution of Thermal problems for Ex-Vessel stage Analysis) is a module of SOCRAT system code package, which is used for modeling melt-structures interaction after core meltdown at both in-vessel and ex-vessel stages. HEFEST-EVA can also function in a stand-alone mode. The numerical procedure is based on finite element method (FEM) solution of 2D energy equation along with material relocation algorithms and chemistry models. The treated configurations are: debris in lower plenum, normally or inversely stratified molten pool in VVER lower head with its heat erosion, normally or inversely stratified molten pool in VVER core catcher, the melt in VVER reactor pit - MCCI.

The validation of code capabilities for modeling main MCCI phenomena was performed on a number of 1D experiments of SURC, SWISS and ACE series and 2D experiments of COMET and BETA series. These experiments allowed validation of models of wide range of phenomena: concrete ablation and 2D cavity profile, melt temperature evolution, gas and vapor release, oxidation reactions in the melt, effects of top flooding including top crust formation, effects of iron rebar, fission product release.






10.09.2013 10:40 Poster session 1

Severe accidents - 417

CFD Predictions of a Light Gas Dissolution and Stratification in a Safety Containment

Mouza Al Hebsi, Yacine Addad

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

yacine.addad@kustar.ac.ae

 

A 3-dimensional CFD code, which solves the Unsteady Reynolds Averaged Navier-Stokes (URANS) equations, with a turbulence model to close the system, is to be used for analysis of the light gas dissolution in a safety containment. To achieve this, a number of numerical studies have been conducted to validate the CFD codes (Fluent, Visser et al. (2012) and GasFlow II, Royl et al. 2006) or examine the predictions of RANS models (k-eps and RSM models by Zirkel (2010), and more recently the scale-adaptive k-omega SST model by Gärtner 2012). However, these studies are always limited to qualitative comparisons with the experimental data or include many complex physical phenomena making it very difficult to isolate the influence of the turbulence modeling part on the predicted results. In addition, newly developed, advanced models, exist in the commercial codes that account for laminar to turbulent transition such as transition model developed by Langtry (2006); or Reynolds stresses anisotropy such as the RSM models.

Hence, a detailed validation study is required on a simplified test case (as considered in the present study) to asses and examines the capability of these models to predict transient steam-hydrogen distributions, especially if these codes are to be used for power plants safety analysis such was the case for the work reported in Kim et al. (2004) for the APR1400 power plant.Thus, the aim of this paper is to test and validate the CFD code Star-CCM+ on a simplified containment geometry and compare with existing experimental data available from the German ThAI facility and an LES simulation of the same case which is being conducted in parallel to the present study.






10.09.2013 10:40 Poster session 1

Severe accidents - 418

Preservation of Thermalhydraulic and Severe Accident Experimental Data Produced by the European Commission

Patricia Pla1, Luca Ammirabile2, Alessandro Annunziato3, Ghislain Pascal2

Universitat Politecnica de Catalunya, C. Jordi Girona, 31, 08034 Barcelona, Spain1

European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands2

European Commission, Joint Research Centre, Institute for the Protection and Security of the Citizen, Via E. Fermi, 2749 , 21027 Ispra (VA), Italy3

patricia.pla-freixa@ec.europa.eu

 

The experimental data recorded in Integral Effect Test Facilities (ITFs) are traditionally used in order to validate Best Estimate (BE) system codes and to investigate the behaviour of Nuclear Power Plants (NPP) under accident scenarios. In the same way, facilities dedicated to specific thermal-hydraulic Severe Accident (SA) phenomena are used for the development and improvement of specific analytical models and codes used in the SA analysis for Light Water Reactors (LWR).

The extent to which the existing reactor safety experimental databases are preserved was well known and frequently debated and questioned in the nuclear community. The Joint Research Centre (JRC) of the European Commission (EC) has been deeply involved in several projects for experimental data production and experimental data preservation.In the area of ITFs the JRC LOBI facility and its project produced data of 70 experiments simulating different accidents and transients in PWR. The JRC was engaged during decades in relevant SA experimental projects: The FARO, KROTOS facilities simulated Melt Fuel Coolant Interaction (MFCI) phenomena, considering either in-vessel (quenching) and ex-vessel (spreading) experiments and potential situations for steam explosions. The STORM facility simulated experiments in the area of aerosol transport.The STRESA (Storage of Thermal REactor Safety Analysis Data) web-based informatics platform was developed by JRC-Ispra with the main objective to disseminate documents and experimental data from large in-house JRC scientific projects, and it is extensively used in order to provide a secure repository of experimental data, exploiting computer information technologies for access and retrieval of the information. At present the JRC STRESA databases are hosted and maintained by JRC-Petten.The paper is presenting these large EC initiatives on the production of experimental data and its storage in the JRC STRESA node (http://stresa.jrc.ec.europa.eu/stresa/). FARO, KROTOS and STORM data are accessible also through the JRC SARNET-STRESA portal (DATANET) (http://stresa.jrc.ec.europa.eu/sarnet/) which is connected to several other STRESA institutions nodes with SA experimental data. The objective of the paper is to further disseminate and promote the usage of the database containing these experimental data and to demonstrate long-term importance of well maintained experimental databases.The present JRC NURAM Action is engaged in the development of a new STRESA tool in the framework of the NUGENIA network for SA data. The target is to keep the main features of the existing STRESA structure but using the new informatics technologies that are nowadays available and providing new capabilities. A future activity will be also to promote the new database as a secure EU storage for SA experimental data and calculations.






10.09.2013 10:40 Poster session 1

Post Fukushima actions - 1201

ANSALDO NUCLEARE and the Fukushima Issues

Federico Fortunato, Fulvio Fardi

Ansaldo Nuclear S.p.a., C.so F.M. Perrone 25, 16152 Genova, Italy

federico.fortunato@ann.ansaldo.it

 

The stress tests, performed In response to the Fukushima nuclear accident on 11 March 2012, resulted in reports outlining the reassessment of the safety margin provided by the plant design and configuration against “beyond-design-events” and the identification of enhancements and protective measures capable of increasing the robustness of a NPP at the occurrence of those events. Based on the results of the stress tests, the entire nuclear community began studying products and new solutions to solve the issues outlined and implement the enhancements.

Based on its technical capabilities and the experience gained over the years as a during nuclear plant designer, Ansaldo Nucleare has launched a research and development program to develop such new products/solutions that can help facing/mitigating post-Fukushima issues. The As a result of this program are the following products described below have been developed:• New Design Concept for Diesel Generator Station• Mobile Diesel Generators• Flexible Connections• Ultimate Heat Sink• Cooling of Spent Fuel Pool






10.09.2013 10:40 Poster session 1

Post Fukushima actions - 1202

The Preparation of the Slovenian Post-Fukushima National Action Plan

Siniša Cimeša, Andrej Stritar, Djordje Vojnovič, Matjaž Podjavoršek

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

sinisa.cimesa@gov.si

 

As it was agreed in the ENSREG group, the Slovenian Nuclear Safety Administration (SNSA) prepared a complementary National Action Plan (NAcP) of improvements, which is based on the lessons learned from Fukushima accident in March 2011. The Slovenian NAcP and the NAcPs of other nuclear countries of the European Union, as well as Switzerland and Ukraine were reviewed in the framework of NAcP peer review workshop organized by the ENSREG in April 2013.

The core of the Slovenian NAcP and post-Fukushima improvements in Slovenia represents the planned Krško nuclear power plant’s Safety Upgrade Program (SUP), which includes the installation of passive autocatalytic recombiners, containment filtered venting system, establishment of the emergency control room and relocation of the technical support centre (i.e. emergency control centre) into a bunkered and severe accident protected building, installation of alternative ultimate heat sink and additional pumps for injecting into steam generators, the reactor coolant system and spent fuel pool, all designed for the design extended conditions.Beside the implementation of the Safety Upgrade Program the SNSA identified 11 additional actions that could further enhance nuclear safety in Slovenia, either indirectly by changing the legislation, hosting additional peer reviews, performing additional studies, or directly by improving the nuclear power plant and regulatory body processes, enhancing of emergency preparedness and nuclear safety infrastructure or improving the safety culture of both the operator and the regulatory body.Most of the measures which are part of the Safety Upgrade Program will be implemented in medium term till 2016. The rest of the measures are scheduled until 2018.Specific attention will be given to the description of design extension conditions, which will represent the design bases for the SUP system and structures.






10.09.2013 10:40 Poster session 1

Post Fukushima actions - 1204

Improved protection against external flooding at Tihange NPP site

Dimitri Vanbellinghen, Pierre Latteur, Loic Villers, Gauthier Polet

TRACTEBEL Engineering, GDF SUEZ, Avenue Ariane 7, B-1200 Brussels, Belgium

dimitri.vanbellinghen@gdfsuez.com

 

Electrabel, part of energy world leader GDF-SUEZ, operates the seven nuclear reactors (pressurized water reator type) in Belgium with a total capacity of almost 6000MWe. Four nuclear units are located in Doel and three units in Tihange. Doel Nuclear Power Plant (NPP) lies on the bank of the Scheldt and Tihange NPP along the Meuse River.

After the earthquake event at Fukushima Daiichi in Japan, the safety margins of European nuclear power plants have been analysed and reassessed considering extreme natural events and their circumstances on the units. Periodic safety review and conclusions of the Belgian Stress Tests indicate the need for improved protection of Tihange NPP against external flooding. It has been decided to adopt the 10,000 years flooding as the new design basis condition of Tihange NPP. The associated improvements consist of implementation of successive and independent protections based on in-depth-defense principles.These protections against external flooding are intended to ensure the fulfillment of the fondamental safety fonctions i.e. control of the reactivity, cooling of the core and of the spent fuel and containment integrity. The main protection is composed of a peripherical and volumetric protection of the NPP whose function is to maintain a dry site during the flooding period, by taking into account the revised design flooding. It includes the construction of a new concrete wall surrounding the site (2km long), of huge isolation structures in the intake channel and in the various discharge channels, and of other facilities. Complementary equipment and structures have to be installed to evacuate accumulated water on site during the flooding period.Additional in-depth-defense protections are implemented in order to face potential failure of the main protection or higher flooding level i.e. beyond design conditions. It consists of improvement of reliability, capacity and robustness of mobile and fixed ultimate means.As engineering and consultancy services company, Tractebel Engineering, part of GDF-SUEZ, is mandated by Electrabel to study, develop and implement the construction of the technical solutions related to these external flooding protections.The paper to be submitted after the present abstract acceptance will describe and detail the modifications performed or to be performed for the improved protection against external flooding at the Tihange NPP.






10.09.2013 10:40 Poster session 1

Nuclear fusion - 1401

Reaction of Oxygen Plasma with Hydrogenated W-C Deposits

Alenka Vesel1, Miran Mozetič1, Marianne Balat – Pichelin2

Institut "Jožef Stefan", Jamova 39, 1000 Ljubljana, Slovenia1

CNRS-PROMES, Laboratoire Procédés, Matériaux & Energie Solaire, 7 Rue du Four Solaire, 66120 Font-Romeu Odeillo, France2

alenka.vesel@ijs.si

 

Removal of hydrogenated carbon from mixed W-C films was studied for two different initial compositions (approx. 20 and 50 at.% of W). The deposits were prepared by sputter deposition in a mixture of Ar and C2H2. Samples were exposed to oxygen plasma created in a quartz tube by a microwave discharge at the power of 1000 W. The evolution of the coating composition was monitored versus plasma treatment time by AES depth profiling. Furthermore, XRD characterization was performed to reveal crystalline structure of the samples, while SEM characterization was performed to observe morphological differences. The results showed that almost entire hydrogenated carbon could be removed from the mixed W-C films as long as the initial concentration of tungsten is reasonably low (e.g. 20 at.% of ). At high tungsten concentration (50 at.% of W), a passive WO3 layer was formed on the surface of the samples preventing interaction with hydrogenated carbon even after prolonged plasma treatment. XRD characterization revealed that the oxide film was highly crystalline. Successful removal of carbon from W-C films with low W content was explained by migration of particles through porous films as observed on SEM images.






10.09.2013 10:40 Poster session 1

Nuclear fusion - 1403

Towards Possible Control of Plasma Outflow in Fusion-Relevant Devices via Employing Virtual Terminating Surfaces

Nikola Jelić1, Leon Kos2

University of Innsbruck, Department of Theoretical Physics, Plasma and Energy Physics Group, Technikerstrasse 25, A-6020 Innsbruck, Austria1

University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta. 6, 1000 Ljubljana, Slovenia2

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

 

Plasma-boundary interaction can manifest via either "clean" and/or "impure" features and effects. The term "clean" here means that no new kind of atoms, ions or molecules appear near and/or at boundaries, except those which primarily originate from the plasma. Some basic such effects (besides diverse particle-particle interactions) are particle-field-particle and particle-surface transfers of moments. Second "clean" effect is neutralisation of ions at the wall and their free reflection back to the plasma, and/or the ion absorption/implementation into the solid surface. Depending on plasma density, reflected neutrals can be again ionised, i.e., to become the "recycled" ions. In fusion devices increased recycling may lead to formation of a region of plasma located near the main ion outflow surface (divertor), well "detached" from the material, in which volume processes, start to dominate dissipation of kinetic ion energy at the surface.

The recycling and detachment regimes are highly desirable in fusion devices, since the plates are better protected from damages (such as arcing and erosion), and the plasma is better protected against contamination, (e.g., sputtering and dilution originated from the material). Physics description related to the plasma-wall transition and interaction during low (sheath-limited -- radially 1D) and high (conduction-limited -- parallel to flow 1D) recycling, layer detachment (intense -- parallel to flow 1D, conduction-limited, with strong parallel "cooling", i.e., energy detachment) and "flame"-detached (complete detachment, intrinsic 2D, complete momentum detachment with strong reduction of particle flux to plate is extremely demanding. Simulations of tokamak SOLs via 2D codes such as SOLPS-B2 require additional CPU-expensive modules for properly simulate neutral transport represented by e.g., 3D Monte-Carlo IRENE code.Finally, engineering aspects appearing with installing divertor plates are very untrivial and become even more and more demanding with each new "movement" of the plates further from X-point (e.g., in very promising snowflake divertor geometry [Ryutov 2008]). Therefore the idea emerges to get rid of the plasma-divertor interaction problems in fusiondevices, and also to reduce contamination in e.g., technology plasmas via simple proposing a "virtual plate", i.e., a DL-structure, as the main surface for outflow of ions. Possibilities of establishing conditions for their formation are investigated.






10.09.2013 10:40 Poster session 1

Nuclear fusion - 1406

Modelling hydrogen-metal surface interactions – the integral study

Anže Založnik1, Iztok Čadež1, Sabina Markelj1, Vida Žigman2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Univerza v Novi Gorici, Vipavska cesta 13, 5000 Nova gorica, Slovenia2

iztok.cadez@ijs.si

 

Neutral hydrogen interaction with dominantly metallic surfaces of plasma facing components in tokamaks plays an important role in determining plasma wall interaction in main vessel and especially in bottom part of a tokamak, divertor. Detailed modelling of edge plasma requires detailed knowledge of state specific reaction rates and cross sections for relevant processes. Therefore, vibrational excitation and de-excitation, dissociation from both ground and vibrational states on hot surfaces and atom/ion recombination on cold surfaces are all considered significant for understanding and controlling the edge plasma characteristics. We have been studying these reactions involving neutral atoms and molecules experimentally by operating a hydrogen source cell for vibrationally excited molecules in a low pressure regime (to ensure the dominance of particle-surface interactions) and theoretically by physics-based modelling of the above listed interactions, taking into account the particular cell characteristics [1,2].

The linear forward kinetic model at the current stage includes molecules in all vibrational states (ground and 14 excited states), dissociated atoms and atoms adsorbed on the wall, and thus describes a system of 17 particle species in total, undergoing dominant particle-metal wall interactions. In order to drive the model, a dataset of state specific transition probabilities for all the above specified interactions are compiled using theoretical and empirical data from literature. Successful predictive modelling relies heavily on the ingested data sets and this makes the basis of our procedure: to infer the appropriate/optimal transition probabilities, through concerted modelling and continued data analysis. We present the performance of the model using databases which are needed to both drive the model and to validate its output results, and complement it with the first results of the corresponding inverse modelling technique. [1] Markelj,S., and Čadež, I.,. J. Chem. Phys. 134 (2011)124707-1-17. [2] Žigman, V., Nucl. Eng. Des. 241 (2011) 1272-6.






10.09.2013 10:40 Poster session 1

Nuclear fusion - 1408

Kinetic Simulations in Support of Probe Measurements of COMPASS scrape-off-layer

Jernej Kovačič1, Tsviatko Popov2, Miglena Dimitrova3, Tomaž Gyergyek4, Milan Čerček1, Renaud Dejarnac5, Michael Komm5, Jan Stöckel5, Radomir Panek5

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Faculty of Physics, St. Kliment Ohridski University of Sofia, 5 James Boulcher Blvd., 1164 Sofia, Bulgaria2

Bulgarian Academy of Sciences, Emil Djakov Institute of Electronics, 72, Tzarigradsko chaussee blvd, 1784 Sofia, Bulgaria3

University of Ljubljana, Faculty of Electrical Engineering, Tržaška 25, 1000 Ljubljana, Slovenia4

Institute of Plasma Physics, Czech Academy of Science, Za Slovankou 3, 18200 Prague, Czech Republic5

jernej.kovacic@ijs.si

 

Electrical probes remain one of the important diagnostic tools for edge and SOL tokamak plasma, even as we steadily approach the era of big fusion reactors. The measured I-V characteristic of a simple Langmuir probe can provide a variety of plasma parameters with an excellent spatial and temporal resolution. However, interpretation of the obtained data is often a subject of oversimplification, leading to imprecise results. One such peculiarity in tokamak probe measurements is the use of the part of the I-V characteristic only below the floating potential Vfl and assuming a Maxwellian electron energy distribution function (EEDF), without, in fact, measuring it. Such classical approach can therefore overestimate the overall temperature of the electron population due to small population of fast electrons, which can result in significant errors in e.g. plasma potential ?pl evaluation. An advanced method has been developed, which utilizes the electron part of the characteristic even in strongly magnetized plasma and has been applied in edge and SOL tokamak plasmas with success [1]. This progressive approach i.e., first-derivative technique, is used here to evaluate the real EEDF in the area of divertor probes close to the strike points, which - in return - significantly decreases the classically predicted density of high energy electrons.

In our work we then made use of a fully-kinetic code BIT1 [2] to simulate a single flux tube inside the SOL region during a D-shaped L-mode discharge. The flux tube is bounded between two divertor plates and volumetric injection of particles represents the cross-field diffusion as the plasma source. Injected electrons have high energy Maxwellian distribution that is “reshaped” through collision interactions and “measured” at the divertor sheath, where a high energy tail to the Maxwellian distribution is expected to still be present. In the first-derivative method this is modelled via bi-Maxwellian EEDF configuration, which leads to the possibility of a direct comparison between the EEDF obtained from divertor probe measurements to the results from PIC simulation. The qualitative agreement between the results from two approaches is good.REFERENCES[1] Tsv. K. Popov, P. Ivanova, J. Stöckel, R. Dejarnac, “Electron energy distribution function, plasma potential and electron density measured by Langmuir probe in tokamak edge plasma”, Plasma Phys. Contr. Fusion, 51, 2009, pp. 065014 (15 pages)[2] D. Tskhakaya, A. Soba, R. Schneider, M. Borchardt, E. Yurtesen, J.Westerholm, PIC/MC code BIT1 for plasma simulations on HPC, Proceedings of the 18th Euromicro Conference on Parallel, Distributed, and Network-based Processing, IEEE, (2010), 476-481






10.09.2013 10:40 Poster session 1

Nuclear fusion - 1410

W-SiC composite material for potential fusion application

Aljaž Ivekovič, Saša Novak

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

aljaz.ivekovic@ijs.si

 

The proposed W-SiC composite material fabricated by active filler controlled pyrolysis of tungsten or tungsten-silicon carbide powder mixtures is a promising material for potential applications in fusion environment. High temperature stability, together with low activation of both W and SiC make the material an excellent candidate for in-vessel components. The material has already been recognised as a potential joining material and is being investigated as a potential He permeation barrier for SiCf/SiC composites.

In this work the effect of composition and different fabrication temperatures (1500-2000 °C) on final properties of the material has been investigated. A series of samples with 0-60 vol. % of tungsten in the initial W-SiC powder mixture, densified by polymer infiltration and pyrolysis has been fabricated and characterised. Samples with high tungsten content (>50 vol. %), exhibited high density and good mechanical properties (?=> 350 MPa, > 1000 HV) with relatively low room temperature thermal conductivity of 10-20 Wm-1K-1, which was increased to 25-35 Wm-1K-1 at 1000 °C. Samples with 60 vol. % of W fabricated at 1700 °C were also subjected to thermal shock tests at different base temperatures (RT, 400°C, 1000°C) with power density close to the damage threshold of pure tungsten (around 250 MW/m2) and exposure to 100 to 1000 ELM-like thermal shock events.






10.09.2013 10:40 Poster session 1

Nuclear fusion - 1412

Flow and Heat Transfer Characteristics of Multiple Impinging Jets: Large Eddy Simulation

Martin Draksler1, Boštjan Končar1, Leon Cizelj1, Bojan Ničeno2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Paul Scherrer Institut (PSI), Nuclear Energy and Safety Department, OVG/421, CH-5232 Villigen PSI, Switzerland2

martin.draksler@ijs.si

 

Jet impingement technique is characterized by high heat removal capability, and therefore recognized as one of the most efficient cooling methods among single-phase convection flows. As such, it has been proposed also for cooling of the divertor, a plasma facing component of the future fusion reactor DEMO. Due to the divertor application where multiple jets are used to achieve the uniformity in heat transfer over wider area, our interest is focused to configurations of multiple, circular, and highly turbulent jets in hexagonal arrangement.

In this paper the flow and heat transfer characteristics of impinging jets are analysed numerically, by the means of Large Eddy Simulation. Numerical study was performed with the open-source code PSI-Boil. The explicit Wall-Adapted Local Eddy-viscosity (WALE) subgrid-scale model was used to model the unresolved scales. Numerical results were validated against the available experimental data for downstream locations which are farther than half diameter from the nozzle exit. Since velocity profile at the nozzle exit was not measured in the experiment, the boundary conditions at the nozzle exit represent the largest source of uncertainty in our simulations.Validation of the preliminary results against experimental data revealed that a flat profile at the nozzle exit is a rather weak approximation of the experimental conditions, and therefore additional “fitted” profiles, obtained by prior steady-state (RANS) simulation were tested. In addition, a pseudo-random generator by Kraichnan was used to impose predefined velocity fluctuations at the nozzle exit.The main objective of the present study is therefore to analyze the influence of boundary condition at the nozzle exit on the formation and dynamics of coherent structures and heat transfer characteristics. The simulation results with different inlet conditions are compared and analyzed.






10.09.2013 11:20 Invited lecture 3

Invited lectures - 104

Better reactors grow from better simulations

Emilio Baglietto

Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, 24-107, Cambridge, MA 02139, USA

emiliob@mit.edu

 

The evolution of nuclear reactor design is a slow but continuous progression from the reliance on inflexible empirically derived guidelines to the more versatile numerical analyses of next-generation tools. The adoption of the new analysis and simulation techniques is pushing the operational limits of reactors and fuel, with large impact on the economics of the plants, while simultaneously improving the safety standards. One of the most recent and important developments has been the adoption and adaptation of computational fluid dynamics (CFD). Used alone or in combination with neutronics, structural, material performance and system analysis tools, the analysis software has enabled nuclear operators and vendors to greatly enhance the capacity and availability of current reactors.

A tight coupling of the thermal-hydraulics, structural and neutronics phenomena will not only demonstrate that there is more margin of safety to increase power, but also may give insight into issues beginning to emerge due to longer operations, such as crud and grid-to-rod fretting. It has helped that the advanced simulation tools can provide a predictive means of avoiding these problems in the design phase. The net result is going to be a new generation of analysis capability that can provide better insight into how to improve designs and how to introduce new materials that will yield benefits in plant upgrading and life extensions. The industry will move from the application of conventional methods that rely on experimental correlations to using CFD on small components, larger components, and full-plant.






10.09.2013 12:00 Thermal-hydraulics 1

Thermal-hydraulics - 201

Long term station blackout analyses of two loop PWR using RELAP5/MOD3.3

Andrej Prošek, Leon Cizelj

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

andrej.prosek@ijs.si

 

Stress tests required evaluation of the consequences of loss of safety functions from any initiating event (e.g., earthquake or flooding) causing loss of electrical power, including station blackout (SBO), loss of the ultimate heat sink or both. SBO scenario involves a loss of offsite power, failure of the redundant emergency diesel generators, failure of alternate current (AC) power restoration and the eventual degradation of the reactor coolant pump (RCP) seals resulting in a long term loss of coolant. Long term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs) leaks assumed) to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS). For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.






10.09.2013 12:20 Thermal-hydraulics 1

Thermal-hydraulics - 203

Thermal-Hydraulic Evaluation of the Advanced Safety Design Features of APR+

Young Seok Bang, Sweng-Woong Woo, Andong Shin, Gonghee Lee

Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

k164bys@kins.re.kr

 

The APR+ standard design is an evolutionary development of the proven APR1400 design being constructed at the Shin-Kori Nuclear Generating Station in Korea. APR+ incorporates a variety of engineering and operational improvements designed to provide additional reliability and safety margins when compared to the APR1400 design.

The Passive Auxiliary Feedwater System (PAFS) and improved Emergency Core Cooling System (ECCS) in fully independent four trains are the most outstanding safety features. In addition, ECCS Core Barrel Duct (ECBD) and Safety Injection Tank (SIT) with Fluidic Device (FD) are adopted in APR+ to reduce the bypass of ECCS water during the loss-of-coolant accidents (LOCA). Those advanced features may have an impact of the plant response and system performance following design basis accidents (DBA) and Beyond DBA. In the present Paper, performance of the advanced design features was evaluated through the thermal-hydraulic analysis. For this purpose, Phenomena Identification and Ranking Tabulation (PIRT) process was applied to select the important accidents under the advanced safety design features, especially PAFS and advanced ECCS. Main Feedwater Line Break (MFLB) and Large Break LOCA were identified as the important accidents to be investigated through expert panel discussion. Also lists of specific thermal-hydraulic phenomena were identified for those two accidents, in which the condensation in horizontal heat exchanger of the PAFS and ECCS flow division in reactor vessel downcomer and ECBD were included. System thermal-hydraulic responses were calculated for MFLB and LBLOCA, respectively, using the regulatory auditing codes, MARS-KS, which has been validated for the available experiments, PASCAL in KAERI, Korea. Adequacy of the predicted phenomena which was highly-ranked at PIRT process such as natural circulation flow over piping of PAFS and their influences to the safety criteria were addressed. Based on the predicted results of two DBA, the degree of enhancement of safety margin was also discussed.






10.09.2013 12:40 Thermal-hydraulics 1

Thermal-hydraulics - 215

Thermal Hydraulic System Codes Performance in Simulating Bouyance Flow Mixing Experiment in ROCOM Test Facility

Eugenio Coscarelli, Sergii Lutsanych, Francesco D'Auria

University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Livornese 1291, 56122 Pisa, Italy

eugenio.coscarelli@ing.unipi.it

 

The MSLB (Main Steam Line Break) accident scenario is one of the severe abnormal transients that might occur in a NPP. The main concerns of the MSLB are the potential return to power condition and the occurrence of PTS (Pressurized Thermal Shock) as a consequence of both rapid depressurization of the secondary circuit and the entrainment of cold water into the core region. Assessment of these issues is the main objective of integrated experimental tests carried out in the PKL-III and ROCOM facilities. The first test rig is aimed to simulate thermal-hydraulic phenomenology at the system level whereas supporting ROCOM test facility is focused on the coolant mixing phenomenon took place in the Reactor Pressure Vessel (RPV). Combination of these two typologies of experiments (integral effect test (IET) and separate effect test (SET)) provides appropriate experimental data for CFD and TH-SYS (Thermal Hydraulic-SYStem) codes validation against the relevant thermal hydraulic phenomena that occur during the MSLB.

