05.09.2016
16:20 Invited Gilles Bignan
Invited
lectures - 102
The Key role of Research
Reactors in support to the development
of nuclear energy: example of the JHR
Project, a new Material Testing
Reactor working as a European and
International Users Facility in
support to Research Institutes and
Nucl. Industry
Gilles
Bignan
CEA
France, CEN Saclay ORE/SRO, France
European Material
Testing Reactors (MTR) have provided an
essential support for nuclear power
programs over the last 50 years within the
European Community. However, the large
majority of these Material Test Reactors
(MTRs) are more than 50 years old, leading
to the increasing probability of some
shutdowns for various reasons
(life-limiting factors, heavy maintenance
constraints, possible new regulatory
requirements…). Such a situation cannot be
sustained in the long term.
On the other hand, associated with hot
laboratories for the post irradiation
examinations, MTRs remain key structuring
research facilities for the European
Research Area in the field of nuclear
fission energy.
MTRs address the development and the
qualification of materials and fuels under
irradiation with sizes and environment
conditions relevant for nuclear power
plants in order to optimize and
demonstrate safe operations of existing
power reactors as well as to support
future reactor design:
- Nuclear plants will follow a long-term
trend driven by the plant life extension
and management, reinforcement of the
safety, waste and resource management,
flexibility and economic improvement.
- In parallel to extending performance and
safety for existing and power plants to
come, R&D programs are taking place in
order to assess and develop new reactor
concepts (Generation IV reactors) that
meet sustainability purposes.
- In addition, for most European
countries, keeping competences alive is a
strategic cross-cutting issue; developing
and operating a new and up-to-date
research reactor appears to be an
effective way to train a new generation of
scientists and engineers.
This analysis was already made during the
previous decade by a thematic network of
EuratomFramework Program, involving
experts and industry representatives,
confirming the need for a new Material
Testing Reactor (MTR) in Europe.This is
the genesis of the Jules Horowitz Reactor
(JHR),a new Material Testing Reactor (MTR)
currently under construction atCEA
Cadarache research center in the south of
France. It will represent a major research
infrastructure for scientific studies
dealing with material and fuel behavior
under irradiation (and isconsequently
identified for this purpose within various
European road maps and forums; ESFRI,
SNETP…).The reactor will also contribute
to medical Isotope production.
06.09.2016
08:30 invited Hamid Ait Abderrahim
Invited
lectures - 103
Role of Nuclear Energy in
the Future Energy Mix and Needs for
R&D in Closing the Fuel Cycle
Hamid
Ait Abderrahim
SCK.CEN,
Av. Herrmann Debrouxlaan 40, 1160
Brussels, Belgium
Presently, the European
Union produces 30% of its electricity by
Gen.II and IIInuclear reactors. This leads
to the production of 2500 t/y of used
fuel, containing 25 t of Plutonium, and
High Level Wastes (HLW) such as 3.5 t of
minor actinides (MA), namely Neptunium
(Np), Americium (Am) and Curium (Cm) and 3
t of long-lived fission products (LLFPs).
The used fuel reprocessingfollowed by the
geological disposal or the direct
geological disposalare today the envisaged
solutions depending on national fuel cycle
options and waste management policies. The
Partitioning and Transmutation (P&T)
has been pointed out in numerous studies
as the strategy that can relax constraints
on the geological disposal, and reduce the
monitoring period to technological and
manageable time scales. Transmutation
based on critical or sub-critical fast
spectrum transmuters should be evaluated,
in order to assess the technical and
economic feasibility of this waste
management option.
After nearly twenty years of basic
research funded by national programmes and
EURATOM framework programmes, the research
community needs to reach a position of
being able to quantify indicators for
decision-makers, such as the proportion of
waste to be channelled to this mode of
management, but also issues related to
safety, radiation protection, transport,
secondary wastes, costs, and scheduling.
From 2005, the research community on
P&T within the EU started structuring
its research towards a more integrated
approach.This resulted during the FP6 into
two large integrated projects namely
EUROPART dealing with partitioning and
EUROTRANS dealing with ADS design for
transmutation, development of advanced
fuel for transmutation, R&D activities
related to the heavy liquid metal
technology, innovative structural
materials and nuclear data measurement.
This approach resulted in a European
strategy given in introduction based on
the so-called “four building blocks” at
engineering level for P&T.
The MYRRHA project contributes heavily to
the third building block of this European
strategy and in this paper we will focus
on the ADS programme in the EU through the
MYRRHA project.
In this seminar we will present the EU
strategy for P&T and the status of the
MYRRHA project as by End-2015 concerning
the technical design, the pre-licensing
and the projected implementation scenario
for the realisation of the MYRRHA
facility.
07.09.2016
08:30 invited Böck Helmuth
Invited
lectures - 104
Five Decades of TRIGA
Reactors
Helmuth
Böck
Vienna
University of Technology, Atominstitut,
Stadionallee 2, 1020 Vienna, Austria
The concept of TRIGA
(Training, Research, Isotopes, General
Atomics) has been developed immediatly
after the Geneva Conference on Peaceful
Uses of Atomic Energy in 1955. The aim of
the TRIGA design was a reactor that “could
be given to a bunch of high school
children to play with, without any fear
that they would get hurt” and it should
include inherent safety features. The
basis of this inherent safe feature for
all TRIGA type reactors is the U-Zr-H fuel
with its strong negative temperature
coefficient.
During the past 50 years many different
types of TRIGA fuel elements with
different uranium content and different
enrichment have been developed but the
fuel basis remained always the same.
In the late 1950ties to the end of the
1960ties TRIGA reactors were mainly
commissioned in the US and Europe while
later Asian countries as well as Latin
America followed. Most of these reactors
were used for the formation of engineers
and scientist to develop a national
nuclear program, at universites for
academic training or at hospitals for
radioisotope production.Totally 66 TRIGA
reactors have been built , some were
converted from MTR type fuel to TRIGA
fuel. The following decades during the
1990ties and beyond are characterised by
TRIGA recators being shut down or
decommissioned due to changes in the
national nuclear programs, under
utilization or simply lack of funds.
In addition the US fuel return program
started in 1996 put pressure on many TRIGA
reactors to return any HEU fuel to the US.
In many countries this program initiated
TRIGA shut down processes due to reasons
mentioned above.Today 38 TRIGA fueled
reactors remain operational.
Presently the main concern of the TRIGA
community are the continuous supply of
TRIGA fuel, presently suspended due to
necessary safety and security investment
at the fuel factory located at
Romans,France . Other concerns are costly
refurbishments due to new safety and
security requirements and under
utilization.
After a brief history of TRIGA reactors
the paper covers the present situation of
the TRIGA community and gives an outlook
of problems to be solved during the next
decade for further successful TRIGA
operation.
08.09.2016
08:30 invited Simon Pinches
Invited
lectures - 105
The ITER Integrated
Modelling Programme
Simon
Pinches
ITER Organization,
Cadarache Centre, Building 519, 13108 St.
Paul lez Durance, France
ITER is based on the
“tokamak” concept of magnetic confinement,
in which the fusion (deuterium-tritium)
fuel is contained in a toroidal vessel.
The ITER reactor is designed to generate
500 MW of fusion power for periods of 300
to 500 seconds with a fusion power
multiplication factor, Q, of at least 10.
ITER will also aim at demonstrating long
fusion power production pulses, of at
least 1000 seconds, with a fusion power
multiplication factor of at least 5 and,
ultimately, of 1 hour duration (limited
only by hardware design) when full
non-inductive operation is demonstrated.
A major element of the ITER
Physics Research Programme is the
establishment of an integrated modelling
programme, including benchmarking and
validation activities. Whilst this
activity is co-ordinated by the ITER
Organization, it is being developed using
expertise and existing co-ordination
structures within the ITER Members’ fusion
programmes. The overall aims of this
programme are to meet the initial needs of
the ITER project for more accurate
predictions of ITER fusion performance and
for efficient control of ITER plasmas, to
support the preparation for ITER operation
and, in the longer term, to provide the
modelling and control tools required for
the ITER exploitation phase.
Support of Plasma
Operations requires a set of
computationally efficient, robust, physics
modelling tools that are executed
systematically prior to operation for
pulse validation, during the pulse for
plasma control and live display, and
post-pulse for comprehensive
reconstruction of the plasma from the full
collection of diagnostic measurements.
They should capture the macroscopic
behaviour of the plasma with a level of
fidelity that improves as ITER operation
explores the new physics domain of burning
plasmas. Collectively, these modelling
tools comprise the Integrated Modelling
Analysis Suite (IMAS).
Support of Plasma Research
requires a much more extensive set of
modelling tools to be employed both prior
to operation and post-operation. These
tools may examine microscopic behaviour,
investigate more rigorous theoretical or
computational behaviour, or explore new
physics. They are the primary basis for
model improvement and validation. They may
be applied to selected pulses, segments or
time slices, and may often require
significant high performance computing
capabilities.
IMAS, coupled with the more
extensive array of physics codes in the
domestic programmes, is expected to evolve
toward a more self-consistent description
as the ITER Research Programme progresses.
One of the first applications for
prototyping the IM infrastructure and
developing the tools required for pulse
preparation is the capability to undertake
co-simulations involving the Plasma
Simulator (PS) and the Plasma Control
System Simulation Platform (PCS-SP).
In this presentation, an
introduction to ITER and an overview of
the Integrated Modelling Programme and
IMAS will be presented.
The views and opinions
expressed herein do not necessarily
reflect those of the ITER Organization.
05.09.2016
17:20 Research reactors
Research
reactors - 200
Fifty years of neutron
activation analysis in Slovenia
Borut
Smodiš
Institut
"Jožef Stefan", Jamova cesta 39, 1000
Ljubljana, Slovenia
borut.smodis@ijs.si
The TRIGA Mark II
reactor of Jožef Stefan Institute became
critical in the year 1966. Soon after its
commissioning, the installation has become
to be utilized for neutron activation
analysis (NAA). Early applications were
dedicated to development of radiochemical
procedures for determining trace elements
in the environment and in human health.
Particular emphasis was devoted to
studying the effects of the Idrija mining
and milling activities onto the
environment and man. The mine and
distillation plant, the second largest in
the world, had been in operation since
discovery of mercury in 1490. Due to its
long history of discharge, the nature of
the environment, the low population
mobility and the reliance on local
supplies of food, it represented a
challenging opportunity to study the
mercury transport in the environment, and
its effects on biota and man.
Along with the research on mercury, the
JSI scientists focused their work on
radiochemical procedures for the
determination of microgram and nanogram
amounts of essential and toxic elements in
both organic and inorganic matrices. The
numerous procedures developed were largely
based on solvent extraction, ion exchange
and volatilization processes, and included
essential and toxic elements that were
difficult to be determined. Procedures
comprised either isolation of a single
element or simultaneous separation of a
group of elements followed by their
individual isolation and subsequent
measurement. Radiochemical NAA (RNAA)
procedures for the determination of
numerous elements were developed and
successfully applied in characterizing
Standard Reference Materials prepared by
NBS, a predecessor of the National
Institute of Standards and Technology. The
quality of analytical measurements
developed in the laboratory resulted in
long-term collaboration with eminent
international organizations producing
reference materials, such as USA National
Bureau of Standards (NBS) – nowadays
National Institute of Standards and
Technology (NIST), International Atomic
Energy Agency (IAEA), EU Community Bureau
of Reference – nowadays Institute for
Reference Materials and Measurements
(IRMM) and Japanese National Institute for
Environmental Studies (NIES). Many
reference materials were analysed during
that time either for certification
purposes or simply as contribution to
international high quality data.
Simultaneously, procedures for the
determination of long-lived radionuclides
by combination of NAA and other
radiometric methods have been developed
and applied.
Along with the development of nuclear and
gamma spectrometric equipment in late
seventies and early eighties, INAA has
attracted ever more applications,
gradually replacing the RNAA procedures,
whenever applicable. Further decline in
the application of RNAA occurred as
consequence of introduction of other
modern analytical techniques.
Instrumental neutron activation analysis
in its relative mode had been used as soon
as the laboratory received its first
Ge(Li) detector in the seventies. In the
eighties, the k0 –based NAA was
introduced; soon after its introduction,
its potential for characterizing certified
standard materials was tested and
confirmed. The k0 –NAA gradually replaced
the relative method of INAA, eventually
resulting in its accreditation as a
routine analytical tool in 2009. Nowadays,
the k0 –NAA is used as primary analytical
tool.
In spite of ever-growing market of new
and/or improved analytical techniques for
elemental analysis, neutron activation
analysis still plays a significant role in
the preparation of reference materials due
to its many favourable features.
In the presentation, the main success
stories over the years are shown, the
educational aspects are outlined, and the
contributions towards improved quality of
analytical measurements are discussed.
05.09.2016
17:40 Research reactors
Research
reactors - 201
Neutron Radiography and
SSNTD's at Ljubljana Triga Research
Reactor: Almost 50 years of developing
the methods, facilities and of
research and applications
Jožef
Rant
Institut
"Jožef Stefan", Odsek za reaktorsko
fiziko, Jamova cesta 39, 1000 Ljubljana,
Slovenia
joze.rant@ijs.si
The idea to introduce
neutron radiography (NR) at the Ljubljana
TRIGA Mark II research reactor was put
forward in 1968 by the present author.
First NR facility ( neutron collimator
with beam filter, beam shutter and
shielded exposure room) was constructed in
the tangential beam tube and first neutron
radiographs were produced in summer 1969.
Later the NR facility was removed to
thermal column and it was in operation
until the spring 2015 when an irradiation
experiment was installed in the thermal
column. Already at the beginning it was a
demanding challenge to develop a
microneutronography (MNR) as a high
resolution (~ 10 µm) neutron radiography
using rather thin (few µm ) Gd metal
converter screens and fine grained Kodak
Maximum Resolution photographic plates as
a complementary method to conventional
X-ray microradiography for the inspection
and characterization of thin metallurgical
or geological samples . Film based MNR
requires high neutron exposure fluences (
1010 – 1012n/cm2) and later a vertical
neutron beam was constructed through the
reactor water tank down to the reactor
core which enabled much shorter and
practical exposure times in comparison
with the exposures in the tangential
neutron beam. The vertical beam tube was
also used for neutron induced
autoradiography (NIAR or NCAR- neutron
capture autoradiography) with solid state
nuclear track detectors (SSNTD's) which
were introduced in 1974 by the present
author as a fallout of the 1973 BNES
Birmingham conference and with the support
of R. Barbalat and G. Farny (CEA CEN,
Saclay) and of J. Barbier (Kodak Pathe').
The introduction of SSNTD's (track etch
techniques) was a success and later
significant research work and applications
were conducted in the newly established
Laboratory for SSNTD's under the
leadership of R. Ilić. Early applications
of MNR and NIAR/NCAR in metallurgy have
been reviewed. The early development of NR
and NIAR , the characterization of
facilities and the use of various neutron
beam tubes of te Ljubljana TRIGA have been
presented elsewhere. Early applications of
NR and NIAR7NCAR have been reviewed in
1981. Major refurbishment of the NR
facility in the thermal column was
achieved in 1995 and the photoluminescent
imaging plates (IP) as a highly efficient
neutron image detectors were introduced
with the support of J. Stade (BAM, Berlin)
in 1996. Some important applications of NR
include examination of deformed irradiated
TRIGA fuel elements, a study of diffusion
of hydrogenous liquids into porous
materials and of processes of impregnation
of building materials. In the last decade
applications of NR were in the field of
archaeology and in the preservation of
cultural heritage . Detection and maping
of boron in the histological samples using
NCAR with SSNTD's or IP's in developing
boron neutron capture therapy for cancer
was a topic of many studies. A significant
achievement was the development of
selective autoradiography on the basis of
computerized automatic analysis and
characterization of nuclear tracks by
Skvarč et al.. A systematic and
comprehensive analysis of image quality
and of image transfer response functions
in radiography and autoradiography with
SSNTD's was performed by Ilić and Najžer.
The role of backdiffused beta radiation in
imaging with beta-rays emitting converter
screens was evaluated by Rant et al. and
the possibility to use backdiffused
beta-ray radiation for simple examination
of surface layers was demonstrated.
Characteristic for the past research work
at Ljubljana TRIGA was, that it was not a
part of the mainstream of the scientific
funding, it was more the result of
enthusiastic endeavour of a few
collaborators and was enabled through the
support within the international
cooperation. The experimental
possibilities and versatility of the
Ljubljana TRIGA reactor was fully
exploited.
05.09.2016
18:00 Research reactors
Research
reactors - 202
Spallation Target Design
for Converting the Isfahan MNSR
Reactor to an Accelerator Driven
System
Mohsen
Kheradmand Saadi, Kimia Mokhtari
Department
of Nuclear Engineering, Science and
Research Branch, Islamic Azad University,
Tehran, Iran, 1477893855, Iran
mohsen.kheradmand@gmail.com
There are many
research reactors around the world that
have been installed from many years ago
and must be decommissioned sooner or
later. However, most of the initially high
enriched uranium has not been exploited
yet and the reactor core has much fission
products as well as actinides. The reactor
conversion to an Accelerator Driven System
(ADS) is one of the novel ideas for minor
actinide utilization and reducing the
spent fuel radio toxicity. Usually, the
reactor target design is considered as a
first step toward ADS design. The target
performance plays an important role in ADS
design and is characterized by some
parameters including the spallation
neutrons yield, neutron energy spectrum,
deposited energy in target and the angular
distribution of spallation neutrons. The
main objective of this study is dedicated
to spallation target design for the
Miniature Neutron Source Reactor (MNSR)
core conversion to an ADS one. The MNSR is
a pool type research reactor, which was
developed by china and installed in
Isfahan nuclear technology center in 1994.
After more than 20 years reactor
operation, the fuel burn-up could not be
compensated more by adding plate shims and
the present reactor core must be removed
sooner or later. The sub-criticality in
MNSR reactor was attained by entirely
removing the top plate shims as well as
control rod and installing the spallation
target in interior space of guide thimble.
Different state of the art targets such as
Tungsten, Lead, Bismuth, and LBE have been
investigated and the target parameters
were evaluated using MCNPX2.6 in both
proton and neutron mode. The results
showed that the neutronic performance of
Tungsten is somewhat greater than its
competitors. However, the Tungsten has an
extraordinary thermal properties and
thermal analysis of these targets is
suggested for further investigations.
05.09.2016
18:20 Research reactors
Research
reactors - 203
Laboratory of fast
neutron generators of the NPI
Mitja
Majerle
Nuclear
Physics Institute of the CAS, v. v. i.,
Řež 130, 250 68 Řež, Czech Republic
majerle@ujf.cas.cz
The Nuclear Reactions
Department of the NPI operates neutrons
sources with neutron energies extending up
to 35 MeV. The cyclotron U-120M provides
protons in the energy range of 20-36 MeV.
These are directed to a thin Li foil
(quasi-monoenergetic neutrons) or to a
thick Be target (continuous neutron
spectrum). The available neutron fluxes
are up to 10^9 n/cm^2/s for QM neutrons
and 10^11 n/cm^2/s for continuous neutron
spectrum.
The produced neutrons are used in a wide
scale of activities connected to ADS
technologies and fusion. This contribution
focuses on cross-section measurement and
benchmarks, the measurements of the
produced neutron spectra, the development
of the (n,cp) chamber and online gamma
measurements.
06.09.2016
09:10 Reactor physics I
Reactor
physics - 301
I2S-LWR pressure vessel
fast fluence calculations
Mario
Matijević, Dubravko Pevec, Krešimir
Trontl
University
of Zagreb, Faculty of Electrical
Engineering and Computing , Unska 3, 10000
Zagreb, Croatia
mario.matijevic@fer.hr
The I2S-LWR concept
(Integral Inherently Safe Light Water
Reactor) is a high-power (1000 MWe) LWR
with improved inherent (passive) safety
features. This new reactor concept, led by
Georgia Institute of Technology (USA), is
based on the integral primary circuit
configuration, a new type of the
fuel/cladding system, and a novel steam
generation system. This paper presents
shielding studies of the I2S-LWR reactor
model using SCALE6.1 code package to
identify the fast neutron fluence rate
distribution. The reactor pressure vessel
(RPV) fast fluence calculation with
uniform fission source showed that
significant fast fluence of 2e19 n/cm2 was
not reached, so risk from pressurized
thermal shock at RPV is not impacting the
reactor design for operating lifetime of
100 years. The SCALE6.1/MAVRIC shielding
sequence was used to optimize neutron
fluence results in the complete RPV for
energies E > 0.1 MeV, E > 1.0 MeV
and for the complete neutron spectra. The
CADIS and FW-CADIS methodologies, based on
forward-adjoint discrete ordinates (SN)
solution via Denovo solver, were used to
accelerate the final Monte Carlo
calculation with Monaco code. Denovo
utilizes Koch-Baker-Alcouffe parallel
transport sweep algorithm over the XYZ
meshes covering the problem domain and
Krylov iteration on multigroup equations
giving space-energy dependent fluxes. Such
hybrid shielding methodology with
mesh-based variance reduction parameters
is very efficient for complex shielding
problems, where particle flux is
attenuated by many orders of magnitude.
The same shielding methodology was used in
the second part of the paper, where we
calculated the radial neutron reflector
heating and RPV power-level monitor's
feasibility. The MAVRIC/FW-CADIS was
successfully used to produce well
converged neutron fluence rate (in fast
and thermal region) over the reduced
I2S-LWR model, extending from the core to
the biological shield exterior. Obtained
results were then used to optimize
calculations involving inelastic neutron
scattering on 28Si and 12C, since these
isotopes comprise the SiC type detector
which will be used for power level
monitoring. Visualization of the obtained
results in 3D was done using VisIt code
from the Lawrence Livermore National
Laboratory.
06.09.2016
09:30 Reactor physics I
Reactor
physics - 302
Variance reduction of
fusion and fission neutron transport
problems using the ADVANTG hybrid code
Bor
Kos, Ivan Aleksander Kodeli
Jožef
Stefan Institute, Reactor Physics
Department , Jamova cesta 39, 1000
Ljubljana, Slovenia
bor.kos@ijs.si
Hybrid methods in an
optimal way combine the best attributes of
Monte Carlo (MC) and deterministic
methods. Such hybrid computational
radiation transport codes thusly expand
the potential for solving large, complex
real-world problems. The complementary use
of both methods opens the way for the
simulation not achievable with “analog”
Monte Carlo simulations such as deep
penetration or very large and complex
streaming geometries.
The Automated Variance Reduction Generator
(ADVANTG) code is a MC/Deterministic
Hybrid transport code developed by ORNL
(Oak Ridge National Laboratory), using the
Denovo deterministic neutron transport
code and MCNP a widely used Monte Carlo
transport code. Its approach for combining
deterministic and Monte Carlo transport
methods is based on the Consistent Adjoint
Driven Importance Sampling (CADIS) method.
The fundamental concept is to generate an
approximate importance function from a
fast-running deterministic adjoint
calculation and use the importance map to
construct variance reduction parameters,
more specifically weight window
parameters, which can accelerate tally
convergence in the MC simulation.
ADVANTG’s reliability and consistent
performance has to be tested on a variety
of different example problems. The use and
performance of ADVANTG on three different
examples encompassing a variety of
neutronics applications will be presented
in this paper.
Firstly the use of ADVANTG for
accelerating MC simulations of deep
penetration benchmark experiments such as
the NESDIP3 and JANUS1 experiments will be
presented. In connection with this the
importance of reducing statistical
uncertainty of MC simulations when
validating new nuclear data will be shown.
Secondly ADVANTG coupled with MCNP will be
used to determine neutron flux and neutron
dose in a very large streaming geometry -
the JET tokamak. More specifically the
acceleration of the simulation in
accordance to the NEXP benchmark
experiment. The aim of this experiment is
to measure the neutron streaming through
ducts and the dose rates outside of the
JET Torus Hall.
Lastly the use of ADVANTG for accelerating
MC simulations on a newly developed
detailed model of the Krško nuclear power
plant will be presented. The aim of this
simulations is to determine neutron dose
fields in the steam generator and reactor
coolant pump cubicles where “analog”
simulations are difficult because of large
attenuation of neutrons between the
reactor core and cubicle.