The main purpose of this study is to evaluate the capability of two TH-SYS codes TRACE V5 and CATHARE2 V2.5 to predict reasonably buoyancy driven mixing phenomena that affects the IVF (In-Vessel Flow) and the distribution of coolant temperature at the core inlet using 3-D porous media approach. Test 1.1 that had been carried out in ROCOM facility was selected to investigate the coolant mixing inside the RPV under flow conditions typical for a MSLB scenario. Averaging analysis of integral behaviour of the experimental and calculated temperature distributions inside the RPV has been performed.






10.09.2013 14:30 Invited lecture 4

Invited lectures - 103

EU DEMO Design and R&D Studies

Gianfranco Federici, Jon Harman and EFDA PPPT Team

EFDA Power Plant Physics & Technology, Boltzmann str.2, Garching 85748, Germany

gianfranco.federici@efda.org

 

Demonstrating the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle, in a DEMOnstration Fusion Reactor is viewed by many of the Nations engaged in the construction of ITER as the remaining crucial step towards the exploitation of fusion power after ITER. This talk reviews the recent DEMO design and R&D work in progress in Europe, being implemented in the EFDA Power Plant Physics and Technology (PPPT) Programme and provides an outlook on the activities that are expected to be launched as part of the new EU fusion roadmap in the period 2014-2018 [1].

ITER represents an indispensable step and significant advances in technology and physics are going to be acquired by its construction and operation. However, for a DEMO reactor, there are still some outstanding physics and engineering challenges with potentially large gaps beyond ITER where R&D and design work must be urgently launched. For example, a solution for the heat exhaust is needed and there is a risk that the divertor concept to be used by ITER cannot be extrapolated to a DEMO fusion reactor. Hence, a parallel, aggressive programme on alternative solutions for the divertor is necessary. Similarly, the R&D to ensure tritium self-sufficiency and efficient power extraction in a breeding blanket should be strengthened. As a risk mitigation strategy, the evaluation, and potentially, the development, in addition to the two TBM designs based on the use of helium as coolant, of parallel lines such as a water-cooled lithium lead design, or dual-coolant design is recommended. It is expected that the DEMO breeding blanket selection will be made taking into account also the constraints on coolant and breeder arising from the choice of an efficient Balance of Plant.The recent EU fusion roadmap (Ref. [1]) advocates for a pragmatic approach and considers for the initial design integration studies a pulsed “low extrapolation” DEMO that could be delivered in the short to medium term. This should be based on mature technologies and reliable regimes of operation to be, as much as possible, extrapolated from the ITER experience, and on the use of materials adequate for the expected level of neutron fluence (see Ref. [2]). It argued that by waiting to design DEMO for the ultimate technical solutions in each area would postpone the realization of fusion indefinitely. Since the mission requirements of a near-term DEMO put more emphasis on solutions with high technical readiness levels and realistic performance and component reliability, rather than on high-efficiency, the R&D priorities in the Roadmap are defined to achieve these goals. Nevertheless, one has to recognize that these goals remain very ambitious and many technology advances and innovations would be required. More advanced technological solutions could also be developed but on a longer timescale and be part of a parallel long term R&D.A broad but integrated design-oriented approach is viewed as essential during this early concept design stage: (i) to better understand the problems and evaluate the impact of uncertainties and technical risks of foreseeable technical solutions; (ii) to identify design trade-offs and constraints to address the most urgent issues in physics, technology and system engineering integration; and (iii) to prioritize the R&D needs. Ensuring that R&D is focused on resolving uncertainties in a timely manner and that learning from R&D is used to responsively adapt the technology strategy will be crucial to the success of the DEMO Programme.[1] F. Romanelli, The European Fusion Roadmap, EFDA_D_2M8JBG v1.0 - https://user.efda.org/?uid=2M8JBG[2] D Stork et al, Materials R&D for a timely DEMO: key Findings and Recommendations of the EU MaterialsAssessment Group, Dec. 2012. EFDA_D_2MJ5EU v1.0 - https://user.efda.org/?uid=2MJ5EU.






10.09.2013 15:10 Nuclear fusion

Nuclear fusion - 1407

Thermal Analyses of the Breeder Unit of the Helium Cooled Pebble Bed Test Blanket Module

Marigrazia Moscardini, Simone Pupeschi, Donato Aquaro

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

marigrazia.moscardini@ing.unipi.it

 

This paper deals with the research activity concerning the Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM), developed in the frame of the EU-TBM Consortium of Associates. The HCPB Breeding Blanket represents a reference design concept for the first demonstration fusion reactor DEMO. HCPB concept will be tested in ITER.

HCPB is a breeding blanket in which Lithium Orthosilicate (OSI) is selected as tritium breeder, while the Beryllium (Be) is adopted as neutron multiplier. OSI and Be are used in form of pebble beds and are packed within a box structure (made of a Reduced Activation Ferritic Martensisitc Steel) actively cooled by helium. The Breeder Units (BU) are located within the TBM box in the spaces defined by the structure of the stiffening grids. The temperature distribution in the Breeder Units is one of the key design aspects in order to ensure a safe and efficient operation of the TBM. The aim of this study is to investigate the BU thermal performance in order to guarantee its reliable and efficient operation, taking into account design temperature limits and recommended operating temperature ranges for each material.Steady state and transient thermal analyses were carried out implementing a FEM model of the TBM-BU by ANSYS® code. Uncertainties on the input data, such as those related to the material properties, thermal loads and boundary conditions have been considered. Parametric sensitivity studies have been carried out in order to determine the influence of input data possible range of variations on the maximum and average temperatures of BU structural and functional materials. Sensitivity studies have been completed by simulating the worst possible combination of the considered input parameters. Steady state and transient thermal analyses have shown that, in all analyzed materials, temperatures fulfill the allowable limits. The results of the transient analyses highlighted that Be and OSI average temperature are not sufficient to guarantee an efficient operation of the TBM. Future design improvements will be necessary to achieve higher average temperatures so to assure lower tritium residence time and inventory in the Breeder Zone. The sensitivity studies showed that the OSI and Be thermal conductivity strongly influence the BU maximum temperature values.






10.09.2013 15:30 Nuclear fusion

Nuclear fusion - 1405

Deuterium thermal desorption from mixed layers relevant for ITER

Sabina Markelj1, Iztok Čadež1, Primož Vavpetič1, Primož Pelicon1, Corneliu Porosnicu2, Cristian P. Lungu2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

National Institute for Laser, Plasma and Radiation Physics, P.O. Box MG36, Magurele-Bucharest, Romania2

sabina.markelj@ijs.si

 

Mixed material deposits are formed by material migration during operation of fusion devices. Fuel retention in these deposits during their formation and after is important for predicting the overall fuel retention. The materials that will be used in ITER are beryllium for the inner wall and tungsten in the divertor area so that the mixed material deposits will mostly consist of these two metals and possibly also of seeding (nitrogen) and intrinsic impurities such as carbon. The formation of mixed material deposits and deuterium co-deposition is extensively studied by in situ and post mortem analyses in tokamaks but also by depositing such layers using laboratory plasmas (e.g. magnetron sputtering) with addition of deuterium.

In this work we have studied deuterium thermal desorption from mixed material C:W and C:W:Al (Al used as Be substitute). The layers were obtained using Thermionic Vacuum Arc (TVA) method [1] under deuterium atmosphere in the vacuum chamber during the deposition process. Thermal desorption was studied in situ by Nuclear Reaction Analysis (NRA) and mass spectrometry during linear sample heating. The deuterium desorption was correlated by decrease of the NRA signal at single 3He probing beam energy (2.5 MeV) and increase of mass 4 in the mass spectrometer both recorded during heating. Main deuterium desorption takes place at relatively high temperature, 1000 K, what is higher than typical desorption temperatures for W [2] and amorphous deuterated carbon films [3]. The deuterium depth profile in the layer was measured before and after heating. Deuterium atom absorption in multilayer material (200 nm of C on top of 500 nm W (graphite substrate)) was also studied by in situ NRA, where layer was exposed to deuterium atom beam for one day and results will be presented. [1] C. P. Lungu et al., Phys.Scr. T128 (2007) 157–161[2] O. Ogordnikova et al., J.Nucl. Mater. 313-316, 469 (2003) and O.Ogordnikova et al., J. Appl. Phys. 109, 013309 (2011).[3] E. Salancon et al., J.Nucl. Mater. 376, 160 (2008).






10.09.2013 15:50 Nuclear fusion

Nuclear fusion - 1402

Unified Approach to Visualizations within the European Integrated Tokamak Modelling Framework

Leon Kos1, Hans-Joachim Klingshirn2, Pablo Luis Garcia Müller3, Frederic Imbeaux4

University of Ljubljana, Faculty of Mechanical Engineering, LECAD Laboratory, Aškerčeva cesta. 6, 1000 Ljubljana, Slovenia1

Max-Planck-Institut für Plasmaphysik (IPP), Boltzmannstr. 2, D-85748 Garching, Germany2

CIEMAT, Avda. Complutense, 22, 28040 Madrid, Spain3

CEA - Cadarache, DER/SERA Bat 212, CE Cadarache 13108 St. Paul Durance, France4

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

 

Diverse visualization approaches are used within integrated fusion simulations in the frame of by the European Integrated Tokamak Modeling Task Force (ITM-TF).

In an effort to provide diversity and at the same time universality of visualizations for different backends, ITMVis library provides a common description of the specialized plots contributed by users. Besides dedicated plots many "standard" plots are used and are described within ITM-TF database via XSD schema that contains plot representation tags. From the representation tags one can generate standard plots for a specific computer language and visualization tool in use. Such approach is used for the VisIt visualization tool, where a C++ code for all possible plots in the database is generated with XSLT transform. Given that the same translation is needed by other tools in different languages to provide standard visualizations an intermediate plot description with XML, which is easily interpreted, was introduced. With such unified approach standard and custom plots are available for different backends such as VisIt and matplotlib.






10.09.2013 16:40 Thermal-hydraulics 2

Thermal-hydraulics - 208

Pool Boiling Critical Heat Flux of Oxidized Zircaloy Surface in Saturated Water

Chi Young Lee, Chang Hwan Shin, Dong Seok Oh, Tae Hyun Chun, Wang Kee In

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

chiyounglee@kaeri.re.kr

 

The researches on an increase in CHF (Critical Heat Flux) have been highlighted, and extensively performed for the safety of nuclear reactor. In the present experimental study, the enhancement of CHF in saturated water pool was investigated using the oxidation of zircaloy surface. Four kinds of zircaloy specimens were prepared, and tested. One was a mechanically polished (i.e., non-treated) surface, and the others were the surfaces oxidized at 300, 450, and 600 °C for 10 min, respectively. The CHF value measured in the non-treated surface was in good agreement with previous correlations proposed by Zuber (1958) and Kandlikar (2001). On the other hand, all correlations tested under-predicted the CHF value of oxidized surfaces, which became much higher than that of the non-treated surface. This was because the oxidized surfaces appear the smaller water contact angles, as compared with the non-treated surface. Based on the present study, it was found that a convenient and cost-effective oxidation process proposed in this work improves the surface wettability (i.e., decreases the water contact angle), and consequently, contributes to enhancing the CHF.






10.09.2013 17:00 Thermal-hydraulics 2

Thermal-hydraulics - 228

Pump intakes - numerical simulations and experimental model solutions

Aljaž Škerlavaj, Rok Pavlin

Turboinstitut d.d., Rovsnikova 7, 1210 Ljubljana, Slovenia

aljaz.sk@gmail.com

 

Design of pump intake is crucial for stable operation of vertical pumps, such as circulating water (CW) pumps. Unstable operation occurs because of flow inhomogeneity in axial velocity distribution, due to swirling of flow or because of strong vertices. As a result the velocity distribution at the impeller eye is time-dependent. Consequences of such unstable operation range from small (e.g. noise or small vibrations) to large ones (pump failure).

Turboinštitut has been involved in providing solutions for incorrectly designed pump intakes for more than two decades. A relatively recent project was to prevent cracking of impeller blades in CW pumps of NPP Krško. The paper will present some numerical simulations, as well as some experimental case studies with hydraulic models.






10.09.2013 17:20 Thermal-hydraulics 2

Thermal-hydraulics - 212

Modelling of a Passive Catalytic Recombiner for Hydrogen Mitigation by CFD methods

Antoni Rożeń

Warsaw University of Technology, Faculty of Chemical and Process Engineering, Waryńskiego 1, 00-645, Poland

a.rozen@ichip.pw.edu.pl

 

Hydrogen, generated during normal operation of light-water reactors and emergency situations such as overheating of a reactor core, cumulates in a cooling system and in a reactor safety containment. An increase of hydrogen concentration in a gas phase can lead to uncontrolled ignition of a hydrogen-air mixture, followed by detonation and breach of the reactor safety containment [1]. One of methods of reducing of an explosion risk is to remove hydrogen from the gas phase by recombining it with oxygen in a passive autocatalytic recombiner (PAR) [2]. Hydrogen and oxygen react on surfaces of metal plates covered by platinum or palladium catalysts, mounted inside the PAR, generating steam and heat. Hot gas flows upward due to natural convection and is replaced by fresh hydrogen-air mixture. Such recombiners are self-adaptive devices, which require no external power input or supervision.

The aim of this work was to test different turbulence closure hypotheses in modelling of gas flow, heat and mass transport inside a PAR and compare their predictions with results obtained for fully laminar flow. Numerical simulations were conducted by means of Ansys Fluent 14.5 in a simplified 2D geometry and in a full 3D geometry of the PAR test device used Reinecke et al. [2]. It was assumed that the catalytic recombination of hydrogen proceeded according to Kasemo kinetics [3]. Forced convection, conduction and radiation within the PAR as well as heat conduction in the recombiner metal housing and heat exchange between the housing and surroundings by natural convection and radiation were taken into account in modelling of heat transport. It was also assumed that species transport in a gas phase occurred due to forced convection, molecular and thermal diffusion.Reynolds averaged Navier-Stokes equations of gas flow were solved by four different models of turbulence: k-omega model, intermittency (transition shear stress transport) model, k-epsilon model and Reynolds stress model (RSM). The main difficulty in resolving a gas flow inside the PAR was caused by flow laminarization in its central section, comprising several parallel steel plates covered by a platinum catalyst. As a result all four turbulence models predicted different temperature and species concentration profiles as well as the efficiency of hydrogen removal from the gas stream and the amount of heat generated due to a recombination reaction. The highest PAR efficiency in hydrogen removal was predicted by k-? model, while the lowest by the intermittency model. Predictions of the intermittency model were very close to those obtained for a limiting case of the laminar flow. On the other hand no major differences were found between results of 2D and 3D simulations except for amount of heat transferred from PAR external walls to the surrounding. In the simplified 2D geometry this heat stream was approximately twice smaller than in the full 3D geometry.Further research will focus on comparing results of RANS simulations with Large Eddy Simulations (LES) and available experimental data in order to chose the proper turbulence closure hypothesis. This work was financially supported by The National Centre of Research and Development in Poland (grant no. SP/J/7/170071/12).REFERENCES[1] IEAD-TECDOC-1661, “Mitigation of Hydrogen Hazards in Severe Accidents in Nuclear Power Plants”, IAEA, Vienna 2011.[2] E.A. Reinecke, I.M. Tragsdorf, K. Gierling, Nucl. Eng. Des., 230, 2004, pp. 49-59.[3] E. Fridel, A. Rosen, B. Kasemo, Langmuir, 10, 1994, pp. 699-708.






11.09.2013 09:00 Invited lecture 5

Invited lectures - 105

Perspective of industry on modelling of materials ageing: the multi-scale approach

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

 

One of the major challenges in the management of nuclear power plants lifetime is to justify properly that all components affected by an ageing mechanism remain within the design and safety criteria. Indeed, properly understanding the performance of materials relevant to structural components and the effect of ageing mechanisms on their performance are key issues from the start to the end of life of each NPP as recognised within NUGENIA [1] roadmap.

The high variability of ageing and degradation mechanisms necessitates predictive tools to allow transferability and interoperability of the knowledge gained from limited experimental/empirical data (surveillance, in-field monitoring…). In this regard, one long term aim is to develop fully validated multi-scale based models that link the nano scale through to the macro (i.e. structural) scale based on a multi-disciplinary approach(e.g. FP6-PERFECT and FP7-PERFORM60 [2]). In this lecture, in addition to a very concise overview of the main scientific and technical challenges identified by NUGENIA association, the main talk will be concentrated on the use of the multi-scale approach to address industrial issues, such as irradiation assisted stress corrosion cracking (IASCC) of reactor vessel internals. This topic will be used as example to demonstrate the robustness and to illustrate the complexity of the multi-scale modeling approach when applied to nuclear materials from an industrial perspective.[1] www.nugenia.org[2] www.perform60.net






11.09.2013 09:40 Materials, integrity and life management

Materials, integrity and life management - 303

K/J value estimation of specimen extracted from mock-ups containing dissimilar metal welds

Oliver Martin, Gangadhar Machina, Igor Simonovski

European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands

oliver.martin@jrc.nl

 

Primary piping systems of light water reactors (LWRs) contain a certain number of dissimilar metal welds (DMWs) connecting ferritic and austenitic stainless steels (e.g. at reactor pressure vessel nozzles, connection primary piping to pressurizer). DMWs contain different material zones, i.e. ferritic zone, austenitic zone, weld material, heat affected zones, which differ significantly in their material properties and their fracture behavior. The aim of the FP7 project MULTIMETAL is the development of a test standard for fracture resistance measurement of multi-metallic specimen and the development of harmonized procedures for DMW brittle and ductile integrity assessment. The project involves an extensive test program in which standard geometry specimens (CT, SENB, SENT) are used. These are cut from mock-ups containing DMWs resembling real DMWs from NPPs in terms of geometry, material and weld procedure. Part of MULTIMETAL are numerical analyses on the above three specimens with the aim of estimating K/J values and extraction of ? factors for different crack positions and crack lengths. In this paper the results of the numerical analyses performed by JRC so far are presented.






11.09.2013 10:00 Materials, integrity and life management

Materials, integrity and life management - 305

Automatic Fatigue Monitoring based on real Loads

Jouan Benoit, Steffen Bergholz, Juergen Rudolph

AREVA GmbH, Paul-Gossen-Straße 100, 91052 Erlangen, Germany

benoit.jouan@areva.com

 

The ageing management of power plants is nowadays a main issue for all nuclear industry actors: states, regulatory agencies, operators, designers or suppliers. Consequently, lots of operators have to deal with demanding security requirements to ensure the operation of power plants. Regarding with fatigue assessment of nuclear components, stringent safety standards are synonymous of new parameters to take into account in the fatigue analysis process, for instance: new design of fatigue curves, consideration of environmental parameters or stratification effects. In this context AREVA developed within the integral approach AREVA Fatigue Concept (AFC) new tools and methods to live up to operators’ expectations. Based on measured thermal loads, the Fast Fatigue Evaluation (FFE) process allows for highly-automated and reliable data processing to evaluate time-dependant cumulative usage factors of mechanical components. Calculation and management of results are performed with the software FAMOSi, thus impact of operating cycles on components in terms of stress but also with regard of fatigue can be taken into account to plan an optimized decision related to the plant operation or maintenance activities.

This paper mainly describes the calculation methodology used to perform a Fast Fatigue Evaluation, but also application examples with relevant results to point out the benefits of this method to the ageing management of mechanical parts.






11.09.2013 10:20 Materials, integrity and life management

Materials, integrity and life management - 309

Mechanical properties of the steel T91 in contact with lead

Jakub Klecka, Fosca Di Gabriele, Anna Hojna

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

jakub.klecka@cvrez.cz

 

The ferritic/martensitic steel T91 is among the materials selected for components to be used in the Gen IV LFR (Lead Fast Reactor) concept. Interaction of the steel with the liquid Pb has been studied for a few years. However, issues of main concern are the corrosion of T91 in Pb and its sensitivity to Liquid Metal Embrittlement, LME. The last is a phenomena usually observed at low temperature (close to the melting point of the liquid metal), over yielding point and when there is perfect wetting of the steel from the Pb. In particular, the complete absence of an intermediate oxide layer at the interface T91/liquid metal (result of accidental conditions) promotes the LME. In literature, LME is a phenomena observed mainly for the couple of materials T91/PbBi. However, preliminary studies of the interaction of T91 with Pb, under loading conditions did not show that LME could represent a problem.

The experimental cell CALLISTO was designed and manufactured in the aim of carrying out mechanical testing of materials immersed in the liquid metal. Several tensile tests were carried out in CALLISTO, with Pb. Experimental variables considered were the surface state, temperature and strain rate. No LME was observed in most of the cases. However, when specimens were notched and the wetting was induced, tests revealed the typical features of the embrittlement induced by LM. Results are discussed in terms of effect of the environment on the mechanical properties of the steel T91 and the brittle features observed in the fracture surface.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 301

Recent Experience of ANSALDO NUCLEARE on PLEX Projects

Federico Fortunato, Francesco Benvenuto

Ansaldo Nuclear S.p.a., C.so F.M. Perrone 25, 16152 Genova, Italy

federico.fortunato@ann.ansaldo.it

 

With the energy policy currently in place, the Argentine government has decided to continue to developing the country’s nuclear program and financing the construction of a second unit at Atucha, as well as to extend the operating life of Embalse and increase its capacity. The Embalse nuclear plant will complete its normal period of operation at the end of 2013, at which point an upgrade and refurbishment program will be introduced to support twenty-five more years of safe and stable operation in terms of both security and production.

To optimize refurbishment activities and make modernization work attractive from an economic point of view, it was decided to repower the plant from the current 650 MW to about 680 MW.Against this backdrop, Ansaldo Nucleare have been awarded a contract to perform all the equipment and system modifications required to obtain the requested power rating.The turnkey project involves work on the turbine set, electric generator and thermal cycle in order to increase the global efficiency of the plant and generate more electric power with a small increase in nuclear power. The contract involves both manufacturing and system engineering work. The manufacturing work, mainly to build turbine and generator set components, will be performed by our parent company Ansaldo Energia, while systems engineering will be the responsibility of Ansaldo Nucleare.Ansaldo Nucleare have been also awarded a contract to perform the complete replacement of the Stand By Diesel stations, to upgrade the system to the new safety requirements and guarantee the capability of the system to supply all the required loads.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 302

The Simulated Damage of Zircaloy Fuel Cladding Tubes with Brittle Hydride Blisters

Zhengxiang Chen, Leon Cizelj, Mitja Uršič

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

zhengxiang.chen@ijs.si

 

The reactivity initiated accidents in the light water reactors involve unwanted increase in fission rate and reactor power. One of the consequences may be the heating of the fuel cladding material, Zircaloy. The heated Zircaloy tubes may undergo the high temperature oxidation with the steam, resulting in a release of hydrogen. The hydrogen might then diffuse in the Zircaloy tubes in areas with high thermal gradient, which leads to the development of hydride blisters in the areas of localized high hydrogen concentration. The hydrides are less dense and more brittle than Zircaloy. Therefore, the failure of the cladding may originate from the brittle hydride blisters.

The aim of this work is to develop numerical model of the progressive damage originating from the hydride blisters. The simulation results will be analyzed and discussed in comparison with the experimental results available in literature.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 304

The European Network for Inspection and Qualification (ENIQ)

Oliver Martin1, Etienne Martin2, Russ Booler3, Tony Walker4

European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands1

Electricité de France, Direction Production Ingénierie, Ceidre, 2 rue Ampere, 93206 St Denis Cedex 01, France2

AMEC Clean Energy Europe, Walton House, Birchwood Park, Risley, Warrington WA3 6GA, United Kingdom3

Rolls-Royce Submarines, Derby, Derby, United Kingdom4

oliver.martin@jrc.nl

 

The European Network for Inspection and Qualification (ENIQ) is a nuclear operator driven network that works towards a harmonized European approach on reliable and effective in-service inspection (ISI). ENIQ was founded in 1992 and nearly all nuclear operators in the European Union member countries and Switzerland are members of ENIQ.

ENIQ has a Steering Committee (SC) and three task groups: a Task Group on Qualification (TGQ), which works on issues related to the qualification of in-service inspection (ISI) systems, and a Task Group on Risk (TGR), which is focused on risk-informed ISI (RI-ISI). In 2012 the ENIQ SC decided to establish a third task group dedicated to issues facing inspection qualification bodies (IQBs). The new task group with the name Task Group for Inspection Qualification Bodies (TGIQB) should be a forum for IQBs to exchange experience and discuss common issues facing them and had its inaugural meeting in March 2013. Since its establishment ENIQ has issued the ENIQ Methodology document and a number of recommended practices, which are all recognised by nuclear regulators as guidance documents. Of particular importance for the network were the two pilot studies and the participation of member organisations in benchmark projects like the RISMET project on RI-ISI organised by OECD-NEA. Recently ENIQ joined NUGENIA, the newly established European association for R&D on Gen II & III reactors, making ENIQ the 8th technical area (TA) of NUGENIA (out of 8). NUGENIA combines the activities of four previous networks and technical working groups (TWG). Beside ENIQ these are the network for severe nuclear accidents SARNET, the network for plant life management NULIFE and the TWG on Gen II & III reactors of the Sustainable Nuclear Energy Technology Platform (SNETP). The accession of ENIQ to NUGENIA required a revision of the ENIQ roadmap to highlight more clearly the R&D and harmonisation projects planned for the future. Also a number of new projects are currently under preparation in TGQ and TGR. In 2012 a questionnaire was circulated among TGQ members for preparing a study on the mutual recognition of qualification procedures between countries. Experienced personnel from IQBs or utilities of each country provided answers and on the basis of these a study is planned to compare different stages of the qualification processes of countries. Another project planned within TGQ is a comprehensive investigation of the performance of computed and digital radiography. TGR is planning a project on quantitative modelling methods for probability of detection (PoD) data for RI-ISI applications and a project on risk reduction through ISI. The aim of this paper is to describe the development of ENIQ since its establishment and the future challenges.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 306

Cohesive zone modeling of intergranular cracking in polycrystalline aggregates

Igor Simonovski1, Leon Cizelj2

European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

igor.simonovski@ec.europa.eu

 

Understanding and controlling early damage initiation and evolution are amongst the most important issues in nuclear power plants, occurring both in austenitic steels and nickel based alloys. In this work a meso-scale approach to modeling initiation and evolution of early intergranular cracking is presented. Finite element modeling is used to explicitly model both the grains and the grain boundaries. Spatial Voronoi tessellation is used to obtain the grain topology. In addition, as measured topology of a 0.4mm stainless steel wire is used. Anisotropic elasticity and crystal plasticity is used as a constitutive law for the grains. Grain boundaries are modeled using the cohesive zone approach, implemented through the cohesive surfaces. Two cases of grain boundary damage initiation: a) initiation due to normal displacements and b) initiation due to a combination on normal and shear displacements. The evolution of the intergranular cracks is compared for the two cases and differences between the spatial Voronoi tessellation and as-measured stainless steel structure are highlighted. In addition, macroscopic responses under external load are compared for the different cases during the intergranular damage initiation and evolution.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 307

Integrated multielement probe for non-destructive evaluation of nuclear reactor pressure vessel head penetrations

Marko Budimir, Nikola Pavlović, Renato Gracin, Matija Kekelj

INETEC-Institute for Nuclear Technology, Dolenica 28, 10 250 Zagreb, Croatia

marko.budimir@inetec.hr

 

A novel type of integrated non-destructive evaluation probe comprising piezoelectric effect-based ultrasound transducers and an eddy current element has been virtually designed and optimized, constructed, assembled and then tested in laboratory conditions. The probe is intended for non-destructive testing of penetrations in nuclear reactor pressure vessel heads. During a NDE of the penetrations, the available probe testing trajectory is determined by a gap between the base material of the pressure vessel head and a penetration pipe. The gap size of 3mm – 7mm is a critical constraint for the probe dimensions and the delivery method choice. The thickness of the tested area of up to 16mm is another demanding technical specification that influences the multielement spacing, ultrasound beams central frequencies, incident angles and widths.