06.09.2016
09:50 Reactor physics I
Reactor
physics - 303
Effects of the neutronic
and thermohidraulic simplifications on
the neutronic power
Nicolás
Olmo-Juan1, Teresa María
Barrachina Celda2, Rafael
Miró Herrero3, Gumersindo
Verdú4
1Instituto de
Seguridad Industrial, Radiofísica y
Medioambiental (ISIRYM), , Spain
2Department of
Chemical and Nuclear Engineering,
Polytechnic University of Valencia, Camí
de Vera sn, 46022 Valencia, Spain
3Universitat
Politecnica de Catalunya, C. Jordi Girona,
31, 08034 Barcelona, Spain
4Universidad
Politecnica de Valencia, Departamento de
Ingeniería Química y Nuclear, Camino de
Vera s/n, 46022 Valencia, Spain
nioljua@iqn.upv.es
In neutronic
calculations, the approximation of the
Boltzman neutron transport equation by
diffusion equation is widely accepted.
However, there are certain cases in which
this approximation does not take into
account the heterogeneity of the reactor
core and therefore the errors in the
results are not acceptable.
In such cases, other methods to solve the
Boltzman neutron transport equation have
to be used as for example the Simplified
Spherical Harmonics, known as SPn
equations.
In this paper, the analyses of the
application of SP3 approximation
implemented in the neutronic code PARCS in
some cases, are presented. The purpose is
to analyze the influence of the
homogenization process of the cross
sections in the results studying the
influence of the Assembly Discontinuity
Factors (ADF’s) in the accuracy of the SP3
approximation results.
Another study carried out using PARCS in
stand-alone mode is presented in this
paper. When performing calculations with
coupled codes, usually the thermahydraulic
model of the reactor is a simplified model
in which the fuel assemblies (FA) are
grouped. The procedure used to group the
FA affects the accuracy of the results. To
analyze this influence different cases are
run in PARCS stand-alone to capture the
effects of the simplification of the
thermalhydraulic model using different
external thermalhydraulic conditions, that
is, fuel temperatures and moderator
densities.
06.09.2016
10:10 Reactor physics I
Reactor
physics - 304
On-the-fly towards pure
Monte-Carlo transient reactor core
analysis
Antonios
Mylonakis1, Melpomeni
Varvayanni2, Nicolas Catsaros2
1National Centre
for Scientific Research “Demokritos”,
Institute of Nuclear & Radiological
Sciences & Technology, Energy &
Safety, Nuclear Research Reactor
Laboratory, Agia Paraskevi Attikis,
P.O.Box 60037, 153 10 Athens, Greece
2National Center
for Scientific Research “DEMOKRITOS”
Institute of Nuclear and Radiological
Sciences and Technology, Energy and Safety
Research Reactor Laboratory, PO Box 60228,
15310 Agia Paraskevi, Attiki, Greece
mylonakis@ipta.demokritos.gr
In the field of reactor
physics the transient behavior of the
reactor core is mainly analyzed using
deterministic algorithms. However,
deterministic algorithms make use of
various approximations mainly in geometric
and energetic domain which may induce
inaccuracy. On the other hand Monte-Carlo
analysis, which generally does not require
significant approximations, is currently
very extensively used in static problems
but not in transient analysis. Since
nowadays the available computational
resources are continuously increasing, the
potential use of the Monte-Carlo
methodology in the field of reactor
transient analysis seems quite attractive.
This work performs an investigation of
this possibility by developing a
Monte-Carlo transient solver on the
open-source Monte-Carlo static code
OpenMC. The obtained results are
encouraging giving motivation for further
investigation and development.
06.09.2016
11:10 Reactor physics II
Reactor
physics - 305
Analysis of ARC system
for gas fast reactor
Filip
Osuský1, Lenka Dujčíková1,
Stefan Cerba1, Gabriel Farkas2,
Branislav Vrban1, Jakub Lüley1
1Slovak University
of Technology, Faculty of Electrical
Engineering and Information Technology,
Institute of Nuclear and Physical
Engineering, Ilkovičova 3, 812 19
Bratislava 1, Slovakia
2Slovak University
of Technology Faculty of Electrical
Engineering and Information Technology
Department of Nuclear Physics and
Technology, Ilkovičova 1, 812 19
Bratislava, Slovakia
filip.osusky@stuba.sk
The paper is focused on
application of assembly reactivity control
(ARC) system within gas fast reactor
(GFR). The ARC system provides negative
reactivity feedback without damaging the
neutron economy. Liquid/liquid system is
used and the idea is that the separate
liquid pushes 6Li in to the core region
after temperature increase. Potassium is
current best choice for the expansion
liquid with low solubility with lithium,
large thermal expansion coefficient, low
neutron absorption cross-section, low
corrosion with the cladding materials and
is chemically stable under irradiation.
The main idea is replacement of one or
more fuel pins by ARC injection rods with
minimal change to fuel assembly. Liquid
reservoir is located in the upper part of
fuel assembly with neutronically
transparent liquid. The ARC injection rod
consists of two concentric tubes where the
inner tube is filled with potassium and
outer tube with argon. The lower reservoir
contains dual-layer of liquids with
floating 6Li on potassium. The absorber in
the form of 6Li is pushed in to the outer
tube with the temperature increase by
thermal expansion of potassium. Different
speed of control system actuation can be
achieved by changing of diameter for inner
and outer tube. Second recriticality of
fast reactor core is discussed based on
the steady state neutronics calculations.
It is assumed that the molten core is
relocated within fixed core boundaries and
new core compaction is responsible for
second recriticality of the nuclear
system. The purpose of the ARC system is
to mitigate such event and to overcome the
issue of too positive coolant temperature
feedback and too large positive coolant
void worth. The analysis provides
reactivity worth of system with different
number and type of ARC rods within the
fuel assembly by SCALE code. The
investigated cases are during normal
operation and during voiding of coolant.
06.09.2016
11:30 Reactor physics II
Reactor
physics - 306
Delayed gamma ray
modeling around irradiated JSI TRIGA
fuel elements by R2S method
Klemen
Ambrožič, Luka Snoj
Jožef
Stefan Institute, Reactor Physics
Department, Jamova cesta 39, 1000
Ljubljana, Slovenia
klemen.ambrozic@ijs.si
The Jožef Stefan
Institute (JSI) TRIGA reactor is a 250 kW,
pool type reactor with fuel elements
arranged in an annular configuration,
which is equipped with numerous
irradiation facilities with well
characterized neutron fields (Snoj,
Žerovnik, & Trkov, 2012) and has
become a reference center for neutron
detector testing for Atlas experiment,
CERN (Cindro, Kramberger, & Mandić,
2005).
Prompt gamma ray production can already be
calculated using existing Monte Carlo
particle transport codes such as MCNP
(Goorley, 2012) and nuclear data libraries
such as ENDF/B-VII.1 (Chadwick, 2011).
However delayed gamma generation and
isotopic changes are commonly not yet
supported. To this end, Rigorous two-step
(R2S) method codes have been developed and
incorporated with different degrees of
accuracy (Batistoni, Angelone, Petrizzi,
& Pillon, 2002). In this article, an
in-house developed R2S method code is
described, and results of its application
for calculation of delayed gamma-ray flux
and dose from activated nuclear fuel, as
well as modifications to the isotopic
concentrations and their contributions to
gamma dose are presented.
Activated nuclear fuel is the dominant
source of delayed gamma rays in the
reactor, and is commonly used as a source
of gamma rays for gamma sample
irradiation.
An R2S method couples Monte Carlo particle
transport codes with neutron activation
and transmutation codes, superimposing a
3D mesh over Monte Carlo model geometry,
where neutron spectrum and total flux are
calculated in each voxel of the mesh.
Delayed gamma ray spectra and intensities
are then calculated in all voxels using
neutron activation code, and input into
the Monte Carlo particle transport model
as gamma sources respectively.
References:
Batistoni, P., Angelone, M., Petrizzi, P.,
& Pillon, M. (2002). Benchmark
Experiment for the Validation of Shut Down
Activation and Dose Rate in a Fusion
Device. Journal of Nuclear Science and
Technology, 39, 974-977.
doi:10.1080/00223131.2002.10875263
Chadwick, H. O. (2011, December).
ENDF/B-VII.1: Nuclear Data for Science and
Technology: Cross Sections, Covariances,
Fission Product Yields and Decay Data.
Nuclear Data Sheets, 112(12), 2887-2996.
Cindro, V., Kramberger, G., & Mandić,
I. (2005). Irradiation studies for the
ATLAS inner detector opto-electronic
readout system for SLHC. 11th Workshop on
Electronics for LHC and Future Experiments
(pp. 311-315). Heidelberg, Germany: CERN.
Goorley, J. T. (2012, December). Initial
MCNP6 Release Overview. Nuclear
Technology(180), 298-315.
Snoj, L., Žerovnik, G., & Trkov, A.
(2012, March). Computational analysis of
irradiation facilities at the JSI TRIGA
reactor. Applied Radiation and Isotopes,
70(3), 483-488.
06.09.2016
11:50 Reactor physics II
Reactor
physics - 307
Generation of Transport
Equivalent Multi-Group Cross Sections
and Diffusion Coefficients for
Neutronic Analysis
Şamil
Osman Gürdal, Mehmet Tombakoglu
Hacettepe
University, Nuclear Engineering
Department, 06800 Beytepe, Ankara, Turkey
mtombak@hacettepe.edu.tr
In this study,
generation of transport equivalent
assembly averaged macroscopic cross
section set using Monte Carlo technique is
discussed for graphite and light water
moderated reactors. One of the
contributions of this study is
demonstration of cell averaging technique
to find an expression for direction
dependent diffusion coefficient using the
simulation results of MCNP5 code with
analytical results obtained by using
diffusion theory. It should be noted that,
reaction rates and flux shapes obtained by
using diffusion theory becomes equivalent
to transport theory results for two group
transport equivalent cross section set and
diffusion parameters. The results are also
compared with the lattice cell code
results for benchmark problems defined in
literature.
06.09.2016
12:10 Reactor physics II
Reactor
physics - 308
Evaluation of criticality
and reaction rate experimental
benchmark in spherical geometry
Tanja
Kaiba1, Gašper Žerovnik1,
Luka Snoj2
1Institut "Jožef
Stefan", Odsek za reaktorsko fiziko,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2Jožef Stefan
Institute, Reactor Physics Department,
Jamova cesta 39, 1000 Ljubljana, Slovenia
tanja.kaiba@ijs.si
Critical and reactor
physics experiments involving aqueous
uranyl fluoride (UO2F2) solutions were
performed at the Oak Ridge National
Laboratory (ORNL) between 1958 and 1960.
In order to determine under which
conditions the aqueous solutions of
intermediate enriched uranium (37 wt%
235U) can be made critical and to
determine basic physical parameters.
Second experimental part was performed to
evaluate reaction rate distribution inside
the sphere. Radial fission rate profile
was measured using two fission chambers,
one was used as a static counter inside
the sphere, while other was positioned
inside guide tube and was moving
vertically through the sphere. Radial
profile was measured relative to the
static counter and relative to the center
of the sphere. The computational model of
the experiments was made in the Monte
Carlo neutron transport code MCNP based on
experimental reports and logbooks. The
model and the MCNP code were than used to
perform the evaluation of experimental
uncertainties of the measured quantities,
i.e. k-eff and fission rate profile,
according to the methodology proposed by
ICSBEP (International Criticality Safety
Benchmark Evaluation Project). Different
contributions to the overall uncertainty
were studied, such as: uncertainties in
solution volume, enrichment, uranium
concentration, solution impurities,
departure from sphericity, surrounding
structure etc. the most important for the
criticality evaluation being uncertainty
in enrichment. It has been found that the
experimental uncertainty is low enough to
consider evaluations to be published in
the ICSBEP and IRPhE (International
Reactor Physics Benchmark Experiment
Evaluation Project). The criticality
benchmark has already been published in
the ICSBEP handbook under the identifier
IEU-SOL-THERM-005, while the evaluation of
reaction rate measurements is still in
progress.
06.09.2016
14:00 Severe Accidents
Severe
Accidents - 801
CoreSOAR Update of the
Core Degradation State-of-the-Art
Report: Status September 2016
Tim
Haste1, Marc Barrachin1,
Georges Repetto2, Martin
Steinbrück3, Paul David
William Bottomley4
1Institut de
Radioprotection et de Sureté Nucléaire,
Bât. 702 Centre de Cadarache, BP 3-13115
Saint Paul lez Durance, France
2Institut de
Radioprotection et de Sureté Nucléaire
(IRSN) Centre d’ Etrudes de Cadarache,
Cadarache B.P 3, Batiment 702, F-13115
Saint Paul-les-Durance CEDEX, France
3Karlsruhe
Institute of Technology, P.O. Box 3640,
76021 Karlsruhe, Germany
4JRC-Directorate
for Nuclear Safety and Security ,
Hermann-von-Helmholtz-Platz 1, P.O. Box
2340,, 76125 Karlsruhe, Germany
tim.haste@irsn.fr
In 1991 the
Commitee on the Safety of Nuclear
Installations (CSNI) published the first
State-of-the-Art Report on In-Vessel Core
Degradation, which was updated to 1995
under the European Union (EU) 3rd
Framework Programme. These covered
phenomena, experimental programmes,
material data, main modelling codes, code
assessments, identification of modelling
needs, and conclusions including the needs
for further research. This knowledge is
relevant to such safety issues as
in-vessel melt retention of the core
(IVMR), recovery of the core by water
reflood, hydrogen generation and fission
product release.
In the following 20 years, there has been
substantial progress in understanding,
with major experimental programmes
finished, such as the integral in-reactor
Phébus FP tests, and others with many
tests completed, e.g. in the integral
ex-reactor QUENCH series on reflooding
degraded rod bundles, and LIVE, on melt
pool behaviour in the lower head. These
are accompanied by separate-effects tests
to study individual phenomena in more
detail. A similar situation obtains
regarding integral modelling codes such as
MELCOR (USA) and ASTEC (Europe) that
encapsulate current knowledge in a
quantitative way. After the two EU-funded
projects on the SARNET network of
excellence, now continuing in the NUGENIA
association, it is timely to take stock of
the knowledge gained.
The CoreSOAR project, in the
NUGENIA/SARNET framework, draws together
the experience of 11 European partners
with the aim of comprehensively updating
the state of the art in core degradation,
over the next two years. The review of
available data has begun, and this paper
as an example indicates recent progress in
small-scale tests involving material
interactions that provide data to cover
the gaps in knowledge identified in the
integral tests such as Phébus FP. This is
complemented by detailed examination of
post-test samples to elucidate the
mechanisms involved. It also considers
advances in thermodynamic databases
necessary to model the formation and
relocation to the lower head of the
complex corium compositions involved, as
well as phenomena related to in-vessel
retention. The report will serve as a
reference point for ongoing research
programmes in NUGENIA, in other EU
research projects such as in Horizon2020,
and in CSNI, e.g. the Fukushima benchmark
BSAF.
06.09.2016
14:20 Severe Accidents
Severe
Accidents - 802
Nordic collaboration:
Impact of Ag and NOx compounds on the
transport of ruthenium in the primary
circuit of NPP in a severe accident
Teemu
Kärkelä1, Ivan Kajan2,
Unto Tapper1, Leena-Sisko
Johansson3, Melany Gouello4
1VTT Technical
Research Centre of Finland, Tietotie 3,
Espoo, 02044 VTT, Finland
2Chalmers
University of Technology, Kemirägen 4,
SE-41296 Goeteborg, Sweden
3Aalto University,
School of Science, P.O. Box 11000, 00076
Aalto, Finland
4VTT, Tietotie 3,
FI-02150 Espoo, Finland
teemu.karkela@vtt.fi
During the
operation of a nuclear power plant (NPP),
a significant amount of ruthenium is built
up in the fuel as a product of the nuclear
fission. The importance of ruthenium from
the radiological point of view is mainly
due to the isotopes 103Ru and 106Ru with
half-lives of 39.35 days and 373.5 days,
respectively. When ruthenium is released
from the fuel to the environment in a
severe NPP accident, these ruthenium
isotopes cause a radiotoxic risk to the
population both in a short and long term
by building-up to the human body and
external exposure to the radiation, thus
possibly leading to a development of
cancer.
The transport of ruthenium through a
reactor coolant system (RCS), after being
released from the fuel, has been
investigated in several experimental
programmes recently. The VTT Ru transport
programme has shown that the release of Ru
from RuO2 powder was dependent on the
oxygen partial pressure in air-steam
atmospheres at 827, 1027, 1227 and 1427
°C. The highest fraction of gaseous RuO4
at the outlet of the model primary circuit
was observed at 1027 °C oxidation
temperature. At higher temperatures,
ruthenium transported mainly as RuO2
aerosol. In the experiments of RUSET
programme it was observed that the
presence of other FPs, e.g. BaO and CeO2,
as mixed with the metallic Ru precursor
when the sample was oxidized at 1100 °C,
decreased the fraction of gaseous RuO4 in
the outlet air over the stainless steel
surface compared to the pure Ru oxidation.
It was also shown that the transport of
RuO4 was dependent on the surface material
in the coolant circuit. In both VTT and
RUSET programmes it was noticed, that the
partial pressure of RuO4 reaching the
outlet of model primary circuit was in the
range of 10-7 to 10-6 bar, which is
significantly higher than what is expected
based on thermodynamic equilibrium
calculations.
As the previous studies have mainly been
conducted in pure air-steam atmospheres,
the current study was dedicated to air
ingress conditions with representative
airborne fission product/control rod (Ag)
and air radiolysis (NOx) species which
were mixed with vaporized Ru oxides. The
aim was to study the impact of these
additives on the transport of ruthenium as
gas and particles through the primary
circuit of nuclear power plant in a severe
accident. As a main outcome, the transport
of gaseous ruthenium through the facility
increased significantly when the oxidizing
NO2 gas was fed into the atmosphere. The
feed of pure silver particles into the gas
flow showed a significant decrease in
gaseous RuO4 reaching the outlet of the
facility. Simultaneously, a noticeable
increase of ruthenium in form of RuO2
trapped on the filter was observed. When
both silver aerosol and NO2 in form of
AgNO3 compound were fed into the
atmosphere, the transport of ruthenium in
gaseous and aerosol forms was promoted.
Based on experiments it was concluded that
the composition of atmosphere in the
primary circuit will have a notable effect
on the speciation of ruthenium transported
into the containment building during a
severe accident.
06.09.2016
14:40 Severe Accidents
Severe
Accidents - 803
Investigation of external
reactor pressure vessel cooling with
ATHLET-CD
Peter
Pandazis1, Sebastian Weber2
1Gesellschaft für
Anlagen- und Reaktorsicherheit
Forschungsgelände, Postfach 12-28, 85748
Garching b. München, Germany
2Gesellschaft für
Reaktorsicherheit (GRS), Schwertnergasse
1, 50667 Köln, Germany
peter.pandazis@grs.de
Severe accident
management (SAM) strategies are coming
into the focus of nuclear safety
investigations after the Fukushima
accident. The main goal of these
strategies is to prevent the release of
radioactive fission products in the
environment. During a severe accident in
light water reactors after SCRAM the fuel
elements and reactor internals may start
to melting due to the decay heat if the
cooling systems fail. The collected molten
fuel and core internals (corium) relocate
into the lower plenum after the collapse
of the grid plate (Pressurized Water
Reactor, PWR) or the control rod guide
tubes (Boiled Water Reactor, BWR) because
of the mass and high temperature of the
corium. Without any further
counter-measure the thermo-chemical attack
of the corium leads to a melt through of
the RPV wall and radioactive materials
release into the environment.
The SAM strategy of in-vessel retention by
ex-vessel cooling has been developed to
minimize the risk of fission product
releases into the containment. According
to this strategy the corium will be
stabilized within the lower plenum by
transferring the decay heat through the
wall into the containment via external
cooling. The external cooling of the RPV
is realized via flooding of the reactor
cavity. Comprehensive works showed that
the success of this SAM concept depends
mainly on the inner thermal load (mass,
composition and decay heat of corium) and
on the coolability of the wall (cooling
channel shape, massflow, roughness, etc.).
In this work a method has been developed
and adopted, using the thermal-hydraulic
system code ATHLET-CD (Analysis of
Thermal-hydraulics of Leaks and Transients
with Core Degradation) developed by GRS,
to investigate the SAM concept of
in-vessel melt retention (IVMR) by
ex-vessel cooling.
The investigations have been performed
using a German generic BWR design and
include the simulations of the transient
corium behavior, the structural response
of the wall as well as the external
cooling process. In the simulations a
Station Black Out accident scenario has
been assumed in addition to the failure of
the core cooling systems. The key-points
of successful cooling have been determined
with the developed model considering the
actual geometry and accident scenario.
Furthermore, the calculations have been
demonstrated the applicability of
ATHLET-CD to perform complex and effective
analyses to evaluate the SAM strategy
IVMR.
06.09.2016
15:00 Severe Accidents
Severe
Accidents - 804
Comparison and analysis
of corium pool behavior in lower head
modeled by MAAP (EDF version) and
PROCOR (CEA) codes
Sophie
Bajard1, Nikolai Bakouta2,
Benoit Habert1, Romain Le
Tellier1, Laurent Saas1
1CEA, Member of
SNETP Executive Committee, Gif sur Yvette
91191, France
2Electricite de
France, 140 Avenue VITON, 13009 Marseille
Cedex 20, France
sophie.bajard@cea.fr
Corium pool formation
in the lower head is likely to occur
during a Severe Accident (SA) in Light
Water Reactors (LWR). In this situation,
the thermal load on the vessel is largely
dependent on the molten pool
stratification in terms of the immiscible
oxide and metal phases.
According to the results given by some
experimental programs (RASPLAV, MASCA,
CORDEB), the number and the positions of
the layers in the corium pool can evolve
within a transient process involving
chemical elements migration. In order to
describe this phenomenology, in-vessel
corium models have been adapted in PROCOR
(CEA) and modified from the original MAAP
models, in MAAP4* and MAAP5* (EDF
proprietary version). So, these codes
share important common features:
- 0D mass and energy conservation
equations for the corium pool layers;
- a kinetic inter-layer mass transfer
model based on a simplified representation
of the miscibility gap in oxide-metal
corium systems and controlled by the
Uranium diffusion in the oxide phase;
- chemically reacting metal and oxide
layers surrounded by an oxide crust;
- a steel layer above the oxide crust.
With the models complexification, the need
for code verification and validation
increases. In this framework, detailed
code cross-comparisons of the most
relevant models with increasing complexity
provide a helpful approach. Accordingly,
several numerical benchmarks have been
constructed by EDF and CEA regarding
in-vessel corium transient modelling
A first set of three calculations have
been performed with MAAP4-EDF and PROCOR,
starting from the simplest (thermal
exchanges of a homogeneous corium pool) to
the most complex (thermochemical and
thermal models with a three layer pool).
For each benchmark, the differences have
been understood and some specific
adjustments (user parameters for example)
were introduced in the codes, in order to
lead to the best estimated results.
A second set of two calculations have then
been performed with MAAP4*, but also
MAAP5* and PROCOR, in order to cover all
the possible corium pool stratifications,
with finally a transient one which can be
representative of a reactor case.
In this paper, first we briefly summarize
the results given by the first set of
calculations, which have already been
presented in ERMSAR 2015 (in Marseille,
France). The causes of the differences
were mainly due to the equation of state
associated to the corium (linking its
enthalpy H to its temperature T) and to
the form of the energy conservation
equation, written in enthalpy in MAAP
whereas it is written in temperature in
PROCOR. Then, the heat transfer
correlations and the physical properties
data also showed some discrepancies. On
this basis, we detail two additional
benchmarks that have been completed
recently. In particular, in the last
benchmark, a more complex transient of the
corium pool in the vessel lower head has
been investigated. This case is more
representative of a reactor case and
involves:
- a transient configuration with a three
layer corium pool simulating the
stratification inversion process;
- a transient migration of the upper steel
layer towards the melt pool, through the
upper surrounding crust.
These benchmark activities have helped
both CEA and EDF to improve their
numerical tool by a better understanding
of the models features and of their
coupling impact on some typical scenarios.