The probe designed in this work combines a creep wave ultrasound transducer of central frequency 4MHz, two perpendicular pairs of ultrasound TOFD configurations working at 6MHz of central frequency, a normal beam ultrasound transducer of central frequency of 2MHz and an eddy current coil working at 400kHz. The probe head housing is made of an acoustically transparent polymer and the delivery method uses an elastic ribbon-like stainless steel probe head holder with a connector to an end-effector and ultrasound and eddy current pulser-receiver systems. The probe has an autonomous supply system of water that is used as the ultrasound couplant.Prior to engineering and constructing the probe building parts and assembling them, a finite element method modelling and optimization of electromechanical properties of the probe were performed (virtual prototyping) in the PZFlex software package. The probe head assembled prototype electromechanical properties were tested both by a high precision system including a 3D positioner, a water tank with a degassing system and a hydrophone, and by using standard calibration stainless steel blocks. The probe prototype including the head and the ribbon-like holder was tested on a reactor pressure vessel mock-up in the INETEC laboratory. The NDE was successfully obtained in the range of specified testing depths and the minimal crack size detected was less than 1 mm.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 308

Accelerating Test Method of Thermal Aging for High Heat Generating Equipment in Nuclear Power Plant

Lim Byung-Ju, Chang-Dae Park, Kyung-Yul Chung

Korea Institute of Machinery and Materials, 171 Jang-Dong, Yusung-Gu, 305-343 Daejeon, South Korea

bzoo77@kimm.re.kr

 

In nuclear power plant, the electrical equipment which is qualified by procedures and rules of IEEE standard performs accident condition test after aging in many normal conditions of heat, radiation, mechanical and vibration. Because install life of the equipment is over 10 or 20 years, the accelerating aging test is applied for normal aging test. Especially in thermal aging test, Arrhenius equation, which defines relationship between activation energy of a material and chemical reaction rate, is used as a tool calculating the accelerated condition of the test. The equation enables to easily convert actual operating temperature in a long time into accelerated test condition with higher temperature of shortened period. However, temperature of components of the electric equipment often increases by heat sources such as power supply, electrical circuit and coil. This heat-rise makes difficulty for deciding the accelerated condition of thermal aging test. In addition, the number of components composing of the equipment and various temperatures of each component are an additional consideration of the difficulty. We proposed the effective method and technology of accelerated thermal aging test for heat-rise equipment in NPP and verified that the method could suggest the reasonable accelerating condition.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 310

Characterisation of coatings methods for HLM applications

Jakub Klecka1, Fosca Di Gabriele1, Carlo Cristalli2, Alessandro Gessi2, Sarka Houdkova3, Dalibor Karnik4, Alessandra Bellucci5, Francesca Nanni6

Research centre Rez, Hlavni 130, 250 68 Husinec-Řež, Czech Republic1

ENEA, Via Martiri di Monte Sole 4, 40129 Bologna, Italy2

Research and Testing Institute, Tylova 1581/46, 301 00 Plzeň, Czech Republic3

ÚJV Řež, a.s., Hlavní 130, Řež, 250 68 Husinec, Czech Republic4

CSM S.p.a., Piazzale s. Benigno 16149, Genova Meici, Italy5

Universita degli Studi di Roma Tor Vergata, Via del Politecnico, 1, 00133 Roma, Italy6

jakub.klecka@cvrez.cz

 

The use of heavy liquid metals (HLM), such as Lead, Pb, is foreseen as a coolant in Generation IV fast Reactors. However, structural materials suffer significant damage when in contact with the HLM, in certain environmental conditions. In the austenitic steels the prevailing corrosion mechanism is dissolution of the alloying elements in the liquid metal The ferritic-martensitic steels are instead more sensitive to phenomena such as liquid metal embrittlement and the formation of Fe3O4 scales that, being very rough and porous, hinder the thermal exchange and show very low stability under mechanical and thermal loads. Then, the recently developing strategy is to interpose a layer of a different chemical composition between the base material and the liquid metal coolant in order to prevent the core metal from corrosive effects. Coatings are proposed as a valid protection against high temperature damage in this environment. Their capability to grow more stable and protective oxides, by introducing the oxide forming elements in higher amount, is proven to be an effective alternative to material engineering. There are different approaches being investigated; surface coatings are in fact meant to assure the maximum protection but they should also fit the requirements in terms of stability during exercise, that is to say limited diffusion towards the lattice below, low composition variation with lack of brittle phases, good adhesion to the bulk and compliance with the underneath material under the different kinds of load ( creep, fatigue, shock ).

Several deposition techniques and compositions have been proposed and tested, some of them reported in this work. High Velocity Oxygen Fuel, HVOF, combined with laser remelting was selected for deposition of FeCrAlY coatings. The combination of the two technologies lead to a compact and adherent coating with an enriched content of Al, enabling the formation of a protective Al2O3 oxide scale. The method was evaluated in terms of the corrosion resistance of the coating and also its effect on the microstructure of the substrate alloy.Another technique is the Physical Vapour Deposition (PVD), for coating compositions such as TiN, FeAl, FeCrAl. The first is an inert material that is meant not to react with Lead; the second and the third are the so-called Al2O3-formers, whose aim is to create a stable Al2O3 layer, once in contact with Lead, even at low Oxygen contents. The results are presented in terms of microstructure examinations.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 311

ARCHER – Advanced system for RPVH inspection and repair

Tomislav Tomašić, Igor Vuković, Ante Bakić

INETEC-Institute for Nuclear Technology, Dolenica 28, 10 250 Zagreb, Croatia

igor.vukovic@inetec.hr

 

The reactor pressure vessel head (RPVH) is an integral part of the reactor coolant pressure boundary. Its integrity is important for the safe and reliable operation of the nuclear power plant (NPP). After detection of the leakage and cracks in French NPP, followed by another that occurred in NPP in USA, methods and frequency of inspection were defined, and are strictly regulated by the US NRC Order EA-03-009 (substituted lately by ASME Code Case N-791-1) since 2003.

Usual scope of inspection from inner side of RPVH comprises of visual inspection of the surface, ultrasonic testing (UT) and eddy current testing (ET) of the penetration nozzle and ET of the J-groove weld and nozzle outside surface below the weld.ARCHER, new INETEC’s manipulator, is designed to provide full scope inspection of the RPVH, by use of various test modules and by performing the surface repair action on J-groove weld. It is adjustable to work with different types of penetration nozzles and thermal sleeves on both VVER and PWR type of NPP. Due to complex geometry each module is specially designed for particular type of examination. Modules are exchanged through the docking system without need for personnel to enter under the head region, thus reducing the personnel’s exposure to the ionizing radiation.The end effectors are used to determine the surface flaws or cracks on inner diameter surface of penetration nozzle gap. It guides a slim sword-like probe which carries a pair of TOFD transducers for detection and sizing of circumferential and axial cracks, an eddy current cross-wounded coil, and a zero-degree UT probe through a gap between the penetration nozzle and thermal sleeve. In case of a non-sleeved penetration nozzle, an open housing module is used.J-groove module is designed to fit geometry of the J-groove weld of penetration nozzle, vent pipe and funnel guide. The whole weld area (2” mm on shell side and 1” on nozzle side) is covered by two specially designed array eddy current probes.Surface flaws, discovered by eddy current examination of J-groove weld, define the scope of the automated surface repair module (ASR module) performed by the grinding method. Specially developed grinding procedure, based on the surface probing and UT results, ensures the treatment doesn’t affect the originally designed structural integrity basis. When compared with the other repairing methods, ASR module significantly reduces the inspection time and radiation exposure of the personnel, and does not introduce residual stress into the structural material.The paper describes the system’s capabilities and features, and its advantages compared to other systems for performing the RPVH inspection and repair activities on PWR and VVER reactors.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 312

FORERUNNER - Efficient and Smart Solution for SG Inspection

Petar Mateljak, Domagoj Liebl

INETEC-Institute for Nuclear Technology, Dolenica 28, 10 250 Zagreb, Croatia

petar.mateljak@inetec.hr

 

Steam generator (SG) is a critical component in the nuclear power plants (NPP) with the largest surface area in the primary reactor coolant system, and its integrity is essential for avoiding possible radioactivity release to the environment. SG tube walls are susceptible to aging, i.e., various degradation mechanisms take place in its structural material, such as volumetric material loss due to fretting wear, stress corrosion cracking (SCC), pitting corrosion, flaw accelerated corrosion, intergranular attack (IGA) etc.

New more strict regulatory requirements request plant management to assure the safety of the public and the environment, as well as better SG life management strategies. Therefore, those requirements forced specialized inspection companies to develop advanced probe technologies, more reliable instruments and robotics, and improve training and knowledge of personnel involved in inspection process.Thanks to intensive evolution of electronics and computers in the last decade, inspection systems have evolved to a much higher level of automation, efficiency and reliability. Tools based on the eddy current examination techniques were subject to continuous development - from a simple detection tools to advanced diagnostic tools that provide input for decision making based on the integrity assessment.FORERUNNER is a part of the INETEC's inspection system for PWR plants, primarily used for quick and accurate positioning of the tube guide on the SG tube sheet, and efficient performance SG tube walls inspection. It is a light mobile robot, adjustible for different tube sheet configurations and inner tube diameters. Integrated electronics based on the EtherCAT technology increases the speed of operation and simplifies the cable managment. Using the strongest grippers currently available at the market, the FORERUNNER is a realible and robust system, highly automated with a machine vision, and built-in smart algorithms for optimal movement.FORERUNNER is controlled by PC-based software, which is synchronized with INETEC EddyOne software package. The complete scope of inspection activities, the planning, examination, data analysis and final report, became a highly automated process, which makes the inspection much easier and more reliable.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 313

Soft missile impact into reinforced concrete wall: comparing simulations to experiments

Matej Bogataj, Samir El Shawish, Leon Cizelj

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

samir.elshawish@ijs.si

 

Containment building in a nuclear power plant protects environment from radioactive releases and also reactor vessel from external influences, such as airplane crash or other missile impact. After September 11th an airplane crash into containment building became even more serious safety question. A lot of work has been done to model such accidents and to predict the response of the containment. Because high-speed impact is highly non-linear event, poor results can be expected in numerical simulations. To better understand the problem, the IRIS 2010 benchmark was organized, where different groups performed numerical simulations of three different experiments. These experiments were small-scale crash tests of pipe-shaped steel missile into reinforced concrete wall. A large scatter of simulation results was observed among the groups due to insufficient material properties data, large set of input parameters and difficulties when quantifying damage. In this study we conduct finite element simulations of one of the experiments in Abaqus and additionally carry out a sensitivity study to define the dominant model parameters and to assess the sensitivity range. In particular, we concentrate on the concrete constitutive model parameters which are very important for prediction of containment building damage.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 314

Study of ions implanted RAFM steels with application of positron annihilation spectroscopy

Stanislav Sojak, Vladimír Slugeň, Martin Petriska, Jana Šimeg Veterníková, Matúš Stacho, Veronika Sabelová

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

stanislav.sojak@stuba.sk

 

Current nuclear power plants (NPP) require radiation, heat and mechanical resistance of their structural materials with ability to stay operational during NPP planned lifetime. Radiation damage much higher, than in current NPP, is expected in new generations of nuclear power plants, such as Generation IV and fusion reactors.

Investigation of perspective structural materials for new generations of nuclear power plants is among others focused on study of reduced activation ferritic/martensitic (RAFM). These steels have good characteristics as reduced activation, good resistance to volume swelling, good radiation, and heat resistance. Our experiments were focused on microstructural changes study of binary Fe-Cr alloys after annealing and irradiation, experimentally simulated by ions implantation. Alloys with 11.62% Cr were examined after helium ions implantation at different doses (0.1; 0.3; 0.5 C/cm2). Thermal annealing at temperatures of 400, 475, 525 and 600 °C was performed after implantation with aim to study change of the defects size/amount. Pulsed Low Energy Positron System (PLEPS) at FRM II reactor in Garching (Munich) was applied for lifetime studies. In case of annealing temperature of 600 °C the positron lifetime parameters decreased and assumptions about the defects size decrease were made and the microstructure recovery was present.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 315

Review on the development of the double-stabilized austenitic stainless steels; mechanical behaviour and metallurgical properties

Carlo Cristalli, Luciano Pilloni, Gianni Filacchioni, Alberto Calza Bini

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

carlo.cristalli@enea.it

 

In the framework of the research and development over structural materials for Fast Reactors the ideal steel to be used in the construction of fuel claddings should be the one with the high temperature properties and corrosion resistance of the austenitic steels and the stability under irradiation which is instead typical of the ferritic-martensitic steels. At the beginning of the ‘80s, within an experimental program carried out at the Saclay Center, the under electrons irradiations (1 MeV HVEM) have shown the effectiveness of the simultaneous presence of Ti and Nb on the swelling resistance of 316 and 15 Cr-15 Ni matrix. This experimental evidence lead CEA and ENEA to start the production and the characterization of the first double stabilized steels (DS). This first generation was widely characterized showing the strong influence of some additional elements on the structural stability of an austenitic matrix. The mechanical properties were largely affected by the precipitation behavior and, in some cases, by the onset of highly undesirable phases. A very critical analysis of the results of the mechanical tests and the structural evolution lead to an optimization of the chemical composition. A 2nd generation has then been realized based on 15 Cr-15 Ni and 15 Cr-25 Ni matrix. The whole tensile and creep program was achieved with test temperatures up to 850°C. The results are extremely promising, particularly the ones of 15 Cr-25 Ni (Ti + Nb).






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 316

Thermal Stresses in Pipes Caused by Randomly Generated Two-dimensional Temperature Fluctuations

Oriol Costa Garrido, Samir El Shawish, Leon Cizelj

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

oriol.costa@ijs.si

 

Thermal fatigue assessment of pipes due to turbulent fluid mixing is a difficult task because of the existing uncertainties and variability of the induced thermal stresses. The thermal stresses arise on three-dimensional pipe structures driven by the random thermal loads generated by this type of fluid phenomenon at the fluid-wall interface. A new approach has been developed for the generation of two-dimensional fields of time-dependent thermal loads. The fields recreate the thermal boundary conditions acting on the inner surface of pipes in turbulent fluid mixing circumstances. The approach uses fluid temperature statistics, such as mean and variance, from experimental or fluid dynamics simulation’s results. Moreover, the random nature of the simulated temperature fields is of great advantage for the study of thermal stresses induced on piping structures. In the paper, the presented approach is used to reproduce fluid temperature fields of a case study from literature. The simulated temperature fields are employed as thermal boundary conditions in heat transfer analyses of a pipe wall. At the end, thermal stresses in the pipe wall are evaluated by employing the resolved pipe wall temperatures in uncoupled mechanical analyses.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 317

M?ssbauer Phase Composition Analysis of the Corrosion Products

Julius Dekan, Vladimír Slugeň

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

julius.dekan@stuba.sk

 

The properties and composition variability of the stainless CrNi and mild steels corrosion products is of such range that, in practice, it is impossible to determine the properties of the corrosion products from the theoretical data only. Since the decontamination processes for the materials of the water-cooled reactor (VVER-440) secondary circuits are in the progress of development, it is necessary to draw the required information by the measurement and analysis of the real specimens.

Phase composition of iron metal corrosion can be reliably determined by applying 57Fe Mössbauer spectroscopy, which is one of the most suitable methods for phase analyses of iron-bearing components. This method is sensitive to the hyperfine interactions and these provide valuable information on the magnetic and electronic states of the iron species. These interactions, so-called Mössbauer parameters, often enable detailed insight into the structural and magnetic environment of the Mössbauer isotope (57Fe in this case). Mössbauer parameters are different and unique for the each iron-bearing compound; therefore, various corrosion products from provided samples can be distinguished.Mössbauer spectroscopy studies of phase composition of corrosion products were performed on real specimens gathered from different NPPs in Slovak and Czech Republic. Specimens were (i) scrapped from water pipelines or (ii) in form of filters deposits. Recent results in our long-term corrosion study confirm good operational experiences and suitable chemical regimes (reduction environment) which results mostly in creation of magnetite (on the level 70% or higher) and small portions of hematite, goethite or hydrooxides. Regular observation of corrosion/erosion processes is essential for keeping NPP operation on high safety level. The output from performed material analyses can influence the optimisation of operating chemical regimes as well as decontamination and passivation of pipelines, or secondary circuit components.






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 320

Experimental Assessment Techniques for CANDU Pressure Tubes Degradation Mechanisms

Bogdan Negulici, Anca Gheorghe

University “Politehnica” of Bucharest, 313 Splaiul Independentei Street, Sector 6, 060042 Bucharest, Romania

negulicibogdan@yahoo.com

 

CANDU which stands for CANada Deuterium Uranium is a nuclear reactor designed by the Candu Energy Inc. the former AECL (Atomic Energy of Canada Limited). It uses natural uranium as fuel and heavy water as both cooling agent and moderator. This type of nuclear reactor is a special one because it doesn`t have a pressure vessel(like the majority of reactors)instead it has 380 fuel channels also known as CANDU Pressure Tubes (PT`s). The PT`s are critical components of this type of nuclear reactor and we have a special inspection program for our degradation mechanisms called fitness-for-service inspections. There are a series of possible degradation mechanisms:delayed hydride cracking (DHC), irradiation enhanced deformation (creep), corrosion, deuterium ingress and changes in material properties: reduction in ductility and fracture toughness. The inspections are visual, dimensional, ultrasonic, radiographic in order to discover this mechanisms. Pressure tube deformation has been managed such that it alone could not cause a safety or structural integrity concern for Cernavoda Nuclear Power Plant(NPP) fuel channels. Assessment methodology and prediction capacity enables Cernavoda NPP safe operation with certain conservatism. This paper presents the experimental assesment techniques for CANDU Pressure Tubes degradation mechanisms for the Cernavoda Nuclear Power Plant.

References[1] Canadian Standards Association(CSA) N285.4-09(2009)[2] IAEA, Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes IAEA, VIENNA, 1998 IAEA-TECDOC-1037






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 321

The Delayed Hydride Cracking (DHC) and the irradiation enhanced deformation (creep) of CANDU Pressure Tubes Zr-2.5%Nb from Cernavoda Nuclear Power Plant

Anca Gheorghe, Bogdan Negulici

University “Politehnica” of Bucharest, 313 Splaiul Independentei Street, Sector 6, 060042 Bucharest, Romania

gheorghe.anca86@yahoo.com

 

The Pressure Tubes are critical components of a CANDU (CANada Deuterium Uranium) Nuclear Power Plant. These components are part of the primary pressure boundary and are subjected to several degradation mechanisms which must be prevented and/or controlled otherwise they could lead to radioactive emissions. For preventing and/or controlling these degradation mechanisms we have several type of methods and inspections: specific and random inspections, inaugural and periodic inspections. The areas that need focus are: rolled joints, feeders conections and end fittings . The inspections are visual, dimensional, ultrasonic, radiographic in order to discover the possible degradation mechanisms: corrosion, erosion, wear, leakage, defects, fractures. The most important degradation mechanisms are: irradiation enhanced deformation (creep) and the delayed hydride cracking (DHC). The limited knowledge regarding the causes of the degradation may lead to susceptible areas that are not inspected. The scope and frequency of these inspections are determined based on the results of a fitness for service assessment and taking into account the relative susceptibility of the Pressure Tubes to each specific degradation mechanism. This paper presents the methods used at the Cernavoda Nuclear Power Plant for detecting, preventing and controlling irradiation enhanced deformation (creep) and the delayed hydride cracking (DHC), which affects the global deformation and/or the rupture of the CANDU Pressure Tubes.

References[1] Canadian Standards Association(CSA) N285.4-09(2009).[2] IAEA, Assessment and management of ageing of major nuclear power plant components important to safety:CANDU pressure tubes IAEA, VIENNA, 1998 IAEA-TECDOC-1037






11.09.2013 10:40 Poster session 2

Materials, integrity and life management - 322

Elasto-plastic analyses of selected standard fracture mechanics specimens

Raphaël Connes1, Leon Cizelj2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

raphael.connes@ijs.si

 

This paper describes the elasto-plastic numerical simulations of three different specimens commonly used for the determination of the fracture toughness: CT (Compact Tension), SENT (Single Edge Notch Tension) and SENB (Single Edge Notch Bending). For each specimen, two different crack lengths are presented. The J-integral has been selected as the parameter representing the crack driving force in all specimens. Standard experimental procedures have been simulated for the material properties used in the simulations are representative for Inconel 52 at 300°C. A parametric study has been attempted to investigate the influence of mesh size on the global response of the specimen assuming both plane strain and plane stress. Results are compared with analytical and 3D numerical solutions. Good agreement between results of different simulation models indicates robustness of the finite element solver and meshing software (ABAQUS/CAE) used in the analysis.






11.09.2013 10:40 Poster session 2

Probabilistic safety assessment - 501

Barriers and Operational Risk Assessment of Incidents and Accidents occurring in the Transport of Radioactive Materials

Thomas Breznik1, Borut Smodiš2, Marko Gerbec3

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia3

thomas.breznik@ijs.si

 

Keywords: Risk analysis, RAM release, loss of containment, safety barrier, organisational factors.

Investigations of major incidents and accidents in the transport of radioactive materials (TRAM) show that technical, human, operational, as well as organisational factors influence the incident rate sequences in everyday TRAM performance. In spite of these facts, quantitative or semi-quantitative risk analyses of human error beside technical factor as main cause of all emergency events in the transport chain should be performed. Therefore, the main focus should be on generation, performance and evaluation of human organisational systems such as generic safety barriers, failure modes and their control measures. The paper presents a method called Barrier and Operational Risk Analysis (BORA-TRAM) in the qualitative and semi-quantitative risk analysis in the general performance of TRAM. Application of BORA-TRAM for analysis of the loss of containment barrier evidently presents a more detailed risk picture than the traditional quantitative risk analyses (QRA), since no analyses of causal factors of RAM release are carried out in the existing QRA.By using BORA-TRAM it is possible to analyse the effect of safety barriers introduced to prevent loss of containment and control (radiation releases and radiation exposition). Furthermore, it reveals how platform specific conditions of technical, human, operational, and organisational risk influencing factors affect the barrier performance in final risk evaluation of TRAM. BORA-TRAM comprises the following main steps: 1) System Identification (development of a basic risk model including release scenarios; this is often done by task analysis);2) Modelling the performance of safety barriers (literature review). This is the key point that can incorporate human organizational and operational risk influencing factors (RIFs) into the barriers and then into the initiating events;3) Assignment of generic data and risk quantification based on these data; establishment of a fault tree. This is used for analysis of barrier performance. All of basic events in TRAM should be analyzed;4) Development of risk influence diagrams. This should cover representative scenarios, and it usually consists of initiating event, barriers and outcomes. This barrier block diagram can also be converted into an event tree;5) Scoring of RIFs;6) Weighting of RIFs;7) Adjustment of generic input data;8) Recalculation of the risk in order to determine the platform specific risk related to radioactive material-RAM package loss or release. The various steps in BORA-TRAM will be presented and discussed. Final results from a case study where BORA-TRAM is applied and revision of the proposed method in risk assessment in TRAM will also be presented.






11.09.2013 10:40 Poster session 2

Probabilistic safety assessment - 502

Independent off-site water storage connection to nuclear power plant

Blaže Gjorgiev1, Andrija Volkanovski1, Duško Kančev1, Marko Čepin2, Ljubo Fabjan1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Fakulteta za elektrotehniko, Tržaška cesta 25, 1000 Ljubljana, Slovenia2

blaze.gjorgiev@ijs.si

 

Reliable off-site power supplies are of special importance for the nuclear power plant safety where the active safety systems are necessary to bring the plant in safe shutdown state. Sufficient amount of water for cooling purposes is required. In the conventional generation II nuclear power plants with pressurized water reactors and active safety system, the water supplies for the primary loop is established by the refueling water storage tanks.

In this paper, a connection between off-site water reservoir and a nuclear power plant is considered. The reservoir may be a part of on-river accumulation, i.e. hydro power plant reservoir or any other artificial or natural accumulation. For the established connection a pipe under pressure is considered. The gravity is the main force that drives the water from the reservoir towards nuclear power plant target systems. In such conditions this connection is to be considered as an off-site passive safety system. The established connection can be used in at least two scenarios. One is the loss of all offsite and onsite alternating current power supply, which is known as station blackout event. The other is the large loss of coolant accident. The objective of this paper is to analyze the changes in core damage frequency due to consideration of the connection between the water reservoir and the nuclear power plant. The probabilistic safety assessment of the nuclear power plant is considered as a conservative approach taking in account only the on-site safety systems. The presented approach in this paper extends the probabilistic safety assessment from its original use in order to consider other, off-site non-safety systems, in addition. The fault tree/event tree model of the selected nuclear power plant is modified in this sense. The reservoir and the connection with the nuclear power plant is modeled using the fault three approach with a top event defined as a possible failure to deliver water to the nuclear power plant site. The obtained results show decrease of core damage frequency, which indicates improvement of safety if the water reservoir connection is introduced as off-site passive safety system.






11.09.2013 10:40 Poster session 2

Probabilistic safety assessment - 503

Component unavailability uncertainty and the safety systems unavailability

Andrija Volkanovski, Blaže Gjorgiev, Duško Kančev

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

andrija.volkanovski@ijs.si

 

Fault tree analysis is deductive approach used in the Probabilistic safety analysis (PSA) for assessment of system unavailability. The basic events are the ultimate parts of the fault tree, representing the failures of components or undesired events. The increased and extended application of the PSA requires appropriate consideration of uncertainties in analyses and interpretation of the results. Inadequate treatment of uncertainties may lead to poorly supported or even wrong conclusions whose final consequence is a loss of adequate level of safety.

Epistemic uncertainty results from the imperfect knowledge or incomplete information regarding values of parameters of the underlying model. It is also called state-of-knowledge uncertainty. Epistemic uncertainty is considered in the models by probability distributions associated with uncertain parameters. Probability distributions associated with uncertain parameters represent the state of knowledge about the right values of the parameters and are therefore very often derived from expert judgment.This paper presents the results of the analysis of the introduction of probability distributions associated with component unavailability parameters, on the overall unavailability of the analyzed system. The normal and lognormal distributions are introduced as probability distributions associated with component unavailability. The auxiliary feedwater system of nuclear power plant is selected as test system for assessment of the uncertainty propagation. Six case scenarios are developed for assessment of the implications on introduction of different probability distributions for different number and sets of components. Obtained results show that the probability density function of the top event depends on type and parameters of uncertainty distributions as well as importance of the basic events with considered uncertainty. Introduction of lognormal distribution for uncertainty characterization of basic events can result in heavy tails of probability density function and increased likelihood of having top event probability larger than the mean value. The current approaches for consideration of the uncertainties in risk-informed decision making are discussed and the need for their appropriate consideration in risk-acceptance guidelines is emphasized.






11.09.2013 10:40 Poster session 2

Probabilistic safety assessment - 504

Sensitivity study of a new model for assessing time-dependent risk in ageing NPP

Duško Kančev1,2, Blaže Gjorgiev2, Andrija Volkanovski2

European Commission, DG JRC, Institute for Energy and Transport, P.O. Box 2, NL-1755 ZG Petten, Netherlands1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

dusko.kancev@ec.europa.eu

 

The nuclear power plants risk reduction by application of the probabilistic safety assessment is one of the main focuses of the nuclear safety today. The goal is the reduction of the unavailability of the nuclear power plants safety systems. On the other hand, the world NPP fleet is ageing fast. Equipment ageing has gradually become a major concern in the nuclear industry since the number of safety systems components, that are approaching their wear-out stage, is rising fast.

A previously developed model for assessing time-dependent unavailability of ageing safety equipment is briefly presented and discussed herein. One of the essential features of the model is that it simultaneously considers the effects of performing surveillance testing as well as preventive maintenance, corrective maintenance and overhaul. Also, the ageing-implicated adverse effects on component availability are being explicitly considered as an integral part of this time-dependent unavailability model. Subsequently, the model can be coupled to commercial software for system- and plant-level modelling.This paper is aimed towards performing sensitivity analysis of the developed model. A component level resolution is selected as the basis for performing the analysis. Investigation of the influence of different component-relevant input parameters on the calculated equipment unavailability is the goal of the analysis. The influences of the periodic testing interval as well as the preventive and corrective maintenance intervals on the calculated component unavailability are of particular interest. The results are presented and discussed.






11.09.2013 10:40 Poster session 2

Radiation and environment protection - 802

Transfer of Ra-226 to Chinese cabbage (Brassica pekinensis Rupr.) from soil contaminated with U-mill tailings

Marko Černe, Borut Smodiš, Radojko Jaćimović

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

marko.cerne@ijs.si

 

Environmental contamination due to radionuclide releases is a common phenomenon in uranium mining and milling areas. Disposal of uranium ore processing materials may result in soil contamination if inappropriate waste remediation is applied. Radionuclides can be retained in soil by soil colloids and subsequently taken up by plants. The uptake process depends mainly on concentration of the radionuclides in soil, soil properties and physicochemical conditions and microbial activity of soil. Some plants from the Brassicaceae family are known to have higher metal-accumulation capacity and are therefore suitable for phytoremediation of the U-waste-contaminated soil. In the present study, a transfer of Ra-226 to crops was evaluated from soil contaminated with U-mill tailings. Radium plant-to soil concentration ratios(CR)for the Chinese cabbage (Brassica pekinensis Rupr.) were calculated. A pot experiment was carried out in greenhouse conditions in order to provide the growing conditions controlled as far as possible. The field experiments are less appropriate due to higher CR variability in natural conditions. Different levels of soil contamination were applied under various growing conditions to simulate differnet contamination scenarios. Ra-226 was measured in the leaves of a tested plants to calculate CR values for various levels of soil contamination. The macro-elements in plants were also considered due to their role in the uptake process. Pedological parameters were taken into account as they have an impact on bioavailability of Ra-226 in the soil. Preliminary results of 106 ± 32 and 426 ± 172 Bq kg-1 dry mass for 226Ra in cabbage leaves for lower and higher content of U-waste in the soil, respectively, indicated increased accumulation of radium in more contaminated soil. In the presentation, CR values for different levels of contamination under various growing conditions are shown and discussed. Measurement results are presented and the radium transfer from soil to plants is critically evaluated.






11.09.2013 10:40 Poster session 2

Radiation and environment protection - 805

Spectral dose rate calculation using whole spectrum processing approach within energy range up to 10 MeV

Matúš Stacho, Róbert Hinca, Stanislav Sojak, Vladimír Slugeň

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

matus.stacho@stuba.sk

 

Gamma ray spectra are usually used for source identification and activity calculation, but could be also used for spectral dose rate calculation. Dose rate spectra calculations in our work are based on whole spectrum processing (WSP) approach, which allows us to calculate initiating gamma ray spectra. A principal step for the WSP application is building up the suitable response operator. Problems are put in an appearance when suitable standard calibration sources are unavailable. It may occur in the case of large volume samples and/or in the analysis of high energy range. Combined experimental and mathematical calibration may be a suitable solution. This paper is focused on building up appropriate response operator for 2” x 2” NaI(Tl) detector with energy range up to 10 MeV using MCNP code for response calculation.






11.09.2013 10:40 Poster session 2

Radiation and environment protection - 806

Impact of construction of the Hydro Plant Brezice on the NPP Krsko

Tea Bilić-Zabric1, Igor Zabric2, Aljaž Škerlavaj3, Milko Janez Križman4

INKO svetovanje, d.o.o., Kolezijska 5a, 1000 Ljubljana, Slovenia1

Elektroinštitut "Milan Vidmar", Hajdrihova 2, P.P. 285, 1000 Ljubljana, Slovenia2

Turboinstitut d.d., Rovsnikova 7, 1210 Ljubljana, Slovenia3

Independent qualified expert for RP, Ljubljana, Ljubljana, Slovenia4

inko@siol.net

 

An employment of new industrial objects in the vicinity of the NPP generally requires a detailed analysis of the potential impacts of the new facility on the safe operation of the plant and on potential changes of the original plant design parameters. Slovenia has an ambiguous plan of the construction of several hydro power plants on the Sava river. All new facilities planned for construction, including new hydro power plants, have to adapt to the requirements and conditions of the existing facilities in operation on the Sava river.

The employment of a new power facility in the spatial area of the NPP requires the assessment of the different aspects of its influence: as a first on a plant safety, its environmental impact, and economic, and social aspects.This article provides with main aspect of impact of the new hydro plant at Brezice on the safe operation of the NPP Krsko and indirectly also on the changed environmental impact of the NPP Krsko.Impact of the new hydro power plant is in short provided through the description of the impact on (i) the ultimate heat sink (impact on volume and temperature of the UHS); (ii) necessary modifications of the NPP Krsko systems due to placing of the new facility in the vicinity of plant; (iii) the environmental impact (impact on groundwater levels and watercourses, temperature, biota) and (iv) change impact of radioactive liquid discharges and related radioactivity monitoring).






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1002

Medium Voltage Cable Measurements

Klemen Grozina, Franjo Boh

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

franjo.boh@elmont-kk.si

 

Elmont d.o.o. Krško – We have spread our main scope of services from electrical maintenance, modifications implementations and quality control to cable testing area. The main reason for expanding our scope was to support NEK’s Cable Management program and world trend of LTE (life time extension) in nuclear power plants.

Scope of work – We are identifying potential downgraded conditions for safety related and operational important cables in special areas (heat, water, radiation) of power plant. Path from identification to repair: VISUAL CONTROL, TESTING, ANALYSIS, ACTION (spare, re-qualification, rerouting, repair, replace). We are using tan delta method for testing of medium voltage cables which is also preferred method used in other plants worldwide.Tan delta test – For the purpose of the evaluating the insulation of the medium voltage cables we are using the diagnostic test called tan delta (TD) at Very Low Frequency (VLF). The test is performed at 0.1Hz at 2U0 rated voltage according to IEEE standard – IEEE 400.2 Guide for Field Testing of Shielded Power Cable Systems Using Very Low Frequency. This is a non-destructive method and is mainly used for detection of “Water Trees”. Benefits of using the VLF method are the possibility to test cables in range of km with a small-portable unit, because it takes 500 times less energy to charge cable with 0,1Hz than to charge it with 50Hz. Withstand test – After the successful performance of diagnostic test we perform the withstand test. With this method we test if the cable will fail during operation. For the purpose of the test we apply 3 U0 nominal voltage to the cable for 60 minutes (according to standard). The result is the resistance of the cable insulation. The cable can either pass or fail. In case of a failed cable the week spot has to be localized and repaired.Results – The main key is to have a fingerprint of new installed cable and use that for future reference. If you are doing the first measurement on an old cable you can also get a very good overview of the cable insulation by comparing the three phases. If the cable is damaged and in a very wet area (cable duct filled with water, underground, …) the measured error value for that cable is going to be extremely high and you have to take additional steps to localize the fault.Future Plans – We are continuously involved in Cable Aging Management in NEK. We are also spreading our radius of services to other Nuclear Power Plants in EU and USA. We are gaining experience and spreading our results which are very important for future analysis. Our idea is to go with stream of new technologies in field of cable testing. We are currently involved in Partial Discharge method of medium cable testing for which we are forming a team of specialists.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1003

Combined Heat and Power Production in NPP Krško

Robert Bergant1, Tomaž Ploj1, Luka Štrubelj1, Gregor Androjna1, Stanko Manojlović1, Peter Tomažin2

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia1

SIPRO Inženiring d.o.o., Cesta krških žrtev 135c, 8270 Krško, Slovenia2

robert.bergant@gen-energija.si

 

The district energy systems, especially combined heat and power production (CHP) technologies, play an increasingly important role in local energy supply. On 25 October 2012, the EU adopted the Directive 2012/27/EU on energy efficiency, which establishes a common framework of measures for the promotion of energy efficiency within the Union in order to ensure the achievement of the Union’s 2020 20 % headline target on energy efficiency and to pave the way for further energy efficiency improvements beyond that date.

The paper discusses the possibilities of using technically, economically and environmentally justified heat from the NPP Krško for the purpose of the district heating of Krško and Brežice area. Different steam extractions on secondary site of NPP Krško were analysed. The feasibility study showed that steam extraction just behind the high pressure turbine is the most optimal heat source from the technical, spatial and economic aspects. In addition, the possibility of using low temperature heat on tertiary side with 2-stages heat pumps was analysed.The feasibility study also analysed alternatives in the Krško - Brezice region, such as heat production from biomass, biogas, natural gas plant (existing and new), geothermal energy, alternative without action, etc.Among all analysed heat sources, the extraction on the secondary side of NPP Krško showed the most promising results. The solution would be cheaper for the majority of the end users in Krško and Brežice. A realisation of the CHP production from NPP Krško would achieve not only economic, but also great environmental benefits. The analyses have shown that more than a half of the produced heat for the heating purposes in that area is based on fossil fuels, which produces 20.000 tonnes of CO2 emissions per year. The rest of the heat is mainly produced from wood, which does not contribute to the CO2 emissions but it can be problematic from the “particulates larger than 10 µm” (PM10) emission point of view.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1004

Krško NPP approach to integrated Quality Assurance Program

Igor Fifnja, Romeo Bišćan

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

igor.fifnja@nek.si

 

Since the beginning of Krško NPP construction, the overall Quality Assurance program and its applicable procedures were in place to assure that all planned and systematic actions necessary to provide adequate confidence that an item or service will satisfy given requirements to quality, are in place. The overall requirements for quality as one of the major objectives for Krško NPP operation are also set forth in the Updated Safety Analyses Report, Chapter 17.2, the document that serves as a base for operating license.

The Krško NPP Quality Assurance plan (QD-1, Rev. 6) incorporates various changes and improvements resulting from regulatory requirements (JV-5), international standards (IAEA GS-R-3, ISO 14001, ISO 17025, BS OHSAS 18001) and revised international guidelines (WANO, INPO…).The integrated approach to Krško NPP Quality Assurance Program was established in the following way: Krško NPP has established an integrated quality assurance program, combining all requirements from various references in one single plan (QD-1); all Quality Assurance activities are performed in the manner to provide holistic and thorough evaluation of plant activities, all Quality Assurance activities are conducted and coordinated through a single point (QA Superintendent), Quality Assurance department reports on all activities and issues through QNOD Director…Krško NPP Quality Assurance Plan defines the expectations for the implementation of following activities:- Internal audits are performed periodically in two-year cycles in accordance with international practices. Audits cover various plant processes and areas (operations, maintenance, engineering…). - Supplier audits are performed periodically in three-year cycles in accordance with international practices. Local and mostly EU-based suppliers are being audited directly by Krško NPP. Suppliers from US are audited in cooperation with NUPIC organization.- Preparation and implementation of modifications are being verified and approved through continuous QA involvement. Each SR or AQ classified modification involves a QA engineer as a team member. - New or revised plant procedures classified as SR or QR are being verified and approved within QA.- Documentation for purchasing SR or AQ items or services (technical specifications, purchase orders, contracts…) are being verified and approved within QA,- QA engineers are regularly involved in observations at the technological part of the plant (surveillance, maintenance…),- QA engineers are involved in the review and approval of outage documentation (pre-outage packages, entrance meetings, outage reports, NCRs …),- QA engineers are involved in the oversight of production and testing of modified or new plant components,- … Based on the Krško NPP experience, the QA involvement in plant processes has fulfilled its important role and expectations in achieving overall quality goals. Krško NPP will continue to perform internal quality assurance processes and activities in the future. The most important objective of the entire organization – to ensure the safe and efficient power plant operation, will continue to be the most important goal of the QA program.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1005

Design Verification of Solenoid Operated Valve for Nuclear Applications

Changdae Park, Lim Byung-Ju, Kyung-Yul Chung

Korea Institute of Machinery and Materials, 171 Jang-Dong, Yusung-Gu, 305-343 Daejeon, South Korea

parkcdae@kimm.re.kr

 

Solenoid operated valve (SOV) is widely used in many applications due to its fast dynamic response, cost effective, and less contamination sensitive characteristics. All of instrumental SOV used in nuclear power plant (NPP) in Korea are imported from foreign companies and not localized mainly because of lack of design technology and reliability. In this paper, we have established detailed procedure of designing SOV for special applications such as NPP where are in most harsh environmental conditions. The design process suggested have been verified with theoretical relations and experimental results with some prototypes of the SOV, which include physical dimension of solenoid coil and electromagnetic properties such as coil resistance and attraction force of the solenoid actuator. Good agreement of designed results with the verified parameters gives the reliability of our design methodology. We believe these design and verification are useful for the localization of the SOV for NPP especially for greatly shortening of development time and useful for cost-down for a new model of the SOV.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1006

Computational Analysis for Development of Solenoid Operated Valve Using in Nuclear Power Plant

Kyung-Yul Chung, Lim Byung-Ju, Chang-Dae Park

Korea Institute of Machinery and Materials, 171 Jang-Dong, Yusung-Gu, 305-343 Daejeon, South Korea

bzoo77@kimm.re.kr

 

In Korea, all of the instrumental solenoid operated valves (SOVs) used in nuclear power plant (NPP) have been not localized and imported from foreign companies. The SOV is an important equipment to supply the air into the control equipment as air operated valve. When safety related equipment such as the SOV in NPP are in trouble, advance preparations for rapid maintenance and replacement are required. Equipment with dependence on import is very difficult to supply the parts at right time and is above the market price. In this paper, we performed computational analysis for development of solenoid operated valve using in NPP. Various analyses such as electromagnetic, thermal, fluid and structure analysis are performed with ANSYS. We verified the SOV design is possible to be applied for environmental condition of NPP.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1007

Experience of Nuclear Power Plant NEK Modeling in Computer Code APROS

Samo Fürst1, Luka Štrubelj1, Tomaž Žagar2, Ivica Bašić3

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia1

Agencija RS za radioaktivne odpadke, Celovška cesta 182, 1000 Ljubljana, Slovenia2

APoSS d.o.o., Repovec 23b, 49210 Zabok, Croatia3

samo.furst@gen-energija.si

 

GEN energija, as an interested investor in new NPP at Krško site, would like to develop its own engineering capacity for modeling. A modeling tool capable to model nuclear power plant on different scales from whole nuclear power plant, system and component scale is preferred. Apros, as a multifunctional software for modeling and dynamic simulation of processes of different power plants, including PWR is used. Model in Apros can be built as a complete model of a whole plant, with a complete thermo hydraulic, electrical and regulation systems. Such a model could be used in several stages, with its first use for new NPP design verification, possible optimization of NPP systems, for training purposes as an engineering simulation tool and also a model for full scope simulator.

The NEK’s model of primary system developed in Relap is used as a basis and will be upgraded with additional systems, such as electrical system and secondary system. Results of Relap model will be used for validation of primary system model built in Apros. The pressurizer model and other primary components or subsystems development, verification and validation will be presented in the paper. The experience of modeling in comparison with Relap code will also be presented.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1011

Lessons learned from the Operational safety review missions OSART

Miroslav Lipar

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

m.lipar@iaea.org

 

The IAEA Operational Safety Review Team (OSART) programme provides advice and assistance to Member States in enhancing the operational safety of nuclear power plants (NPPs). Careful design and high quality of construction are prerequisites for a safe nuclear power plant. However, a plant’s safety depends ultimately on the ability and conscientiousness of the operating personnel and on the plant programmes, processes and working methods. An OSART mission reviews a facility’s operational performance against IAEA Safety Standards and proven good international practices.

OSART reviews are available to all countries with nuclear power plants in operation, and also approaching operation, commissioning or in earlier stages of construction (Pre-OSART). Most countries have participated in the programme by hosting one or more OSART missions or by making experts available to participate in missions. Follow-up visits are a standard part of the OSART programme and are conducted between 12 to 18 months following the OSART mission.This presentation is summarizing mission results so that all the aspects of OSART missions are gathered. It also includes the results of follow-up visits. This presentation highlights the most significant findings. First part summarizes the most significant observations made during the missions and follow-up visits between 2010 and 2012. Second part describes the mains trends on issues and good practices that were identified in the period covered. Third part describes the assessment of overall OSART mission results and conclusions.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1012

Reactivity Management Program in NPP Krsko

Barbara Grobelnik, Bojan Kurinčič

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

barbara.grobelnik@nek.si

 

In order to improve nuclear safety all activities and conditions which can affect core reactivity or stored nuclear fuel have to be carefully controlled and monitored. The reactivity management program provides the guidance to ensure that all plant evolutions affecting reactivity are controlled in a safe and conservative manner that is consistent with fuel design and operating limits. The program defines the roles, responsibilities and measures for monitoring and controlling activities which can affect reactivity. One part of the program is to identify reactivity management events and issues. Reactivity management events and precursors are monitored and evaluated thoroughly, therefore knowledge and experience regarding safe and successful operation are continuously improving. In the article the reactivity management events and corresponding corrective actions in the past few years are presented. Additionally, some recommendations for further improvement are discussed. One of the goals for the future is to raise an awareness of the importance of Reactivity management to all site personnel.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1013

Control of reactivity in PWR reactors

Andrija Volkanovski, Ljubo Fabjan

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

andrija.volkanovski@ijs.si

 

The prime purpose of the nuclear safety is prevention of the release of radioactive materials formed in the fuel, ensuring that the operation of nuclear power plants does not contribute significantly to individual and societal health risk.

Safety functions are performed in nuclear power plants to control the sources of energy in the plant and the radiation hazards. Three main safety functions that should me assured in nuclear power plant in all situations are: 1. Control of reactivity, 2. Removal of Decay heat to the ultimate heat sink and 3. Containment of radioactive materials.The safety function “Control of reactivity” prevents uncontrolled reactor power increase and shuts reactor when needed. Current pressurized water reactors have at least two independent reactivity control systems of different design principles. The General Design Criterion 26 - Reactivity control system redundancy and capability of 10 CFR 50, App. A, require one of the systems to be capable of holding the reactor core subcritical under cold conditions.In this paper results of the analysis of the reactor reactivity change resulting from xenon decay in shutdown core for extended station blackout plant conditions will be presented. Analysis will be done for different modes of operations at the beginning and end of fuel cycle. The scenarios requiring boron addition will be identified and discussed. The probability of failure of boron addition will be analyzed and quantified. The current regulatory requirements considering reactivity controlled will be reviewed and potential improvements will be suggested.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1014

Passive safety systems of VVER-type NPPs and their substantiation

Gennady Taranov, Mikhail Maltsev, Ilya Kopytov

JSC "Atomenergoproekt", Bakuninskaya 7, 105005 Moscow, Russian Federation

taranov_gs@aep.ru

 

For the purpose of safety increasing of Russian new generation NPPs the diversity principle is applied in the structure of safety systems performing critical safety functions. From the technical point, this principle is realized by using of active and passive safety systems. No need in power source for operation is the principal feature of passive systems. The systems operated in this way are:

- Passive heat removal system , PHRS- System of second stage hydroaccumulators, HA-2- Passive filtering system of leakages from containment, PFS- Passive system for hydrogen removal from containment- System for corium retention.Passive systems PHRS, HA-2 and PFS prevent the transition of the beyond design basis accidents into severe stage, provide the localizing properties of containment and limit the radioactive release from the plant.The diversity principle in the structure of safety systems was adopted during the development of the basic design of NPP-92 with VVER-type reactor unit. This design became the basis for the subsequent development of NVAES-2 and VVER-TOI designs in Russia, Kudankulam NPP in India, Belene NPP in Bulgaria, Akkuya NPP in Turkey and others.Due to availability of passive systems, the NPPs maintain high safety level under extreme natural conditions. This feature of NPP is reflected in the report by presentation of the safety analyses for accident conditions similar to NPP Fukushima.The report provides information on the engineering solutions adopted in the designs of passive safety systems and the results of research engineering works, carried out for substantiation of quantitative characteristics of safety systems.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1017

Analysis of Steam Generator Tube Plugging in a PWR. Influence in the Emergency Operating Procedures.

Patricia Pla1,2, Francesc Reventós1, Manuel Martin Ramos2, Ismael Sol3

Universitat Politecnica de Catalunya, C. Jordi Girona, 31, 08034 Barcelona, Spain1

Joint Research Centre of the European Commission, Westerduinweg 3, 1755 ZG Petten, Netherlands2

Associació Nuclear Ascó - Vandellos II, L’Hospitalet de l’Infant, Tarragona 43890, Spain3

patricia.pla-freixa@ec.europa.eu

 

A number of Nuclear Power Plants (NPPs) with Pressurized Water Reactors (PWR) in the world have replaced their steam generators (SG) due to degradation of the SG tubes caused by different problems (stress corrosion cracking, fretting, wear, pitting, denting, material problems, etc). Several methods were attempted to correct the defects of the tubes, but eventually the most permanent solution was found to be their plugging. The consequences of plugging the tubes is the subsequent loss of heat transfer surface, reduction of the primary system mass flow, the consequential reduction of reactor nominal power and finally economic losses.

The objective of this paper is to present an analysis with Relap5/mod3.3 patch03 for the Spanish reactor ASCÓ-2, a 3-loops 2940.6 MWth Westinghouse PWR, in which steam generator tubes are simulated to be plugged in order to find the limit for the adequate operation of the plant. Several steady state calculations were performed with different fractions of plugged SG tubes, by modeling the reduction of the primary to secondary heat transfer surface and the reduction of the primary coolant mass flow area in the tubes as well. The results of the analysis yield that plugging 12% of the SG tubes is around the limit for adequate reactor operation. This limit considers also the proper operation of the turbine valve avoiding its complete opening.To complete the study an event in which the steam generators are used to cool-down the plant was simulated to find out if the plugging of SGs tubes could influence the efficiency of the operator actions described in the Emergency Operating Procedures (EOP) to handle this kind of events. The selected events have been the rupture of one tube of a SG and a small LOCA. Two cases with no plugging at all (0%) and 12% SG tube plugging were performed. The actions of the corresponding Emergency Operating Procedures for SG tube rupture and small LOCA were coded in the calculations.






11.09.2013 10:40 Poster session 2

Nuclear power plant operation - 1018

Evaluation of Reactor Coolant System specific activity to determine fuel integrity at NPP Krško

Martin Chambers, Dejvi Kadivnik, Bojan Kurinčič

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

martin.chambers@nek.si

 

In recent fuel cycles (cycles 20-23, 25-26) NPP Krsko experienced some leaking fuel rods that degraded during the operational cycle. The leaking was tight in nature thus overall Fuel Integrity Index was not significantly impacted. In all cycles, leaking fuel was initially determined by the monthly evaluation of Reactor Coolant System (RCS) isotopic specific activity. Typically, the first indication of leaking fuel is observed in RCS specific activity of noble gas isotopes, particularly the Xe-133 isotope that will increase by more than an order of magnitude when a fuel rod is leaking. Using RCS isotopic activity evaluation during the cycle, it is usually possible to determine other properties of the leaking fuel rod(s) such as the fuel rod hole size, failure mechanism, relative power, core position and burnup. Here, we present the RCS chemistry evaluation and explanation of how the attributes of leaking fuel were determined. The RCS chemical evaluation results are compared to number of leaking fuel assemblies determined by fuel examinations during core offload.






11.09.2013 10:40 Poster session 2

Regulatory issues and legislation - 1101

NPP Krško 2 Time Schedule for Licensing an Permitting Processes

Tomaž Ploj, Samo Fürst, Danijel Levičar

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

tomaz.ploj@gen-energija.si

 

The paper will present the results of an external study developed for the investor of the proposed new NPP Krško 2. The aim of the study was to clearly define the time schedule for licensing and permitting processes, to identify legal and other obstacles in the procedures and, on the basis of practical experiences, as well as experiences from foreign practices, develop optimization of procedures in order to enable the implementation of the project of NPP Krško 2 from the strategic decision through siting process to plant start-up and operation in the optimal timeframe.

The initial results of the study showed that within the current legislative framework and with the integrated Environmental Impact Assessment (EIA) into National Spatial Plan (NSP) procedure the time schedule from the strategic decision to commercial operation would be around 14 years. This is not in line with the foreign practice where the investment is carried out in 10 years starting from the strategic decision. The general opinion is that the project is interesting and commercially viable if the siting procedure from the strategic decision up to the first concrete can be accomplished in less than 7 years.Based on further fine-tuning of the initial results we were able to develop a realistic plan that accounts for all the necessary permits, provides extra time for critical activities, allows for simultaneous proceedings and minimizes critical path dependency. The result is a 10-year project time schedule that includes spatial planning and construction of NPP Krško 2, where the Environmental Impact Assessment (EIA) is treated separately from the National Spatial Plan (NSP) procedure. Provisions for proposed approach are already within the current legislation and are seen as the logical way to obtain NPP Krško 2 permits.






11.09.2013 10:40 Poster session 2

Regulatory issues and legislation - 1102

Evaluation of Applicability of LWR General Design Requirements to Very High Temperature Gas-Cooled Reactor

Jin-Hyuck Kim, Changwook Huh

Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

huhcw1@naver.com

 

In Korea, design concepts of very high temperature reactor (VHTR) are developed since 2000 by the Korea Atomic Energy Research Institute (KAERI). The 255th Atomic Energy Commission of Korea, held in Dec. of 2008, determined a long-term research and development plan for future nuclear systems including the VHTR. According to the plan, it is scheduled to submit an application for licensing of a VHTR demonstration reactor in 2022. The construction of demonstration reactors is planned to verify the design performance and economics before the commercialization of the VHTR. In order to prepare the licensing of the demonstration reactor, the general design requirements (GDRs) for the VHTR should be established because they are used as rules in the safety evaluation of the design for the licensing by the regulatory body. The VHTR design is characterized by 1) the refractory triple isotropic layers coated fuel particles (TRISO CFP) which can retain the fission products and then provides a unique robustness of the first barrier for the fission products, 2) the inert, single phase helium gas as coolant and graphite with high temperature stability and long response times as moderator, 3) negative temperature coefficient of reactivity and 4) passive core cooling and decay heat removal by natural process, etc. Due to these design differences current GDRs, which were developed based on the design of the LWR, are not well suited for the safety evaluation of VHTR. In this paper, the applicability of the current GDRs for the LWR to the design of the VHTR was evaluated and current GDRs were classified in the following categories: (1) the requirements to be newly added due to the new systems adopted in the VHTR, (2) the requirements to be modified due to the design differences between VHTR and LWR, (3) the requirements not applicable to the VHTR due to the design differences, and (4) the requirements applicable to the VHTR as it is. At this evaluation, it has been recognized that new GDRs related to specific features of the VHTR design such as low-pressure vented containment building, reactor cavity cooling system (RCCS), etc., need to be developed earlier in the design process of the VHTR to reduce the licensing risk.






11.09.2013 10:40 Poster session 2

Regulatory issues and legislation - 1103

Regulatory View of Hydrogen Management at the Krško NPP

Tomi Živko1, Srđan Špalj2, Andreja Peršič1

Uprava Republike Slovenije za jedrsko varnost, Litostrojska ulica 54, 1000 Ljubljana, Slovenia1

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia2

tomi.zivko@gov.si

 

The Krško NPP and Slovenian Nuclear Safety Administration (SNSA) have been for a longer time paying attention to the problem of hydrogen in the containment of NPP. Findings of the IAEA's RAMP mission concerning hydrogen were included in the first Periodic Safety Review of the Krško NPP. It resulted with studies which showed that there was no danger due to nonuniformity of hydrogen distribution in the case of accident. The Fukushima disaster demonstrated the importance of containment protection. Accordingly, ENSREG requested from national regulators to urgently consider prevention of hydrogen explosions.

On the basis of SNSA regulatory orders and additional analyses, the Krško NPP decided to install passive autocatalytic recombiners (PARs) in 2013. The PAR system will protect the containment in the case of design as well in the case of beyond design basis accidents (BDBA). This will be the first case of licencing of BDBA systems at the Krško NPP and therefore also a challenge for the regulator.