It has also helped to define the most
important R&D tasks to be carried out
for their enrichment or improvement.
MAAP4* / MAAP5*: MAAP4 or MAAP5 version,
modified by EDF for R&D needs
06.09.2016
15:20 Severe Accidents
Severe
Accidents - 805
Detailed
Thermal-Mechanical Modelling of
Cylindrical Core Support Plate During
Severe Accident in PWR
Maciej
Skrzypek, Eleonora Klara Skrzypek
National
Centre for Nuclear Research, ul. Andrzeja
Sołtana 7, Otwock-Świerk, Poland
maciej.skrzypek@ncbj.gov.pl
In the field of severe
accident analysis of nuclear reactors many
calculations focus on corium propagation.
Molten material can drain laterally or
axially to the lower plenum of Reactor
Pressure Vessel (RPV). In case of axial
draining corium can be kept on the core
support plate or drain through, depending
on the thermal and mechanical loads
(strains). Detailed calculations in this
area are necessary to precisely predict
possible rupture of the vessel and time,
because any experimental data exist. Those
parameters can be calculated using Finite
Element Method which can help to obtain a
database of certain analysis.
Implementation of the plate model to fast
running parameter code allows to better
understand phenomena under severe
accidents and determine rupture of the
vessel, taking into consideration also
mechanical analysis, which is very often
omitted by code developer.
06.09.2016
15:40 Posters I
Research
reactors - 204
3D model of Jožef Stefan
Institute TRIGA Mark II Reactor
Anže
Jazbec1, Luka Snoj2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Jožef Stefan
Institute, Reactor Physics Department,
Jamova cesta 39, 1000 Ljubljana, Slovenia
anze.jazbec@ijs.si
Computer assisted
design (CAD) methods have already become
standard in the area of engineering.
However the geometrical data about the
majority of the research reactor is still
in the form of simple drawings and
blueprints. In addition computational
modelling of the reactor geometry for the
purpose of neutron transport calculations
or thermal hydraulics calculations still
heavily relies on manual conversion of
blueprints into the computer format.
Recently, activities were initiated to
develop a 3D model of the completer
reactor including the reactor building in
a 3D CAD format. The motivation for this
was the need to calculate gamma and
neutron dose fields across the whole
reactor hall, reactor basement and
possibly inside control room.
As that attenuation of neutron and gamma
fields is large, therefore standard analog
Monte Carlo methods would not be very
efficient. We either satisfy ourselves
with large deviances or run calculation
for a long time. Since none of the options
is acceptable, a variance reduction
technique will be used. This will be
achieved by calculating weight windows
with the so called CADIS method
implemented in the ADVANTG package [2].
In the paper, the development of the TRIGA
3D model is described and utilisation of
the model discussed.
[1] S.W. Mosher et. al., ADVANTG – An
Automated Variance Reduction Parameter
Generator, ORNL/TM-2013/416, Oak Ridge
National Laboratory, 2013.
07.09.2016
11:30 Thermal Hydraulics II
Research
reactors - 205
Coolant Temperature
Measurements in the core of TRIGA
Research Reactor
Romain
Henry1, Marko Matkovič2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
romain.henry@ijs.si
TRIGA Mark II research
reactor at the "Jožef Stefan" Institute in
Ljubljana is an open-pool type reactor
cooled by demineralised light water. It
has been used in various applications such
as Neutron Activation Analysis, Neutron
Radiography and Tomography and also for
training personnel. Substantial amount of
studies have been done with regard to the
neutron physics, however, only few
experiments related to reactor’s thermal
hydraulics have been performed so far,
which makes the validation of neutron
physics and reactor thermal hydraulics
coupling attempt very difficult. In this
light, two measurement campaigns were
performed.
The first one aims to describe the natural
convection process in the pool of the
reactor [1].The second one focuses on
axial coolant temperature profile
measurement within the reactor core.
Indeed, axial coolant temperature profiles
along the narrow water column confined
with hot fuel elements were acquired
during various modes of reactor operation.
For this purpose, specially tailored
support structure was designed and built
to accommodate 10 thermocouples in a
vertical column within the core. Special
attention was paid to: first, shield the
temperature sensors from the hot surfaces
of the fuel elements, second, keep the
sensor’s tips in contact with local
coolant circulation, and third, generate
reduced amount of activated material.
The obtained experimental results were
properly analysed and compared with CFD
simulations. In fact, the only
measurements of-a-kind were essential as
they produced unique experimental data
suitable for validation of the TRIGA’s
reactor core CFD model, which will further
on be used for coupling attempt with the
neutron physics code.
[1] Henry, Romain, Matkovič, Marko,
Temperature distribution in the pool of
TRIGA mark II reactor during heating
transient. V: 24nd International
Conference Nuclear Energy for New Europe -
NENE 2015, Portorož, September 14-17.
06.09.2016
15:40 Posters I
Research
reactors - 206
TRANSURANUS Code
Performance under Fuel Melting
Conditions: the HEDL P-19 Experiment
Rolando
Calabrese1, Paul Van Uffelen2,
Arndt Schubert2
1ENEA, Via Martiri
di Monte Sole 4, 40129 Bologna, Italy
2European
Commission, Joint Research Centre,
Institute for Transuranium Elements,
Hermann-von-Hermoltz-Platz 1, 76344
Eggenstein-Leopolshafen, Germany
rolando.calabrese@enea.it
A reliable
simulation of fuel pin behaviour is
challenging due to interacting phenomena
such as fission gas release, fuel/cladding
swelling, thermal conductivity
degradation, actinides and oxygen
redistribution. Under fast reactor
conditions, the description of the central
void formation/closure is even more
complex when power rating and geometrical
conditions lead to a partial melting of
fuel.
The HEDL P-19 experiment was conducted in
the EBR-II reactor addressing the
relationship between fuel/cladding gap
width and power-to-melt. The experiment
was focused on the behaviour of MOX fuel
rods without preconditioning irradiation.
During the experiment, power was increased
up to the reactor design level (62.5 MW)
which was maintained for about ten
minutes. Post irradiation analyses
provided information regarding the axial
extension of fuel melting, central void
formation, columnar grain region
formation, and fuel/cladding gap width.
The experiment was conducted on sixteen
encapsulated pins containing MOX fuel with
an enrichment in plutonium of about 25
wt.%. The diametral fuel/cladding gap
width was tailored in each pin (0.086 mm
to 0.250 mm) while the cladding outer
diameter was either 6.35 mm or 5.84 mm.
The peak linear rating reached during the
HEDL P-19 experiment was estimated to be
in the interval 538 - 679 W/cm.
After a review of literature, the
experiment was modelled by means of the
TRANSURANUS code (2015 version) aiming to
assess the performance under FBR
conditions in particular considering the
fuel melting correlation recommended for
MOX. In addition, further information
about fuel restructuring such as central
void formation and plutonium
redistribution models have been analysed.
06.09.2016
15:40 Posters I
Research
reactors - 207
Triga Reactor Simulator
Jan
Malec1, Dan Toškan1,
Luka Snoj2
1Institut "Jožef
Stefan", Odsek za reaktorsko fiziko,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2Jožef Stefan
Institute, Reactor Physics Department,
Jamova cesta 39, 1000 Ljubljana, Slovenia
jan.malec@student.fmf.uni-lj.si
A real time TRIGA
reactor simulator was developed at the
Jožef Stefan Institute. It’s primary goal
is to help educate students and future
reactor operators, especially in
developing counteries with no access to a
research reactor. The behaviour of the
reactor simulated resembles that of a
TRIGA reactor. Its output is simulated by
numerically solving point kinetics
equations in the 6 group approximation and
is validated by analizing the simulator’s
output with a digital reactivity
meter[1][2]. A simple thermodynamic model
has been implemented to simulate negative
temperature effects on reactivity. The
reactor provides multiple modes of
operation. In manual mode, the operator
has full control of the control rods. In
Automatic mode, the control positions are
automatically adjusted to maintain a
desired reactor power. In square wave
mode, the reactor rods are partially
inserted and
ejected periodically and in pulse mode, a
control rod can be quickly ejected to
simulate a fast transient. The reactor
behaviour in the pulse mode is simulated
in the Fuchs-Hans approximation. A
graphical user interface enables the user
to operate the reactor, visualize and
analyze data such as temperature, power,
reactivity, and concentration of delayed
neutron precursors.
References
[1] A. Trkov, Digital Reactivity Meter
DMR-043, September 13, 2004, “Jožef
Štefan”
Institute, Ljubljana, Slovenia
[2] LENGAR, Igor, TRKOV, Andrej, KROMAR,
Marjan, SNOJ, Luka. Digital meter of
reactivity for use during zero-power
physics tests at the Krško NPP ISSN
1855-5748.
[Tiskana izd.], feb. 2012
06.09.2016
15:40 Posters I
Reactor
physics - 309
Neutron streaming
analysis and shielding determination
for the Krško nuclear power plant
Bor
Kos1, Marjan Kromar2,
Žiga Štancar1, Peter
Klenovšek3, Luka Snoj4
1Jožef Stefan
Institute, Reactor Physics Department ,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2Institut "Jožef
Stefan", Odsek za reaktorsko fiziko,
Jamova cesta 39, 1000 Ljubljana, Slovenia
3Nuklearna
elektrarna Krško, Vrbina 12, 8270 Krško,
Slovenia
4Jožef Stefan
Institute, Reactor Physics Department,
Jamova cesta 39, 1000 Ljubljana, Slovenia
bor.kos@ijs.si
At the Nuclear Power
Plant Krško it was decided to evaluate
possibilities of minimizing neutron
streaming radiation dose rates from the
reactor core to reasonably low levels to
reduce staff radiation exposure entering
and working in the reactor building and
also to reduce long term effects on
radiation-sensitive equipment exposed to
neutron radiation. This action is
particularly important for locations where
significant neutron streaming is present.
The purpose of the shield determination is
to reduce the neutron dose for personnel
in SG (steam generator) and RCP (reactor
coolant pump) cubicles during reactor
operations low as reasonable achievable.
A detailed geometrical model of the Krško
nuclear power plant for Monte Carlo
neutron transport calculation was made
based on CAD (computer assisted design)
models, blueprints, technical drawings and
other available data. Monte Carlo
simulations to determine the absolute and
relative dose fields are performed with
the general purposes Monte Carlo neutron
transport code MCNP. Monte Carlo
calculations are coupled with
deterministic neutron transport codes to
determine optimal variance reduction
parameters, such as cell importances.
In the paper basic (conceptual) design of
shields placed at locations around RCS
(reactor coolant system) hot and cold leg
piping entering SG and RCP cubicles to
reduce neutrons streaming through the RCS
loop pipe penetrations is presented. In
order to determine optimal position and
dimensions of the shield, a thorough
parametric analysis is performed including
assessing the importance of neutron
streaming through the reflective
insulation of the RCS piping. A conceptual
and physics analysis of shielding is made
in order to enhance understanding in the
neutron transport from the reactor core to
the cubicle and to determine the neutron
paths and most important components from
neutronic point of view.
06.09.2016
15:40 Posters I
Reactor
physics - 310
SCALE 6.1.3 and Serpent
2.1.24 criticality safety analysis of
a Fukushima Daiichi-like Spent Fuel
Pool
Antonio
Guglielmelli1, Federico
Rocchi2, Giacomino Bandini2
1Italian National
Agency for New Technology, Energy and
Substainable Economic Development, Via
Martiri di Monte Sole, 4 - Bologna ,
40129, Italy
2ENEA, Via Martiri
di Monte Sole 4, 40129 Bologna, Italy
antonio.guglielmelli@gmail.com
The Fukushima Daiichi
nuclear power plant accident has
highlighted the risk of criticality safety
problems in the spent fuel pools (SFPs)
used to store fresh and/or burnt fuel
assemblies of a nuclear reactor under
specific conditions. In a SFP the fuel
arrangement and the geometrical
configuration must be designed to keep
such a system with a given subcriticality
margin to ensure criticality safety under
both operational and credible accidental
conditions. In the framework of the
Nugenia+ AIR-SFP Project, and with the aim
to verify the existence of the proper
critical safety margin under accidental
conditions in a SFP similar to that of the
Fukushima Daiichi Unit 4, a series of
calculations have been executed with the
Montecarlo code KENO VI of SCALE 6.1.3
package by means of both continuous-energy
and multigroup cross sections based on the
ENDF/B-VII.0 library. The criticality
simulations have been performed both on a
single unit-cell of a rack and on a whole
3x10 SFP rack. The fuel assembly
considered was a fresh 9x9-9 BWR FA
equipped with 12 Gd-doped pins whose
material and geometrical description have
been taken from the specifications of the
OECD/NEA Burn-up Credit Criticality
Benchmark Phase IIIC. Preliminarily, the
criticality of the initial safe state
(isothermal @ 25 °C , water density 1
g/cm3) and that of accidental conditions
with uniform properties (isothermal @ 100
°C, water density between 1 and 0.1
g/cm3); have been evaluated. Subsequently
- employing the results of some RELAP5
thermo-hydraulic calculations – the
criticality safety margin for more
realistic accidental conditions
(non-isothermal fuel, water density
distribution) has been estimated. The
thermo-hydraulic simulations have been
achieved assuming a loss of coolant
accident (LOCA) with a partial fuel
uncover, a “rod bundle” correlation, and a
fuel decay heat of 9.0 kW corresponding to
a decay of about 15 days after shutdown
and at burn-up of 12 GWd/MTU. The effect
of a burn-up of 12 GWd/MTU and of the
corresponding depletion of gadolinium on
the criticality margin has been taken into
account a-posteriori using a lumped,
pre-estimated reactivity increase.
Finally, it has also been realized a
sensitivity analysis to estimate the
effect on criticality of different rack
unit interspacing.
06.09.2016
15:40 Posters I
Reactor
physics - 311
Analysis of operational
history of the JSI TRIGA for the
purpose of benchmarking burnup
calculations
Anže
Pungerčič1, Luka Snoj2
1Jožef Stefan
Institute, Reactor Physics Department ,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2Jožef Stefan
Institute, Reactor Physics Department,
Jamova cesta 39, 1000 Ljubljana, Slovenia
anze.pungercic@student.fmf.uni-lj.si
The TRIGA reactor
at the JSI started operating on 31st May
1966. Since then 320 different fuel
elements were used, arranged into 218
reactor cores. Our goal is to simulate 50
years of operation by using deterministic
(TRIGLAvW) and stochastic (SERPENT)
neutron transport and burnup codes and
validate the calculations by experimental
and operational data. In order to perform
this a complete reactor operational
history had to be analysed and put into a
computer readable format. First important
quantity, directly connected with burnup
through energy released from fission of a
nucleus (e.g. Uranium-235), is the thermal
energy generated in the reactor. This data
is stored in reactor operation logbooks,
which are written by hand, therefore
digitalization is required. The logbooks
also contain information regarding the
fuel shuffling between different reactor
cores and excess reactivity, which is
measured every Monday since the beginning.
Fuel element burnup could be determined
with well-known methods: reactor
calculations, gamma ray spectrometry of
irradiated fuel, measurement of the
elements relative reactivity worth. Last
two methods cannot be performed for
majority of the fuel elements as more than
200 elements were shipped back to USA in
1999.
The main purpose of this paper is to
describe the analysis of all information
required for burnup simulations and show
the quantity of reactor cores used in
calculations with Monte Carlo. The fuel
element burnup accumulated during
1966-2016 will be calculated with two
codes; the TRIGLAV fuel management
two-dimensional multigroup diffusion code
and the Monte Carlo neutron transport code
SERPENT. Burnup modules in both codes will
be compared against each other and
calculated excess reactivities will be
compared against the measurements.
06.09.2016
15:40 Posters I
Reactor
physics - 312
Analysis of the primary
water activation in a typical PWR
Andrej
Žohar1, Luka Snoj2
1Jožef Stefan
Institute, Reactor Physics Department ,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2Jožef Stefan
Institute, Reactor Physics Department,
Jamova cesta 39, 1000 Ljubljana, Slovenia
andrej.zohar@student.fmf.uni-lj.si
In pressurised water
reactors the cooling water in primary loop
is radioactive due to activation of the
water itself, activation of corrosion
products, migration of fission and
activation products through the cladding.
We are going to analyse activation and
activity of water in the model of a
typical two loop PWR nuclear power plant.
We will focus mostly on the most important
nuclides, i.e. activation of different
oxygen nuclides, especially O-16, due to
high natural abundance (99.76%) and high
energy gamma radiation of activated
product N-16 (~ 6 and 7 MeV). Other
important oxygen nuclides are O-17 and
O-18, due to high energy neutrons from
N-17 decay (~ 1 MeV) and high energy gamma
from O-19 decay (~ 0.2 and 1.4 MeV).
The neutron spectra and reaction rate
calculations is performed using the Monte
Carlo neutron transport code MCNP. In
parallel water activation is calculated by
using activation codes FISPACT and ACAB.
The calculations will be performed with
various up to date nuclear data libraries,
such as ENDF/B-VII.0, TENDL-2015, JEFF-3.2
and EAF-2010. It can be observed that some
nuclear data libraries lack the cross
section energy dependence for activation
of O-18. Among the previously mentioned
nuclear data libraries, library
ENDF/B-VII.0 lacks this cross section
energy dependence. In addition, some cross
section energy dependences from different
libraries differ significantly between
each other, especially for the activation
of O-18.
The neutron spectrum calculations are
performed for hot zero power conditions at
various positions inside the reactor
pressure vessel.
Calculations of the time dependence of
activity for each presented activated
nuclide will also be carried out. The time
dependence of activity is going to include
behaviour at the start of the reactor,
behaviour at changes of power of the
reactor and behaviour after shutdown of
reactor. The results will then serve to
describe a gamma ray source in subsequent
Monte Carlo photon transport calculations
to evaluate dose fields around the primary
loop.
06.09.2016
15:40 Posters I
Reactor
physics - 313
Analytic function
expansion nodal method (AFEN) for
solving SP3 and diffusion equations in
hexagonal geometry
Mohammad
Hasan Jalili Bahabadi1, Ali
Pazirandeh2
1Department of
Nuclear Engineering, Science and Research
Branch, Islamic Azad University, Tehran,
Iran, 1477893855, Iran
2Islamic Azad
University, Science and Research Campus
Islamic Azad University, Hesarak St,
Ashrafi Isfahani Boulevard, Tehran
1454696111, Iran
jalili.mohammadhasan@gmail.com
In this paper, two
nuclear codes named MGHANSP3 and HexDANM
are introduced. In MGHANSP3 and HexDANM
codes, the AFEN method was utilized to
solve SP3 and diffusion equations in
hexagonal geometry respectively. This
method represents a multidimensional intra
nodal flux distribution in terms of
analytic basis functions at any points in
the node. the surface averaged partial
currents in half of the surfaces of the
hexagon adopted at the nodal boundary
coupling conditions. Finally, the IAEA
benchmark problem was used to comparison
of SP3 and diffusion theory. The numerical
results show that the MGHANSP3 code more
accurate and also slower than HexDANM.
06.09.2016
15:40 Posters I
Reactor
physics - 314
Validation of the ADVANTG
for neutron fields in three-section
concrete labyrinth experimental
benchmark from Cf-252 neutron source
Domen
Kotnik
Jožef
Stefan Institute, Reactor Physics
Department , Jamova cesta 39, 1000
Ljubljana, Slovenia
domen.kotnik@student.fmf.uni-lj.si
ADVANTG, Automated
Variance Reduction Generator [1], is a
code developed by the Oak Ridge National
Laboratory that aims to automate the
process of generating variance reduction
parameters for fixed source MCNP
calculations. As it was released recently
it has not been tested on majority of
available experimental benchmarks.
The purpose of this paper is to validate
the use of ADVANTG on the ICSBEP
(International Criticality Safety
Benchmark Evaluation Project) shielding
benchmark ALARM-CF-AIR-LAB-001, i.e.
neutron fields in three-section concrete
labyrinth from Cf-252 neutron source [2].
Experimental investigation of the neutron
flux in a large three-section concrete
labyrinth was done in the summer of 1982
in an open area at the Institute of High
Energy Physics at Protvino, near Serpukhov
(Moscow Region), Russia. The source of
neutrons for these experiments was
spontaneous decay of 252Cf. The source was
installed at the center of the doorway
aperture of the labyrinth. The experiments
were performed with “unfiltered” radiation
from the bare source and with “filtered”
radiation from the source surrounded by a
30.5-cm-diameter polyethylene sphere with
a 4-cm-diameter spherical central cavity.
Neutron flux was measured by the Bonner
sphere method at different points inside
each section of the labyrinth. The
influence of different coverings of the
labyrinth wall on the neutron flux in
remote sections of the labyrinth was
investigated. The aim of the experiments
was to obtain benchmark data for
validation of the computer codes used for
estimation of doses from the neutrons that
penetrate the shielding, via numerous
leaks, of the acceleration-storage ring of
the Large Serpukhov Proton Accelerator.
The calculations are divided in 2-steps
process because of the geometry of the
labyrinth and consequently extremely low
number of 6Li (n, ?) reactions in the
detectors crystal. First step is
calculation of the response function
(sensitivity) of the detectors (Bonner
spheres) and then calculation of the
neutron flux at points of the measurement
positions.
Results from a MCNP calculations fits very
well with the experiment. In order to get
statistical reliable results MCNP need
long computational time because of the
extremely low number of neutrons which
toward to detectors {reactions}. Hence
ADVANTG is used to speed up calculation.
References:
[1] S.W. Mosher, A.M. Bevill, S.R.
Johnson, A.M. Ibrahim, C.R. Daily, T.M.
Evans, J.C. Wagner, J.O. Johnson, R.E.
Grove. 2013. ADVANTG-An Automated Variance
Reduction Parameter Generator. Oak Ridge
National Laboratory, Oak Ridge.
[2] E.A. Belogorlov et al. 1984. Neutron
fields investigation in three sectional
concrete labyrinth from Cf-252 source,
ALARM-CF-AIR-LAB-001. Serpukhov, IHEP.
06.09.2016
15:40 Posters I
Reactor
physics - 315
Current Developments of
the VVER Core Analysis Code KARATE-440
György
Hegyi1, András Keresztúri2,
Csaba Maráczy2, Emese
Temesvári2, István Panka3
1Retired, nn, nn,
Hungary
2Centre for Energy
Research Hungarian Academy of Sciences,
P.O.Box 49, H-1525 Budapest, Hungary
3Hungarian Academy
of Sciences Centre for Energy Research,
Budapest 114, P.O. Box 49, Hungary,
H-1525, Hungary
gyorgy.hegyi@energia.mta.hu
Due to the new challenges
(more heterogeneous and higher enriched
fuel assembly, the safety requirements)
the updating of nodal codes for steady
state and transient core analysis is
carried on continuously to enhance the
accuracy and robustness. The fuel
modifications and the upgraded regimes
requiring more accurate calculations have
necessitated the further development and
validation of the KARATE code system.
On the other hand even the calculations
have reached a high quality level; it is
very important to take into account the
uncertainties of the calculations,
especially the uncertainties of the input
parameters related to the applied models
which cannot be eliminated. A realistic
estimation of these uncertainties is
necessary for judging the reliability of
the simulation results. Recently there is
a tendency to use best estimate plus
uncertainty methods in the field of
nuclear energy. This implies the
application of best estimate code systems
and the determination of the appropriate
uncertainties.
Taking account of these goals, the
following improvements were implemented
into the KARATE-440 code system:
• more detailed parametrization of the few
group constants (making the accurate
calculations of the cores containing fuel
relaxed for a longer time possible),
corresponding renewal of the multigroup
libraries and the parametrized few group
constants,
• application of the more accurate
calculation of the power distribution at
the core periphery by using albedo
matrices from Monte Carlo calculations,
• capabilities to handle the uncertainties
of the basic nuclear data and the
technological parameters.
The updated code has been verified by some
standard calculations made for a VVER-440
core.
06.09.2016
15:40 Posters I
Reactor
physics - 316
2-D reflector modelling
for VENUS-2 MOX Core Benchmark
Dusan
Ćalić1, Andrej Trkov2
1ZEL-EN razvojni
center energetike, Hočevarjev trg 1, 8270
Krško, Slovenia
2International
Atomic Energy Agency, Wagramerstr. 5,
P.O.Box 100, A-1400 Vienna, Austria
dusan.calic@zel-en.si
The choice of the
reflector model is important issue in full
core calculations. In 2015 [1] new
approach was developed where the existent
WIMSD code for lattice cell calculations
was replaced with Monte Carlo code Serpent
2 in order to have new reference full core
calculation. However the Serpent-GNOMER
simulation code uses simplified 1-D
reflector model. In order to develop 2-D
reflector model the VENUS-2 benchmark was
proposed as a reference model. The main
aim of this paper is to present the
development of the 2-D reflector model
based on VENUS-2 benchmark.
[1] Ćalić D., Trkov A., Monte Carlo
Simulation on KRŠKO NPP. In: PHYSOR-2016.
Sun Valley, Idaho, May 1-5, 2016.
06.09.2016
09:50 Reactor physics I
Reactor
physics - 317
3D Cartesian TRIGA
reactor model quality assessment by
radial power distribution
Vid
Merljak1, Andrej Trkov2
1Jožef Stefan
Institute, Reactor Physics Department ,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2International
Atomic Energy Agency, Wagramerstr. 5,
P.O.Box 100, A-1400 Vienna, Austria
vid.merljak@ijs.si
Numerical simulations
provide useful insight into the behaviour
of neutron population in a nuclear
reactor. Indeed, defining the problem
geometry is a crucial point where biases
are introduced to a bigger or lesser
extent. Recently, a Cartesian 3D
geometrical model of the Jožef Stefan
Institute’s TRIGA Mark II research reactor
has been developed for use with the GNOMER
diffusion code. This model suffers from an
inherent deficiency since the true reactor
geometry is cylindrical and a Cartesian
approximation had to be used to comply
with the GNOMER’s capabilities.
Nevertheless, one can find good use of it.
Experiments such as measurements of
relative quantities (e.g. control rod
reactivity worth) can easily be simulated.
As a continuation of assessing the quality
of the geometrical model in question, this
paper presents comparison of radial power
distribution as calculated by three
different computer codes using various
levels of model complexity: a cylindrical
2D model for TRIGLAV, a 3D Cartesian model
for GNOMER and both Cartesian and detailed
cylindrical 3D models for MCNP. Results
are separated into two sections, namely
determining the error due to geometry
simplifications and the error due to
approximations used in underlying theory
(e.g. diffusion equation vs. Monte Carlo
stochastic approach, the procedure of
generating the nuclear cross-sections,
etc.). It can be concluded that the GNOMER
3D Cartesian model is adequate for
qualitative purposes. This is further
confirmed by explaining most of the
observed discrepancies. Combined with
previous research, this paper represents a
firm foundation for the model’s practical
use, particularly with respect of its
strengths and weaknesses.
06.09.2016
15:40 Posters I
Reactor
physics - 318
Simplified In Core Fuel
Management Software for Education and
Training
Erhan
Şenlik, Mehmet Tombakoglu
Hacettepe
University, Nuclear Engineering
Department, 06800 Beytepe, Ankara, Turkey
mtombak@hacettepe.edu.tr
In this study,
simplified in-core-fuel management
software was developed to model one and
two dimensional in core fuel management
problems.
This study consists of six computer
programs. These programs are based on one
and two dimensional core loading pattern
neutronic solvers and genetic algorithm
optimization software. Neutronic
parameters of Almaraz II Nuclear Power
Plant data is utilized to perform power
and burnup dependent full core
calculations.
Simulation platform uses 1- and 2-D burnup
dependent neutronic solver and they are
coded in FORTRAN to acquire results
quickly. To perform constraint and
unconstrained optimization, genetic
algorithm was developed and it is also
coded in FORTRAN.
Remaining programs are graphical user
interface programs which were coded in
Python programming language. Calculation
programs are; 1-DNodal, RPM-HUNEM and
RPM-Genetic, graphical user interface
programs are; Py1DNodal, PyRPM,
PyRPM-Genetic.
1DNodal software is based on 1 dimensional
core loading pattern. Py1DNodal software
is a graphical user interface for loading
pre-chosen fuel types as an input to the
1DNodal software.
PyRPM code is used to specify fuel loading
pattern, burn-up and power for RPM-HUNEM
calculation software. PyRPM-Genetic
graphical interface is designed for
supplying number of fuel assemblies and
required input parameters used in genetic
algorithm software. The developed software
has been tested using the benchmark data
of Almaraz II Nuclear Power plant with
different inputs and all are open for
further development.
06.09.2016
15:40 Posters I
Reactor
physics - 322
Determination of the
Computational Bias in Criticality
Safety Validation of VVER-440/V213
Branislav
Vrban1, Jakub Lüley1,
Štefan Čerba1, Filip Osuský2
1B&J NUCLEAR
ltd., Alžbetin Dvor 145, 90042 Miloslavov,
Slovakia
2Slovak University
of Technology, Faculty of Electrical
Engineering and Information Technology,
Institute of Nuclear and Physical
Engineering, Ilkovičova 3, 812 19
Bratislava 1, Slovakia
filip.osusky@stuba.sk
The key issue in any
criticality safety problem is to estimate
and to predict the deviation of
calculation from reality. If the
calculated value is not equal to its true
value bias occurs. In criticality
calculations the computational bias is the
difference between the computed and the
actual value of keff. The fundamental
assumption is that the computational bias
is mostly caused by errors in the
cross-section data. In addition the use of
random variables in the calculation
introduces a non-random bias in the
computed result as well. The American
National Standards are utilized to predict
and bound the computational bias of
criticality calculations. These standards
require the validation of the analytical
methods and data used in nuclear
criticality safety calculations to
quantify the computational bias and its
uncertainty. This paper presents a method
for determining the computation bias and
bias uncertainty for VVER-440/V213
reactor. For this analysis a SCALE KENO 3D
core model was developed by B&J
NUCLEAR ltd company. This model is based
on technical data and operational history
of NPP Jaslovské Bohunice provided by the
Slovenské elektrárne a.s. The operational
conditions were defined for the end of
campaign for which a fuel assembly-wise
isotopic compositions in one sixth
symmetry were calculated. The
concentration of the boric acid was below
1 g per kg of water and the sixth group of
control assembly was 18 cm below the upper
position. Several calculation steps are
used to address bias estimation method
including sensitivity analysis,
uncertainty analyses and cross section
adjustment method. In addition the
neutronic similarity of VVER-440/V213 core
to several hundred critical benchmark
experiments is evaluated by the use of
three integral indices. The database of
the benchmark experiment is based on the
selection and processing procedure VALID
provided by the Oak Ridge National
Laboratory and specified in the ICSBEP
Handbook. Systems with similar
sensitivities to nuclear data
uncertainties are expected to be computed
to comparable accuracy so identification
of similar integral experiments supports
the accuracy of the determined
computational bias. In cases where
experimental benchmarks are available to
validate specific nuclides, sensitivity
and uncertainty analysis are used to
project biases observed in the benchmarks
to biases appropriate for the safety
system. The results of all analyses
performed are given and discussed in the
paper.
06.09.2016
15:40 Posters I
Reactor
physics - 323
Recent development and
examples of the use of the Windows
interface environment XSUN-2016 for
transport and sensitivity-uncertainty
analysis
Ivan
Aleksander Kodeli, Slavko Slavič
Jožef
Stefan Institute, Reactor Physics
Department , Jamova cesta 39, 1000
Ljubljana, Slovenia
ivan.kodeli@ijs.si
In 2013 the first
version of the Windows interface XSUN-2013
facilitating the deterministic radiation
transport and cross-section
sensitivity-uncertainty calculation was
developed and submitted to OECD/NEA Data
Bank Computer Code Collection. The package
allows the preparation of input cards,
rapid modification and execution of the
complete chain of codes including TRANSX,
ARTISN and SUSD3D in a user-friendly way.
Recent updates of the code utility and
several examples of its use will be
presented, including cases such as:
- Transport, sensitivity and uncertainty
analysis of the MYRRHA accelerator driven
system (ADS), including both keff and
__eff parameters;
- Analysis of several benchmark
experiments from the IRPhE and ICSBEP
databases (SNEAK-7A & -7B, JEZEBEL,
FLATTOP-Pu, etc.
The performance of the code system will be
compared with those of other codes such as
TSUNAMI, SERPENT and MCNP6.
06.09.2016
15:40 Posters I
Reactor
physics - 324
Experimental and
calculated data on criticality of
uranium-zirconium hydride systems with
45% enriched Uranium-235
Svyatoslav
Sikorin1, Siarhei Mandzik2,
Andrei Kuzmin2, Tatsiana
Hryharovich3
1The Joint
Institute of Power and Nuclear
Research-Sosny of the National Academy of
Sciences of Belarus , PO BOX 119, 220109
Minsk, Belarus
2POWER, Izpolni
naslov!, USA
3The Joint
Institute of Power and Nuclear Research -
Sosny of the National Academy of Sciences
of Belarus , PO BOX 119, 220109 Minsk,
Belarus
sikorin@inbox.ru
The critical facilities
“Rose”, “Edelweis”, “Liliya”, “Astra”,
GFS, “Crystal” and “Giacint” of the Joint
Institute for Power and Nuclear Research –
Sosny of the National Academy of Science
of Belarus have been used for over 45
years to generate and investigate more
than a hundred uranium-containing critical
assemblies with different material
compositions, structures and targeted use,
including uranium-water, uranium-alcohol,
uranium-polyethylene and uranium-zirconium
hydride multiplication systems and systems
without moderators, with fuel rods and
fuel assemblies with 10, 21, 36, 45, 75
and 90% enriched uranium-235, as well as
with natural and depleted uranium. Also
were researched uranium-water critical
assemblies with rotating annular vortex
core based of small diameter fuel
particles with 90%-enriched uranium-235
and others multiplication systems. The
uranium-zirconium hydride experiments were
performed with 21, 36 and 45% enriched
uranium-235. Beryllium, zirconium hydride
and stainless steel were used in the
reflector.
The paper presents criticality data
produced at the critical facility
“Crystal” for several uranium-zirconium
hydride systems, representing non-uniform
multiple zones heterogeneous
uranium-zirconium hydride lattices
comprising hexagonal fuel assemblies with
cylindrical fuel rods, absorbing plates
and rods, zirconium hydride and steel side
and end reflectors. The critical
assemblies represented the cores collected
from three types of fuel assemblies with
different structure, surrounded by
assemblies and units of a side reflector.
The core included channels for the
regulating rods. The moderator – ZrH1.89.
The fuel composition – UO2-Ni-Cr with 45 %
uranium-235 enrichment. The absorber in
plates – B with 85 % boron-10 enrichment.
The absorber in rods – Eu2O3. The results
of experiments on critical facilities with
zirconium hydride have been analyzed by
creating detailed calculation models. The
analyses used the MCNP and MCU computer
programs. The paper presents
configurations of the studied
uranium-zirconium hydride critical
assemblies, as well as the experimental
and calculation results.
06.09.2016
15:40 Posters I
Reactor
physics - 326
Construction of a Monte
Carlo Benchmark Pressurized Water
Reactor Core Model
Žiga
Štancar, Marjan Kromar, Bor Kos, Luka
Snoj
Jožef
Stefan Institute, Reactor Physics
Department , Jamova cesta 39, 1000
Ljubljana, Slovenia
ziga.stancar@ijs.si
The main goal of the
presented project was to construct a
geometrical Monte Carlo model of a
typical pressurized water reactor core.
The motivation behind the creation of
such a detailed model was to obtain a
computational tool which could be
utilized to perform reactor calculations
in support of core design, analyze
reactor core parameters and produce an
accurate description of the reactor core
as a source of neutrons. The latter
could be used for computationally
intense Monte Carlo neutrons transport
simulations in the surroundings of the
reactor pressure vessel, namely at the
positions of ex-core detectors and steam
generator cubicles located behind the
concrete biological shield.
The basic element of the core model is
represented by the fuel pin – it
comprises 10 axial fuel and water
coolant slices, Zircaloy cladding, inert
gas, a safety spring and simplified
straps. The pin cells are arranged into
a 16 × 16 square lattice, which,
together with the integral fuel burnable
absorber (IFBA) rods, control rod guide
thimbles and instrumentation guide
tubes, forms a fuel assembly. At the
upper and lower part of an individual
assembly, stainless steel nozzles are
added. The 121 fuel assemblies are
arranged into the reactor core and are
enclosed with the baffle wall and a
water gap separating the core from the
pressure vessel thermal shield.
Additionally fuel burn-up data is also
implemented into the model using the
output of the CORD-2 code package,
intended for deterministic reactor core
design calculations. The data needed for
modelling of burnup in a Monte Carlo
model is the position of the fuel pin,
fuel axial slice, fuel temperature, fuel
density and its isotopic composition
composed of 158 isotopes. In the paper
the constructed model is presented
together with the results of a
preliminary testing of the model, which
allows for a check-up of the geometry,
computational demand of the model and
the design of a subroutine capable of
automated Monte Carlo input generation
using burnup data. Upon the completion
of the initial model trials the addition
of control rods, analysis of the effect
of the neglected fission products in the
burnt fuel on the effective
multiplication factor of the reactor
model, the preparation of the neutron
source for ex-core computations and
possible comparison with measured core
physical parameters will be performed.
06.09.2016
15:40 Posters I
Reactor
Operation - 402
Developing a New Neutron
and Reactivity Monitoring System for
Paks NPP
Sándor
Kiss1, Sándor Lipcsei1,
Gábor Házi2, Zoltán Dezső2,
Tamás Parkó3, István Pós3,
Miklós Ignits3, László
Hományi4
1Centre for Energy
Research, Hungarian Academy of Sciences ,
Konkoly Thege M. út 29-33, H-1121, Hungary
2Centre for Energy
Research Hungarian Academy of Sciences,
P.O.Box 49, H-1525 Budapest, Hungary
3MVM Paks Nuclear
Power Plant Ltd., P.O. Box 71, H-7031
Paks, Hungary
4KFKI-Regtron Ltd,
Hungary, Address, ZIP, Hungary
lipcsei.sandor@energia.mta.hu
The Reactivity
Monitoring System and the Refuelling
Neutron Monitoring System of Paks NPP are
aged and need to be reconstructed. Since
both systems are based on neutron flux
measurements, the new system is to be
served by the same detectors and
measurement instrumentation. In order to
provide data during refuelling, start-up
and at full power, a full-range system is
required, i.e. the detectors and the
connected instrumentation should span the
full range of neutron flux measurements
from 0% to 100% of the reactor power.
Additionally, the new system is required
to operate continuously, to build a
measurement archive, and to provide data
for the Process Computer and the VERONA
core monitoring system. In order to span
the full neutron flux range, Photonis
CFUL08 type ionisation chamber was chosen.
The interface electronics will serve all
three operation modes of the detector:
impulse, Campbell (AC) and current (DC)
modes. In order to obtain high reliability
and dependability, the system will be
built from independent and redundant
components.
06.09.2016
15:40 Posters I
Reactor
Operation - 404
DMReS, Digital Reacivity
Meter of the new Generation
Slavko
Slavič1, Andrej Trkov2,
Bojan Žefran1
1Jožef Stefan
Institute, Reactor Physics Department ,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2International
Atomic Energy Agency, Wagramerstr. 5,
P.O.Box 100, A-1400 Vienna, Austria
slavko.slavic@ijs.si
A Digital Meter of
Reactivity (DMR) was developed, which
solves the point kinetics equations taking
into account the source term. It is
composed of a programmable picoameter,
AD/DA converters, a PC computer and the
accompanying software. From the start, the
DMR proved to be superior to other similar
devices, due to its ability to provide
correct results in a very broad range of
reactor operating conditions, including
measurements in a deeply subcritical
reactor. Special solutions was implemented
in the DMR in order to cover broad range
of cases.
The rod-insertion method makes full use of
the special features available in the DMR.
It is used for bank-worth measurements,
during which a control bank is inserted
into the core with the control rod drive
mechanism at normal speed. No reactivity
compensation is required; thus, the method
is much faster than other available
methods. The method has routinely been
used during the start-up tests at the
Krško NPP since 1990; it allowed to
shorten the time required for the start-up
tests from several days to only 12 hours.
After the success in NEK, the
rod-insertion method was adopted by others
and is today used in several power plants
around the world.
The DMR is used also for other
measurements during start-up tests, e.g.
for the isothermal temperature coefficient
determination, perhaps the most important
parameter that has to be satisfied at all
times during a power reactor operation.
The DMR device is indispensable for its
determination. The temperature coefficient
is defined as the change in reactivity
with respect to the change of the reactor
core temperature and must be negative
throughout the cycle. During the
measurement, the temperature of the
coolant water is slowly lowered by
approximately 2°C within half an hour and
then raised again to the initial value.
The reactivity is closely monitored. Since
the absolute value of the coefficient is
usually small at the beginning of each
cycle, the accurate DMR has to be used for
reactivity recording. The method enables
very fast evaluation of data and the
results are available immediately after
the measurement.
The Windows version of the program, DMReS,
was created with new graphical user
interface (GUI). DMReS program uses a
completely new way of reading data using
DLL routines. Graphical representation of
the results is completely new, making use
of all the advantages of the Windows
environment. Validation of the new DMReS
code will be demonstrated in the paper.
06.09.2016
15:40 Posters I
Reactor
Operation - 405
Non-Destructive Testing
of Reactor Pressure Vessel Nozzle
Petar
Mateljak
INETEC-Institute
for Nuclear Technology, Dolenica 28, 10250
Zagreb, Croatia
petar.mateljak@inetec.hr
The reactor pressure
vessel (RPV) is an integral part of the
reactor coolant pressure boundary.
Ensuring safe operation of nuclear steam
supply installation is a main and
obligatory condition for the operation of
all power units. One of the most important
measures in fulfilling these requirements
is periodical inspection of the condition
of base metal, welded joints and RPV
austenitic steel overlaying welding. As
key part of reactor pressure vessel
structural integrity, RPV nozzle sections
in nuclear reactor pressure vessels are
classed as critical components, requiring
regular inspection to verify their
integrity. Early detection of cracks is
essential, however inspection costs are
high. A plant shutdown costs operators an
estimated €800,000 per day and a typical
outage takes around 20 days to complete.
INETEC has recently developed new system
for inspection of reactor vessel nozzle
from the inside. Taking into account
increasing requirements to the safety
enhancement during plant operation,
shortening of the inspection time,
radiation exposure to examination
personnel and cutting the total inspection
costs, new system is designed as
underwater automated robotic system with
integrated NDT equipment. This paper
describes the system’s capabilities and
features with focus on recent design
evolutions.
Key words: reactor pressure vessel nozzle,
non-destructive testing, automated
inspection
06.09.2016
15:40 Posters I
Reactor
Operation - 406
A feasibility study on
in-core fuel management via Quantum
Particle Swarm optimization
Francesca
Giacobbo1, Gabriele Tavelli2,
Antonio Cammi2, Marco Cauzzi2
1Politecnico di
Milano, Dipartimento di Energia, Via
Ponzio 34/3 , 20133 Milano, Italy
2Politecnico di
Milano - department of energy, Via La Masa
34, 20156 Milano, Italy
gabriele.tavelli@mail.polimi.it
Nowadays the increasing
needs of optimizing the operations of fuel
loading in a nuclear reactor core have
been calling for efficient and reliable
methods to determine suitable
configurations of fuel assemblies able to
maximize or minimize required neutronic or
engineering features (i.e. multiplication
factor keff, reactivity swing k, power
peaking factor).
The current paper leads an investigation
over already existing and widely employed
optimization algorithms, aiming to target
and further develop a promising method to
tackle in-core fuel management
optimization problems. Comparisons between
algorithms were focalized on the
fundamental goals of maximizing final
result’s accuracy and minimizing
computational time required.
Given the intrinsic complexity of the
issues under consideration, it looked
reasonable to resort to global
optimization methods to account for the
possible existence of local optima. To
this aim, Genetic Algorithms (GA) and
Particle Swarm (PS) were selected, tested
and eventually compared using an
analytical multi-variable continuous test
function with plenty of local optimum
points.
The capabilities of both algorithms to not
get stuck into local optima but to
converge towards the global optimum were
examined. Results obtained showed Particle
Swarm to have better performance with
respect to both accuracy and machine time.
Therefore, the successive analyses were
focused on particle swarm based algorithms
such as Discrete Particle Swarm (D-PS)
(Kennedy and Eberhart, 1995) and its
updated version, Discrete Quantum Particle
Swarm (D-QPS) (Sun et al., 2004, 2005).
Given a number of fuel assemblies with
different enrichments, the optimization of
neutronic and engineering features is
strictly related to the design of a
precise core fuel spatial configuration.
Discrete Particle Swarm and a modified
version of the original Discrete Quantum
Particle Swarm algorithm, here proposed by
the present authors, were tested with
respect to their efficiencies and
performances on a reactor physics case
study.
The adopted case study regards an
extremely simplified PWR core made of 5x5
fuel assemblies: 16 enriched with a
content of U235 up to 3% (weight) and 9
enriched with a content of U235 up to 5%
(weight). The goal was to optimize fuel
assemblies disposition in order to obtain
the highest value of the multiplication
factor, keff. Given the extreme simplicity
of this case study the global solution is
a priori known. The calculations were
performed coupling Particle Swarm based
algorithms with Serpent Monte Carlo code.
Obtained results showed a clear
superiority of the Quantum Particle Swarm
based algorithm proposed by the authors
which speaks in favour about its
application to more complex in-core fuel
management optimization cases.
References
Kennedy, J., Eberhart, R., 1995. Particle
swarm optimization. In: IEEE International
Conference on Neural Networks (ICNN, 95),
vol. 4. IEEE, Perth, Western Australia.
Sun, J., Feng, B., Xu, W.B., 2004.
Particle swarm optimization with particles
having quantum behavior. In: IEEE
Proceedings of Congress on, Evolutionary
Computation, pp. 325–331.
Sun, J., Feng, B., Xu, W.B., 2005.
Adaptive parameter control for quantum
behaved particle swarm optimization on
individual level. In Proceedings of the
2005 IEEE International Conference on
Systems, Man and Cybernetics, Piscataway,
NJ, pp. 3049–3054.
06.09.2016
15:40 Posters I
Reactor
Operation - 407
Possibility of nuclear
cogeneration development in the region
of Paks
Török
Szabina, Börcsök Endre, Talamon Attila
Neznana
organizacija, Unknown Organisation ,
Address, ZIP, Slovenia
torok.szabina@energia.mta.hu
Almost half of GHG
emission of world’s energy sectors is
related to heat generation. The
development of nuclear cogeneration offers
a convenient possibility for emission
reduction; however examination of economic
constrains is essential. This study
focused on heat demand of households in
the vicinity of Paks NPP and compares
economic and environmental aspects of
domestic heating alternatives. In first
part of our work we analyze
competitiveness of nuclear cogeneration in
district heating sector and in the second
part we consider the optimal heating
alternatives for different building
typological groups taking into account
economic and environmental aspects,
distance from Paks NPP and heat demand
density. We found that development of
nuclear cogeneration is valuable above 17
€/tCO2 price and with alrady existing
district heating network. In districts
with high heat demand density nuclear
cogeneration based district heating can be
competitive with stand-alone heaters if
environmental externalities are also
considered.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 508
Thermal-hydraulic
Analysis Code for Plate-type Fuel
Nuclear Reactors
Duvan
Alejandro Castellanos Gonzalez, Pedro
Carajilescov, Jose Maiorino
Universidade
Federal do ABC - PROGRAMA DE PÓS-GRADUAÇAO
EM ENERGIA, Av. dos Estados, 5001. Bairro
Bangu. Santo André - SP , 09210-580,
Brazil
duvan.castellanos@ufabc.edu.br
The use of plate-type
fuel assembly, in nuclear reactors, are
mostly associated to researched reactors
and naval propulsion reactors (aircraft
carriers and submarines), bringing
immediate benefits in security and
thermal-hydraulic performance of the
reactor. Computational codes are used to
calculating the thermal-hydraulic core
behavior. This project presents the
development of thermal-hydraulic code for
reactors with plate-type fuel elements,
written in FORTRAN. According to geometric
input data, operational and boundary
conditions, the code involves the analysis
of steady state flow and power regime,
solving the conservation equations for
mass, momentum and energy. Furthermore, it
performs the calculation of minimum DNBR,
based on the analysis of critical channel.