11.09.2013 10:40 Poster session 2

Regulatory issues and legislation - 1104

New Basic Safety Standards within the EU

Helena Janžekovič

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

helena.janzekovic@gov.si

 

Regulatory framework is one of the main parameters determining not only investment in new nuclear installations and decommissioning of old ones but also other phases of nuclear installations, e.g. maintenance. The framework has also a deep influence on other nuclear activities, e.g. transport of nuclear fuel, research and development in nuclear area. Three basic directives are in the core of radiation and nuclear safety in the European Union (EU):

• Directive laying Down Basic Safety the Protection of the Health of Workers and the General Public against the Dangers arising from Ionizing Radiation (Basic Safety Standards Directive), 1996• Nuclear Safety Directive, 2009• Spent Fuel and Radioactive Waste Directive, 2011.Today two of three fundamental directives are under revision. They are all based on the EURATOM Treaty from 1957. The Basic Safety Standards Directive is under the revision for some years and it is expected to be adopted in a near future. After the Fukushima accident in 2011 the European Commission started with the preparation of the revision of the Nuclear Safety Directive which is going to be the first revision of the text. At present no revision of Spent Fuel and Radioactive Waste Directive is under way.The impact of the revised Basic Safety Standards Directive on radiation and nuclear safety in EU countries could be substantial. Today the Directive is supplemented by five other specific Directives and numerous other acts reflecting the complexity of issues related to radiation safety area. The proposal of the revision combines the Basic Safety Standards Directives from 1996 with the majority of requirements given in four of five directives mentioned into one directive. The preparation of the revision started well before the Fukushima accident and was initiated by numerous factors, e.g.:• revision of a system of radiation protection given by the International Commission on Radiological Protection in 2007, i.e. introduction of so-called planned, existing and emergency exposure situation,• development of scientific knowledge in radiation safety,• harmonisation of some areas, e.g. control of contaminated building materials and requirements for exemption and clearance of the materials.In addition, strengthening of emergency preparedness as well of education and training are given in the revised directive. The protection of the environment was also heavily discussed during its preparation.The preparation of the revised directive Basic Safety Standard Directive which took more than five years reflects the complexity of the issues related to the safety standards. The paper discussed new solutions in the regulatory regime of safety within EU countries focusing on its impact on nuclear installations including decommissioning. The impact on uranium mines is also discussed. The paper analyses main burning issues which were addressed among experts, regulators, industries and others involved in its preparation. The link of the directive proposal to the Nuclear Safety Directive and Spent Fuel and Radioactive Waste Directive is enlightened.






11.09.2013 10:40 Poster session 2

Regulatory issues and legislation - 1105

NPP Krško SAMG Upgrade

Mario Mihalina

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia

mario.mihalina@nek.si

 

Nuclear Power Plant Krško (NEK) has decided to take steps for upgrade of safety measures to prevent severe accidents, and to improve the means to successfully mitigate their consequences. The content of the program for the NEK Safety Upgrade is consistent with the nuclear industry response to Fukushima accident, which revealed many new insights into severe accidents. Therefore, new strategies and usage of new systems and components should be integrated into current NEK Severe Accident Management Guidelines (SAMG’s). SAMG’s are developed to to arrest the progression of a core damage accident and to limit the extent of resulting releases of fission products.

Krško new SAMG’s revision major changes are made due to: replacement of Electrical Recombiners by Passive Autocatalytic Recombiners (PARs) and the installation of Passive Containment Filtered Vent System (PCFV); to handle a fuel damage situation in Spent Fuel Pit (SFP) while a core damage event is simultaneously occurring into the reactor; and to assess risk of core damage situation during shutdown operation.New SAMG revision will be in use after NEK 2013 Outage, after new equipment will be installed.






11.09.2013 10:40 Poster session 2

Regulatory issues and legislation - 1108

Regulatory Issues Concerning the Siting of Nuclear Facilities in the Case of NPP 2, Krško, Slovenia

Aleš Janžovnik, Tamara Tepavčević, Andrej Špiler

SAVAPROJEKT, d.d., Cesta krških žrtev 59, 8270 Krško, Slovenia

ales.janzovnik@savaprojekt.si

 

The siting and construction of new nuclear facilities presents a great challenge, since it is a complex and demanding project. Nuclear power plant is a highly specific works for the siting and construction, and there is yet no experience of its implementation under the current legislation in the Republic of Slovenia, as the existing Krško Nuclear Power Plant was built at the time of other social and economic conditions and within a completely different legislative framework. On the territory of the Republic of Slovenia only major infrastructural facilities of national importance, such as motorways, hydroelectric power plants, transmission lines, gas pipelines, etc., have been sited under the current legislation. For these works it has been shown that the procedures for their siting are long-lasting, which poses a question about the actual timing of the construction of the new nuclear power plant and whether the existing procedure of siting and construction can be reduced. This is particularly important because of the way of approaching to the implementation of financing of such a project and the selection of the supplier of technology. The question is also relevant because of the sensitivity of the topic related to safety and because of related public consensus on the acceptability of construction of such facilities.

The purpose of the article is to examine the current Slovenian legislative framework for the siting of nuclear facilities and to identify problems and obstacles resulting from the existing legislation. The aim of the article is to stimulate thought and discussion on the adequacy of the existing legislative framework governing the siting and construction of works of national importance, and on possible changes of the current legislation, which will enable fast, effective and transparent implementation of key national projects.






11.09.2013 10:40 Poster session 2

Nuclear security - 1501

Assessing the attractiveness of nuclear power plants as terrorist attack targets

Davor Šinka

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

davor.sinka@enconet.hr

 

Nuclear power plants (NPPs) have long been recognized as potential targets for terrorist attacks. A successful attack could have widespread consequences on both public health and the environment. Following the September 11 events and more recent Fukushima accident, the adequacy of the protection measures to defend NPPs against terrorist attacks have been questioned. In order to develop optimal protection strategies it is important to understand what, how and why terrorists attack. However, while the number of studies on the causes of terrorism is vast, not too many studies explore the subject of terrorist targeting preferences.

This research investigates the factors which determine whether a terrorist organization will make a decision to attack nuclear power plant. In general, the attractiveness of potential target depends on two parameters: (1) the motivation of the terrorist organization and (2) its capabilities. In the first step of the research the factors which determine the level of motivation and capabilities were identified and analyzed. Such factors include terrorist organization's characteristics (ideology, organizational structure, resources, etc.), perceived target's characteristics (function, profile, inherent hardness, level of protection, potential damage, location, etc.) and other factors (security environment, presence of armed conflict, historical context, etc.). As the second step of the research a set of hypotheses have been developed to help with assessing the attractiveness of various terrorist attack modes against NPPs. In the final, third step the hypotheses were tested by examining historical data on terrorist attacks against NPPs. In this step the data from the Global Terrorism Database (GTD) and the RAND Database of Worldwide Terrorism Incidents (RDWTI) were used, as well as the data from the Global Chronology of Incidents of Chemical, Biological, Radioactive and Nuclear Attacks compiled by Mothadi et al.






11.09.2013 10:40 Poster session 2

Nuclear security - 1502

Generic Design Basis Threat - Possible Solution for Security by Design at new NPPs

Miroslav Gregorič

Miroslav Gregorič S.P., Martinčeva 30, 1000 Ljubljana, Slovenia

miroslav.gregoric@gmail.com

 

The paper presents the development of Design Basis Threat - DBA, based on IAEA guidance and research of past security events at nuclear as well as non nuclear facilities, challenging the extrapolation of threat characteristics data into generic DBA that could be used in the concept known as security by design. There are only few vendors of NPPs worldwide and a generic DBT might be a good way in establishing high level of nuclear security on one hand and providing possibility of transparent comparison of different NPP designs by investors or regulators. In addition, knowing generic DBT at the design phase of an NPP in terms of intent, motivation, opportunity and capability, the synergies of safety and security could be exploited into an optimal design.






11.09.2013 10:40 Poster session 2

Nuclear security - 1503

The Nuclear Power Plants New Methodology for Universal Vulnerability Assessment of Terrorism Threats and Natural Disasters Analyses and Predictions

Venceslav Gospodinov Kostadinov

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

venceslav.kostadinov@gov.si

 

National emergency systems in the past, based on our in-depth analysis and assessments, explicitly did not include vulnerability assessments of the critical nuclear infrastructure as an important part of a comprehensive preparedness framework. After the huge terrorist attack on 11.09.2001, decision makers became aware that critical nuclear infrastructure could also be an attractive target to terrorism, with the purpose of using the physical and radioactive properties of the nuclear material to cause mass casualties, property damage, and detrimental economic and/or environmental impacts.

For assurance of safe nuclear facilities operation and national security the necessity to evaluate critical nuclear infrastructure vulnerability to threats like human errors, terrorist attacks and natural disasters, as well as preparation of emergency response plans with estimation of optimised costs, are of vital importance.In the article presented new universal methodology and solution methods for nuclear power plants (NPPs) vulnerability assessment can help the overall national energy sector to identify and understand the terrorist and natural disaster threats to and vulnerabilities of its critical infrastructure. Furthermore, adopted methodology could help national regulators and agencies to develop and implement a vulnerability awareness and education programs for their critical assets to enhance the security and a safe operation of the entire energy infrastructure. The new methods presented in the article can also assist nuclear power plants to develop, validate, and disseminate assessment and surveys of new efficient countermeasures. Consequently, concise description of developed new quantitative method and adapted new methodology for nuclear regulatory vulnerability assessment of nuclear power plants are presented.An important new additional original contribution is also presentation of initial quantitative vulnerability assessment estimations for three different nuclear power plants from which one power plant is Fukushima in Japan. Particularly important is the use of new methodology for the qualitative and quantitative “case study” assessment of the vulnerability of nuclear power plant in Fukushima to natural disaster which has occurred.






11.09.2013 11:20 Invited lecture 6

Invited lectures - 106

The Detection of Hydrogen flakes in the Belgian Doel 3/Tihange 2 Reactor Pressure Vessels – Overview of Technical Developments to support Restart Justification

Eric van Walle

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

eric.van.walle@sckcen.be

 

In the summer of 2012, unexpected indications were discovered in the base metal of the Reactor Pressure Vessel of the Doel 3 and Tihange 2 Nuclear Power Plants at the occasion of a voluntary UT inspection. Seen the uncertainty on the nature and origin of the indications, and in agreement with the requirements of the Belgian Safety Authorities, the operator Electrabel kept these plants in cold shutdown and built a comprehensive and convincing Safety Case to justify the RPV integrity in the presence of these hydrogen flakes and to guarantee safe reactor operation in all operational conditions.

To this aim, a Technical Roadmap was built that makes use of innovative technical research and development in various fields: UT inspection techniques, metallurgy and material sciences, material testing, structural analysis and fracture mechanics. This work was mostly performed by Belgian experts and laboratories. The results of the Safety Case were already made public by the Belgian Safety Authority and the Doel 3/Tihange 2 reactors were allowed to restart this summer.This presentation will give an overview of the technical challenges that had to be overcome to obtain the solutions that led to the approved Safety Case. We will put emphasis on some of the innovative technical research and development jobs, especially on the material testing that was mainly performed at SCK•CEN, the Belgian Nuclear Research Centre.






11.09.2013 12:00 Nuclear power plant operation

Nuclear power plant operation - 1009

Using “Process-to-Part” Techniques in Large Nuclear Power Plant Component Manufacturing to Improve Supplier Competitiveness

Joshua Barnfather

Nuclear AMRC, Advanced Manufacturing Park, Brunel Way, Catcliffe, Rotherham, S60 5WG, United Kingdom

j.barnfather@namrc.co.uk

 

Large components in nuclear power plants can reach 20m in height and 6m in diameter, weighing up to 750 tons. This makes the setup of manufacturing facilities highly capital intensive, particularly when considering the cost of the necessary foundations and floor space in addition to the large tools. From an operational perspective this is also costly due to conventional procedures involving lifting these components and their sub-sections between work zones. This creates further health and safety challenges and is something particularly seen between machining and inspection operations. The methods typically used in large nuclear power plant component manufacture therefore lack flexibility and represent a significant proportion of component cost and increase lead-time. This is a major barrier for new entrants to the nuclear supply chain and it highlights an opportunity to improve the economic competitiveness of nuclear power in the form of “process-to-part” machining and inspection using portable tools.

The purpose of this work is therefore to review procedures currently used in the manufacture of reactor pressure vessels, steam generators and pressurisers and the equipment used for this from both a technical ability and lifecycle cost perspective to set a base case for comparing alternatives against. These alternatives are selected by reviewing “process-to-part” techniques used in other industries to determine which are technically best suited for adaptation to the nuclear industry. This also considers what the best equipment would be to implement this in terms of the ability of portable solutions to achieve the tolerances offered by conventional machinery. Following this, an analysis of the estimated equivalent lifecycle costs is made and compared against the base case, allowing conclusions to be made on the costs benefit of using “process-to-part” manufacturing processes in the production of heavy plant components. In addition, this work highlights benefits that this approach can offer from a health and safety and supply chain competitiveness perspective.






11.09.2013 12:20 Nuclear power plant operation

Nuclear power plant operation - 1015

Leveraging Specific Plant Features to Manage Internal Hazards

Harri Tuomisto

FORTUM, Power Division, P.O. Box 100, FIN-00048 FORTUM, Finland

harri.tuomisto@fortum.com

 

As a consequence of the Fukushima accident the Management of External Hazards and Severe Accident Management have been subject to the increased attention at nuclear power plants. The lesson learnt was that there might be cases of paying too little attention to external hazards in comparison to the risk they might pose for the plant. Since the approaches chosen to Severe Accident Management vary significantly among the nuclear power plants, the stress test and regulatory processes have identified further need to reinforce mitigation of severe accidents.

Earlier we have published extensively of the capability to utilize plant-specific features of the Loviisa VVER-440 units to develop and implement a consistent Severe Accident Management approach to respond to the plant specific vulnerabilities. Respectively, the internal hazards and accident progression are often determined by plant-specific features. A typical feature of internal hazards and a special concern of accident progression are that they challenge simultaneously more than one level of the Defence-in-Depth concept or penetrate more than one of the physical barriers of the fission product releases.Internal hazards may themselves be initiating events, such as common cause failures, internal fires or floods, missiles, inhomogeneous boron dilution, or primary-to-secondary leakage accident (PRISE). In many cases they are hazards that are created during the accident progression such as pressurized thermal shock, loop seal issue, boron crystallization, containment sump clogging, or inherent boron dilution mechanisms.The aim of this paper is to discuss the bases, how the management approach was developed and to present how the resolutions were implemented for some of the internal hazards identified for the Loviisa VVER-440 units. The Loviisa plant configuration is in many respects quite unique, since the original VVER-440 design has been added with ice condenser containment, specific reactor coolant pumps and many other features. In many cases we have applied integrated deterministic and probabilistic analyses to study various possibilities and plant capabilities to resolve the raised issues. Most of the presented work has been carried out already many years ago. The reason to revisit these developments is to bring further insights and perspective to the current work done as Post-Fukushima actions on external hazards, extensive damage conditions and severe accident mitigation.






11.09.2013 12:40 Nuclear power plant operation

Nuclear power plant operation - 1008

Needs and Challenges for “Non-Baseload Operations” in Nuclear Power Plants

Peter Schimann, Zoran V. Stošić

AREVA NP GmbH, Koldestraße 16, D-91052 Erlangen, Germany

peter.schimann@areva.com

 

There is no direct way to store electrical energy in bulk, so all electrical power systems require that all electrical power generated is adjusted continuously to closely match the electrical demand as it changes and also to control system frequency. This requires that generating units are able to change their generated output as needed, so that the total generation matches the variations in total electrical demand.

It may also be necessary to change the output of certain generating units in order to control power flows on the transmission system. In particular, countries that have an electrical power system connected to the electrical systems of their countries will also need to adjust the electrical power generated in order to control power flows across the connections to these other networks. In cooperation with renewable energy and the transnational integrated network it will be increasingly necessary for a great percentage nuclear power plants to be adjustable within a very short response time. We need flexibility at unit and fleet level. A NPP capable of flexible operation is a power unit able to: - stabilize power output at any value from the minimum operating level to the maximum operating level;- ramp up and down the power between the minimum and maximum at a defined ramp rate;- Deliver a minimum frequency response corresponding to a defined frequency deviation.- Stabilize power outputA fleet of NPPs capable of flexible operation is a set of power units able to: - Offer a variable power capacity that can be dispatched by varying individual unit outputs in a predictable, stable and timely manner.- Participate to the frequency control of its synchronous areas by delivering a minimum power frequency characteristic corresponding to a defined frequency deviation; - offer a minimum power capacity in a predictable, stable and timely manner;Changes in power reactor, even if small and limited have an impact on: Neutron flux, Number of solicitations, Number of RCDM moves, Core physics and Insensitivity of primary temperature control.In the following, this status will be further highlighted.






11.09.2013 14:30 Other related topics

Regulatory issues and legislation - 1107

Regulatory Perspective on Validation of Computer Codes Used in Safety Analysis

Janusz Kowalski

Canadian Nuclear Safety Commission (CNCS), P.O. Box 1046, K1P 5S9 Ottawa, ON, Canada

janusz.kowalski@cnsc-ccsn.gc.ca

 

Validation is essential final step in qualifying any computational method. The purpose of validation is to provide confidence in the ability of a code to predict, realistically or conservatively, the values of the safety parameter or parameters of interest. Canadian Nuclear Safety Commission (CNSC) requires that all computer codes must be validated for their application in safety analysis [1, 2 and 3].

A considerable amount of resources has been devoted in Canada during the past three decades for establishing and conducting validation of the computer codes used in safety analysis. The validation has been carried out using eight discipline-based Validation Matrices, which describe phenomena relevant to CANDU reactor and experimental data sets available for their validation. At that time, the validation effort was primarily phenomena-based and did not fully quantify the code prediction accuracy of key parameters. In recent assessment of validation performed to support the use of Canadian computer programs in licensing analysis, the CNSC has indicated the need to conduct validation using important parameters. In this new approach, the identification of the important phenomena and parameters associated with the acceptance criteria and Figures of Merit (FOM) is an essential step of parameter-based validation process. For each safety analysis case, the key FOM parameters and phenomena governing the transient are identified, including both operating and modeling parameters. The key parameters and safety phenomena are then ranked based on importance to the safety analysis results for each acceptance criterion. Once the relevant phenomena are identified and the FOM or trip related parameters specified, the available tests need to be assessed for suitability for use in validation. The applicability of the validation results may differ depending on the type of safety analysis. For the Best Estimate And Uncertainty (BEAU) analysis, the bias and variance in bias in key parameters need to be determined. However the bias component may be only applied for the conservative analysis. This paper outlines the CNSC regulatory approach to data qualification and means of comparing code calculation and experimental data. Our experience, practices and expectations are provided, pertaining to selection and processing of the data and methodology used for determining the code accuracy. Also, regulatory views are presented on applicability of code accuracy in conservative and BEAU type safety analysis.1. Canadian Nuclear Safety Commission, RD-310, “Safety Analysis of Nuclear Power Plants”, Ottawa, 2008. 2. Canadian Nuclear Safety Commission, GD-310, “Guidance on Safety Analysis of Nuclear Power Plants”, Ottawa, 2008. 3. Canadian Nuclear Safety Commission, GD-149, “Computer Programs Used In Design and Safety Analyses of Nuclear Power Plants and Research Reactors”, Ottawa, 2000.






11.09.2013 14:50 Other related topics

Regulatory issues and legislation - 1106

Nuclear Renaissance in Finland: The Role of the State and Stakeholders in an Industry-Led Policy Process

Victoria Tuomisto

London School of Economics and Political Science, Houghton Street, London WC2A 2AE, United Kingdom

victoria.tuomisto@gmail.com

 

Energy security and climate change are of increasing concern and salience in Europe. In the 2000s, this spurred a renewed interest and activity in nuclear construction, production and energy politics in Western Europe. A period of Nuclear Renaissance in Finland began with a decision to build a fifth nuclear power plant in 2002, first Western country to do so in over a decade.

Traditionally, the state is seen as a prominent leader in coordinating nuclear energy and industrial policy. Yet Finland’s nuclear development operates on a different model: the industry leads the nuclear planning and determines the scale of production, whilst the state has the ultimate role in approving construction and operation. This research ascertains how Finland, with a backseat role of the state, is able to actively build new nuclear plants.This study is conducted from a socio-political economic perspective, applying six factors that drive nuclear development. It is based on the framework introduced by Sovacool and Valentine (2012), which implicitly relies on a state-led model, implying a strong role of the state that coordinates the stakeholder network, subordinates politicalopposition and seeks to undermine the views of the general public. Instead, the following properties are identified in the Finnish case: energy security and independence, pragmatism and trust in technology, limited interventionism of the state, stakeholder networking through industrial cooperation, inclusion of opposition and social participation in decisionmaking.This research analyses how the Finnish case is different by evaluating how the nuclear new build policy process relates to these properties. It is based on stakeholder interviews and concurrent newspaper articles to acquire views of the role of the state and the industry as well as the level of stakeholder participation and cooperation in the process.Finland’s alternative industry-led model entails a limited state role that is defined by the Nuclear Energy Act. This role is however ambiguous in the sense that the state is a stakeholder in one of the companies and consortia. The nuclear companies plan and lead the scale and processes and are empowered by the Mankala ownership model that enablescost and risk sharing in a small, liberalised energy market, as well as a concentration of interests between different industries and public bodies. Thus it allows coordination of mutual interests and lobbying efforts during the policy process.The role of the state in Finland still entails a decision-making role for the government and parliament. However, instead of taking the state ‘out’, this process mostly takes politics and inherent instability out of nuclear expansion. The lack of an established nuclear policy, amendable by government, gives the industry the stability to plan, construct and operate new NPPs.Taking a socio-political economic approach to the Finnish nuclear policy process during the nuclear renaissance gives a needed insight into determining how a politically sensitive industry is able to pursue further operation in time of deep political, economic and international public uncertainty.






11.09.2013 15:10 Other related topics

Post Fukushima actions - 1203

Status of the Spent Fuel in the reactor buildings of Fukushima Daiichi 1-4

Bernd Jaeckel

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

bernd.jaeckel@psi.ch

 

The ratio of the radio nuclides Cs-134g and Cs-137 deduced from measurements of liquid samples from the spent fuel pools in Fukushima Daiichi 1-4 is used to interpret the status of the spent fuel assemblies in the pools of the damaged reactor buildings. The different nature of the production of Cs-134g (neutron capture product of Cs-133) and Cs-137 (cumulative fission product from mass chain 137) and the different half lives (2.06 years and 30.17 years) requires a complicated calculation of the mass and activity of the two nuclides. These masses depend on the local burn up of the fuel, the burn up history and the radioactive decay. For the calculation of the radionuclide masses with ORIGEN-2 the user has to use different cross section tables depending on the maximum burn up of the fuel. These tables are based on experience from activity measurements after irradiation of fissionable materials. Especially for the neutron capture product Cs-134g the calculation is rather complex, because the production of Cs-133 (stable cumulative fission product from mass chain 133) has to be taken into account. The neutron capture cross section for Cs-133 was fitted from a gamma scan of spent fuel rods in a hot cell. The method of the calculation of the nuclide activities and the interpretation of the gamma measurements of the spent fuel pool samples from Fukushima Daiichi 1-4 are described in detail.






12.09.2013 09:00 Invited lecture 7

Invited lectures - 107

Technical and scientific nuclear safety expertise: a rare resource, crucial for safety and security enhancement across Europe

Jacques Repussard

Institut de Radioprotection et de Sureté Nucléaire, 31, avenue de la Divison Leclerc, 92260 Fontenay Aux Roses, France

jacques.repussard@irsn.fr

 

After a short introduction to IRSN, the French public institution responsible for research and expertise in the field of nuclear and radiological risks, and to its role as a TSO within the French nuclear safety, security and radiation protection system, the conference will develop the following key issues:

1. Nuclear safety regulatory oversight requires state of the art expertise in many fields. It is usually the case that nuclear safety regulatory bodies do not have within their own internal organization the necessary scientific expertise to investigate in appropriate depth the safety demonstrations which are proposed by operators, or to optimize public policy decisions for the protection against ionizing radiation. Reliance on external expertise is therefore a common practice. However this raises issues about the necessary independence of such external expertise from possible conflicts of interest, and about the proficiency of such external expertise capabilities. The development of formal TSO structures and their reunion in the ETSON European association has been a response to these challenges. 2. Although many EU Member States have common needs of access to expertise, such capabilities are not evenly distributed between them. Moreover, the burden of maintaining, and expanding where necessary the body of knowledge and expertise, through research and systematic analysis of experience feedback is also unequally distributed, in part because of the scarce availability of experimental facilities and of R&D development resources. This situation leads also to insufficient sharing of existing knowledge and technical expertise tools between European national organizations. The creation of ETSON aimed at enhancing the sharing of technical knowledge and experience feedback in the field of safety among the European national experts.3. The current EU legislative initiative to reinforce the European coherence of nuclear regulatory systems, through amendments of the EU directive on nuclear safety, could therefore also be an appropriate framework to encourage the reinforcement of European cooperation in the field of nuclear safety expertise, through ETSON and dedicated EU R&D Platforms.4. ETSON has now become a representative and trustworthy body bringing together 13 TSO’s, built on sound principles and values, which has started to produce concrete deliverables contributing to enhancing nuclear safety on a European scale. This initiative is backed up by the development of ENSTTI, a European TSO based structure dedicated to knowledge management, training and professional tutoring for nuclear safety, security and radiation protection expertise.5. If it is to survive in the long term in highly demanding and fast evolving European energy markets, nuclear energy needs to be at the same time safe (and perceived as adequately so by society, including for waste management solutions), and economically competitive. Safety needs to be achieved to the highest standards because Europe could not afford a major nuclear accident. Competitiveness concerns not only the industry, but also through the organization of safety supervision and licensing processes. So the TSO’s, the research teams, the regulators, the training institutes must all work closely together, and support each other across Europe to enhance nuclear safety to the highest standards, and to optimize processes that are aimed to achieve this goal.






12.09.2013 09:40 Sustainability, education, training and public relations

Sustainability, education, training and public relations - 1316

The role of international collaboration in knowledge development in creation of TSO

Tomasz Marian Jackowski, Eleonora Grodzicka, Michał Spirzewski

National Centre for Nuclear Research, ul. Andrzeja Sołtana 7, Otwock-Świerk, Poland

tomasz.jackowski@ncbj.gov.pl

 

National Centre for Nuclear Research has been created on September the 1st 2011 by the decree of the Polish Government with a clear goal to form Technical Support Organization for Polish regulator and public administration. It was done by merging two institutes, the former Institute of Atomic Energy POLATOM and the former Andrzej Sołtan Institute for Nuclear Studies, which worldwide reputation and successful research in various fields of nuclear power-related studies are well known. The fields of activity are widely focused on the nuclear physics, cosmology, electronics as well as detectors, accelerators, material research and many more. The main factor of National Center for Nuclear Research development was the participation in the European Structural Founds program "Świerk Computing Centre" from the beginning. In the frame of this particular program, the collaboration with the IAEA, NEA OECD, Euratom and research and TSO organizations from different countries began. The international collaboration plays the key role in the development of the Technical Support Organization expertise, which aim is to become the institution able to provide experts’ support for decision-makers in nuclear power industry in Poland. Expertise and knowledge is expanded by the engagement in the code users international trainings, by being involved in various projects and benchmarks. The National Centre for Nuclear Research participates in, among the others, EURATOM projects such as NURESAFE (creation of BE Codes platform), PRACE and EXASCLAE (both for mass calculations), in collaboration with Institute of Nuclear Chemistry and Technology – ASGARD and PELGRIMM (Transmutation of spent fuel). Moreover NCBJ is involved in EURATOM FP7 projects such as NC2I-R (Cogeneration Initiative), ASAMPSA_E (Advanced PSA), and ALLIANCE as part of the ALLEGRO project. In most of these groups our centre is taking active part in development as well as in management activities. Thanks to experiences gained and with work on research we are effectively expanding knowledge, experience and expertise to meet future’s demands as a Technical Support Organization for first Nuclear Power Plant in Poland.