The code has maximized the radial mesh
with the use of the chain or cascade
method for two stages: in the first stage,
the core is subdivided in sub channels
with size equivalent to a fuel assembly
and the second stage, the hot fuel
assembly is subdivided in sub channels
with size equivalent to the one channel
that comprise. For the program validation,
it was considered the research reactor
CARR (China Advance Research Reactor), and
the LABGENE reactor (Brazilian reactor of
naval propulsion). The code yields
detailed information of reactor core as
the change of the static pressure in the
channel, flux distribution, variation of
coolant temperature and coolant
velocities, quality and local flux heat in
the critical channel. The analysis showed
good agreement compared to the results
obtained for CARR reactor and for a
typical reactor power PWR.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 509
The influence of imposed
gas velocity profile on wave dynamics
in the simulation of vertical
air-water churn flow
Matej
Tekavčič1, Boštjan Končar1,
Ivo Kljenak2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Jožef Stefan
Institute, Reactor Engineering Division ,
Jamova cesta 39, 1000 Ljubljana, Slovenia
matej.tekavcic@ijs.si
A three-dimensional
transient simulation of isothermal churn
flow of air and water in 19 mm internal
diameter vertical pipe was performed. The
churn flow regime in vertical pipes can be
viewed as a transitional regime between
slug flow and annular flow and is often
related to the onset of the flooding
phenomena, which is of particular interest
for safety analyses of the loss-of-coolant
accident in light water nuclear reactors.
Single-fluid interface capturing approach
based on the volume of fluid method with
interface compression was used to model
the gas-liquid interactions. A short
section of a vertical pipe with perforated
wall was simulated representing the region
near a typical liquid inlet section in
experiments, where large waves of liquid
travelling upwards can typically be
observed in the churn flow regime. One of
our previous attempts to model such
flooding type liquid waves with
computational fluid dynamics approach
showed a systematic over-prediction of
wave frequencies compared to the values
reported from experiments available in the
open literature.
The influence of different gas velocity
profiles on the wave dynamics in the pipe
is investigated in the present paper.
Three different types of gas velocity
profiles imposed at the bottom of the
vertical pipe computational domain were
chosen to represent a model for gas flow
predicted by k-omega SST turbulent model
and two extreme deviations from it. The
flow conditions – the values for mass flow
of gas and liquid - are taken from
experiments from the literature with gas
Reynolds number between 7000 and 10000.
The effect of the gas velocity profile on
the calculated frequency of waves is
presented. Results for time-averaged axial
and radial profiles of pressure, velocity,
and liquid volume fraction are presented.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 510
Analysis of the MSIV
Closure Transient Simulation in APROS
Tadeja
Polach1, Ivica Bašić2,
Luka Štrubelj3
1ZEL-EN razvojni
center energetike, Hočevarjev trg 1, 8270
Krško, Slovenia
2APoSS d.o.o.,
Repovec 23b, 49210 Zabok, Croatia
3GEN energija
d.o.o., Vrbina 17, 8270 Krško, Slovenia
tadeja.polach@zel-en.si
The Slovenian Krško
Nuclear Power Plant (NEK) model was built
in using APROS - Advanced PROcess
Simulation environment. The basis for the
this model was the RELAP5/MOD3.3
Engineering Handbook, the model was
updated to the 26th cycle and also
includes the upflow conversion
modification.
A detailed model nodalisation was created
for each system and every system was
separately validated. The current model
covers the primary circuit with the core
kinetics model, the secondary circuit and
their control systems. The steady state
model already having been validated the
plan is to validate the model for some
transients and design basis accidents. In
this article the plant behaviour after the
Main Steam Isolation Valve (MSIV) closure.
Two scenarios of the closure are
performed. In the first both MSIVs are
closed at the full power operation and in
the second only one MSIV is closed, again
at full power operation.
Upon initiation of the of the MSIV closure
the control system signal actuations and
their times were followed and the
responses of different affected systems
were being observed. All those recorded
values were then compared with the
identical transient performed on the
similar NEK model with the RELAP5/MOD3.3
system code. This procedure allowed to
bring the current APROS NEK model one step
forward towards being assured to have
accurate calculations.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 511
Spectral Element Direct
Numerical Simulation Of Sodium Flow
Over A Backward Facing Step
Jure
Oder1, Jernej Urankar2,
Iztok Tiselj1
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
jure.oder@ijs.si
In this paper we
present the direct numerical simulations
of a turbulent flow of a liquid metal past
the backward-facing step (BFS) with finite
dimensions. The BFS geometry can be
visualised as a channel, where one of the
walls has a shape of a step. The flow is
flowing from the narrower part to the
wider part. The simulations are performed
in three dimensions.
For the inflow boundary condition over the
BFS, a fully developed turbulent velocity
field is used. To obtain this fully
developed turbulent inflow, a separate
domain is constructed. It has a geometry
that corresponds to the geometry of the
channel before the step, the narrow part.
The boundary conditions within this
subdomain are set to be periodic in the
direction of general flow. A velocity
field in a plane from this subdomain is
then used as an inflow boundary condition
for the flow over the BFS. The flow in the
subdomain is calculated simultaneously
with the simulation of the flow over the
step.
Simulations are performed with the NEK5000
code. The most notable feature of this
code is the use of spectral elements to
solve for velocity, temperature and any
other passive scalar. It is an open source
code developed by the Argonne National
Laboratory.
Spectral element method is a hybrid method
between finite element method and a
collocation spectral method. The method
divides the computational domain into
finite elements, within which a spectral
method is used to solve for variables.
This method allows for the use of spectral
method in irregularly shaped geometries
and to perform direct numerical
simulations in such geometries.
The main purpose of this work is to test
the numerical set-up to later perform
calculations with temperature field as a
passive scalar. Dimensional walls with
internal heating will be added to simulate
the heat production in the walls.
This work is part of work that is
performed within the SESAME project of
Horizon2020 research programme and is a
continuation of research at our
department.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 512
Development of Turbulent
Mixing Layer in Horizontal Confined
Two-Component Flow
Rok
Krpan1, Boštjan Končar2
1Jožef Stefan
Institute, Reactor Engineering Division,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
rok.krpan@amis.net
Interaction of two
fluid liquid streams with different
densities and temperatures is of
particular interest in nuclear industry.
Notable examples are turbulent mixing in
junctions of primary coolant piping or
boron mixing in reactor core of the
pressurized water reactor. The turbulent
flow phenomena in the wake mixing zone, in
which two streams interact, develop
differently depending on the physical
properties of the two liquids involved.
The purpose of this work is to predict the
development of turbulent mixing layer due
to mixing of two water streams with
slightly different densities in a
horizontal square duct. The mixing of such
flows can be modelled as the flow of two
components, where the concentration of one
component in the mixing zone can be
described as a passive scalar. Velocity
field remains common over the entire
computational domain and is affected by
density difference due to concentration.
Different CFD codes and turbulence models
will be used for simulations and the
results will be compared with experimental
data. The main goal of the study is to
demonstrate the capabilities of different
codes and modelling approaches to predict
the differences in turbulent flow
phenomena in the wake mixing zone between
the single liquid case and two-component
liquid case. Computational results
obtained with OpenFOAM and ANSYS FLUENT
code will be compared with GEMIX
experimental data, obtained from Paul
Scherrer Institute in Switzerland.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 513
Experiments on bubbly to
slug flow transition in a vertical
cylindrical tube
Matic
Kunšek1, Daisuke Ito2,
Yasushi Saito2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Kyoto University
Research Reactor Institute, Izpolni
naslov!, Osaka, Japan
matic.kunsek@ijs.si
Two-phase fluid flows
can be found in numerous industrial
installations, such as nuclear power
plants, chemical, biological and
biomedical reactors, within different
industrial fields. A special case of
two-phase flow, which is excellently
suited for experimental work, is air-water
two-phase flow. Both fluids are easily
accessible and, with their use, adiabatic
system of steam-water is easily simulated.
This system enables one of the key
processes in nuclear and other power
plants as well. Steam-water two-phase flow
can be found in the cores at working
conditions in boiling water reactors, in
all steam generators and in the cores of
pressurized water reactors during some
hypothetical accidents. Because of that,
the knowledge of behavior of those systems
is very important. Although two-phase
flows can be found all around us and that
research field exists for decades, a lot
of questions still remain unanswered
because of complexity of interactions
between phases.
In the proposed paper, measurements of
void fraction in bubbly and bubbly to slug
flow transition regimes are presented.
First, the description of wire mesh sensor
construction is shown. Then, the
measurements of two-phase air-water flow
at 20 different flow conditions (liquid
and gas flow rate) are described. In the
end, the processing of data is explained
and the results are presented, analyzed
and the bubbly to slug flow transition is
identified.
The work for the proposed paper was done
at the Heat transfer laboratory of Kyoto
University Research Reactor Institute
(KURRI) in Osaka prefecture in Japan.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 515
Evaluation of
Non-condensable Gas Effect on the
Operation of Emergency Core Cooling
System during LBLOCA
Seunghun
Yoo, Kwang-Won Seul, Young-Seok Bang
Korea
Institute of Nuclear Safety, 34 Gwahak-ro,
Yuseong-gu , Daejeon 305-338, South Korea
k720ysh@kins.re.kr
The gas accumulation in
the piping of diverse fluidic systems has
occurred since first commercial operation
of nuclear power plants. If the gas is
accumulated in the Engineered Safety
Features (ESF) such as Emergency Core
Cooling System (ECCS), the gas can
increase the potential to damage the pipe
and components and it may cause the
condition that the ESFs is inoperable
during the Design Basis Accident. On the
basis of the pump which is an important
active component in the ESFs, if the gas
is accumulated in the pump suction, it can
induce the pump cavitation or influence
the other pumps sharing common suction
pipe. If the gas is accumulated in the
pump discharge pipe, it can cause the
water hammering and damage the pipe and
its associated systems. Despite of the
significance of the gas accumulation
consequences, the gas transport and its
effect on the Design Basis Accident were
not comprehensively considered in the
current Final Safety Analysis Report.
Therefore it is necessary to understand
and clarify the consequences of the gas
accumulation in the ESFs and during the
Design Basis Accident.
In this study, the gas accumulation effect
on the operation of the ECCS during Large
Break Loss of Coolant Accident (LBLOCA)
was evaluated as dividing two analyses: a)
the effect on High and Low Pressure Safety
Injection Pumps (HPSIP and LPSIP) with the
condition that the gas is located in the
pump suction, b) the gas transport
behavior in the ECCS during LBLOCA.
Shin-Kori unit 1 and 2 was selected as a
target plant for this study. RELAP5/Mod
3.3 Patch 4 was used to model HPSIP, LPSIP
and ECCS. For the first analysis, the
single phase homologous curve and the two
phase head multipliers for the pumps were
newly modified to reflect the
characteristics of Shin-Kori’s SIPs. The
hypothetical non-condensable gas was
injected into the pump suctions and the
changes of the pump suction void fraction,
suction flow regime, pump velocity, head
and flow rate were evaluated to quantify
the degradation of the pumps. And the flow
condition in the pipe was evaluated as
calculating the Froude Number. For the
second analysis, the precise piping system
for the pump upstream which was covered
from Refueling Water Tank to each pump
suction was modeled as reflecting minor
losses caused by geometrical change of
pipe, valve or orifice, etc. As assuming
that the gas can be randomly accumulated
in the ECCS, the gas transport trajectory
for the randomly accumulated points and
finally transported locations in the ECCS
were identified. On the basis of the
analysis, the gas transport behavior for
the the hypothetical amount of gas which
located in the final accumulation points
wad evaluated as simulating the LBLOCA
condition that modeled by the cold leg’s
pressure and temperature changes using the
time dependent volume. Moreover, the flow
condition for the ECCS during LBLOCA was
analyzed by calculating the Froude Numbers
for diverse locations.
As a result, we have quantified the
degradation of the Shin-Kori’s SIPs due to
non-condensable gas. When the gas is
randomly accumulated at the diverse points
of the ECCS, the finally transported
locations were identified. And the gas
transport behavior in the ECCS during the
LBLOCA was analyzed and vulnerable points
were identified. The flow conditions for
the whole ECCS locations, which may help
to judge whether the gas is transported or
not, was evaluated as the Froude Number.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 517
Modeling of NEK Steam
Line Break analysis in computer code
Apros 6
Jure
Jazbinšek1, Luka Štrubelj2,
Klemen Debelak2, Ivica Bašić3
1ZEL-EN razvojni
center energetike, Hočevarjev trg 1, 8270
Krško, Slovenia
2GEN energija
d.o.o., Vrbina 17, 8270 Krško, Slovenia
3APoSS d.o.o.,
Repovec 23b, 49210 Zabok, Croatia
jure.jazbinsek@zel-en.si
Model of Nuclear power
plant Krško (NEK) developed in computer
code Apros 6 was upgraded with reactor
containment building and compared to Krško
NPP RELAP and GOTHIC code models developed
by FER.
Best estimate Apros 6 analysis of
double-ended Main Steam Line Break (MSLB)
transient was simulated. MSLB transient
between Steam Generator (SG) outlet and
Main Steam Isolation Valve (MSIV), so the
blowdown of affected SG could not be
prevented, was assumed.
When the break occurs, the rapid steam
flow to containment building occurs from
affected SG, causing rapid cooling and
pressure drop of Reactor Coolant System
(RCS). Containment absolute pressure,
temperature and liquid void fraction in
Apros 6 simulation will be compared to
results of GOTHIC code. The discharged
liquid and gas volume from MSLB event and
corresponding main control room variables
in Apros 6 simulation will be compared to
RELAP5 results.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 518
Assessment of
Condensation Heat Transfer Models of
MARS-KS and TRACE Codes Using PASCAL
Test
Kyung
Won Lee1, Aeju Cheong2,
Andong Shin2
1Korea Institute of
Nuclear Safety , 62 Gwahak-ro, Yuseong-gu,
Daejeon, 34142, South Korea
2Korea Institute of
Nuclear Safety, 34 Gwahak-ro, Yuseong-gu ,
Daejeon 305-338, South Korea
leekw@kins.re.kr
The advanced power
reactor plus (APR+) is a GEN-III+ nuclear
power plant, the standard design of which
is currently being developed in Korea. The
passive auxiliary feedwater system (PAFS)
is one of the advanced safety features
adopted in the APR+ and is design to
replace the conventional active auxiliary
feedwater system. During the plant
transient, PAFS cools down the secondary
side of steam generator, and eventually
remove the decay heat of the reactor core
by condensing steam in nearly-horizontal
U-shaped tubes (passive condensation heat
exchanger, PCHX) submerged inside the
passive condensation cooling tank (PCCT).
In order to validate the operational
performance of the PAFS, Korea Atomic
Energy Research Institute (KAERI) has
performed the experimental investigation
using the PASCAL (PAFS Condensing heat
removal Assessment Loop) facility. The
PASCAL simulates a single U-shaped tube
with the volumetric scaling ratio of
1/240. The dimension and material of the
tube are the same as the prototype. The
inner and outer diameter of the PCHX are
44.8 mm and 50.8 mm, respectively. The
tube length is 8.4 m. The width and depth
of the PCCT are 6.7 m and 0.112 m,
respectively. The height of the PCCT is
11.484 m.
In this study, we assess the applicability
of the condensation heat transfer models
of MARS-KS and TRACE codes to the
nearly-horizontal condensing tube using
the PASCAL SS-540-P1 test. In the MARS-KS
input model, the PCHX, is modeled using
the PIPE component with the 28 cells. The
steam-supply line and the
condensate-return line are modeled using
the time dependent volumes. The PCCT is
modeled with the multi-D component. The
HTSTR component is used to model the heat
transfer between the PCHX and the PCCT.
The TRACE input model has the same
nodalization as the MARS-KS input model.
The PCCT is modeled with the 3-D vessel
component.
The calculation results of heat flux,
steam and condensate temperatures are
compared with the experimental data. The
results show that MARS-KS slightly
under-predict the heat fluxes. However,
TRACE over-predicts the heat fluxes at the
tube entrance region and under-predicts
the heat fluxes at the tube exit region.
When compared to MARS-KS results, TRACE
provides more reasonable condensate
temperatures.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 519
The Effect of Tube
Arrangement and Turbulence Models for
Steady Flow Past Tube Bundles
Ali
Tiftikci, Cemil Kocar
Hacettepe
University, Nuclear Engineering
Department, 06800 Beytepe, Ankara, Turkey
alitiftikci@hacettepe.edu.tr
In present work,
three-dimensional flow in tube bundles for
different tube configurations is studied.
In-line and staggered type bundle
arrangements are used. The tube
configurations are 3.6x1.6 and 3.6x2.1 for
staggered arrays and 3.6x2.1 for in-line
arrays. The lattice-Boltzmann method is
used for numerical calculations. The
validation processes are made by using the
experimental data available in ERCOFTAC
Database (Case 80). The turbulence models
LES (Smagorinsky-Lilly) and VLES (k-omega)
are selected for comparison. Also, mesh
sensitivity analyses are made. Post
processed quantities such as axial and
transverse mean velocity and corresponding
rms (root mean square) velocity profiles
at different locations are compared with
the experimental data. The simulation
results of LES and VLES are in good
agreement with the experiment for
different tube bundle arrangements.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 521
Simulation of a station
blackout transient using TRACE5.
Application to ATLAS facility.
Maria
Lorduy1, Jara Turégano Lara2,
Sergio Gallardo1, Gumersindo
Verdú1
1Universidad
Politecnica de Valencia, Departamento de
Ingeniería Química y Nuclear, Camino de
Vera s/n, 46022 Valencia, Spain
2Department of
Chemical and Nuclear Engineering,
Polytechnic University of Valencia, Camí
de Vera sn, 46022 Valencia, Spain
sergalbe@iqn.upv.es
The purpose of this
work is to test the capability of the
TRACE5 code in the simulation of a Station
Black Out (SBO) transient with delayed
asymmetric secondary cooling in the frame
of the OECD-ATLAS project. In this
proposal, test A1.1 is analysed.
A TRACE5 model of the ATLAS facility has
been developed in order to simulate both
steady state conditions and the SBO
transient. This facility is designed to
simulate transients and accidents in
APR1400 reactors, operating under the same
prototypic pressure and temperature of its
reference plant. Due to the especial
characteristics of this transient
regarding to the possible asymmetries, it
is necessary to study the 3D behaviour of
coolant in the vessel. A 3D-vessel
component has been used to simulate the
real vessel of the facility. In this work,
a detailed analysis of some components of
the model is performed, especially of the
pressurizer and the steam generators
relief valves. Furthermore, main steam
lines and the line to the refuelling water
tank is analysed to best estimate the
water discharge performance.
The test simulates a prolonged station
black out with delayed asymmetric
secondary cooling through the supply of
auxiliary feedwater only in one steam
generator. The main goal of this test is
to analyze the primary cool-down
performance by delayed asymmetric
secondary cooling as an accident
mitigation measure. The transient
comprises two phases: the first phase
simulates the station blackout transient
without auxiliary feedwater, and the
second phase simulates the activation of
auxiliary feedwater only in one steam
generator.
The transient is analysed by means of the
main thermal hydraulic variables: primary
and secondary pressures, mass flow rates,
discharged coolant inventory through the
pressurizer and steam generators relief
valves, collapsed water level in the
vessel and in the pressurizer, fluid
temperatures in hot and cold legs, etc.
The results show that the TRACE5 code is
able to successfully reproduce the main
physical phenomena observed during the
experiment.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 523
UHS Cooling Pond
Evaluation using NUREG-0693
Methodology
Davor
Grgić1, Nikola Čavlina2,
Tomislav Fancev1
1University of
Zagreb, Faculty of Electrical Engineering
and Computing , Unska 3, 10000 Zagreb,
Croatia
2Fakultet
elektrotehnike i računarstva Zagreb, Unska
3, 10000 ZAGREB, Croatia
davor.grgic@fer.hr
The main objective of a
Ultimate Heat Sink (UHS) is to provide
cooling water for nuclear power plant
safety related systems. It dissipates
residual heat after reactor shutdown and
after an accident through cooling
components of the Essential Service Water
(ESW) System and the Component Cooling
Water (CC) system. All accident analyses
described in SAR Chapter 15 are performed
taking into account some boundary
conditions directly or indirectly based on
UHS temperature. From licensing (US NRC)
point of view UHS role is covered by three
General Design Criteria (GDC) of Appendix
A to 10 CFR Part 50, GDC 2 "Design bases
for protection against natural phenomena",
GDC 5 "Sharing of structures, systems and
components", and GDC 44 "Cooling water".
Main UHS licensing requirements are
formulated in Regulatory Guide (RG) 1.27
"Ultimate Heat Sink for Nuclear Power
Plants", and acceptable supporting
analysis, at least for the case when
cooling pond is used as UHS, is described
in NUREG-0693, "Analysis of Ultimate Heat
Sink Cooling Ponds".
NPP Krsko is using Sava river as one
ultimate heat sink with dual role. In
normal situation the river flow is
directly providing cooling water for ESW
heat exchanger (the requirements are in
the form of minimum river flow rate
(guaranteed NPSH for ESW pumps) and
maximum allowed water temperature), and in
highly unlikely situation that river flow
is stopped, remaining water is forming
cooling pond upstream of the river dam
(the requirements are in the form of
minimum initial required water volume and
its temperature during mentioned
interval).
In this paper the analysis using
NUREG-0693 methodology was performed to
check behavior of NPP Krsko cooling pond
temperature and volume during time
interval of 30 days after DBA LOCA
coincident with loss of Sava flow.
NUREG-0693 uses a simple mathematical
model of a cooling pond to scan weather
data to determine the period of the time
for which the most adverse pond
temperature or rate of evaporation would
occur. Once the most adverse conditions
are available, the peak pond temperature
is determined for given heat load coming
from the plant.
06.09.2016
15:40 Posters I
Thermal
Hydraulics - 524
Estimation of SFDS Cask
Heat-up after Blockage of Ventilation
Openings
Davor
Grgić, Siniša Šadek, Vesna Benčik
University
of Zagreb, Faculty of Electrical
Engineering and Computing , Unska 3, 10000
Zagreb, Croatia
davor.grgic@fer.hr
Spent Fuel Dry Storage
(SFDS) is becoming, especially after
Fukushima accident, popular alternative to
spent fuel pool storage of spent fuel
elements. Usual requirements for SFDS
casks are structural robustness in all
conditions, guaranteed subcriticality of
the content, low contact dose rate and
adequate passive cooling of the fuel. Two
basic designs of the SFDS casks exist, one
using metal and other using concrete
casks. One of the usual assumptions in
safety analysis of the concrete cask
design is loss of natural circulation
between cask’s insert and body of the cask
due to inlet and outlet openings blockage.
It is required to demonstrate how long it
takes till reaching limiting fuel cladding
temperature in adverse conditions. The
simple calculation model for GOTHIC code
was developed based on publically
available data (Safety Analysis Report)
for HOLTEC vertical concrete cask system.
For given heat loading steady state
temperature distribution is calculated as
well as subsequent heat-up during extended
period of time (7 days) after closing top
and bottom ventilation holes. In the
specific case it took 5 days to reach the
targeted temperature of 570 oC (with more
conservative assumptions it could be
less), but taking into account that only
approximate data were available for
geometry of the cask, the model is mainly
used for estimation of heat-up trends and
for quantification of different heat
transfer mechanisms relative importance.
06.09.2016
15:40 Posters I
Materials
- 601
Microstructural
evaluation of creep behavior in
hydrided E110 cladding
Hygreeva
Kiran Namburi
Research
centre Rez, Hlavni 130, 250 68
Husinec-Řež, Czech Republic
hygreeva.namburi@cvrez.cz
Creep is considered as
one of the dominant damage mechanisms in
zirconium based fuel cladding's during
reactor operation and spent fuel in dry
storage. This paper emphases on the
microscopic examination results from the
E110 cladding in its virgin state and
after high temperature creep test at dry
storage conditions. Test specimen was
oxidized in an autoclave at a pressure of
10.7 MPa and temperature of 425 °C (150
ppm by weight H). Creep test was performed
at UJP , Praha in a horizontal furnace
under the conditions: internal pressure -
4 MPa, temperature -530 °C, exposure time
-30 hours.