The IT infrastructure of "Świerk Computing Centre" is located in newly adapted building in Swierk which is specially dedicated for the headquarters of the Center. The work is still ongoing and the target production installation is the 30 blade servers each with 64 computing cores, 256 GB RAM and two 900 GB SAS local disks for each server, along with casing, redundant power supply. This Center will prepare the competence base capable of providing advanced data processing services for domestic nuclear power engineering and conventional power industry, simulations of fuel processes, simulation and monitoring of radiological hazards as well as to conduct scientific and developmental research in related fields.






12.09.2013 10:00 Sustainability, education, training and public relations

Sustainability, education, training and public relations - 1314

Enhancing Cooperation in EURATOM FP7 Projects

Petre Ghitescu, Gabriel Lazaro Pavel

University “Politehnica” of Bucharest, 313 Splaiul Independentei Street, Sector 6, 060042 Bucharest, Romania

gabriel.pavel@gmail.com

 

States who wish to start and develop a nuclear program are considered to be New Member States (NMS) in the industry or simply newcomers. They are supposed to benefit from the cooperation with experienced states, or Old Member States (OMS), and each of the parties has some expectations as a result of their collaboration. An important outcome that NMS is looking for as a result of the cooperation is the expertise record and visibility at international level. Building a successful nuclear energy program is based on correct and in-time shaping of E&T demands.

All types of agreements like bilateral, regional or interregional combined with programs between experienced states and NMS are the basis of efficient knowledge transfer. One key aspect in fruitfull collaboration is represented by the regional and interregional agreements between OMS & NMS.In Romania, as a result of strong international networking in projects like REFIN (Romanian Network of Excellence in Nuclear Physics and Engineering) ENEN, ENEN-II, ENEN-III, NEPTUNO, TRASNUSAFE, EURECA!, ENEN-RU, EUJEP, NEWLANCER it has been developed an efficient, flexible and modern training scheme which answers the needs of most of nuclear industry present in the country. A thorough analysis of the above stated projects and their outcomes with respect to E&T will be made.Future projects like EAGLE and ALFRED (Gen 4 reactor) it’s expected to represent a good oportunity to enhance cooperation between OMS and NMS. Using the experience gained in over a decade of international succesfull collaboration in different research programs (starting with EURATOM FP5) the paper will also comprise an analysis of tools to be used for Romania’s E&T integration methodology in a Gen 4 program. Vision of a complete nuclear research program includes:• a good national strategy and support on the topic, • strong research laboratories supported by good personnel,• education component to provide sustainable and qualified workforce, • national/international interest from stakeholders and governments and a well informed society that needs to be aware of the benefits such program brings.






12.09.2013 10:20 Sustainability, education, training and public relations

Sustainability, education, training and public relations - 1308

Kingdom of Saudi Arabia’s Nuclear Program: A future Prospect

Salah Ud-Din Khan1, Shahab Ud-Din Khan2

King Saud University, Sustainable Energy Technologies Center, P.O.Box 800, Riyadh 11421, Saudi Arabia1

Chinese Academy of Sciences, Institute of Plasma Physics , P.O.Box 1126, 230031 Anhui, Hefei, China2

drskhan@ksu.edu.sa

 

The growing energy demands interconnected with oil assets consumption and growing population warn the Kingdom of Saudi Arabia (KSA) to make a strategic plan for energy. It is estimated that the KSA will lose its oil market within couple of decades as the domestic consumption is also increasing enormously in line with industrial needs. The total reserves of the KSA have been calculated as 267 billion barrels/day while the current situation leads to a decline of 3.4 million barrels/day in 2009 to 8.3 million barrels/day by the year 2028. This increase consumptions will definitely affects the energy production and according to Saudi government, it is estimated that domestic consumption for oil and gas resulting into the decrease of country’s export. The use of fossil fuel for electricity production is not a favourable option in order to compete the economic and political crises. Keeping in mind the entire current situation, the KSA government has increased its subsidies of energy to 100 billion dollar in 2011 only for the domestic consumptions. Considerably, the electricity demand is estimated to increase from 75GWe by 2018 to more than 120GWe by 2030. This demand cannot be compensated by only fossil fuel and the decline recourses of oil. This situation compels the leaders to think about any other resources that could pave the way to a better future. In these criteria, the development of nuclear reactors is considered to be an obvious choice and the KSA government is agreed to work on the nuclear energy. This paper deals with the KSA current situation of installing nuclear facilities and nuclear hybrid energy system.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 603

Use of lattice code Dragon in reactor calculations

Dušan Ćalić1, Marjan Kromar2, Andrej Trkov2

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

dusan.calic@ijs.si

 

A computer code Dragon is a free deterministic code developed by various organizations however it is a property of École Polytechnique de Montréal. Dragon contains a collection of models at each level which can describe the neutronic behavior in a given geometry of a unit cell, assembly or in a reactor core. To obtain the final solution it is necessary to link together different modules at each step and any compromise at any level can lead to poor final result. For a nuclear engineer it is crucial to maintain the accuracy with the reduced computational time. In the past [1] we have analyzed the advanced self shielding models that have been incorporated in the Dragon code Version 4 however the final verdict was that the computation time in the burnup calculations was too long to use a code on a daily basis. With the additional research and analysis we have obtained the satisfied results that maintain the accuracy and reduce the computational time. In this paper the results will be presented and the results are compared to the reference results obtained with the Monte Carlo codes.

[1] D. Ćalić, M. Kromar, A. Trkov, Use of Monte Carlo and deterministic codes for calculation of plutonium radial distribution in a fuel cell, Nuclear Energy for New Europe 2011, Bovec, Slovenia, 2011






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 605

Analysis of tritium production in TRIGA Mark II reactor at JSI for needs of fusion research reactors

Anže Jazbec1, Gašper Žerovnik2, Luka Snoj2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

anze.jazbec@ijs.si

 

Tritium (T) and deuterium (D) constitute the fuel for the existing (e.g. JET) and future (e.g. ITER) fusion reactors. Commercial fusion reactors will produce tritium inside the vessel out of tritium breeding blanket that will contain lithium. Experimental reactors like ITER will have to obtain tritium from external sources. It is predicted that tritium consumption of ITER will be around 1.5 kg per year. Altogether, there is only about 4 kg of tritium available in nature globally, so it will have to be produced artificially. In the paper we examine the possibility of producing tritium in small experimental light water reactors such as the TRIGA Mark II at the Jožef Stefan Institute. We use the Monte Carlo neutron transport code MCNP to calculate production of T by irradiation of Li targets inside various irradiation channels in the TRIGA reactor.

We find that the tritium production is the largest if all irradiation channels in the reactor’s reflector are used for the irradiation of Li. When targets are inserted for irradiation, the reactivity decreases by 200 pcm.In the second part of the analysis we determined how long the reactor can operate with the current fuel supplies. Calculations were made with the TRIGLAV computer code. We found that TRIGA can operate at full power for at least 2860 days, during which 151 mg of tritium could be produced. We conclude that small TRIGA reactors can not produce significant quantities of T required by the future fusion reactors.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 606

Analysis of a void reactivity coefficient of the JSI TRIGA Mark II reactor

Anže Jazbec1, Luka Snoj2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

anze.jazbec@ijs.si

 

The TRIGA reactor at Jozef Stefan Institute is among others used for training of students and trainees at the Nuclear training centre. They perform various practical exercises in the field of reactor physics.

In 2012 we decided to establish some new exercises and improve the existing ones. Currently the exercises and experiments are treated as temporary modifications of the reactor. In order to get approval from the Slovenian Nuclear Safety Administration for the modification, a safety assessment has to be performed. In the paper we describe the safety assessment for the experiment entitled measurement of the void reactivity coefficient. In the past this measurements were done by inserting a 6 mm thick and ~30 cm long Al rod into the core to simulate void and measure the change in reactivity. This was then changed by establishing a system for producing air bubbles in the core. The safety assessment of installation of the system was done by calculating the void reactivity coefficient to make sure it is negative, by analyzing the doses and power transient in case of system failures. We used the Monte Carlo neutron transport code MCNP to simulate and to calculate the activation of air going through the core. The results show that the system was safe and that its operation under normal and accident conditions does not lead to violation of operational limits and conditions of the reactor.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 609

Capabilities of TRANSURANUS code in simulating inception of melting in FBR MOX fuel

Davide Rozzia1, Alessandro Del Nevo2, Alessandro Ardizzone3, Mariano Tarantino2, Pietro Agostini2

Universita degli Studi di Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione, Largo Lucio Lazzerino 1, 56100 Pisa, Italy1

ENEA CR Brasimone, Localita Brasimone, 40032 Camugnano (BO), Italy2

Dipartimento di Energetica, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino, Italy3

daviderozzia@libero.it

 

The capability of the fuel to operate at high power without melting is of interest to FBR's because reactor design limits normally require that there be a low probability of fuel melting during steady-state operation including overpower conditions. This requirement has a direct effect on the steady-state power limit of the fuel pin and the reactor. Optimization of this power capability is important to reactor thermal efficiency and economy of operation.

The HELD-P-19 experiment was conducted during ‘70 in the EBR-II reactor to investigate the effects of initial fuel-to-cladding diametral gap sizes, from 0.086 to 0.25 mm, on the linear-heat-rate needed to cause incipient fuel melting at beginning-of-life. The test included 16 FBR fresh MOX fuel rods clad with cold worked type 316 stainless. The aim of the activity is to summarize the main results obtained after the simulations of 16 FBR fuel rods included in the above mentioned database by means of TRANSURANUS code. Particular emphasis is given to the main variables which influence the prediction of the fuel temperature profile and its related phenomena during power excursion. The activity is based on the comparison between experimental data collected at the end of the test and simulated trends (i.e. axial and radial extent of melting, columnar grain radius, central void formation, gap size). The importance of the sensitivity analysis, as tool to address the relevance selected parameters and code models on the results is discussed as well as the impact of the knowledge of the boundary conditions in the simulations.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 612

Uncertainty Analysis of Infinite Homogeneous Lead and Sodium Cooled Fast Reactors at Beginning of Life

Risto Vanhanen

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

risto.vanhanen@aalto.fi

 

The proven uranium reserves and estimated resources are enough to fuel the current open fuel cycle for 250 years. The thermal reactors produce minor actinides, which are more radiotoxic than natural uranium for hundreds of thousands of years. The same amount of uranium would last for 16 000 - 19 000 years with breeder reactors and closed fuel cycle. Fast breeder reactors do not produce any minor actinides and could also transmute the already produced minor actinides.

The present work concentrates on lead and sodium cooled fast breeder reactors. In addition to energy production fast breeder reactors have two main functions: in the long term, production of fissile material, and in the short term, transmutation of minor actinides. Compared to thermal reactors the materials in fast reactors suffer from high radiation damage.The objective of the present work is to estimate breeding ratio, minor actinide transmutation rate and radiation damage rate of infinite homogeneous lead and sodium cooled fast reactors. Uncertainty analysis is performed to estimate uncertainties caused by nuclear data uncertainty and uncertainty in the composition of the reactors. Uncertainty caused by the computational method itself is not taken into account. We restrict the work to beginning of reactor life.ENDF/B-VII.1 nuclear data is converted into multigroup form using NJOY nuclear data processing system. The Bondarenko method is used to acquire effective cross sections for the homogeneous reactor. The method relies on narrow resonance approximation, which is well valid within fast energy spectrum. However, the resonance interference effect is lost.The adjoint sensitivity analysis is performed to calculate generalized adjoint fluxes for the quantities of interest. The generalized adjoint fluxes are used to calculate first order sensitivities of the quantities of interest to input data. The acquired sensitivities are used to propagate uncertainties in the input data to find out uncertainties in the quantities of interest.Typically only the uncertainties in nuclear data is taken into account and the presumably small uncertainty of reactor composition is neglected. We expect the uncertainty in nuclear data to dominate the uncertainty in the composition of the reactors but calculate it to quantify its magnitude.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 613

Assessment of Double Heterogeneity Treatment Capability in SCALE

Gwanyoung Kim, Jin-Hyuck Kim, Changwook Huh

Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu , Daejeon 305-338, South Korea

k722kgy@kins.re.kr

 

A HTGR fuel is composed of a graphite matrix and a large number of tiny tri structural-isotropic (TRISO) fuel particlesembedded randomly in a graphite matrix. And each particle consists of a fuel kernel in the center, coated with several layers. This double heterogeneity (DH) provides a difficulty to process cross sections and geometry, and it requires methods different from those used for a typical LWR fuel.

In recent years, the SCALE system, which has been used for various light water reactor applications, has introduced numerical techniques and approximations that handle of problems specific to HTGRs including the treatment of the double heterogeneity of the fuel. In order to utilize the SCALE as HTGR core analysis code, the validation for the DH treatment capabilityhas been validatedmany times, and most validations have been carried out by comparing SCALE calculations with Monte-Carlo calculations based on the geometric modelling of repeated and latticed fuel particles. However, that modelling is far from the realistic and results in incorrect reference values. In order to correctly assess the DH treatment capability in SCALE system, it should be compared with Monte-Carlo calculation based on the modelling of the explicitly random-distributed fuel particles which is more close to real reference. In this study, the McCARD code, which is Monte-Carlo neutron-photon transport simulation code developed by Seoul National University, is utilized to obtain reference values based on the geometric modelling of the explicitly random-distributed fuel particles. For assessing the intrinsic DH treatment effect, the DH effect factor, which is calculated from a DH model calculation and a homogeneous model calculation, is introduced and the DH treatment capability in SCALE system is assessed by comparing the DH effect factors from SCALE calculations with those from McCARD calculations. For the wide-ranging assessment, the assessment is carried out from pin problems to assembly problems for various packing fractions. The results will be included in the full paper.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 614

Using TRIGA Mark-II Research Reactor for Irradiation with Thermal Neutrons

Aljaž Kolšek, Vladimir Radulović, Luka Snoj

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

aljaz.kolsek@gmail.com

 

The TRIGA Mark II research reactor at the Jozef Stefan Institute features several ex-core irradiation facilities that can be used for different applications, e.g. neutron radiography, radiation damage studies, etc. Recently a few test irradiations were performed for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. The percentage of fast neutrons must be very low, in the range of 0.01 % and the thermal neutron fluence should be about 1E15 neutrons/cm2s.

The MCNP Monte Carlo neutron transport code is used to calculate the neutron fluxes and spectra in the major TRIGA ex-core irradiation facilities. The Radial Beam Port (RBP) and the Radial Piercing Port (RPP) extend radially from the outer and inner boundaries of the graphite reflector, while the Tangential Channel (TangCh) and the Elevated Piercing Port (EPP) are tangential to the reactor core. The Thermal Column, a graphite stack that extends from the graphite reflector to the outer concrete wall of the reactor, thermalizes the neutrons leaking from the reactor core.In the paper the design and optimization of an irradiation device, placed in one of the TRIGA reactor’s ex-core irradiation facilities, is presented. Calculations of neutron flux and energy spectra were made in the irradiation facilities to optimize the position of the device. Optimal conditions were found in the thermal column, while the irradiation channels can’t be used for our application. An irradiation device, consisting of an aluminium vessel filled with heavy water and with the opening for sample insertion, was placed in the TRIGA Mark-II MCNP model. Finally, thermal to fast neutron ratio and neutron fluence was calculated in the irradiation position, using this irradiation device.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 615

MCNP Calculations Supporting the Start-Up of the New Core at the TRIGA Vienna

Thomas Stummer1, Rustam Khan2, Mario Villa1, Helmuth Böck1

Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna, Austria1

Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 44000, Pakistan2

stummer@ati.ac.at

 

During 2012 the TRIGA Mk II reactor in Vienna was converted from a highly heterogeneous core which included HEU fuel elements to an all LEU core. Because after 45 years of operation a simple exchange of the HEU with new LEU elements would have been insufficient to reach the necessary excess reactivity, most old pins were sent back to the US. As replacement 77 slightly used (burn up < 1.3%) 20% enriched stainless steel fuel pins were leased from DOE, which together with the 13 retained elements should ensure another 20 years of operation. The previously developed MCNP model was adapted for the new core and used to calculate the basic core load. As it turned out the reactivity was considerably overestimated. This paper presents a discussion of these results together with the measured values and changes in the neutron flux at several experimental positions compared to the 2011 values.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 616

Effect of fuel cooling on isotopic inventory of the NPP Krško fuel

Dušan Ćalić1, Marjan Kromar2, Andrej Trkov2

ZEL-EN razvojni center energetike, Hočevarjev trg 1, 8270 Krško, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

dusan.calic@ijs.si

 

Accurate determination of the isotopic composition is of outermost importance in the design calculations of the reactor core. Isotopic inventory is a function of burnup and operational parameters such as power, temperatures, water density etc. At the beginning of each fuel cycle reactor core is loaded with fresh and partially burned fuel assemblies. Employed burned fuel experiences cooling from a few weeks to several years, if fuel assemblies from previous cycles are used. In this paper the effect of cooling time is analyzed for a typical fuel used in the NPP Krško using models from CORD-2 package. Since the cooling can not be automatically performed in the current state of the CORD-2 package, a manual depletion calculation using a lattice code WIMS is executed. A sensitivity analysis of individual isotopes on the reactivity is performed. Findings will be used to enhance accuracy of the current isotopic library concept in the CORD-2 package.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 617

Active channel for silicon carbide composite testing

Martina Mala, Otakar Frybort, Michal Koleska

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

mlm@cvrez.cz

 

Currently new types of fuel cladding mainly for Gas-Cooled Fast Reactors are being developed and studied. Among all these experiments and studies there are not all necessary data. That was the reason to start a project oriented on silicon carbide testing in GFR conditions in the Research Centre Rez Ltd. The intention was to create a feasibility study and an engineering design of the active channel for nondestructive and destructive testing of fuel cladding made of silicon carbide composite. The channel was designed for utilization in the existing high temperature helium loop. The main parameters of the loop after modification are the maximal temperature 1000°C and the helium pressure 9 MPa, the thermal source for the testing is the core of the research reactor LVR-15. The new channel will be equipped with instrumentation for temperature measurement and pressurizing system with pressure sensors, which enable mechanical loading of the tested samples. The proposed materials of the channel and its internals are the stainless steel 08CH18N10T, Inconel 600 and SiCf/SiC. The paper describes all parts of the channel design, such as the 3D model of the active channel that served mainly for neutronic calculations in MCNP for two positions in the core of LVR-15, and thermo-hydraulic calculations in Fluent. The design of a test section for silicon carbide composite testing was a part of 3D model. The last part of the project was the strength calculations of proposed design in OpenFOAM.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 618

Computer environment integrating several deterministic transport codes distributed by the NEA DB

Ivan Alexander Kodeli, Slavko Slavič

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

slavko.slavic@ijs.si

 

Several high quality deterministic neutron transport codes are available from the OECD/NEA Data Bank, such as ANISN, DORT-TORT, DANTSYS, PARTISN etc. The development of many of these codes dates back in the 60-ies and 70-ies and use the computational standards of these times. Nowadays the use of these tools diminished largely, on one side due to the advances in the Monte Carlo codes, and on the other due to the archaic input formats used by these old transport codes. However, many possible applications of the deterministic codes still exist, be it for the sensitivity and uncertainty analyses, deep penetration problems and validation of Monte Carlo calculations.

The objective of the present paper is to present the development of a modern computer environment for the preparation and execution of these codes in a user friendly way adopted for today standards and types of users.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 619

Re-evaluation of the EURACOS Sodium and Iron Integral Shielding Experiments

Gašper Žerovnik1, Ivan Alexander Kodeli1, Andrej Trkov1, Anže Jazbec2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

gasper.zerovnik@ijs.si

 

A series of activation measurements with different detectors in a large irradiation chamber behind a large fission plate in the thermal column of the TRIGA Mark II reactor in Pavia, Italy, has been performed in the 1980s. Measurements have been done in three configurations: reference measurements in an empty irradiation chamber, and mesurements at different depths in the iron and sodium blocks. In addition to the activation measurements, proton recoil spectra, from which neutron spectra can be derived by unfolding procedures, have been measured at different positions in the metal shields.

The experiments were already used in the past for the validation of Na and Fe cross sections. The objective of the re-evaluation undertaken recently was to the determine if these benchmarks can be useful for the validation of the modern nuclear data evaluations. In the scope of the re-evaluation the older MCNP models of the experiments have been extended, refined and adopted to the recent MCNP version. New detectors have been introduced into the model, and variance reduction optimized. Also, tallies for neutron spectra calculations have been added. Sensitivity of the reaction rate distributions to different effects have been studied, with focus on the definition of the neutron source, which is the cause of the largest uncertianties of the experiments.The results show good agreement with the experiment only for the high-threshold S-32(n,p) reaction detectors. Comparison of the lower threshold Rh-103(n,n') and In-115(n,n') reactions reveals large discrepancy between the experimental and calculated spatial distributions for these reaction rates. On the other hand, the unfolded neutron spectra are in relatively good agreement with the calculated ones. As a conclusion, the EURACOS Na and Fe experiments cannot be considered as benchmark quality. The main problem is the uncertainty resulting from unsatisfactory information on the neutron source. In some cases, the associated uncertainty can reach an order of magnitude (up to a factor of 10). Nevertheless, some activation measurements from the EURACOS sodium and iron experiments such as the high threshold reactions and the spectra measurements are suitable for modern nuclear data and computer code benchmarking.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 620

Quality Asssessment of Evaluated Experiments Nesdip-2, Nesdip-3, Janus-1, Janus-8

Alberto Milocco1, Andrej Trkov2, Ivan Alexander Kodeli2

Universita degli Studi di Milano-Bicocca, Piazza dell'Ateneo Nuovo, 1, 20126 Milano, Italy1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

alberto.milocco@mib.infn.it

 

The paper describes the re-analysis of the benchmark experiments Nesdip2, Nesdip3, Janus1 and Janus8 performed in the scope of the joint OECD/NEA Data Bank and ORNL/RSICC international project SINBAD, the Shielding Integral Benchmark Archive Database. The project includes the experimental data (radiation shielding and dosimetry) and the computational models relative to integral benchmark experiments relevant for shielding applications. The database proves to be very useful for benchmarking evaluated nuclear data and computational models.

A discussion has been initiated in the SINBAD community to establish the criteria for assessing the quality of the experiments. A preliminary attempt has been carried out in the fusion neutronics section of the database that led to a final note on the quality of the experiments. The present extension of such commitment to the reactor physics section of SINBAD is the occasion for suggesting or revising the criteria that allow to rank a shielding experiment as of benchmark quality or not.The experiments Nesdip2, Nesdip3, Janus1 and Janus8 were performed in the 80ies in the experimental facility ASPIS at the NESTOR research reactor in Winfrith, UK, in the framework of an extensive program for the improvement of the RPV surveillance dosimetry. An independent review of the experimental information is carried out to identify which experimental data are incomplete, inconsistent or inaccurate. The uncertainties associated with the use of the experimental data in the benchmark models (benchmark data) are evaluated or even re-evaluated if original assessment is considered questionable. A suite of computational models have been developed for the MCNPX(5) code .they are for the benchmark analysis of the experiments.The conclusions consist of a short note on the quality of each experiment that provides an easy and quick interpretation from the side of the SINBAD users.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 621

Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0

Mario Matijević, Dubravko Pevec, Krešimir Trontl

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

mario.matijevic@fer.hr

 

A 3D simulation model of typical pressurized water reactor (PWR) primary loop components for effective dose rates calculation based on hybrid deterministic-stochastic methodology was created. Shielding calculations have been performed using MAVRIC/MONACO sequence of SCALE6.0 code package. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility are based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed dose rates calculations of reactor pressure vessel (RPV) and upper reactor head irradiation have been performed. The (n,p) nuclear reactions of fast neutrons with selected isotopes of carbon steel (RPV) and with oxygen (coolant) were examined. The efficiency of particle transport for obtaining global Monte Carlo distributions was further examined and quantified with flexible adjoint source positioning in phase space. Shielding calculations were also performed for reduced PWR facility model which included reactor core and adjacent concrete structures with steam generator and pump. It was demonstrated that definition of air as a global adjoint source gave lower uncertainty of SCALE6.0 photon transmission simulation. Activation of primary loop coolant, which becomes additional gamma source in working reactor, was considered. The gamma dose rates were presented for brief and detailed MAVRIC calculations. Advantage of using FW-CADIS methodology over CADIS and manual variance reduction techniques inside the MAVRIC shielding sequence was quantified. Control and numerical parameters of the MAVRIC shielding sequence with hybrid methodology have been optimized for effective Monte Carlo deep penetration shielding calculation.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 622

Time to Boil – Software for Bulk T&H of SFP

Rok Bizjak1, Slavko Slavič2, Dejvi Kadivnik1, Marjan Kromar2, Bojan Kurinčič1

Nuklearna elektrarna Krško, Vrbina 12, 8270 Krško, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

robidz@gmail.com

 

To support Nuclear Power Plant (NPP) Krško operation after event in Fukushima Daiichi (2011) the need was expressed to develop the software that would be able to anticipate (via calculation) thermo-hydraulic degradation of Spent Fuel Pool (SFP) in NPP Krsko following loss of power to heat exchangers (loss of cooling capability) or rupture of the pool (loss of cooling inventory) in real time.

For that purpose the computer code called “Time to Boil”, that uses exact plant data, was developed in collaboration between NPP Krsko and Institute Josef Stefan. It is an iteration based calculation software that takes into account exact geometry and material information, decay heat of spent fuel assemblies, SFP water temperature and levels, the location and size of a potential SFP crack and calculates times needed for water to reach its boiling point and evaporate to certain level.The program is now used to predict the thermo-hydraulic phenomena in NEK SF pool in the case of a beyond design condition or how much time operators have to implement predetermined recovery actions.The purpose of this paper is to explain the functional design, the capability and limitations of the code and to demonstrate how it works in some specific cases.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 624

The rod-insertion technique at the TRIGA reactor using signal from multiple fission cells

Igor Lengar1, Vid Merljak2

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia1

Fakulteta za matematiko in fiziko, Jadranska 19, 1111 LJUBLJANA, Slovenia2

igor.lengar@ijs.si

 

A digital meter of reactivity is applied for the measurements of physical parameters of the reactor cores of the TRIGA reactor and in the Nuclear power plant Krško (NEK). One of the outstanding features is the measurement of the control rod cluster worth with the rod-insertion method.

The DMR currently uses uncompensated ionization cells in order to obtain the neutron flux signal. The drawback of such measurements is that noise due to gamma rays from activation and fission products is added to the signal. Measurements with the DMR can be significantly improved by using the signal from fission cells, sensitive to neutrons only. In this way the gamma background subtraction, performed by DMR algorithms, can be omitted. The correctness of the new measurements is verified and with their help the current algorithms. At the TRIGA reactor only one ionization cell is currently used for flux measurements. During the insertion of one control rod the neutron flux distribution is significantly altered affecting the flux measurements of inserting different control rods. The problem is presently solved by assigning a correction factor to each control rod what introduces an additional uncertainty.The larger number of detectors will reduce the flux redistribution effects on the signal during individual control rod movements. The reduction of the error in the new measurements is analyzed.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 625

Base Irradiation Simulation and its Effect on Fuel Behavior Prediction by TRANSURANUS Code: Application to RIA Condition

Oleksandr Lisovyy, Marco Cherubini, Davide Lazzerini, Francesco Saverio D`Auria

University of Pisa, San Piero a Grado Nuclear Research Group (GRNSPG), Via Livornese 1291, 56122 Pisa, Italy

o.lisovyy@gmail.com

 

The purpose of the present paper is to investigate the impact of the base irradiation simulation when fuel behavior under RIA conditions is going to be predicted. A RIA is a scenario challenging the fuel integrity and consequently, devoted experimental campaigns and related code simulations have been extensively performed. In all experiments in which irradiated fuel is tested, the experiment is preceded by in reactor period, i.e. the base irradiation. In the present paper the considered RIA experiment is CIP3-1 performed in CABRI reactor (part of the OECD/NEA WGFS benchmark), a discussion about the relevance of the base irradiation simulation is presented. Such a work is conducted by sensitivities calculation in which a single parameter, among a preselected set, is changed. The range of variation of such parameters is either supplied within the selected RIA test specification or is taken from typical values available in the open literature. All mentioned calculations have been performed developing a specific model in TRANSURANUS code.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 626

Validation and Verification of WIMS/TRACE/PARCS Code System for Westinghouse AP1000TM Nuclear Reactor

Mohamed A. Elsawi, Amal S. Bin Hraiz

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

mohamed.elsawi@kustar.ac.ae

 

The purpose of this paper is to report on the validation and verification activities, currently underway at Khalifa University (KUSTAR), of the U.S. NRC coupled code system TRACE/PARCS for Westinghouse AP1000TM nuclear reactor, as a representative of modern pressurized water reactors. The cross section libraries were generated using the lattice physics code WIMS (release WIMS9A). Nine different assembly types were analyzed in WIMS to generate the two-group cross sections needed by the PARCS core simulator. The nine assemblies were identified based on the distribution of the discrete burnable absorbers (Borosilicate glass) and the integral fuel burnable absorbers (IFBA) in each fuel assembly. The generated cross sections were processed through the cross section interface program GenPMAXS, to generate the cross sections in the PMAXS format, recognized by the core simulator PARCS. The thermal-hydraulics feedback is handled through TRACE code with its live connection with PARCS. A detailed TRACE model of the AP1000 is constructed and the neutronics/thermal hydraulics data are exchanged between the PARCS and TRACE codes by defining a job stream inside the Symbolic Nuclear Analysis Package (SNAP) interface. Our work used the data available in the AP1000 Design Control Document (DCD) as a source of benchmark data to the coupled code system. In addition, Monte Carlo simulations, using SERPRNT code, with its unique reactor physics capabilities, including cross section generation for full-core simulations, was used to provide another source of benchmark data to the model developed using the WIMS/TRACE/PARCS code system. The work reported here is new in the following aspects: 1) it uses Westinghouse’ AP1000 TM reactor as a subject of the benchmark with its somewhat challenging core configuration (distributed discrete burnable absorbers and integral fuel burnable absorbers), and 2) SERPENT Monte Carlo model of the AP1000 is used to provide more benchmark data for the lattice and full-core physics parameters.






12.09.2013 10:40 Poster session 3

Reactor physics, fuel cycle and research reactors - 627

Possibilities of pin power density determination

Michal Kostal, Marie Svadlenkova, Jan Milcak, Vojtech Rypar, Vlastimil Juricek, Martina Mala

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

mlm@cvrez.cz

 

The pin wise density is an important quantity. It has to be monitored due to adequate assessment of fuel and core internals conditions. In case of VVER type reactors it is also important for pressure vessel radiation embrittlement calculations.

The paper is focused on the description of techniques used in LR-0 reactor operated by Research Center Rez Ltd. for power density determination.The power density can be dealt as fission density due to very low disproportionality between them. The fission density can be determined by means of gamma spectroscopy of irradiated fuel. The approach is based on proportionality between fission product activity and fission density. An experimental normalisation of the apparatus, which would allow an absolute determination of activities of selected fission products, is very difficult, if not impossible, due to the complex geometry of the source and mainly to the absence of standard with relevant gamma lines. For this reason, the calculated value should be directly net peak area (NPA) which can be measured by means of HPGe detector. The comparison between them is the direct calculation to experiment comparison. Due to the proportionality between power density and HPGe response the same rate of agreement as in NPA can be expected in case of power density. Net peak areas (NPA) of selected fission products induced in the fuel during its irradiation are measured by means of semiconductor gamma spectrometry with an HPGe coaxial detector in a streamline horizontal configuration (Ortec GEM70, resolution approximately 2.1 keV at E? = 1333 keV) and a multichannel analyzer DSA2000 (Canberra). The HPGe detector is placed in a thick Pb cylindrical shield with various types of collimator. Especially in axial measurement the thin collimator (2x1cm) has to be used. In case of high pin activity and resulting high level of the dead time, the Pb-Cu plate (3.3 mm of Pb followed by 1 mm of Cu) placed between the measured pin and the HPGe front end cap can be used for it suppression. Measured gamma spectra were analyzed with the Genie 2K software (Canberra).There are several nuclides which can be used as fission density monitors. Their application depends on irradiation conditions. The half lives of most dominating fission products is similar to the time of irradiation. Brief comparison between the evolution of short living (Te-134 with half life 41.8 minutes) and longer living (Zr-97 with 16.75 hours) isotopes during irradiation and after reactor shut down is presented in the paper. The short living isotopes reach the saturation level shortly after the irradiation startup, but also decay fast. For example if the irradiation time is reduced by half the induced Te-134 activity will be 75% of the activity after 150 minutes. But on the other hand the activity decays fast, thus they are suitable for the measurements of only for few pins. Oppositely the Zr-97 is suitable for much more pins, because before the number of emitted photons is reduced by half ~40 of pins can be measured, when expecting 15-30 minutes measuring time.






12.09.2013 10:40 Poster session 3

New reactor technologies - 701

Startup of “CANDLE” burnup in Gas-cooled Fast Reactor using Monte Carlo method

Mohsen Saadi Kheradmandsaadi

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

mohsen.kheradmand@gmail.com

 

During the past decade, the CANDLE burnup strategy has been proposed as an innovative fuel cycle and reactor design for complete utilization of uranium resources. In this strategy the shapes of neutron flux, nuclide densities and power density distribution remain constant but the burning region moves in axial direction. The feasibility of this strategy has been demonstrated widely by using the diffusion technique in conjunction with nuclide transmutation equations. On the other hand since the Monte Carlo method provides the exact solution to the neutron transport, the Monte Carlo technique is becoming more widely in routine burnup calculations. The main objective of this work is startup of CANDLE burnup in Gas cooled Fast Reactor using Monte Carlo burnup scheme. In this case only natural or depleted uranium is required for fresh fuel region. However, the construction of the first CANDLE core is faced with a big problem. In equilibrium state the burning region contains a spectrum of fission products as well as higher actinides. These isotopes are not easily available for constructing the initial CANDLE core. The solution is startup of a special reactor using the enriched uranium in starter zone. At the end of core life the fuel for the next core is produced with the composition close to the equilibrium state. An originally MCNP-ORIGEN linkage program named as MOBC has been used for criticality and isotopic evaluation of the core. The results of analysis showed that the use of burnable absorber rods for positive reactivity swing offset is mandatory. In this regard the multiplication factor changes are limited in acceptable margin where the maximum reactivity swing is 655 pcm during transient period and 180 pcm during equilibrium state. On the other hand, although the power shape changes drastically in the earlier stage of core operation but remains constant during the equilibrium state. In summary, by using the enriched uranium as a starter fuel the equilibrium state achievement is possible.






12.09.2013 10:40 Poster session 3

New reactor technologies - 703

Business Process Flexibility Analyse on Nuclear Power Plant New Build Project JEK2

Gregor Androjna, Aleš Buršič

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

gregor.androjna@gen-energija.si

 

The purpose of the business process flexibility analyse of Nuclear Power Plant new build project JEK2 is to point out the problem of continuous changes on project business processes triggered by the dynamic growth/reductions of various project activities through the different stages of project lifetime and answer the question, what type of information system (IS) could help manage that sufficiently and successfully enough.

Basic fact is that today we can’t manage the flood of electronic documents and information by “old fashion” procedures and processes on the paper. The information and document exchange is simply too fast and “uncontrolled” in the electronic world. Therefore we need to control the processes through the information system but the question is can we manage that by the conventional programs or we need to integrate some kind of Business Process Management system.To answer that, we need to analyse (systematize) one typical project process and analyse its flexibility in different IS. We will do that on the case of nuclear power plant construction project considered to be built as the second unit at Krško site (JEK2). Process analyse will include modelling of existing and new optimised process as well as analyse of communication and documentation flow resulting in proposition for its optimisation. Process flexibility analyse will stress out key identified or expected process and documentation changes and analyse modification methods of those in conventional IS (regular programing) or in Business Process Modelling System (“graphical’’ programing). It is expected that in case of BPMS the process management is more efficient and effective but can we step forward and accept (maintain) strong process management culture on such big project as building NPP or can we afford not to.






12.09.2013 10:40 Poster session 3

New reactor technologies - 705

Action Plan with Time Schedule of JEK2 Construction Site Arrangement

Aleš Kelhar, Klemen Debelak, Tomaž Ploj, Aleš Buršič, Danijel Levičar

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

ales.kelhar@gen-energija.si

 

Construction of a nuclear power plant in Slovenia is a unique project which requires effective, efficient and transparent process of preparation and construction of the site. After 30 years of successful operation of the first nuclear power plant (NEK), company GEN energija began the process of planning and preparatory work for construction of the second nuclear power plant (JEK2) in Slovenia. GEN energija combaines with other portfolio goals also maintenance and expansion of nuclear capability as the pillar of sustainable development in Slovenia.

An Action plan with a Time schedule will be presented, which includes all activities required for obtaining a final and binding Building Permit for pre-construction site preparation and all relevant activities providing due start of technological facilities construction.Presented Action Plan has been developed for the western (upstream location/site regarding existing NPP) alternative. The eastern (downstream) alternative was assessed similarly, since only a minor part of certain activities are excluded due to differences between locations. At the upstream site certain infrastructural re-arrangements are required which are not relevant at the downstream site; however, the downstream site is more demanding in context of excavations and relocation of excess dredging material.Activities relating to the pre-construction site preparation are divided into the following sets: Preliminary works, Construction of auxiliary buildings outside the JEK2 construction site area, Construction of residential units for workers accommodation, Phase arrangement of the platform, Phase construction of temporary structures, Performance of construction pit and construction of certain underground facilities.In the Time Schedule, performance of pre-construction site preparation activities for the western location will be shown. According to the overall JEK2 Time Schedule, a period of 3 years is foreseen for the listed scope of activities.






12.09.2013 10:40 Poster session 3

New reactor technologies - 706

Melting temperature of MOX fuel for FBR applications: TRANSURANUS modelling and experimental findings

Rolando Calabrese1, Arndt Schubert2, Jacques Van-De-Laar2, Paul Van Uffelen2

ENEA, Technical Unit for Reactor Safety and Fuel Cycle Methods , Via Martiri di Monte Sole 4, 40129 Bologna, Italy1

European Commission Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz Platz 1, D-76344 Eggenstein-Leopoldshafen, Germany2

rolando.calabrese@enea.it

 

Fuel melting temperature is a key parameter in the design and safety assessment of nuclear systems. This paper focuses on the solidus and liquidus temperatures of MOX fuel for FBR applications also considering the presence of minor actinides. In an advanced fuel cycle strategy, MA-bearing MOX fuel is capable of recycling TRUs accumulated in spent fuel. The discussion deals with an hypothesis of homogeneous recycling where MAs are diluted in the driver fuel up to a concentration of 5 wt.%.

Experimental data published in literature is reviewed and a dataset is compiled for validation purpose. Basing on this information, the accuracy of models available in the TRANSURANUS code as well as of other published models is discussed. 1n this step, attention is given to the choice of a conservative/best-estimate approach as well as to the uncertainties of experimental data and in-pile measurements. Finally, preliminary conclusions concerning the effect of MAs content on this important parameter are drawn.






12.09.2013 10:40 Poster session 3

New reactor technologies - 708

The 4th fundamental safety function for systematical evaluation of existing and new generation nuclear reactors: avoidance of fast exothermic chemical reactions

Riitta Kyrki-Rajamäki, Mariaana Talus

Lappeenranta University of Technology, P.O.Box 20, FI-53851 Lappeenranta, Finland

rkyrki@lut.fi

 

New materials and technological solutions are been used in the design of Generation IV reactors. This means that designing safety criteria for these reactors also needs new ideas. At this moment new technology neutral safety criteria are been developed. These general safety criteria could be used when developing technology specific safety criteria for all kinds of nuclear reactors, regardless of their technological solutions. By using this method high safety level in all new reactors can be guaranteed. Exothermic chemical reactions can't be neglected when the safety of nuclear reactors is been considered. By examining safety criteria used today and proposed technology neutral safety criteria, it can be seen that exothermic chemical reactions are already taken into account in safety criteria, but not in a systematic manner.

The aim of this report is to consider what should be the main principles when developing technology neutral safety criteria where exothermic chemical reactions are systematically been taken into account. It was found that the main principle should be that unwanted exothermic chemical reactions should be prevented, but if nevertheless an exothermic chemical reaction occurs, its consequences should be minimized. Additionally, exothermic chemical reactions can be prevented if there are no materials that can react exothermally, or if those materials can be kept apart. Another way of preventing exothermic chemical reactions is to keep temperatures so low that they cannot occur. As examples a couple of Generation IV reactors are examined to look how exothermic reactions are taken into account in their design, and what sort of accident scenarios would lead to an occurrence of exothermic chemical reactions.






12.09.2013 10:40 Poster session 3

Radioactive waste management - 901

Comparative Study of Japanaise and Serbian Bentonite on the Fraction of 137CS from Cement-Ion Exchange Resins-Bentonite Clay Composition

Ilija Plećaš

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

iplecas@vinca.rs

 

To assess the safety of disposal of radioactive waste material in cement, curing conditions and time of leaching radionuclides 137Cs have been studied. Leaching tests in cement-ion exchange resins-bentonite matrix, were carried out in accordance with a method recommended by IAEA. Curing conditions and curing time prior to commencing the leaching test are critically important in leach studies since the extent of hydration of the cement materials determines how much hydration product develops and whether it is available to block the pore network, thereby reducing leaching. Incremental leaching rates Rn(cm/d) of 137 Cs from cement-ion exchange resins-bentonite matrix after 60 days were measured. In this paper we compared two bentonite clay as sorption componente, from Japan and Serbia. The results presented in this paper are examples of results obtained in a 30-year concrete testing project which will influence the design of the engineer trenches system for future central Serbian radioactive waste storing center.

Key words : Bentonite, Cement, Radioactive Waste, Radionuclide, Leaching, concrete






12.09.2013 10:40 Poster session 3

Radioactive waste management - 903

The possibility of the recycling of minor actinides in pressurized water nuclear reactor

Rok Rožman1, Andrej Trkov2, Gašper Žerovnik2

GEN energija d.o.o., Vrbina 17, 8270 Krško, Slovenia1

Institut "Jožef Stefan", Jamova cesta 39, 1000 Ljubljana, Slovenia2

rok.namzor@gmail.com

 

We study the possibility of the recycling of the spent nuclear fuel in pressurized water nuclear reactors such as the one at the Krško Nuclear Power Plant (NEK). We focus on the recycling of irradiated nuclear fuel and the production of the MOX fuel mixtures. We used a new method, where we introduced two separate regions for the minor actinides and the plutonium within the MOX fuel pellets, aiming to increase the efficiency of recycling. For calculation of the burnup and isotopic composition of the spent fuel WIMSD-5B program [1] was used. Comparison of the reactivity change to the standard UO2 fuel shows significant improvement of the neutron economy. A rough estimation of how much of the actinides is it possible to burn in a pressurized water nuclear reactor is given. With our new method the total mass of minor actinides is reduced by a factor of 2.7 after 37 years of operation compared to the spent standard once-through UO2 fuel after equivalent energy production. Additionally, the natural uranium consumption is reduced by around 1/5.

[1] WIMSD-5B, WIMSD: A Neutronics Code for Standard Lattice Physics Analysis, (AEA Technology, Distributed by the NEA Data Bank, Answers Software Service, United Kingdom, 1997)






12.09.2013 10:40 Poster session 3

Radioactive waste management - 905

Comparison of costs for LILW repository development

Nadja Železnik1, Vladimir Lokner2, Ivica Levanat2

Regionalni center za okolje za srednjo in vzhodno Evropo , Slovenska cesta 5, 1000 Ljubljana, Slovenia1

Agencija za posebni otpad, Savska cesta 41/IV, 10000 Zagreb, Croatia2

nadja.zeleznik@rec-lj.si

 

In the last few years several documents on Low and Intermediate level radioactive waste (LILW) repository have been prepared in Slovenia and in Croatia. Especially in Slovenia due to the adoption of the Governmental decree on spatial planning act of national importance for LILW repository at Vrbina site in Krško municipality, many documents were elaborated, including investment program in several revisions. Also in Croatia, because there have been no agreement on joint LILW repository for NPP Krško radioactive waste between Republic of Slovenia and Croatia, basic investigations on the costs assessment for development of national LILW repository have been performed. Additionally, in the revision 2 of Program of NPP Krško decommissioning and spent fuel (SF) and LILW disposal 5 different scenarios on radioactive waste management (RWM) were elaborated including all possible options of joint or separate RWM, were assessed and analysed in order to obtain real costs for development of LILW repository.

The paper will present the costs of LILW repository development in Slovenia and in Croatia, including expenses for site selection, construction, trial and normal operation, and finally closure. It will also describe the option of one joint repository which is not currently selected and the impact on the costs. Different contributions to costs assessment will be presented (like construction, operation, compensations, contingency and VAT) and their relative contribution to the overall costs. The comparison will be made between different concepts (surface, silo, underground), scenarios and the influences of different factor on the expenses will be presented. At the end also international comparison will be provided based on the NEA study which included analyses and comparison of construction and operating costs for LILW surface and underground repositories for both operating and planned ones.






12.09.2013 10:40 Poster session 3

Radioactive waste management - 908

The Development of Bentonite Gap Filling for High-Level Waste Disposal

Jiri Stastka

Czech Technical University in Prague, Faculty of Civil Engineering, Centre of Experimental Geotechnics, Thákurova 7, CZ-16629 Prague 6, Czech Republic

jiri.stastka@fsv.cvut.cz

 

The main aim of the repository disposal system is to prevent the migration of radionuclides into the biosphere. The natural rock barrier as well as the engineered barrier elements of the system should guarantee the safe disposal of radioactive waste for the required time period. Bentonite clay has proven to be particularly suitable in this respect thanks to its sealing properties (very low permeability in the case of both water and gases). The buffer, which makes up a crucial part of the geotechnical barrier, surrounds the container in the disposal well. The main requirements in terms of the geotechnical properties of the buffer material consist of very low hydraulic conductivity (max 10-12 m/s), high swelling ability (SW> 1MPa), rheological stability, high plasticity, and good thermal conductivity.

The article presents the results of research into the most important geotechnical properties of Czech Ca-Mg bentonites and provides an outline of both the design of the technology and the experimental tests employed to simulate the filling of the gap between the buffer and the rock massif in the disposal hole. The materials used in the gap-filling tests included pellets, granules, powder and mixtures thereof. The means of application studied consisted of the free fall pouring method, with and without vibration, and the spray method and the aim was to verify the way in which they worked with the various types and forms of material. The most important geotechnical parameter of the layer, i.e. dry density, was determined following the application of the materials.It is hoped that the final design of gap-filling technology will be determined from the results of the mock-up testing of the disposal location. Mock-up tests reveal the impact of the dimensions of the gap on the most important geotechnical properties of the material. The article also presents geotechnical results gained from in-situ mock-up model testing (Mock-up Josef, Bentonite 95).






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1302

Sustainability and Nuclear Energy: Main Concepts and the Analytic Network Process (ANP) Approach

Rolando Calabrese

ENEA, Technical Unit for Reactor Safety and Fuel Cycle Methods , Via Martiri di Monte Sole 4, 40129 Bologna, Italy

rolando.calabrese@enea.it

 

The concept of sustainable development addresses the capability to meet present needs without compromising the ability of future generations to meet their own needs. Sustainability is therefore a fundamental approach in the definition of future energy systems to accomplish economical, social, and environmental criteria.

If on the one hand nuclear energy has great capabilities in tackling GHG emissions and human-induced climate changes, on the other hand, this energy source faces drawbacks such as the shortage of natural uranium resources, the steadily increase of spent nuclear fuel stockpiles as well as proliferation risks. To overcome these limitations and pursuing a trade-off between safety and economics, new plant designs are under investigation worldwide. The majority of studied systems operate under fast neutron spectrum. Their introduction should take place towards the middle of the century in parallel with the transition to innovative fuel cycles where plutonium and minor actinides produced under irradiation are recovered and recycled.In the paper, main published results on sustainability and nuclear energy are reviewed and discussed also basing on the calculations of simplified scenarios performed by means of different codes such as DESAE 2.2, NFCSS and MESSAGE.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1303

Nuclear Experts' Perception of Lay Attitudes toward Nuclear Issues

Marko Polič1, Nadja Železnik2, Drago Kos3

Filozofska fakulteta, Aškerčeva cesta 2, 1000 Ljubljana, Slovenia1

Regionalni center za okolje za srednjo in vzhodno Evropo , Slovenska cesta 5, 1000 Ljubljana, Slovenia2

Fakulteta za družbene vede, Kardeljeva ploščad 5, 1000 Ljubljana, Slovenia3

nadja.zeleznik@rec-lj.si

 

From the previous research performed during last years it became clear that the attitudes, opinions and perception regarding nuclear issues in the broadest sense including topics like radioactivity and properties of radiation, influence of ionizing radiation on humans, functioning of nuclear facilities and riskiness of support activities (e.g. transport) differ very much between nuclear experts and lay people. As nuclear experts should be involved in communication with the public on nuclear topics it is important to understand their understanding of the public opinions, because this could influence their attitudes toward the general public and their way of communication with them.

Study presents nuclear experts' perception of lay public attitudes toward nuclear issues and reasons behind them. With the help of internet survey within the members of Nuclear Society of Slovenia the opinions of Slovenian nuclear experts were collected. Especial emphasize was devoted to the consequence of existing experts' opinions regarding communication with lay public. Survey contain questions about different aspects of lay public knowledge and understanding of nuclear issues, ways of communication with the public, attitudes toward public participation in the decision processes, perception of the experts' role in the process, etc.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1304

SWOT Analysis of NMS Participation in Euratom Projects

Nadja Železnik1, Metka Kralj2

Regionalni center za okolje za srednjo in vzhodno Evropo , Slovenska cesta 5, 1000 Ljubljana, Slovenia1

Agencija RS za radioaktivne odpadke, Celovška cesta 182, 1000 Ljubljana, Slovenia2

nadja.zeleznik@rec-lj.si

 

SWOT (Strengths, Weaknesses, Opportunities and Treats) analysis was performed by partners from EU new members states (NMS) in the NEWLANCER project with the aim to provide relevant information for development of the national policies to increase the participation in Euratom programs. SWOT analysis was performed by national expert groups participating in the project and for specific fields of interest in each country like nuclear safety, new reactor generations, radiation protection, radioactive waste management and education and training. Altogether 18 SWOT reports were prepared using common methodology in 6 NMS: Bulgaria, Hungary, Lithuania, Poland Romania and Slovenia.

SWOT analyses showed some common issues in all participating NMS and in all fields regarding their participation in Euratom projects:1. In most countries the number of good experts is not sufficient and is even decreasing due to retirement and insufficient interest of young and talented students for nuclear field. This influence the ability to participate in research projects.2. Investment in the nuclear technology field is appreciated as important factor for promotion of participation in Euratom programs. Some countries have current investments in nuclear research and technology, and also some kind of stable funding provided by special funds, but the risk of sustaining these opportunities in future is great.3. Lack of systematic planning on institutional and national level and bad management in research institutions is the most important weakness/threat described. Consequently, the competitiveness of institutions and research groups from NMS in Euratom tenders is reduced.4. Topics of Euratom tenders are not always consistent with the interests and needs of NMS, NMS with small nuclear program could manage to participate if they were able to participate in smaller projects.Final objective of SWOT analyses in NMS was to propose strategies to reduce influence of identified negative factors and to enhance influence of identified positive factors in respective NMS, and in NMS in general. The results will be presented in the paper.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1306

Public Opinion about Nuclear Energy – Year 2013 Poll

Radko Istenič, Igor Jenčič

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

radko.istenic@ijs.si

 

Public information is one of the important activities of the Nuclear Training Centre at the Jožef Stefan Institute. This year marks the twentieth anniversary of the Information Centre that was established within the Nuclear Training Centre to inform the visitors about nuclear power and nuclear technology in general and about Krško Nuclear Power Plant.

The main target group of information activity are schoolchildren with their teachers. Most of them are from the 8th and 9th grade of elementary school, age 14 to 15. Every year some 8000 youngsters visit the Information Centre. The visit consists of a live lecture about nuclear technology followed by the demonstration of radioactivity and a guided tour of permanent exhibition. Since 1993 we monitor the opinion trends by polling about 1000 youngsters every year. The youngsters are polled before they listen to the lecture or visit the exhibition in order to obtain their opinion based on the knowledge from everyday life. In the paper we will present, summarize and comment the trends over the last 20 years.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1307

Knowledge management - a strategically important task of the nuclear industry

Anna Kazakova, Marina Lapshina, Evgeny Kapralov

Training and Methodological Centre for Nuclear and Radiation Safety, 8 Parkovaya St. 9/15, 105043 Moscow, Russian Federation

marina_tcnrs@mail.ru

 

Authorizing documents, Standards of organizations and other regulatory documents contributing to the perpetuation of hidden knowledge are developed by the «Training and Methodological Center of Nuclear and Radiation Safety» Non-state Educational Institute (TMC NRS NEI) in order to solve the problem mentioned above. As knowledge management system is a subsystem of the Quality Management System (QMS), documents under development are documents of QMS of the organization.

One of the developed and implemented documents since last year is the Authorizing Document of safety expertise "Basics of organization and carrying out of expert examination of designing, engineering, technological documentation and documents substantiating the ensuring of nuclear and radiation safety of nuclear installations, radioactive waste storage and activity in radioactive waste management". Being one of the storage methods of hidden knowledge this document contains information and experience accumulated from practice expert examination for years in Russia. There are up-to-date requirements in the Authorizing Document that are demanded to the process of safety expertise, therefore we can consider this document to be the storage of the explicit knowledge.Accordingly, using standard procedures described in the Authority Document of safety expertise, a young specialist having engineering education but without the operational experience in nuclear power plants can carry out expert examination in safety in the field of atomic energy.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1309

Travelling exhibition Fusion Expo

Tomaž Skobe

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

tomaz.skobe@ijs.si

 

The Fusion Expo is a travelling exhibition where visitors could explore fusion energy. Fusion expo is presenting fusion energy as an environmentally acceptable, safe and sustainable energy source. Fusion research, technology and its future use is presented to the citizen of Europe.

Main target group of this exhibition are mediators such as journalists, teachers, decision makers, NGO’s, students, taxpayers, voters, primary and secondary school pupils.Since beginning of Fusion Expo support actions under EFDA in 2008, the Fusion Expo has been the responsibility of the Slovenian Fusion Association (SFA). Since then we are trying to bring it to visitors on most interactive way.The paper will present past experiences, current activities and challenges of Fusion Expo support actions under EFDA(European Fusion Development Agreement).






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1311

The Young in the World of Energy - Communication Project to Promote and Educate Energy-related Topics Among the Younger Generations

Katja Bogovič, Tanja Jarkovič, Melita Lenošek Kavčič, Garsia Kosinac

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

katja.bogovic@gen-energija.si

 

"The Young in the World of Energy" is a GEN energija-run communication project designed to promote the knowledge of energy and energy-related topics among the younger generations. The history of the project dates back to 2008, when the GEN Group launched its Energy-Efficient School project to encourage Slovenian schools to reduce their electricity consumption. Over the years the project grew and changed shape before finally evolving into the awareness-raising project "The Young in the World of Energy" with the opening of the interactive visitor centre The World of Energy in 2011.

Two contests are being held as part of the project in the 2012/2013 school year:• A nationwide contest for Slovenian primary and secondary schools, which is run in association with the Eco-School (“Ekošola”) programme and is designed to increase energy literacy and raise awareness of sustainable development in terms of energy and energy production; and• The quiz "Young Wizards” on Energy Technologies and Nuclear Energy, organized by GEN energija in partnership with Krško Nuclear Power Plant.The aim of the project is to encourage the young, through modern methods and media, to see the big picture behind energy and to gain an in-depth insight into the basic energy concepts, sustainable energy sources, climate change, radioactivity, and nuclear energy as a sustainable source of energy.The paper will present the activities taking place in the framework of the two contests, the communication media used, and the final outcomes, achievements and findings of the projects.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1312

New generation for better nuclear era

Geambasu Cristiana1, Stefania Popadiuc2

Ministry of European Funds, Bd. Mircea Voda, nr. 44, intrarea C, sector 3, Bucharest, Romania1

University “Politehnica” of Bucharest, 313 Splaiul Independentei Street, Sector 6, 060042 Bucharest, Romania2

ileana.geambasu@fonduri-ue.ro

 

Developing the next generation of nuclear reactor technology is an ambitious goal, even for countries with small nuclear energy research programs. The new generation designs will use fuel more efficiently, reduce waste production, be economically competitive and meet stringent standards of safety and proliferation resistance. Today, experts around the world are collaborating to further advance nuclear technology to meet future energy needs. That's why Romania has been working with international partners to coordinate efforts, resources and schedules to achieve success.