TEM observations were made on creep tested
specimen to reveal interaction of
dislocations, secondary precipitate
particles and hydrides with grain
boundaries owing to different deformed
zones.The material is characterized by
polyhedral grains of ? - Zr phase with
hexagonal lattice and lattice parameters a
= b = 3.136 A, c = 5.039 A, ? = ß = 90 °,
? = 120 ° the grain size ranges from ~ 3-5
microns. Grains are recrystallized grains
with boundaries straight or slightly
curved, and often there are sub-grains and
twinning. Numerous dislocations and
precipitates of secondary phases in ? -
phase were observed.
06.09.2016
15:40 Posters I
Materials
- 603
Macroscopic Validation of
the Micromechanical Model for
Neutron-Irradiated Stainless Steel
Samir
El Shawish1, Leon Cizelj1,
Jeremy Hure2, Benoit Tanguy2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2CEA, Member of
SNETP Executive Committee, Gif sur Yvette
91191, France
samir.elshawish@ijs.si
Severe irradiation
conditions in nuclear reactors may limit
the operational life of internal
structural components supporting the
reactor core. Neutron irradiation of
austenitic stainless steels that
constitute internal structures of reactors
may results in the deterioration of
mechanical and fracture properties. In
combination with corrosive environment of
the primary water, irradiation induced
changes in microstructure and
microchemistry may make those steels more
sensitive to Irradiation-Assisted Stress
Corrosion Cracking (IASCC). IASCC has led
to several failures of baffle-to-former
bolts in pressurized water reactors, by
initiation and propagation of
intergranular cracks. The safe operation
and reliable structural integrity of power
plants require precise prediction of the
initiation and propagation of IASCC to
establish a strategy for inspection and
replacement, especially for long life
operation over 60 years. While predictive
models of irradiation-induced hardening
are available, reliable predictions of
IASCC sensitivity are currently
unavailable.
Recently, the authors from CEA, France,
developed a micromechanical crystal
plasticity model to describe a nonlinear
mechanical response of austenitic
stainless steel subjected to neutron
irradiation. The model, based on
dislocation dynamics inferred mechanisms
and finite strain theory, is able to
capture the irradiation-induced hardening
followed by softening during plastic
deformation. In collaboration between CEA
and JSI, the model was used in finite
element simulations of realistic stainless
steel wire aggregate obtained from X-ray
tomography, leading to distributions of
stresses at grain boundaries. These local
stresses are the driving force of
intergranular cracking and need to be
accurately determined in order to reliably
predict IASCC.
In this study, the existing crystal
plasticity model is validated
macroscopically through a series of
simulation tests of tensile specimens
performed on a wide range of deformations
and irradiation levels. An emphasis is put
on finding and using a converged finite
element aggregate model to adjust
constitutive law parameters by performing
tensile tests up to and beyond necking,
where specimen geometry and boundary
conditions become increasingly important.
In the past, to avoid long simulation
times due to complexity of the crystalline
law, the initial identification of model
parameters was carried out by tensile
tests on a simplified, cubic
polycrystalline aggregate composed of 343
cubic grains with 1 element per grain.
Such an aggregate is too coarse to be a
representative volume element of the
model. In this work, therefore, an
upgraded polycrystalline model is built
with real specimen geometry and realistic
boundary conditions. To study the effect
of grain topology, Voronoi tessellations
with different mesh refining are used and
compared to simpler cubical grain
aggregates. In addition, a novel automatic
fitting approach is introduced where model
parameters are allowed to take only
discrete values in order to speed up the
calibration procedure. The adjustment of
model parameters is done with respect to
measurements on 304L stainless steel.
06.09.2016
15:40 Posters I
Materials
- 604
Drop Test Analysis of
Reinforced Concrete Disposal Container
Miha
Kramar1, Franc Sinur2,
Matija Gams1
1Zavod za
gradbeništvo Slovenije, Dimičeva 12, 1000
LJUBLJANA, Slovenia
2IBE, d.d.,
Hajdrihova 4, 1000 LJUBLJANA, Slovenia
miha.kramar@zag.si
The government of
Slovenia has decided to build a repository
for Low and Intermediate Level Short Lived
radioactive waste (LILW) with the capacity
of approx. 9400 m3 of LILW. The
implementation of the project was assigned
to the Agency for Radioactive Waste
Management (ARAO). Based on a comparative
study of different alternatives, the
option with below-ground silos in the
vicinity of Krško nuclear plant (NEK) was
chosen.
Prior to disposal, LILW will be inserted
into concrete disposal containers of type
N2b with dimensions of 1,95 x 1,95 x 3,3 m
filled with 4 tube-type containers (TCC).
The disposal container is qualified as
IP-2 package. The transport of containers
from NEK to the disposal site will use a
public road. Because of the National act
on the transport of dangerous goods on
public roads, ADR regulations for road
transport and manipulation of LILW
packages will be followed. ADR requires
that containers provide protection against
the hazard (radiation) of the material
under all conditions of transport,
including foreseeable accidents. There
should be no more than a 20 % increase in
the maximum radiation level at any
external surface of the package even in
case of the accident. To demonstrate
compliance with these requirements a drop
test is required: for packages IP-2 with a
weight over 15 t a drop test from 0,3 m
should be performed.
The design of the container has been
carried out by consulting engineering
company IBE, d.o.o. which is also
responsible for the performing an actual
drop test. Different design concepts were
considered and their performance to drop
tests simulated using numerical
simulations. Drop test simulations have
been performed at ZAG (Slovenian National
Building and Civil Engineering Institute)
with a general purpose finite element
program Abaqus using explicit dynamics
procedure (Abaqus/Explicit). The container
has been modelled in detail with 3D solid
elements. Nonlinear material properties
were considered while multiple contact
surfaces were assumed between different
parts of the container (e.g. between the
lid and the container, etc.). The
reinforcement was included in the model
(in the form of embedded nonlinear beams).
Different drop scenarios were investigated
exceeding the requirements of ADR: a drop
from 0,3 m onto the most vulnerable corner
was simulated as well as the overturning
of a container which might follow the
initial collision. The accuracy of the
numerical results was checked by
controlling the impact energy flow and
performing sensitivity analyses to
different parameters (contact properties,
type of elements, stiffness of the base,
etc). Special attention was devoted to
establishing quantifiable acceptability
criterion based on different output
variables such as crack widths,
deformations and cumulative damage.
The analyses have shown that overturning
of container might be more critical than
the drop on the corner which causes mostly
local damage in concrete. In some cases
the damage predicted in the numerical
analysis was substantial indicating that
the container might not fulfil the ADR
requirement. Therefore, it was necessary
to change the design of the container. The
design of the container is still ongoing
and other alternatives will be tested.
Finally, an actual drop test will be
performed for the validation of the
numerical model and implementation of even
more robust analyses.
06.09.2016
15:00 Severe Accidents
Severe
Accidents - 806
Influence of Melt Pouring
on Stratified Steam Explosion
Vasilij
Centrih1, Matjaž Leskovar2
1Institut "Jožef
Stefan", Jamova 39, 1001 LJUBLJANA,
Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
vasilij.centrih@gmail.com
A steam explosion may
occur during a severe reactor accident
when the molten core comes into contact
with the coolant water. An important
condition for the occurrence of a steam
explosion is the initial coarse premixing
of the melt and the water. In nuclear
reactor safety analyses steam explosions
are primarily considered in the melt
jet-water pool configuration, where due to
the melt jet fragmentation the required
premixture is efficiently produced. In
stratified melt-water configuration, i.e.
molten corium layer below water layer, it
was previously assumed that there is no
premixing of the melt and the water and
that a strong explosion thus cannot
develop. The recently performed
experiments in the PULiMS and SES (KTH,
Sweden) facilities with corium simulant
materials however revealed that strong
steam explosions may develop spontaneously
also in stratified melt-water
configuration. The development of a
considerable melt-water premixed layer
above the spread melt was clearly visible
in the tests where an explosion occurred.
In these experiments the melt was poured
into a shallow pool of water. Despite the
shallow water, the melt jet fragmentation
during the short penetration through the
water may possibly importantly influence
the formation of the premixture also in
such a stratified configuration. To
address this issue an underwater melt
release stratified steam explosion
experiment is planned to be performed in
the SES facility in the frame of the EC
SAFEST project.
In the paper the potentially important
influence of the melt pouring on the
stratified steam explosion will be studied
by computer simulations with the MC3D code
in the SES geometry. Since no validated
models have been developed yet which would
adequately describe the observed
underwater melt spreading and the
formation of the observed premixed layer
in stratified configuration, an innovative
approach for the coupling of the premixing
phase calculation, the premixture layer
characteristics and the explosion phase
calculation with available MC3D code
procedures was established. The performed
comparative analysis for two scenarios,
first one with underwater melt release and
the second one with the melt pouring above
the water, will be presented and
discussed. Also suggestions for further
experimental and analytical work will be
given.
06.09.2016
15:40 Posters I
Severe
Accidents - 807
Simulation of natural
circulation experiment in MISTRA
experimental containment facility with
OpenFOAM CFD code
Boštjan
Zajec1, Ivo Kljenak2
1Jožef Stefan
Institute, Reactor Engineering Division,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2Jožef Stefan
Institute, Reactor Engineering Division ,
Jamova cesta 39, 1000 Ljubljana, Slovenia
ivo.kljenak@ijs.si
Various experiments on
the behaviour of non-homogeneous
atmosphere are being performed in
containment experimental facilities, both
to understand the phenomena and to obtain
results, adequate for validating
Computational Fluid Dynamics (CFD) codes
for the purpose of using them to simulate
phenomena in actual plants. Within the
OECD project SETH-2, which lasted from
2007 to 2010, experiments have been
performed in the MISTRA (Commissariat a
l'Energie Atomique et aux Energies
Alternatives, Saclay, France) and PANDA
(Paul Scherrer Institute, Villigen,
Switzerland) containment experimental
facilities.
One of the experiments in the MISTRA
facility (which is a cylindrical vessel
with a volume of 98 m2, with some internal
subdivisions), called NATHCO, consisted in
gradually heating condensers, installed
near the vessel wall, so as to heat the
nearby gas and induce buoyant flow in the
stagnant atmosphere. The influence of this
natural circulation on a previously
established horizontal layer of helium in
the upper part of the vessel was observed.
The experiment NATHCO was simulated with
the open-source CFD code OpenFOAM. A
two-dimensional axisymmetric model of the
MISTRA facility was developed. Simulation
results (time-dependent local helium
concentrations and temperatures) are
compared to experimental data and the
discrepancies are analysed.
06.09.2016
15:40 Posters I
Severe
Accidents - 808
Modelling of debris bed
coolability in bottom reflooding
conditions with MC3D code
Janez
Kokalj, Mitja Uršič, Matjaž Leskovar
Institut
"Jožef Stefan", Jamova cesta 39, 1000
Ljubljana, Slovenia
kokalj.janez@gmail.com
A hypothetical severe
accident in a nuclear power plant has the
potential for causing severe core damage,
including meltdown. To prevent or in the
case of already formed debris bed to limit
the in-vessel core degradation the basic
severe accident management strategies
consider the in-vessel reflooding to
ensure the coolability. Due to the debris
bed porosity, which allows easier coolant
intrusion, the debris bed provides greater
chances for cooling than a pool of molten
corium. When the cooling is not
sufficient, with the continuation of the
scenario the degraded reactor core is
melted and relocated to the lower reactor
vessel plenum. To prevent the ex-vessel
melt release the in-vessel melt retention
strategy could be applied.
The coolability of the debris bed was
recognized as an important nuclear safety
issue in the frame of the EU SARNET-2
(Severe Accident Research NETwork of
Excellence) programme. In the SARNET-2
programme the ex-vessel debris bed
formation due to the fuel-coolant
interaction and the coolability of the
formed debris bed were analysed. Currently
the international research on the debris
bed coolability is under investigation in
the frame of the Horizon 2020 IVMR
(In-Vessel Melt Retention Severe Accident
Management Strategy for Existing and
Future NPPs) project.
The purpose of our research is to
understand the key processes related to
the in-vessel debris coolability in the
bottom reflooding conditions. First, the
recently performed tests in the PEARL
facility (IRSN, France) will be presented.
The PEARL experimental program was
launched to provide experimental data to
validate 2-D and 3-D models for the debris
bed bottom reflooding. Next, the modelling
and analysis of the PEARL experiments
using the MC3D code (IRSN, France) will be
described. The aim of the performed work
was to analyse the uncertainties in the
initial experimental conditions and to
assess the heat transfer modelling
approach in the MC3D code.
06.09.2016
15:40 Posters I
Severe
Accidents - 809
Comparison of CFD and LP
Codes for the Simulation of Hydrogen
Combustion Experiments
Tadej
Holler1, Ed Komen2,
Ivo Kljenak3
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2NRG-Nuclear
Research and Consultancy Group Dept.
Fuels, Actinides and Isotopes, P.O.Box 25,
1755 ZG Petten, Netherlands
3Jožef Stefan
Institute, Reactor Engineering Division ,
Jamova cesta 39, 1000 Ljubljana, Slovenia
tadej.holler@ijs.si
The production and
release of hydrogen into the containment
during a severe accident is an important
safety issue for Light Water Reactors
(LWRs). Combustion of hydrogen may cause
structural damage to the containment and
may compromise its function as final
barrier for release of radioactive fission
products to the environment. To reduce the
hydrogen risk as far as possible, hydrogen
mitigation systems such as Passive
Auto-catalytic Recombiners (PARs) and
igniters can be installed. The risk of
hydrogen deflagration has received
increased attention after the Three Mile
Island accident in the USA back in 1979,
and also most recently following the
Fukushima accident in Japan in 2011, where
hydrogen’s destructive power was
displayed. Computational modeling is
required to demonstrate the adequacy of
the Nuclear Power Plant’s (NPP’s) hydrogen
risk management systems as well as for
their optimal design and the assessment of
the accompanied residual risks of the
presence of hydrogen.
Historically, so-called system codes or
lumped parameter codes were used for the
assessment of hydrogen deflagration risk
in NPPs. With the advancement of computers
and thus increase of computational power
in recent time, complementary to the use
of lumped parameter codes, Computational
Fluid Dynamics (CFD) modeling can be used
for more detailed assessment of the
hydrogen risks in determining the
possibility of a breach of the NPP’s
containment integrity.
This paper presents comparison of
simulation results obtained using the
Ansys Fluent CFD code and the lumped
parameter code ASTEC (Accident Source Term
Evaluation Code). It offers brief
theoretical backgrounds of both modelling
approaches and also focuses on advantages
and drawbacks of both applied methods for
the use in hydrogen combustion risk
assessment.
Three experiments performed with
hydrogen-air mixtures and different
initial hydrogen concentrations in the
medium-scale THAI experimental facility
were used for the presented comparative
analysis. A detailed analysis and
comparison also with the experimental
results of maximum pressure are presented
along with a discussion about the future
of computational analyses in the field of
hydrogen combustion safety in NPPs’
containments.
06.09.2016
15:40 Posters I
Severe
Accidents - 810
Improvement of the melt
relocation modelling in ATHLET-CD
Liviusz
Lovasz1, Sebastian Weber2
1Gesellschaft für
Reaktorsicherheit mbH, Forschungsgelände,
85748 GARCHING, Germany
2Gesellschaft für
Reaktorsicherheit (GRS), Schwertnergasse
1, 50667 Köln, Germany
liviusz.lovasz@grs.de
The accident in
Fukushima pointed out the importance of
severe accidents simulations. The severe
accidental phenomena are not fully
understood due to the lack of experiments,
therefore the continuous development of
the tools capable of simulating severe
accidents is important.
The system code ATHLET-CD (Analysis of
THermal-hydraulics of LEaks and Transients
with Core Degradation) is designed to
describe the reactor coolant system
thermal-hydraulic response during severe
accidents, including core damage
progression as well as fission product and
aerosol behaviour, to calculate the source
term for containment analyses, and to
evaluate accident management measures. The
ATHLET-CD structure is highly modular in
order to include a manifold spectrum of
models and to offer an optimum basis for
further development.
In the module ECORE, which simulates the
degradation phenomena, the core is divided
in concentric rings, the fuel and absorber
rods in every ring are modelled by a
representative rod. The melt relocation is
simulated by rivulets with constant
velocity and cross section (candling
model), starting from the node of rod
liquefaction. The movement of these
rivulets is simulated only in axial
direction in every ring in the last
released version of ATHLET-CD.
Developments were performed to achieve
radial relocation of the melt in case of
blockage formation. Due to these
developments the code is capable of
comparing the heights of melt rivulets in
neighbouring rings above a blockage and of
calculating the radial movement of the
molten fuel, based on the height
differences of the melt rivulets. The new
model takes also the BWR specific elements
into account.
To demonstrate the effects of the
development, a simulation with the
improved and with the original version of
ATHLET-CD is presented on an example of a
hypothetical severe accident in a generic
BWR reactor, with an initial event of a
Station Blackout.
06.09.2016
15:40 Posters I
Severe
Accidents - 811
Analysis of X-Ray Images
in SERENA KROTOS Experiments with
Premixing Simulations
Vasilij
Centrih1, Matjaž Leskovar2
1Institut "Jožef
Stefan", Jamova 39, 1001 LJUBLJANA,
Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
vasilij.centrih@gmail.com
A steam explosion is an
energetic fuel-coolant interaction
process, which may occur during a severe
reactor accident when the molten core
comes into contact with the coolant water.
An important condition for the occurrence
of a steam explosion is the initial coarse
premixing of the melt and the water. The
most distinctive process is the melt jet
fragmentation and the corresponding
premixture formation. To resolve the open
issues in steam explosion understanding in
the field of nuclear safety a number of
activities were carried out within the
recent OECD SERENA project. Lately, the
comprehensive summary of the post
processed x-ray radioscopy data analysis
(CEA, France) was completed for the SERENA
KROTOS experiments. The analysis of the
x-ray images provides an additional
insight into the complex premixing
processes.
In the paper the new information about the
premixture formation provided by the
innovative x-ray radioscopy system will be
studied by computer premixing simulations
with the MC3D code. Because the local data
of the premixing region may be obtained
from the x-ray images, the study is
focused mainly on the lateral distribution
of the premixture along the test section.
The analysis is based on the most
informative x-ray data from the KS-1 and
KS-4 experiments. A parametric analysis
was performed, varying the experimental
conditions, material properties and some
model specific parameters such as the
lateral velocity of the produced droplets
from jet the fragmentation. Especially the
influence of the subcooling and the corium
material on the premixture formation will
be analysed and discussed. Also
suggestions for further experimental and
analytical work will be given.
06.09.2016
15:40 Posters I
Severe
Accidents - 813
Material Influence on
Ex-vessel Steam Explosion
Tomaž
Skobe1, Matjaž Leskovar2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
tomaz.skobe@ijs.si
A steam explosion may
occur, when during a severe reactor
accident the corium melt comes into
contact with the coolant water. Steam
explosions are an important nuclear safety
issue because they can potentially
jeopardize the primary system and the
containment integrity of the nuclear power
plant.
In the paper the material influence of the
oxide and metal corium on the ex-vessel
steam explosion will be presented. A PWR
ex-vessel steam explosion study with oxide
and metal corium was carried out with the
MC3D code. For each type of the corium a
premixing simulation and an explosion
simulation was performed, triggering the
explosion at the time of melt bottom
contact. The premixing simulations were
performed with the global jet breakup
model. Comparing calculations with oxide
and metal corium were performed without
oxidation in the premixing and the
explosion phase. With the comparison of
the oxide and metal corium simulation
results the influence of the material
properties of the melt on the strength of
the steam explosion was analysed.
06.09.2016
15:40 Posters I
Severe
Accidents - 814
Thermal-Hydraulic
Analysis of PHWR Containment using
MELCOR Code in severe accident
Sungchu
Song, Seon Oh Yu
Korea
Institute of Nuclear Safety, 34 Gwahak-ro,
Yuseong-gu , Daejeon 305-338, South Korea
scsong@kins.re.kr
As a major safety
function, a containment of Nuclear Power
Plant (NPP) provides a reliable mean to
confine possible fission products release
during severe accident. However, the
severe accident occurred at Fukushima NPPs
raised the concern on the integrity of
containment building, which was failured
accompanied by the severe accident. Thus
it is of essence to guarantee the
integrity of containment building for any
type of commercial nuclear systems of
Boiling Water Reactors (BWRs), Pressurized
Water Reactors (PWRs), and Pressurized
Heavy Water Reactors (PHWR). Variuos
systems codes of ASTEC, MAAP, and MELCOR
have been developed for the purpose and
among them the MELCOR code developed for
severe accident analysis of PWR and BWR
shows the capability of detailed
thermal-hydraulic behavior inside the PHWR
containment building. In the present work,
the PHWR containment with volume of 48,000
m3 was modelled with 52 control volumes of
containment components, environment with 1
control volume and 118 flow paths. To
quantify the overall risk on the
containment integrity, the H2
concentration, which is the dominant
combustible gas, was evaluated. In
addition, the effectiveness of various
Engineering Safety Features (ESFs) such as
PAR, spray system, igniter, and local air
cooler has been investigated given the
postulated severe accident scenario using
the MELCOR code. The time of containment
failure caused by increased concentration
of hydrogen and the overpressurization has
been also evaluated quantitatively to
provide important scoping time for the
accident management.
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 904
CAD data storage and
access in IDAM
Marijo
Telenta1, Leon Kos1,
Robert Akers2
1University of
Ljubljana, Faculty of Mechanical
Engineering, LECAD Laboratory, Aškerčeva
cesta. 6, 1000 Ljubljana, Slovenia
2EUROfusion
Consortium, JET, Culham Science Centre,
OX14 3DB, Abingdon, United Kingdom
marijo.telenta@lecad.fs.uni-lj.si
Integrated Data Access
Management (IDAM) is data access tool for
analysis, visualisation, and
modelling. It is developed at CCFE for
MAST-U data access. Data access is based
on data objects
from within the files. IDAM is also put
capable. Metadata from raw and analysed
files is written to
the IDAM database. IDAM server uses a
plugin architecture for each data resource
type. The goal of
presented work is to build a workflow
which will access and eventually store CAD
data into IDAM.
CAD data in IDAM has a single source, i.e.
it is stored in single location as CATIA
files. A first step
is to build data resource which will
include a metadata and a catalogue design.
This step includes
categorising the resources and record them
to IDAM. The data resource will provide
CAD data in
STEP format for two workflows. The first
workflow is the engineering one in which
more complex
3D models are required in STEP format to
be read by COMSOL and ANSYS and perform,
for example,
electro-magnetic numerical simulations.
The second workflow is the scientific one,
in which 2D
axi-symmetric section-cuts are performed
on different levels of detail. The section
cut is performed
for specific angle from the horizontal
axis. These section-cuts are then
converted/exported by a
python plug-into VTK/XML and used by a
physics code EFIT++. Attention is given to
definition of
a structure in the STEP file in order to
locate different components needed for the
code. Generally,
it is better to have one file to ensure
the provenance. With such zero-copy
approach, type movement
could be achieved for better efficiency.
The second step is to develop an IDAM
server plugin
to get/put metadata for CAD data objects
into an object store. This will provide to
serve the data
through the plug-in. The storage includes
collection and cataloguing of metadata
during the CAD
data handling. This ensures CAD data
provenance tracking and capture together
with other objects
available in IDAM.
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 905
Effect of multilayer
insulation on thermal loading in DEMO
systems
Ingrid
Vavtar1, Martin Draksler2,
Boštjan Končar2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
ingrid.vavtar.93@gmail.com
This paper discusses
the radiative heat exchange amongst the
major components of the DEMO tokamak.