The future designs will use fuel more efficiently, reduce waste production, be economically competitive and meet stringent standards of safety and proliferation resistance.For more than a decade, international collaborative efforts have been developed next-generation nuclear energy systems that can help meet the world's future energy needs. The advanced systems are designed to meet four overarching goals: sustainability, safety and reliability, economic competitiveness, and proliferation resistance/physical protection. More specifically, our Romanian goals for the next 20 years are to:• provide sustainable energy generation that meets clean energy objectives, promotes long-term availability of systems and utilizes fuel more effectively• minimize nuclear waste and reduce long term stewardship burden• excel in safety and reliability• have a level of financial risk comparable to other energy projectsWith these goals in mind, some experts evaluated the current situation for further research and development, in order to recycle fissionable material and produce less nuclear waste. In University POLITEHNICA of Bucharest some projects are dealing with co-generate heat that could be used for industrial processes. They represent the common research effort of the professors and students together.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1313

Radioactivity experiments for schools

Matjaž Koželj, Radko Istenič

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

matjaz.kozelj@ijs.si

 

Every year thousands of pupils and students visit the Information Centre with a permanent exhibition on nuclear technology at Milan Čopič Nuclear Training Centre of the Jožef Stefan Institute in Brinje near Ljubljana. Most of their teachers are returning visitors who always ask for demonstration in the radioactivity workshop, which is a part of our exhibition. The workshop includes presentation of the basic properties of radiation and also demonstration of natural radioactivity and sources with elevated activity in the environment.

According to their statements, the workshop has become one of the most important reasons for their annual visit since it supports not only our lectures and other parts of exhibition, but also supplements regular school program. Many of teachers would like to have a possibility to perform similar demonstrations in schools, or to use it for introduction to research work, but due to limited information and knowledge, it is their opinion that the realisation of that idea is not possible or, at least, too difficult. In fact, it is possible to list a number of questions which confuse teacher and must be answered before he or she should seriously consider introduction of radioactivity demonstrations. First, what are legal requirements and boundaries for school radioactivity experiments? Second, how dangerous it is for me and my students, and am I going to have any problems? Where to get necessary equipment and sources and how much will it cost? What sources and equipment could be acquired for free? What are properties of those sources? What experiments and demonstrations could be done and where to get proper instructions? Where to get support, if needed?We have already answered some of these questions in informal dialogue with our visitors, but never in extensive and documented way. Therefore we have decided to review and discuss possibilities for practical experiments with radioactivity sources in schools, explain formal requirements for school sources and practices and what can legally be done without any licensing, list some available (and affordable) equipment, describe some available sources, suggest some experiments and demonstrations, and also present our experience with radioactivity demonstration in Milan Čopič Nuclear Training Centre.We hope that the contribution will encourage and enable at least some of teachers to introduce the experiments with radioactivity in the schools. It is our wish to enable and support broader and deeper understanding of natural phenomena and technology then it is delivered to our young generation in schools at the moment.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1315

Tomsk Polytechnic University Experience in Specialists’ Training for Global Nuclear Power Industry

Natalia Shepotenko, Yuliya Falkovich

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

shepotenko@tpu.ru

 

Upon entering WTO Russia has faced up the problem of training specialists who are competitive on the international labour market. Russian educational system has already undergone dramatic changes: Bachelor-Master degree Paradigm gains its momentum. Nevertheless, there is a doubt among Russian education managers whether the programs being developed meet the international standards, indeed. There is thus an urgent need to address the challenges posed in close cooperation with the international partners: Universities, Research organizations, International organizations, business and industrial enterprises.

The presentation will sketch some key dimensions of Tomsk Polytechnic University (TPU) experience in training specialists in a wide range of educational programs that embrace the full nuclear fuel cycle, fundamental and applied physics, nuclear chemistry as well as nuclear engineering services. The competencies that TPU alumni acquire will be presented, the main issue to be solved at TPU that are focused on specialists’ training will be discussed. Besides, the tools of international cooperation in terms of high standards educational programs development will be introduced.






12.09.2013 10:40 Poster session 3

Sustainability, education, training and public relations - 1318

Presentation of NEWLANCER Project of 7FP of EC and Bulgarian project’ activities

Pavlin Petkov Groudev1, Ivan Ivanov2, Petya Vryashkova1

Institute for Nuclear Research and Nuclear Energy, 72 Tzarigradsko chaussee, Blvd., BG-1784 Sofia, Bulgaria1

Technical University of Sofia Electrical Power Dept., 8 Kl. Ohridski Blvd., 1797 Sofia, Bulgaria2

pavlinpg@inrne.bas.bg

 

A C&SA project of EURATOM in the 7th Framework Programme of EC named NEWLANCER-New MS Linking for an AdvaNced Cohesion in Euratom Research is presented with the poster.

This project is started at November 2011 and devoted to find and implement rational and efficient actual solutions for enlarged of the NMS’ participation in the next EURATOM Programmes. The aim is to strengthening the using of the R&D potential of the organizations and cohesion with national institutions, in closer collaboration with Old Member States’ relevant research partners.

There are presented the major subjects and outcomes of the NEWLANCER, the created Bulgarian National Experts Groups (NExGs) and their tasks, activities and main results, including the results from SWOT analyses, on this stage of the project.






12.09.2013 11:20 Invited lecture 8

Invited lectures - 102

Nuclear Power as a Basis for Future Electricity Production in the World

Igor Pioro

University of Ontario, Faculty of Energy Systems and Nuclear Science, Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario L1H 7K4, Canada

igor.pioro@uoit.ca

 

It is well known that the electrical-power generation is the key factor for advances in any other industries, agriculture and level of living. In general, electrical energy can be generated by: 1) non-renewable-energy sources such as coal, natural gas, oil, and nuclear; and 2) renewable-energy sources such as hydro, wind, solar, biomass, geothermal and marine. However, the main sources for electrical-energy generation are: 1) thermal primary coal and secondary natural gas; 2) “large” hydro and 3) nuclear. The rest of the energy sources might have visible impact just in some countries. In addition, the renewable-energy sources, for example, such as wind and solar and some others, are not really reliable energy sources for industrial-power generation, because they depend on Mother nature and relative costs of electrical energy generated by these and some other renewable-energy sources with exception of large hydro-electric power plants can be significantly higher than those generated by non-renewable sources. Therefore, a general overview will be provided of various power plants of the world with emphasis on nuclear energy.






12.09.2013 12:00 New reactor technologies

New reactor technologies - 702

Main Safety Features of the ATMEA1 Reactor and Robustness to Extreme Situations

Eric Mathet, Castello Gerard, Blassel Benoit

ATMEA SAS, Tour AREVA, 1 Place Jean MILLIER, 92084 Paris La Defense, Paris, France

gerard.castello@atmea-sas.com

 

ATMEA1 is an evolutionary GEN III+ Pressurized Water Reactor (PWR) designed by ATMEA, furnished with a three-loop coolant system rated at a thermal power of 3150 MWth. Developed primarily based on the US regulations its safety options have been have reviewed by the French safety authorities who positively concluded that they are compliant with the French safety regulation.

Special care has been taken for protecting the safety systems in the containment, safeguard buildings, fuel building and emergency power source buildings against a wide range of external hazards.Following the Fukushima accident the ATMEA Company has taken a proactive approach to take into account all lessons learnt made available on the accident. An internal safety check has been conducted, and safety demonstration methodology for Beyond Design Basis External Events has been developed. This assessment has until now not revealed any necessary change in the design.As far as seismic level is concerned, the Basic Design of ATMEA1 was done with a peak ground acceleration of 0.3g and a spectrum enveloping the EUR requirements, based on US-NRC guidance. One key feature of the ATMEA1 design to achieve such earthquake resistance, for example, is to put the main buildings housing safety related components on the same large and thick concrete slab and to implement snubbers or equivalent equipment at components’ anchoring to reduce vibrations. Should a site yield possibly higher peak ground accelerations, this large concrete slab itself could be put on seismic isolation pads to reduce even further the effects of an earthquake on the structures and equipment of the plant.After a short summary of the main safety features of the ATMEA1 reactor, the paper will focus on how the design answers and copes with extreme external hazards situation.






12.09.2013 12:20 New reactor technologies

New reactor technologies - 707

Design, Construction and Licensing Certainty for New AP1000 Projects in Europe

Julie Gorgemans, Jake Glavin, Neil Buzzard, Brad Carpenter, Andreas Fristedt Ablad

Westinghouse Electric Company , 1000 Westinghouse Drive, PA 16066 Cranberry Twp, USA

gorgemj@westinghouse.com

 

The AP1000® plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs.

The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the CB&I (Shaw) Group, signed contracts with China’s State Nuclear Power Technology Corporation Ltd., Sanmen Nuclear Power Company Ltd., and Shandong Nuclear Power Company Ltd. for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units. Construction for all four units is largely concurrent. Additionally the United States (US) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) & South Carolina Electric & Gas Company (SCE&G) to construct and operate AP1000 plants at the existing Vogtle & VC Summer sites in Georgia and South Carolina, respectively. Although, construction at both US sites is underway, the first four China AP1000 plants will become operational ahead of the U.S. Domestic AP1000 plants. Westinghouse is also actively engaged in deploying the AP1000 plant design in other regions throughout the world such as Europe. For example, the AP1000 plant design was evaluated by the UK Office for Nuclear Regulation as part of the UK Generic Design Assessment and received a statement of interim Design Acceptance in late 2011. Within this paper, the various factors that contribute to an unparalleled level of design, construction and licensing certainty for any new AP1000 projects in Europe will be described. These include:* How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty.* How the AP1000 passive plant robustness against extreme events that results in large loss of infrastructure further contributes to the licensing certainty in a post-Fukushima regulatory environment.* How new AP1000 plant projects will benefit from the ongoing AP1000 plants under construction in the United States and China and how the lessons learned from ongoing projects will be incorporated to the benefit of the future owner and operator of the AP1000 plant.






12.09.2013 12:40 New reactor technologies

New reactor technologies - 704

Near term commercial applications of small lead fast reactors

Janne Wallenius

Royal Institute of Techology, Div. Of Nuclear Power, Brinellv. 60, S-10044 Stockholm, Sweden

janwal@kth.se

 

In Sweden, the 0.5 MW ELECTRA concept is developed as a facility for research on fast reactor dynamics and education and training. ELECTRA is cooled by natural convection of liquid lead, which becomes feasible thanks to the application of inert matrix (Pu,Zr)N fuel.

Transient analysis shows that ELECTRA is an highly safe concept, and it is suggested to build ELECTRA at the Oskarshamn power plant in the southeast of Sweden.Applying forced convection, the power of ELECTRA may be raised to a level where commercial applications such as production of electricity, medical isotopes, steam and ship propulsion can be considered. In this paper, the near term commercial applicability of small lead fast reactors based on the ELECTRA concept is discussed. In particular, the potential of a Slovenian demonstration project is investigated.






12.09.2013 14:30 Reactor physics, fuel cycle and research reactors

Reactor physics, fuel cycle and research reactors - 602

The core conversion of the TRIGA reactor Vienna

Mario Villa1, Robert Bergmann1, Andreas Musilek1, Johannes H. Sterba1, Helmuth Böck1, Charles Messick2

Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Vienna, Austria1

National Nuclear Security Administration, Office of Global Threat Reduction, Savannah River Site Office, P.O. Box A, Aiken, South Carolina, USA2

mvilla@ati.ac.at

 

The TRIGA Reactor Vienna has operated for many years with a mixed core using Al clad and stainless-steel (SST) clad low enriched uranium (LEU) fuel and a few SST high enriched uranium (HEU) fuel elements. In view of the US spent fuel return program, the average age of these fuel elements and the Austrian position not to store any spent nuclear fuel on its territory, negotiation started in April 2011 with the US Department of Energy (DOE) and the International Atomic Energy Agency (IAEA). The sensitive subject was to return the old TRIGA fuel and to find a solution for a possible continuation of reactor operation for the next decades. As the TRIGA Vienna is the closest nuclear facility to the IAEA headquarters, high interest existed at the IAEA to have an operating research reactor nearby, as historically close cooperation exists between the IAEA and the Atominstitut. Negotiation started before summer 2011 between the involved Austrian ministries, the IAEA and the US DOE leading to the following solution: Austria will return 91 spent fuel elements to the Idaho National Laboratory (INL) while INL offers 77 very low burnt SST clad LEU elements for further reactor operation of the TRIGA reactor Vienna. The titles of these 77 new fuel elements will be transferred to Euratom in accordance with Article 86 of the Euratom-US Treaty. The fuel exchange with the old core returned to the INL, and the new core transferred to Vienna was carried out in one shipment in late 2012 through the ports of Koper/Slovenia and Trieste/Italy.

This paper describes the administrative, logistic and technical preparations of the fuel exchange being unique world-wide and first of its kind between Austria and the USA performed successfully in early November 2012.






12.09.2013 14:50 Reactor physics, fuel cycle and research reactors

Reactor physics, fuel cycle and research reactors - 601

A Comparison of Coarse Mesh Rebalance and Diffusion Synthetic Acceleration Techniques in Discrete Ordinates Calculations

Bilge Özgener, Huseyin Atila Ozgener

Istanbul Technical University, Energy Institute, Ayazaga Campus, 34469, Istanbul, Turkey

ozgenb@itu.edu.tr

 

Although the discrete ordinates method (SN) is capable of producing efficient numerical solutions by its marching algorithm in most transport problems, there are also situations in which SN turns out to be very slowly converging. For one thing, the convergence of inner iterations is especially hampered in cases where c (the ratio of the scattering to total cross section) is close to unity. For another, the outer iterations become especially slow whenever the dominance ratio, (the ratio of the second largest eigenvalue to the largest one) is close to unity. Both the inner and outer iterations could be accelerated by the coarse mesh rebalance (CMR) and the diffusion synthetic (DS) acceleration methods.

In inner iterations, CMR is a multiplicative correction technique. That is it provides acceleration by multiplying the group fluxes in each coarse mesh region by the rebalance factor of that region. The rebalance factors are determined by the solution of a linear system usually in each iteration. On the other hand DS is an additive correction technique. That is DS provides acceleration by adding a correction term to the group fluxes. The correction term is computed by the numerical solution of a diffusion theory fixed source problem. In outer iterations, CMR requires the solution of an eigenvalue-eigenvector problem in each outer iteration. The eigenvalue becomes the new keff estimate and the eigenvector consists of the rebalance factors. Acceleration of the outer iterations by DS requires a priory solution of the adjoint diffusion criticality problem. The diffusion keff and adjoint flux are required for the acceleration of the transport problem. The acceleration of the outer iteration by DS requires also the numerical solution of the equivalent of a fixed source multigroup diffusion theory problem in each outer iteration.In this work, CMR and DS have been implemented in the two variants of our spherical geometry discrete ordinates FORTRAN code, SNSP. Both inner and outer iterations are accelerated by one of the two acceleration techniques in each variant. As a model problem a one-group U-D2O reactor model is selected. Modifications of this base problem are employed for the assessment of the acceleration methods. Our study indicates that CMR is most effective when p (the ratio of the number of fine mesh regions to the number of coarse mesh regions) is approximately five in both inner and outer iterations. As c increases both CMR and DS provide adequate acceleration in inner iterations. But in problems with very high c, DS is found to provide slightly more acceleration than CMR. When the outer iterations are unaccelarated (UA), the SN method becomes increasingly slower as approaches unity. On the other hand, both CMR and DS provide reasonable acceleration for high values. Also the convergence rates of both acceleration methods are found to be only moderately dependent on in contrast to the UA case. The major drawback of DS is met in problems for which the mesh size exceeds 1 mean free path (mfp). The convergence of inner iterations becomes slower as the mesh size is increased towards unit mfp. With DS acceleration, the inner iterations diverge when the mesh size is slightly above 1 mfp.






12.09.2013 15:10 Reactor physics, fuel cycle and research reactors

Reactor physics, fuel cycle and research reactors - 610

Modeling of IFA-409 by Means of TRANSURANUS Code

Davide Rozzia1, Alessandro Del Nevo2, Alessandro Ardizzone3, Pietro Agostini2

Universita degli Studi di Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione, Largo Lucio Lazzerino 1, 56100 Pisa, Italy1

ENEA CR Brasimone, Localita Brasimone, 40032 Camugnano (BO), Italy2

Dipartimento di Energetica, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino, Italy3

daviderozzia@libero.it

 

Inert fission gas atoms have a very low solubility in the UO2 matrix causing two important life-limiting phenomena in the fuel rod: either they remain in the pellets and contribute to the swelling, or they are released from the pellets to the pin free volume. Therefore, the correct prediction of Fission Gas Release (FGR) is an essential tool to assess the behavior of the fuel matrix (that is considered the first barrier against the FP release) under steady state and transient conditions.

OECD/NEA sets up the “public domain database on nuclear fuel performance experiments for the purpose of code development and validation – International Fuel Performance Experiments database”. This database includes the IFA-409 experiments. The objective of these experiments was to investigate the Fission Gas Release (FGR) phenomenon in BWR fuel rods irradiated at high burn-up during normal operation. These experimental data have been distributed in the framework of the International Atomic Energy Agency Coordinate Research Program FUMEX III (2008-2011).The database includes 4 rods. The rods were base irradiated in Halden Boiling Water Reactor in the upper cluster of IFA-409 from May 1973 until June 1985. The time averaged heat ratings are mainly in the range 25-30 kW/m. The final burn-up ranges from 42 to 45 MWd/kgUO2. After the base irradiation, FGR and gap pressurization were measured and then the rods were reconstructed and re-instrumented. Two rods were assembled in IFA-535.5, the remaining rods were assembled in IFA-535.6 and then they were subjected to slow and fast power ramps. The present paper focuses on IFA-409 base irradiation only.The aim of the activity is to summarize the main results obtained after the simulations of 4 BWR fuel rods included in the above mentioned database by means of TRANSURANUS code. Particular emphasis is given to the main variables which influence the FGR phenomenon in normal operation. The importance of the sensitivity analysis, as tool to address the relevance of the knowledge of the boundary conditions, as well as the impact of selected parameters and code options on the results is discussed.






12.09.2013 15:30 Reactor physics, fuel cycle and research reactors

Reactor physics, fuel cycle and research reactors - 604

Fast Reactors Deployment Strategy. Some Constraints and Consequences.

Georgios Glinatsis

ENEA, Technical Unit for Reactor Safety and Fuel Cycle Methods , Via Martiri di Monte Sole 4, 40129 Bologna, Italy

georgios.glinatsis@enea.it

 

One of the main concerns of the public related to the nuclear energy production is the handling of highly radioactive nuclear waste (HLW). A safe management of the HLW, is mandatory in the pursuing a sustainable development of the nuclear energy in both Gen-IV and SNE-TP Initiatives. Of course, generate energy sustainably and promote long-term availability of nuclear fuel, as well as minimize nuclear waste and reduce the long term stewardship burden, are two of the main Generation-IV goals. Assuring energy supply, environmental protection, economic feasibility, proliferation issues and safeguards of the nuclear materials, imply the development and deployment of innovative reactors (Generation III/III+ and Generation IV) and development of innovative concepts for the nuclear fuel cycle closure, with the aim of an acceptable cost energy production, reduction of waste volume and long term radiotoxicity. Taking into consideration the good neutron economy of the Fast Reactors, their development is mandatory for the nuclear fuel cycle closure, even if other options are also under investigation. Moreover, the realization of cycle closure, through Fast Reactors, is tightly connected with their deployment. Some constraints should be taken in consideration and some consequences would be expected. The Pu availability is the main issue, while its lack, during the cycle, can be avoided adopting appropriate operative conditions for the different components of the associated fuel cycle. Reduced cooling time or Minor Actinides management, or the increase of the breeding gain, etc. are some of the possible solutions. Of course optimisation processes are required to avoid bottlenecks, as really short cooling time or proliferation issues.

Key words: Fast Reactors, Core Design, Scenario Codes, Nuclear Data.






12.09.2013 15:50 Reactor physics, fuel cycle and research reactors

Reactor physics, fuel cycle and research reactors - 611

Assessment of MOX fuel behavior, based on recent benchmarks

Alessandro Ardizzone1, Davide Rozzia2, Alessandro Del Nevo3

Dipartimento di Energetica, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino, Italy1

Universita degli Studi di Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione, Largo Lucio Lazzerino 1, 56100 Pisa, Italy2

ENEA CR Brasimone, Localita Brasimone, 40032 Camugnano (BO), Italy3

alessandro_ardizzone@aol.com

 

The commercial use of MOX fueled rods is of great interest both for FR and LWR technology since it allows to re-cycle Pu and Minor Actinides produced in the last decades of NPP operation. Because of the extremely large uranium fuel performance database and the corresponding well-validated codes, it is important to define differences in structure and performance of MOX and uranium fuels. In particular, inert gases generated by fissions cause two important phenomena that limit the discharge burn-up of fuel rods: either they contribute to fuel swelling, or they are released in the rod free volume causing pin pressurization and “poisoning” the gap by reducing its conductance. These phenomena are even more relevant in MOX fuel rods, where He is produced through Pu ?-decays and ternary fissions. The correct prediction of Fission Gas Release (FGR) is therefore essential to assess the reliability of the first two barriers to the release of radioactivity: the fuel matrix and the cladding.

The present work focuses on the assessment of TRANSURANUS code in predicting the FGR phenomenon in MOX fueled rods. Two databases that have been recently benchmarked are modeled to fulfill this objective: the PRIMO MOX (rod BD8) and IFA-597.PRIMO included 16 MOX fuel rods irradiated to burn-ups from 20 to 60 MWd/kgHM. The objective of these experiments was to investigate the thermal-mechanical behavior of PWR MOX fuel rods in steady-state and transient conditions. The current work deals with one rod labeled as rod BD8. This rod belongs to the OECD/NEA International Fuel Performance Experiments database (IFPE) and was power ramped in Osiris after a base-irradiation in the BR2 (Mol) up to 34 MWd/kgMOX.IFA-597 belongs to the Halden Reactor Project and investigated two MOX fuel rods irradiated in the Halden Reactor with the aim to study the thermal and FGR behavior of MOX fuel and to explore potential differences between solid and hollow pellets. The discharge burn-up was 32 MWd/kgMOX.The aim of the activity is to summarize the main results obtained after the simulation of these databases by means of TRANSURANUS code. Particular emphasis is given to the main variables which influence the FGR phenomenon. A detailed sensitivity analysis is carried out to assess the influence of models, fuel design and boundary conditions.






12.09.2013 16:40 Radioactive waste management

Radioactive waste management - 902

European project 'Metrology for radioactive waste management'

Petr Kovar1, Jiri Suran1, Jaroslav Solc1, Dirk Arnold2

Czech Metrology Institute, V Botanice 4, 150 00 Praha, Czech Republic1

Physikalisch - Technische Bundesanstalt, Bundesallee 100, 38116 BRAUNSCHWEIG, Germany2

pkovar@cmi.cz

 

In November 2010 joint research project JRP ENV09 ‘Metrology for Radioactive Waste Management (MetroRWM)’ was accepted for financing within the Call ‘Environment’ EMRP (European Metrology Research Programme). Twelve national metrology institutes or designated institutes, Joint Research Centre and one independent researcher (REG) took part in this project with total budget more than 4 M€. The three years project started in October with the Czech Metrology Institute (CMI) as a coordinator.

The project consists of five scientific/technical work packages:WP1: Development of standardised traceable measurement methods for solid radioactive waste free release (clearance levels verification) and for acceptance of solid radioactive wastes to repositories (acceptance criteria verification), according to international recommendations (EC and IAEA): design of measurement facilities, software, calibration and testing methods. Expected relative combined standard uncertainties typically up to 20 % for mass activity measurement.WP2: Development of novel instruments and methods for in-situ measurements: improved on-site radiochemical analysis, rapid in-situ screening techniques for alpha, beta and gamma emitters, measurement of activity at varying depth. Expected relative combined standard uncertainties typically up to 10 % for activity measurement.WP3: Development of a gaseous effluent monitor/sampler for stored wastes. Rapid, sensitive methods are required to determine rates of bulk gas production, chemical composition (CH4, CO2 or H2) and activity concentrations of key radionuclides (H-3, C-14, Rn-222). Expected relative combined standard uncertainties typically up to 10 % for volume activity measurement. WP4: Development of standards and ‘spiked’ or characterised ‘real’ reference materials for ensuring accurate, traceable radio-assays of materials from sites (concrete, steel, aluminium, cables, wood, insulator and others). Expected relative combined standard uncertainties up to 10 % for activity and mass activity.WP5: Improvements to decay data for selected radionuclides present in nuclear wastes, focusing on half-life measurements of long-lived fission and activation products. Expected relative combined standard uncertainties up to 3 % for half-lives of Ho-166m, I-129 and Sm-151.The goal of this paper is to give general information about the project and enlarge stakeholder and end user groups, who can also register at JRP ENV09 website http://www.radwaste-emrp.eu.






12.09.2013 17:00 Radioactive waste management

Radioactive waste management - 907

How technological and contractual innovations are shaping nuclear fuel management in Europe?

Mustapha Chiguer, François Mazaré, Anne-Charlotte Dagorn

AREVA, 1 place Jean Millier, 92084 Paris La Defense, France

anne-charlotte.dagorn@areva.com

 

European utilities provide their country with competitive and responsible electricity supply. In the regional context of growing electricity demand, energy stakeholders have conducted feasibility studies for new nuclear power plants, including studies to determine the best available nuclear fuel management options.

The objective of this paper is to present different used fuel management options adapted to the European context. It intends to assess the economics going along with each possible option. In addition to mainstream scenarios (closed cycle and open cycle), new scenarios are included in order to present technological and contractual innovations such as Precycling, Enriched Reprocessed Uranium (ERU) and Mixed Oxide Fuel (MOX). To support these analyses, case studies on implemented fuel management strategies are also conducted (example of small fleet countries). The study provides key elements which will help stakeholders to assess further the different options available and take decisions about the fuel management strategy to be implemented in their country.






12.09.2013 17:20 Radioactive waste management

Radioactive waste management - 909

Long Term Safety Assessment for Slovenian LILW Repository

Sandi Viršek

Agencija RS za radioaktivne odpadke, Celovška cesta 182, 1000 Ljubljana, Slovenia

sandi.virsek@gov.si

 

In 2004 Slovenia started a siting procedure for a Low and Intermediate Level Waste (LILW) repository. At the end of 2009 the Slovenian government approved the Krško Vrbina site, the near surface silos concept for the site and the concept for the future Slovenian LILW repository. For nuclear facilities, such as repository, it is very important to establish the confidence in long term safety. This is especially important because we need to show that the repository will remain safe even after the closure and since we are also disposing of some long lived radio nuclides, that we cannot separate from the short lived nuclides, this means tens of thousands or even hundreds of thousands of years. For this purpose we are preparing the Safety case, the important part of which is the safety assessment.

At the beginning of the paper, the methodology that was used for the preparation of the last iteration of the long term safety assessment for planned Slovenian LILW repository is presented. The paper then proceeds with descriptions of the safety assessment context, preparation of different conceptual models, modelling work and finishes with the results and their interpretation.





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Dates to remember
31 Jan, 2013 Call for papers
31 May, 2013 Abstract submittal (extended)
15 Jun, 2013 Abstracts acceptance
30 Aug, 2013 Full length papers
6 Dec, 2013 Proceedings

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