Since additional multilayer insulation at
the warm side of thermal shields can
substantially reduce the heat load to the
magnets and thermal shields themselves,
different shielding configurations,
including those with passive multilayer
insulation were investigated. Numerical
analysis shows that regardless of the
additional insulation layers being used,
the excessively high thermal loads on the
magnet system can be avoided only if the
magnets’ thermal shielding is actively
cooled. Discussion is supported by a
simplified theoretical model, which is in
good agreement with the numerical
predictions.
Based on the CAD model of the baseline
DEMO design a 1/18th section of a tokamak
geometry was created using the ANSYS
DesignModeler software and used for a
thermal radiation analyses with the Finite
Element code ABAQUS. Careful meshing and
input model development was required to
reduce the global numerical error of
radiation analysis to acceptable level.
Based on the numerical analysis, the
thermal loads on individual component as
well as the amount of energy being
exchanged in the system have been
estimated. This allowed us to investigate
the effect of the multilayer insulation,
and to estimate the required power for
active cooling of thermal shields.
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 906
Synthesis of W-based
composite as a plasma facing material
Andreja
Šestan1, Matej Kocen2,
Janez Zavašnik3, Saša Novak4,
Petra Jenuš2, Miran Čeh5
1Jožef Stefan
Institute, Centre for electron microscopy
and microanalysis, Jožef Stefan
International Postgraduate School, Jamova
39, 1000 Ljubljana, Slovenia
2Jožef Stefan
Institute, Department for nanostructured
materials, Jamova 39, 1000 Ljubljana,
Slovenia
3Jožef Stefan
Institute, Centre for electron microscopy
and microanalysis, Jamova 39, 1000
Ljubljana, Slovenia
4Jožef Stefan
Institute, Department for nanostructured
materials, Jožef Stefan International
Postgraduate School, Jamova 39, 1000
Ljubljana, Slovenia
5Jožef Stefan
Institute, Department for nanostructured
materials, Centre for electron microscopy
and microanalysis, Jožef Stefan
International Postgraduate School, Jamova
39, 1000 Ljubljana, Slovenia
andreja.sestan@ijs.si
In previous fusion
experiments (TEXTOR, ASDEX Upgrade and
JET- Joint European Torus) tungsten has
been used in plasma-facing components due
to its acceptable thermo-physical
properties. Compared to previous fusion
reactors, materials incorporated in DEMO
divertor will have to withstand even more
extreme conditions. Despite tungsten’s
favorable properties, there are also
several disadvantages that we will try to
overcome, especially in terms of
mechanical properties at high
temperatures.
It has been suggested that by
reinforcement of W-matrix with carbides
(TiC, ZrC and HfC) refractory particles or
oxide particles (Y2O3 and La2O3), tungsten
mechanical properties can be improved [1].
As an alternative, we use W2C and WC
particles to reinforce W-matrix. For this
purpose, tungsten matrix is being
reinforced with 10 vol. % of carbon
precursor. The starting mixtures were
prepared following the same procedure:
tungsten powder was mixed with phenol
formaldehyde resin. Homogeneous powder
mixture was first uniaxial and then
isostatically pressed. Two different
temperature regimes in high-temperature
furnace were used in order to obtain high
density materials with homogeneous
distribution of W2C particles in W-matrix.
As-sintered samples were analyzed in terms
of phase composition (XRD), microstructure
(SEM) and mechanical properties (room
temperature flexural strength).
The introduction of the two step sintering
regime enabled the phenol formaldehyde
resin to completely degrade into carbon,
which was not achieved with the first
sintering program. In our future work, the
sintering process will be further
optimized in order to obtain composite
with a high density (up to 95 % of
theoretical density).
Reference:
1. Pintsuk, G., 4.17 - Tungsten as a
Plasma-Facing Material A2 - Konings, Rudy
J.M, in Comprehensive Nuclear Materials.
2012, Elsevier: Oxford. p. 551-581.
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 907
Tunnel probe measurements
in a low-temperature magnetized plasma
Jernej
Kovačič1, James Paul Gunn2,
Tomaž Gyergyek3
1Institut "Jožef
Stefan", Odsek za reaktorsko fiziko,
Jamova cesta 39, 1000 Ljubljana, Slovenia
2CEA, IRFM,,
F-13108 Saint-Paul-Lez-Durance, F-13108
Saint-Paul-Lez-Durance, France
3University of
Ljubljana, Faculty of Electrical
Engineering, Tržaška 25, 1000 Ljubljana,
Slovenia
jernej.kovacic@ijs.si
The tunnel probe [1] is
a relatively new type of electrostatic
probe which is suitable for use in
tokamaks. Unlike the Langmuir probe, which
is most commonly used in tokamaks, it has
a concave geometry and works in a way like
an inside-out Langmuir probe. The main
advantage of thedesign is that the
measured current-voltage characteristics
do not suffer from the geometrical effect
of sheath expansion. Therefore, the ion
branch of the characteristic is properly
saturated and can be used for plasma
parameter evaluation without the error of
collection area enlargement. The tunnel
probe is divided into two biased
collection areas, the back plate and the
tunnel. The characteristic of the probe
was extensively modelled using dedicated
particle-in-cell simulation in order to
make a perfect calibration of the probe
measurements for different plasmas. Now,
only by measuring the ratio of ion
saturation currents from both collectors,
ion current density and electron
temperature can be measured simultaneously
in a fast way.
Up until now the tunnel probe has only
been used in tokamaks. We have now
installed one such probe in a
low-temperature plasma of the Linear
Magnetized Plasma Device (LMPD) on Jožef
Stefan Institute. Our goal was to study
the behaviour of the probe in various
plasma conditions to try and expand the
range of the possible use of the probe.
Since LMPD is a versatile machine,
magnetic field intensity and angle, plasma
density, electron energy distribution
function, ion beams etc. can be altered or
added, so we performed a number of
measurements in different conditions. We
were especially interested in the electron
branch of the characteristic, since the
discrepancy between the measured and the
simulated results for that part for
tokamak plasmas was significant. We also
focused on the effect of the magnetic
field inclination on the electron
collection inside the tunnel. In this
paper we shall present the results of the
measurements on the LMPD.
[1] J. P. Gunn, R. Dejarnac, J. Stöckel,
“Simultaneous DC measurements of ion
current density and electron temperature
using a tunnel probe”, Journal of Physics:
Conference Series 700 (2016), 012018.
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 908
Deuterium atom loading of
self-damaged tungsten at different
sample temperatures
Anže
Založnik1, Sabina Markelj1,
Thomas Schwarz-Selinger2,
Klaus Schmid2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Max-Planck-Institut
für Plasmaphysik (IPP), Boltzmannstr. 2,
D-85748 Garching b. München, Germany
anze.zaloznik@ijs.si
During the operation of
a fusion device, 14.1 MeV neutrons will be
created. These high energy neutrons will
produce defects in the bulk of the
material of plasma facing components,
degrading the favorable properties of the
material and enhancing the fuel retention.
In order to mimic the neutron-damaged
tungsten, self-damaged samples i.e.
implanted with high energy W ions, are
used in experiments.
The edge plasma in a fusion device will
consist of ions as well as of neutral
atoms and molecules. The study of neutral
particle interaction with plasma facing
components is important in order to
predict and understand the fuel dynamics
in the divertor and to estimate the
contribution of neutral particles to the
overall retention of the fuel in the wall
of a fusion device. In contrast to ions,
neutral particles cannot penetrate
directly into the bulk of the material,
but are rather adsorbed on the surface.
From the surface they can desorb back to
the vacuum or they can diffuse in the
bulk, where they contribute to the overall
retention. In order to understand the
mechanism of surface to bulk migration, a
series of dedicated experiments was
performed with self-damaged tungsten.
Polycrystalline tungsten samples damaged
by 20 MeV W ions up to 0.25 dpaKP at
maximum peak damage were exposed to
deuterium atom beam at the sample
temperature of 450 K, 500 K, 550 K and 600
K for the same atom fluence. Time
evolution of deuterium depth profile was
followed in-situ and in real time during
the exposure by Nuclear Reaction Analysis
(NRA) technique. Deuterium populates only
20 % of the damaged layer at 450 K,
whereas the whole damaged layer is
saturated in the case of 600 K. Namely,
the trap population is slower for lower
temperature. After the exposure a
thermodesorption spectroscopy (TDS) was
performed in order to obtain information
about the defect concentration in the
sample and desorption energy of deuterium
atoms.
Experimental results were modeled in order
to obtain information about the adsorption
site types and the height of the potential
barrier for diffusion from the surface to
the bulk. The TESSim code [1] for
deuterium trapping and bulk diffusion was
upgraded by a surface model and used for
deuterium depth profile calculation. The
modeling results were fitted to the
experimental data and modeling parameters
were determined.
We found a nice agreement between the
modeling and the experimental results. The
determined values of adsorption energy
were found to agree with the values
reported in literature. The heat of
solution for tungsten, calculated from the
height of the barrier for surface to bulk
diffusion, was 0.335 eV. This is
approximately three times lower compared
to the value of 1.04 eV, reported in
literature. Modeling with the literature
value for the barrier height resulted in a
huge discrepancy between experimental data
and modeling results, indicating the need
for the low value of the barrier height or
a temperature dependence of some other
modeling parameters, which were considered
constant in the current model.
[1] K. Schmid, V. Rieger and A. Manhard,
J. Nucl. Mater. 426 (2012), 247
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 909
Thermal Loading of DEMO
Divertor Cassette During Maintenance
Conditions
Luka
Klobučar1, Boštjan Končar2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
klobucar.luk@gmail.com
The demonstration
fusion power plant DEMO is planned to be
the last major step before the commercial
fusion power plant. DEMO tokamak is
composed of several systems operating at
very different temperatures, either
extremely high (divertor) or extremely low
(superconducting magnets). During the
reactor operation the divertor is
subjected to the incident heat flux of
removed plasma particles with values above
10 MW/m3. Such heat loads may eventually
cause severe damaging and consequently the
need for regular replacement of divertor
cassette that is envisaged at two-year
cycle. The cassette under replacement is
unplugged from the cooling pipes, while
the remaining cassette and the blanket are
still actively cooled. Because of lack of
internal cooling the detached cassette is
heated up due to the decay heat in
activated materials.
This study aims to evaluate the heat load
on the divertor cassette during
maintenance conditions taking into account
the decay heat immediately after the
reactor shutdown and one month after the
shutdown. The possibility for external
cooling of the divertor cassette under
replacement will be investigated as well.
The thermal model of the DEMO reactor
vessel will be developed that includes
actively cooled vacuum vessel, blankets
and divertor cassettes as well as the
detached divertor cassette subjected to
the decay heat and thermal radiation of
neighboring components. The 3D thermal
analysis will be performed with the ANSYS
CFX tool and will be based on the most
recent DEMO tokamak design with 54
divertor cassettes. The geometry will be
prepared in ANSYS Design Modeler, while
ICEM CFD will be used for numerical
meshing. Before carrying out the real case
simulations, the code validation study
will be performed on the simplified
analytical benchmark. The benchmark
geometry will be resemble the 3D model of
tokamak as far as currently possible but
will be simple enough to allow analytical
solutions. Several cases with different
boundary conditions (passive thermal
radiation, imposed temperature
conditions…) on divertor cassette surfaces
will be analyzed and special attention
will be paid to the error assessment.
The main goal of the study is to evaluate
the temperature distribution in the
detached cassette (mainly maximum
temperature in the cassette body and
maximum temperature on the plasma-facing
surface) under different heat load and
external cooling conditions in order to
define the maintenance criteria.
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 910
The first study of
deuterium retention in tungsten
simultaneously damaged by high energy
W ions and loaded by D
Sabina
Markelj1, Anže Založnik1,
Thomas Schwarz-Selinger2,
Mitja Kelemen3, Primož
Vavpetič1, Primož Pelicon1,
Etienne Hodille4, Christian
Grisolia4
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Max-Planck-Institut
für Plasmaphysik (IPP), Boltzmannstr. 2,
D-85748 Garching b. München, Germany
3Institut Jožef
Stefan, Jamova cesta 39, 1000 Ljubljana,
Slovenia
4CEA, IRFM,,
F-13108 Saint-Paul-Lez-Durance, F-13108
Saint-Paul-Lez-Durance, France
sabina.markelj@ijs.si
Tungsten or advanced
tungsten alloys are considered to be the
most suitable material for plasma-facing
components in future fusion reactors such
as DEMO. In these nuclear devices tritium
retention in neutron damaged tungsten will
become more significant issue. In order to
study the influence of material
irradiation by neutrons on fuel retention,
high energy ions are used [1]. It was
shown that fuel retention in self-ion
damaged tungsten is strongly increased as
compared to undamaged tungsten [eg. 2].
Till now all retention studies are
performed by sequential high energy ion
damaging and subsequent loading of the
material with hydrogen isotopes. However,
in a real fusion reactor both implantation
of energetic hydrogen ions and neutrals as
well as damage creation by the neutron
irradiation will take place at the same
time. The consequences for retention are
unknown. It is well known that in some
metals impurities such as hydrogen, change
the behavior of defect creation and
recovery, e.g. on vacancy migration during
recovery stage [3].
In this contribution we present the first
experimental results on simultaneous D
atom beam loading and defect creation by
high energy self-ion implantation in
tungsten. In order to have a good database
to compare with, we also studied deuterium
retention in self-damaged tungsten by the
different procedures of tungsten damaging
by tungsten ions and sequential deuterium
loading by neutral D atoms. To make one
step further towards more realistic
situation we have preformed first study of
simultaneous tungsten irradiation by 10.8
MeV W ions and D atom loading, atom flux
of 4.5x1018 D/m2s, at five different
temperatures from 450 K to 1000 K for 4
hours yielding a maximum 0.5 dpa damage
dose. After the damaging and loading D
depth profiles were measured by NRA using
D(3He,p)4He nuclear reaction. For the 450
K case the atoms hardly penetrated in
depth whereas in the case of 800 K -1000 K
the atoms did diffuse through the damaging
area in 4 h due to the faster diffusion.
In order to determine how many traps were
actually created in the material the
samples were after simultaneous damaging
& loading and NRA analysis exposed to
D atoms for additional 19 h at 600 K,
fluence 3.7x1023 D/m2. As expected the
highest concentration was obtained for the
450 K case, decreasing with damaging
temperature. The results were compared to
different sequential damaging/exposure
experiments. Synergistic effects were
observed, namely, higher maximum D
concentrations were found in the case of
simultaneous damaging and exposure as
compared to damaging at elevated
temperatures without offering D. Therefore
part of the defects that would annihilate
at high temperatures does not due to the
presence of solute deuterium atoms in the
bulk, that stabilize the defects. However,
the deuterium retention is still lower as
compared to sequential damaging at room
temperature and defect annealing.
[1] W.R. Wampler, R.P. Doerner, Nucl.
Fusion 49 (2009) 115023.
[2] O.V. Ogorodnikova, et al., J. Nucl.
Mater. 451 (2014) 379
[3] H. Schultz, Mater. Sci. and Eng. A141
(1991) 149.
[4]A. Založnik, et al., Phys. Scr. T167
(2015) 014031.
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 911
Micro-NRA and micro-3HIXE
with 3He microbeam on samples exposed
in ASDEX Upgrade and pilot-PSI
machines
Mitja
Kelemen
Institut
Jožef Stefan, Jamova cesta 39, 1000
Ljubljana, Slovenia
mitja.kelemen@ijs.si
Tungsten or advanced
tungsten alloys are shown as promising
materials for high heat flux plasma facing
components in tokamak fusion reactors.
During plasma operation of such device,
the wall is subjected to severe physical
conditions, which lead to erosion,
deposition, fuel retention and lattice
damage of the wall. To obtain a depth
profile of retained fuel, i.e. deuterium,
for plasma experiments nuclear reaction
analysis (NRA) is often used via the
D(3He,p)? reaction, making use of a
resonance like cross section at the energy
of 630 keV. A sequence of measurements at
several beam energies is used to deduce
the D depth profile [1].
In the case where the information on
lateral distribution of deuterium is
sought for, equivalent NRA method is
applied at Jožef Stefan Institute (JSI)
with a focused 3He beam [2]. Negative 3He
ion beam is formed in the combination of
duoplasmatron ion source and Li- change
exchange canal. A setup for 3He and 4He
gas mixing was built in the duoplasmatron
housing to spare precious 3He gas. The
3He2+ ions are accelerated with tandem
accelerator to energy of 3.3 MeV. Under
such conditions, we are able to form a 3He
beam with diameter of 10 µm and ion
current of 300 pA. During the
measurements, in total four detectors are
used simultaneously: a thick-depleted
implanted silicon detector for NRA, a RBS
PIPS detector, a HPGe X-ray detector for
detection of particle induced X-ray
emission (3HIXE) and a PIPS detector for
beam dose normalization that detects ions
scattered from the beam chopper. By
scanning the focused 3He beam over an area
of 2.2X2.2 mm2 a good lateral resolution
is obtained, providing information on the
lateral distribution of deuterium and
other elements in the sample.
In the presented study we analyzed two
samples, 10-20 nm W deposited on nominal
and smooth graphite substrate, that were
exposed in ASDEX Upgrade machine. The
purpose was to study the effect of surface
roughness on the net erosion/deposition
patterns of tungsten as well as compare
the deposition of different impurities
(carbon, nitrogen, boron) and deuterium on
them [3]. In addition, a sample exposed to
D plasma in the Pilot-PSI linear machine,
was analysed. The sample had a 1.5 µm W+Y
coating on Mo substrate. The sample was
analysed also by Laser-induced breakdown
spectroscopy (LIBS) in order to
characterize and test the performances of
this technique. In addition the elemental
inventory with secondary ion mass
spectrometry was measured and lateral
profile was created. The results obtained
by the micro beam analysis will be
presented and discussed. Good agreement
between the elemental distributions
obtained by NRA, 3HIXE was obtained when
comparing to LIBS and SIMS measurements.
With these results we provide a new
powerful analytical tool for elemental
inventory of plasma facing materials.
References:
[1] Markelj et al. J. Nucl. Mater. 469
(2016) 133
[2] Pelicon et al, Nucl. Instr. Meth. B
269 (2011), 2317.
[3] Hakola et al. Phys. Scr. T167 (2016)
014026
06.09.2016
15:40 Posters I
Nuclear
Fusion and Plasma Technology - 913
Deuterium Removal from
Self-ion Irradiated Tungsten by
Annealing in Vacuum and Isotopic
Exchange
Olga
Ogorodnikova1, Sabina Markelj2,
V. V. Efimov1, Yu. M.
Gasparyan1
1Moscow Engineering
Physics Institute National Research
Nuclear University, "MEPhI", Kashirskoye
shosse 31, 115409 Moscow, Russian
Federation
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
olga@plasma.mephi.ru
Tungsten (W) is a
material for ITER divertor and primary
candidate for plasma-facing material for
DEMO. Materials for fusion and fission
suffer from high dose neutron irradiation
at high temperatures. The tritium (T)
inventory issue is one of safety and
nuclear licensing as well as cost; in ITER
and in any successor facility there will
be strict limits on the amount of tritium
that may be trapped in the wall material.
Consequently, the removal of hydrogen
isotopes from radiation damage in tungsten
is important from point of view of tritium
safety of fusion reactors. In this work,
Wsamples were first pre-irradiated with
self-ions to generate radiation damage and
then exposed to deuterium (D) plasma at
470 K. To study the removal of hydrogen
isotopes from radiation-induced defects in
tungsten (W) by annealing, the samples
were annealed in vacuum at temperatures of
600, 700 and 800 K for 2
hours.Modification of the deuterium depth
profile in self-ion irradiated tungsten
after annealing was measured by nuclear
reaction analysis (NRA) using 3He+ as the
analysing beam. Kinetics of the D release
from radiation-induced defects of
specimens after different post-annealing
treatments was studied by thermal
desorption spectroscopy (TDS). It is shown
that D removal from radiation-induced
defects in W by annealing in a vacuum is
more efficient at lower displacement per
atoms (dpa). At 0.5 dpa, the D
concentration at radiation-induced defects
decreases by factors of ~2 and 3 at
annealing temperatures of 600 and 700 K,
respectively, and by a factor of ~18 at
800 K. In spite of efficient reduction of
D at radiation defects at 800 K, the D
concentration at radiation defects is
still about two orders of magnitude higher
than that at intrinsic defects in
undamaged material.
TheD removal by the annealing in vacuum
was compared with the D removal by the
isotopic exchange using hydrogen (H)
atomic beam [1].The annealing in a vacuum
at 600 K for 20 hours of samples
pre-exposed to atomic D beam at 600 K
reduces the D concentration at radiation
damage only by a factor of about 1.5. The
efficiency of isotopic exchange at the
same temperature is higher: the D
concentration at a peak damagedecreases
in3 times for 21 hours of the exposure to
H atomic beam.In the present work, we show
analytically that the efficiency of
isotopic exchange increases with
increasing the sample temperature,
incident ion/atomic flux and incident H
energy that consists with the experimental
data.
[1] S. Markelj et al., J. Nucl. Mater. 469
(2016) 133
06.09.2016
15:40 Posters I
Radiation
and Environment Protection - 1004
Assessment of Spent Fuel
Activity in Dose Projection Software
Matic
Pirc1, Borut Breznik1,
Primož Mlakar2
1Nuklearna
elektrarna Krško, Vrbina 12, 8270 Krško,
Slovenia
2MEIS storitve za
okolje d.o.o., Mali Vrh pri Šmarju 78,
1293 Šmarje-Sap, Slovenia
matic.pirc@nek.si
Dose projection
software installed at Krško NPP is using
an on-line calculation of radioactivity of
the reactor core but until recently the
software had no possibility to provide a
quick dose assessment in case of
overheated fuel accident in the spent fuel
pit (SFP). Besides the long term inventory
in the SFP, there is also an unloaded core
during each refueling outage and about
half of the core remains there after the
outage.
The software should be able to provide a
continuous activity calculation based on
the operator input regarding inventory
changes in the SFP. Some possibilities
were verified to get a quick input based
on a simplified approach using
pre-calculated results of the Origen
computer code.
Three different options regarding spent
fuel inventory and occurrence of the
accident were taken into account – before
the unloading of the reactor core, during
the fuel unloading/loading and after the
fuel loading, considering also an annual
recalculation by the Origen.
Radionuclides of the main interest are
noble gases and other volatile elements at
higher temperatures such as radioactive
cesium and iodine which could be released
into the air. Due to overall uncertainty
related to dose calculation, the on-line
calculation of SFP source term could be
helpful in case of emergency calculations.
06.09.2016
15:40 Posters I
Radiation
and Environment Protection - 1005
Radiological consequences
of potential disintegration of U
tailings pile at the former Žirovski
Vrh uranium mine, Slovenia
Tea
Bilić-Zabric
INKO
svetovanje, d.o.o., Kolezijska 5a, 1000
Ljubljana, Slovenia
inko@siol.net
The uranium mining
complex at Žirovski Vrh was in operation
during 1985-1990 and produced 455 tonnes
of yellow cake. About 610.000 tonnes of
technological waste were deposited on the
elevated slope located nearby the mine
complex. After extremely heavy and long
lasting rain in autumn 1990, the upper
part of the slope (together with tailings
pile) started to slide with the progress
of 2-3 cm per month. In spite of performed
technical measures a land-sliding of the
disposal site was not completely stopped
and represents quite a real threat to the
public and the environment.
Two research studies on the consequences
of total disintegration of the tailings
pile were elaborated on the initiative of
the ministry responsible for environment.
The preliminary geological study was
dealing with the spread and distribution
of rubble with radioactive waste material
downstream the local valleys. In the
subsequent radiological study the possible
radiological consequences for the local
population were worked out.
The paper presents firstly main results of
the geological study on rubble deposition
along two local streams for two
simultaneous emergency events: a local
earthquake and extreme precipitation with
a 100 years return period and with a 1000
years return period. The results of this
study – such as location and size of the
areas of transported material, layer
thickness and radioactivity of deposits -
were used as input data for radiation
exposure assessment. The evaluation of
public exposure due to total
disintegration of the disposal site was
done for cases whether restoration takes
place or if it does not. Dose assessment
covered all important exposure pathways
(inhalation, ingestion, external
radiation) originating from the deposited
radioactive material. Since the material
would reach the settlements, i.e. houses,
gardens, fields and grassland, the
expected radiation exposure would be an
order of magnitude higher as in the
operational period of the U-mine. Whether
restoration is not carried out the general
dose limit for population of 1 mSv/y would
likely be exceeded (estimated doses
1.3-4.5 mSv/y). Two main contributors to
the public exposure would be the
inhalation of radon with its short-lived
progeny and external radiation (over 95
%). The ingestion of contaminated garden
crops, milk, eggs, etc. would be of the
order of 0,05 mSv/y and would not be of
concern. On the other side, applied
remediation measures would efficiently
reduce public exposure down to the levels
of 0.1-0.2 mSv/y. For a comparison, the
current public exposure - after the
completed restoration of the U-mining site
- amounts 0.1 mSv/y or less.
06.09.2016
15:40 Posters I
Radiation
and Environment Protection - 1006
Ionization Smoke
Detectors in Slovenia – Current Status
and Future Challenges
Simona
Sučič, Marko Kostanjevec, Tomaž Žagar
ARAO
– Agencija za radioaktivne odpadke,
Celovška cesta 182, 1000 Ljubljana,
Slovenia
simona.sucic@arao.si
Ionization smoke
detectors use an ionization chamber and a
radionuclide such as americium-241 to
detect smoke. In Slovenia, installation of
ionization smoke detectors started at the
beginning of the 1970’s. Due to good
technical performance, ionization smoke
detectors were widely used and can be
found in industrial, service and even in
general public buildings like hospitals or
schools. Significant numbers of ionization
smoke detectors are still in use, however
the last 10 years they are increasingly
being replaced by new technologies. In
practice, major reconstructions and
upgrades of the existing fire alarm
systems usually occur during the
renovation of buildings and in that
process old ionization smoke detectors are
also being removed. According to the
Slovenian legislation removed detectors
are collected and transmitted to the ARAO
- organization responsible for the public
service of institutional radioactive waste
management. Due to the large number of
ionization smoke detectors, it is becoming
a routine practice in their treatment that
the device is dismantled and the
associated radioactive source is recovered
and conditioned for storage. The rest of
non-radioactive materials (plastic, metal
and electronic components) are treated and
prepared for recycling. After such
treatment the radioactive part of smoke
detectors is in more appropriate form for
storage and requires significantly less
space in the storage facility. This paper
presents current status and lessons
learned related with the treatment of
smoke detectors that were collected in
recent years through the public service of
institutional radioactive waste management
in Slovenia. Also, future challenges and
positive effects of performed actions
related with the treatment of ionization
smoke detectors in Slovenia are discussed.
06.09.2016
15:40 Posters I
Education,
Public Relations and Regulatory Issues -
1101
Public Opinion about
Nuclear Energy – Year 2016 Poll
Radko
Istenič, Igor Jenčič
Institut
"Jožef Stefan", Jamova cesta 39, 1000
Ljubljana, Slovenia
radko.istenic@ijs.si
The Information Centre
that was established within the Nuclear
Training Centre at the Jožef Stefan
Institute more than 20 years ago informs
the visitors about nuclear power and
nuclear technology, about radioactivity
and about Krško Nuclear Power Plant.
Information activities are targeted mainly
at schoolchildren from the 8th and 9th
grade of elementary school with their
teachers (in total close to 8000 per
year). The visit consists of a live
lecture about nuclear technology followed
by the demonstration of radioactivity and
a guided tour of a permanent exhibition.
The opinion trends are monitored since
1993 by polling about 1000 youngsters
every year. The poll is conducted before
the youngsters listen to the lecture or
visit the exhibition in order to obtain
their opinion based on the knowledge from
everyday life. Trends over the last 23
years will be presented, summarized and
commented.
06.09.2016
15:40 Posters I
Education,
Public Relations and Regulatory Issues -
1102
Energy for Children
Vesna
Slapar Borišek
Institut
"Jožef Stefan", Jamova cesta 39, 1000
Ljubljana, Slovenia
vesna.slapar-borisek@ijs.si
At the Information
Center that was established within the
Nuclear Training Center at the Jožef
Stefan Institute visitors are informed
about nuclear technology, the basic
properties of radioactivity and protection
against ionising radiation.
Lectures and demonstrations about nuclear
technology and radiation protection are
intended primarily for youngsters in last
couple of grades of primary school and
secondary school students. Occasionally
our visitors are pupils from lower grades
of primary school and even kindergarten
children. For these children lectures
about power plant operation and nuclear
technology are too demanding and should be
therefore appropriately adjusted according
to the level of their knowledge. For this
purpose, we decided to prepare a new
special lecture with demonstrations which
is intended for children of all ages. With
this lecture and demonstrations we want
children to understand the concept of
energy and energy conversion. In addition
we want to show to the children the
technology development from Heron's
Aeolipile, to the steam engine and the
turbine and its use in power plants.
06.09.2016
15:40 Posters I
Education,
Public Relations and Regulatory Issues -
1103
European Decommissioning
Academy (EDA) – 2nd run
Vladimír
Slugeň, Martin Hornáček, Róbert Hinca,
Filip Osuský
Slovak
University of Technology, Faculty of
Electrical Engineering and Information
Technology, Institute of Nuclear and
Physical Engineering, Ilkovičova 3, 812 19
Bratislava 1, Slovakia
martin.hornacek@stuba.sk
According to analyses
presented at EC meeting focused on
decommissioning organized at 11.9.2012 in
Brussels, it was stated that at least 2000
new international experts for
decommissioning will be needed in Europe
up to 2025, which means about 35 per year.
This growing decommissioning market
creates a potential for new activities,
with highly skilled jobs in an innovative
field. A clear global positioning of the
European Union is beneficial and will
stimulate export of know-how to other
countries, especially those having a large
nuclear programme, and promote highest
safety levels.
Having in mind the actual
EHRO-N report from 2013 and 2014 focused
on operation of nuclear facilities [1] and
an assumption that the ratio between
nuclear experts, nuclearized and nuclear
aware people is comparable also for
decommissioning, as well as the fact that
the special study branch for
decommissioning in the European countries
almost does not exist, this European
Decommissioning Academy (EDA) could be
helpful in the overbridging this gap.
European Decommissioning
Academy was created at the Slovak
University of Technology in Bratislava
Slovakia, based on discussion and
expressed needs declared at above
mentioned meetings and reports including
ECED2013 conference.
The main goal of the
Academy is from nuclearized experts
(graduated at different technical
universities and increased their nuclear
knowledge and skills mostly via on-job
training and often in the area of NPP
operation) to create nuclear experts for
decommissioning, which includes the
lessons, practical exercises in our
laboratories, on-site training prepared at
Jaslovské Bohunice and Mochovce sites,
Slovakia. Technical tour via most
interesting European facilities in Swiss
and Italy are part of the course as well
[3].
The first run successfully
passed 15 participants during June, 7 –
26, 2015. After the final exam, there was
an option to continue in knowledge
collection via participation at the 2nd
Eastern and Central European
Decommissioning (ECED) conference in
Trnava (Slovakia) [4].
Based on the lessons
learned during the first run of EDA and
the feedback of the participants we now
plan the 2nd run of the European
Decommissioning Academy. The Academy is
planned in June, 4 – 22, 2017 (including
3rd Eastern and Central European
Decommissioning (ECED) conference in
Trnava, Slovakia in June, 20 – 22, 2017).
Further information except
the references can be also found at
http://kome.snus.sk/inpe/ which is being
actualised.
Acknowledgements
Authors thank to the Slovak
Government grants VEGA 0204/2013 and to
001STU-2/2014-CEPVYJZ. We highly
acknowledge also IAEA for support.
References
[1] European Nuclear
Education Network, [online], available:
<http://www.enen-assoc.org>.
[2] SLUGEN, V., HINCA, R.:
European Academy of decommissioning: In
XXXVI. Days of Radiation Protection Book
of astracts, Poprad 2014, Slovak Republic.
ISBN 978-80-89384-08-2, EAN 9788089384082
[3] SLUGEN, V.: European
Decommissioning Academy (EDA): Ready to
start. In International Journal for
Nuclear Power, ISSN 1431-5254, 2015, vol.
60, no. 2, pp. 82-84.
[4] SLUGEN, V., HORNACEK,
M.: European Decommissioning Academy –
successful 1. run in June 2015. In:
Proceedings of the Eastern and Central
Europe Decommissioning ECED 2015, 23-25
June, 2015, Trnava, Slovak Republic. ISBN
978-80-971498-5-7.
06.09.2016
16:20 Probabilistic Safety Assessment
Probabilistic
Safety Assessment - 701
Shutdown Probabilistic
Safety Assessment – A Case Study for
the Pressurized Water Reactor
Marko
Čepin1, Rudolf Prosen2
1Fakulteta za
elektrotehniko, Tržaška cesta 25, 1000
Ljubljana, Slovenia
2Nuklearna
elektrarna Krško, Vrbina 12, 8270 Krško,
Slovenia
marko.cepin@fe.uni-lj.si
Shutdown probabilistic
safety assessment may be considered as an
extension of power probabilistic safety
assessment applied to the shutdown
conditions of the nuclear power plant. It
is more complex in sense of variety of
models, because the plant shutdown is a
sequence of plant operating states, which
differ among each other so much that the
probabilistic safety assessment models,
which represent them, need to be varied in
order to represent them realistically. The
method is developed, which follows mostly
the steps known in power probabilistic
safety assessment with an exception of
definition of plant operating states,
which are defined in a way that one plant
operating state suits its respective
specific probabilistic safety assessment
model. The number of plant operating
states has to be large enough to consider
the differences between them in sense of
the related plant parameters and
configurations of the plant equipment. At
the same time, the number of plant
operating states has to be small enough,
in order that the number of the
probabilistic safety assessment models
representing them is small enough for the
performance and cost effectiveness of the
related work. The results of a case study
based on a two loop nuclear power plant
with the pressurized water reactor are
presented and discussed. The results show
that the risk of different plant operating
states varies significantly and is in
average lower than the risk of the plant
in full power operation.
06.09.2016
16:40 Probabilistic Safety Assessment
Probabilistic
Safety Assessment - 702
Challenges of external
hazards assessment. ASAMPSA_E project
achievements.
Mirela
Nitoi1, Emmanuel Raimond2,
Yves Guigueno2
1RATEN ICN
Institutul de Cercetari Nucleare Pitesti,
Str. Campului nr.1, 115400 Mioveni,
Romania
2IRSN - Institut de
radioprotection et de sureté nucléaire,
Nuclear Safety Division , BP17, 92262
Fontenay-aux-Roses Cedex?, FRANCE, France
mirela.nitoi@nuclear.ro
One of the main
challenges for the nuclear domain is
represented by an efficient management of
the critical events, events that have an
impact on the safety operation of nuclear
power plants (NPP).
The external hazards constitute a
significant source of perturbations in the
safe operation of a NPP, and for this
reason, to properly investigate them and
to find ways to obtain a better picture
about their consequences, is quite
important.
The Advanced Safety Assessment
Methodologies: extended PSA (ASAMPSA_E)
project aims to examine in detail how
efficient is the Probabilistic Safety
Assessment (PSA) methodology in
identifying any major risk induced by the
interaction between a NPP and its
environment, and to develop some guidance
documents and technical recommendations
for PSA developers and users, dedicated to
improve the quality of the PSA studies.
Launched after the Fukushima accident, the
ASAMPSA_E project pays an increased
attention to the risks induced by the
occurrences of external events and their
combinations.
The paper presents the work packages and
the organizations involved as partners in
the project. The main directions of
actions, the identified priorities for the
project and the challenges encountered in
the attempt to assess the external hazards
are discussed. The results obtained so far
and the recommendations developed in frame
of the project, regarding the assessment
of external hazards, are specified. The
connection with the stakeholders and their
expectations in terms of guidance for the
development and use of extended PSA are
highlighted.
06.09.2016
17:00 Probabilistic Safety Assessment
Probabilistic
Safety Assessment - 703
Assessments of EP&R
provisions in Europe
Nadja
Železnik
Regionalni
center za okolje za srednjo in vzhodno
Evropo , Slovenska cesta 5, 1000
Ljubljana, Slovenia
nzeleznik@rec.org
The Fukushima accident
in March 2011 has intensified European
concerns about off site nuclear emergency
preparedness and response. As this
important aspect of defence in depth was
not included in the EC/ENSREG process of
stress tests, several initiatives took
place afterwards. The HERCA association
formed a working group on “Emergencies”
and started to work on the proposition
leading to a uniform way of dealing with
any serious radiological emergency
situation, regardless of national border
lines. In 2013 DG ENER commissioned a
“Review of current off-site nuclear
emergency preparedness and response
arrangements in EU member States and
neighbouring countries” which provided the
evaluation of the EU EP&R provisions.
In parallel a civil society association
Nuclear Transparency Watch has organised
an assessment of EP&R provisions
across the Europe from civil society point
of view and reported findings.
The findings of all investigations show
that current arrangements and capabilities
for off- site nuclear EP&R appear, on
paper, to be broadly compliant with
current EU legislative requirements and
international guidance. However, more deep
examinations of arrangements in practice
identified a number of gaps and
inconsistencies that need to be addressed,
like not harmonised criteria and cross
-border arrangements, mainstreaming of
nuclear emergency preparedness into civil
protection mechanisms, long term
protective measures and strategies,
involvement of local population and
communication, inclusion of societal
development (new social media, new spatial
and demographic development,…). New Basic
Safety Standard directive, adopted in
3013, and addressing also EP&R
requirement could be a good opportunity to
improve the EP&R arrangements if not
taken only formally. The paper will
present the findings of different
investigations and recommendation for
improvement of EP&R.
06.09.2016
17:20 Reactor Operation
Reactor
physics - 319
Simulation of the Initial
NPP Krško Cycles with CASL Core
Simulator - VERA-CS
Andrew
T. Godfrey1, Fausto
Franceschini2, Mohamed
Ouisloumen2, Marjan Kromar3
1Oak Ridge National
Laboratory, P.O.Box 2008, Oak Ridge,
Tennessee 37831-6162, USA-Tennessee
2Westinghouse
Electric Company LLC, 1000 Westinghouse
Drive, Cranberry Twp 16066,
USA-Pensylvania
3Institut "Jožef
Stefan", Odsek za reaktorsko fiziko,
Jamova cesta 39, 1000 Ljubljana, Slovenia
marjan.kromar@ijs.si
This paper describes
the application of the Virtual Environment
for Reactor Applications (VERA) core
simulator (VERA-CS) under development by
the Consortium for Advanced Simulation of
Light Water Reactors (CASL), to the core
physics analysis of the Krško NPP. VERA-CS
aims at enabling whole-core fuel cycle
depletion deterministic transport analysis
with subchannel thermal-hydraulic
coupling. It uses a three-dimensional
(3-D) whole core transport code MPACT
capable of generating sub-pin level power
distributions. CTF, an improved version of
the COBRA-TF subchannel code, is used for
the calculation of the thermal-hydraulic
parameters needed for the coupled
calculation. This paper is focused on the
application of VERA-CS to the analysis of
the initial NPP Krško cycles. Obtained
results are compared to the measurements
performed during the plant operation. In
addition results obtained from the
Westinghouse PARAGON2/ANC9 system under
development and IJS CORD-2 simulator are
given also.
06.09.2016
17:40 Reactor Operation
Reactor
Operation - 403
Systematic Approach to
Training (SAT) for the design of
Nuclear Power Plant (NPP)
Decommissioning Training in South
Korea
Jeong
Keun Kwak
Korea
Hydro & Nuclear Power Company, Ulsan,
45014, South Korea
bryan.kwak@khnp.co.kr
In 1979, the
unavailability of Main Feedwater
System(MFWS) in Three Mile Island(TMI)
Nuclear Power Plant(NPP) Unit-2 happened.
To make it worse, due to the malfunction
of isolation valve control in Auxiliary
Feedwater System(AFWS), the supply of
cooling water to a Steam Generator(SG) was
delayed approximately 8 minutes compared
to a normal process in Abnormal Operating
Procedure(AOP). In the long run, on
account of deferred heat sink provision to
a SG, the reactor core was melted
partially. It was the first critical event
in the US commercial NPP history.
Therefore, after TMI accident, US Nuclear
Regulatory Committee(NRC) suggested more
than one hundred alternatives. Among them,
one proposal was related to training area
and it was Systematic Approach to
Training(SAT) methodology. Hence, the goal
of SAT is improvement of NPP stability
through training point of view.
Additionally, since the appearance of SAT
in the NPP industry, it has been acquired
the unwavering position in the NPP
training field, so far.
Meanwhile, the significance of NPP
decommissioning has been soared up in
South Korea since the announcement of Kori
NPP Unit-1 decommissioning determination.
According to the proclaimed plan, Kori
Unit-1 is scheduled to be decommissioned
from June, 2017. Under this circumstance,
nurturing sufficient number of proficient
decommissioning engineers are one of the
most urgent issue in South Korean NPP
industry. Hence, to upgrade the efficiency
and consistency of training quality, SAT
methodology can be the reliable solution
for decommissioning training. For this
reason, establishment of SAT based NPP
decommissioning training will be a main
considering factor in my paper.
Key words : Systematic Approach to
Training, SAT, Nuclear Power Plant, NPP,
Decommissioning, US Nuclear Regulatory
Committee, US NRC, Three Mile ISland
Unit-2, TMI Accident
07.09.2016
09:10 Thermal Hydraulics I
Thermal
Hydraulics - 501
Investigation of Thermal
Turbulent Flow Characteristics of
Wire-wrapped Fuel Pin Bundle of Sodium
Cooled Fast Reactor in
Lattice-Boltzmann Framework
Ali
Tiftikci, Cemil Kocar
Hacettepe
University, Nuclear Engineering
Department, 06800 Beytepe, Ankara, Turkey
alitiftikci@hacettepe.edu.tr
In the presented study,
thermal turbulent flow simulations of fuel
pin bundles with helical spacer wires have
been carried out. The lattice-Boltzmann
method (LBM) is used for both fluid flow
and heat transfer calculations. Hence,
WALE and VLES turbulence models are
implemented to the open-source LBM code
and are coupled with the heat transfer
modules. A 7-pin bundle geometry, flow and
uniform heat flux conditions of Indian
Prototype Fast Breeder Reactor (PFBR) are
selected for the simulation purposes. The
simulations are handled for hexagonal fuel
rod bundle with two-helical-pitch-length
geometry. The post-processed quantities
such as velocity and temperature profiles
and Reynolds stresses are compared for
WALE and VLES turbulence models.
Additionally, the Nusselt number and
friction factor obtained from WALE and
VLES are compared with the experimental
correlations. The comparisons show that
LBM simulations are in good agreement with
the experimental data. Under the coarse
(for WALE simulations) Dh/50 lattice
resolution, VLES model gives relatively
better results. In this study, it is also
pointed out that LBM can be used for the
complex fast breeder reactor coolant
geometry and thermal turbulent flow
conditions
07.09.2016
09:30 Thermal Hydraulics I
Thermal
Hydraulics - 502
LOCA spectrum
calculations for PWR by RELAP5 and
TRACE
Andrej
Prošek
Institut
"Jožef Stefan", Jamova cesta 39, 1000
Ljubljana, Slovenia
andrej.prosek@ijs.si
The accident at the
Fukushima Dai-ichi nuclear power plant in
2011 demonstrated that external events
could cause loss of all safety systems. In
the Europe stress tests were performed and
the need was identified to further improve
the safety of the existing operating
reactors. Therefore the safety upgrade
programs were started. The objective of
this paper was to demonstrate that
developed input model of two-loop
pressurized water reactor (PWR) for TRACE
thermal-hydraulic systems code has the
capability for independent assessment of
RELAP5 computer code calculations. For
demonstration the response of PWR to
loss-of-coolant accident (LOCA) was
simulated. The break spectrum consists of
30.48 cm (12 inch), 20.32 cm (8 inch) and
15.24 cm (6 inch) equivalent diameter cold
leg breaks. The initiating event was
opening of the valve simulating the break.
The reactor trip on (compensated) low
pressurizer pressure (12.99 MPa) further
caused the turbine trip. The safety
injection (SI) signal was generated on the
low-low pressurizer pressure signal at
12.27 MPa. On SI signal no active safety
systems started (e.g. high pressure safety
injection pumps and low pressure safety
injection pumps and motor driven auxiliary
feedwater pumps). Only passive components
were assumed available, i.e. accumulators.
All these LOCA scenarios with above
assumptions lead to the core heatup. In
this way the time available before
significant heatup could be obtained.
For calculations the latest TRACE and
RELAP5 computer codes were used: TRACE
Version 5.0 Patch 4 using extension of
Ransom and Trapp critical flow model
(default) and RELAP5/MOD3.3 Patch 4 using
Henry-Fauske critical flow model (default)
and Ransom-Trapp critical flow model
(Option 50).
The results showed that RELAP5
calculations using different break flow
models are rather similar, therefore also
other parameters are similar. The
accumulators discharge was faster in TRACE
calculation than in RELAP5 calculations.
Therefore the calculated TRACE break flow
was also larger than RELAP5 calculated
break flow during this period. It can be
concluded that the break flow seems to be
the largest contributor to the differences
in the results between RELAP5 and TRACE.
07.09.2016
09:50 Thermal Hydraulics I
Thermal
Hydraulics - 503
Prediction of
low-pressure subcooled boiling with
advanced interfacial area source term
modelling
Ronak
Thakrar1, Simon P. Walker2
1Imperial College
of Science, Technology and Medicine,
Prince Consort Rd, London SW7 2BP, United
Kingdom
2Department of
Mechanical Engineering, Imperial College,
Exhibition Road, London, SW7 2BX, United
Kingdom
rkt08@imperial.ac.uk
Subcooled boiling flows
are encountered frequently in the nuclear
industry. There has been increased
interest in the past decade to investigate
low-pressure flows, primarily for advanced
LWR concepts and research reactors.
Mechanistic approaches to modelling
nucleate boiling are in their infancy and
remain a topic of intense research.
Present-day CFD codes continue to rely
heavily on conventional empirical
modelling and have been applied widely
towards the prediction of high-pressure
flows in particular. In the current work,
the Eulerian multiphase approach of the
commercial CFD code STAR-CCM+ is applied
to compute a vertically upward flow of
water in a uniformly heated pipe near
atmospheric pressure. Particular attention
is placed on the prediction of the
diameter of bubbles departing from the
heated wall - a parameter that is referred
to normally as a ‘bubble departure
diameter’. This parameter is an important
modelling closure that affects a strong
influence on the source terms for vapor
generation and interfacial transport, and
consequently on the generated void
profiles. To this end, a more modern
semi-mechanistic approach based on an
intricate analysis of forces acting on the
bubble is introduced into the established
modelling framework. The predictive
capability of this approach is compared
and contrasted against conventional
approaches. Prospects for improvement and
suggestions for future investigation are
outlined subsequently.
07.09.2016
11:10 Thermal Hydraulics II
Thermal
Hydraulics - 504
Crack growth assessment
in pipes under turbulent fluid mixing
using an improved spectral loading
approach and linear elastic fracture
mechanics
Oriol
Costa Garrido, Samir El Shawish, Leon
Cizelj
Institut
"Jožef Stefan", Jamova cesta 39, 1000
Ljubljana, Slovenia
oriol.costa@ijs.si
The turbulent mixing of
fluids with different temperatures inside
of the pipes is a well-recognized source
of thermal fatigue in the safety related
piping of nuclear power plants. The fluid
temperature fluctuations at the fluid-wall
interface, caused by the turbulent mixing,
induce temperature fluctuations in the
surrounding pipe walls. Rather fast
temperature fluctuations at the pipe
surface induce fluctuations of localized
thermal strains which are constrained by
the adjacent material at different
temperature. In this way, the fluid
temperature fluctuations induce stress
fluctuations in the pipe, which may lead,
in some circumstances, to fatigue and
subsequent leakage or even loss of
structural integrity. This phenomenon is
also known as thermal stripping.
This paper estimates the probability of
surface crack growth through the pipe wall
under turbulent fluid mixing conditions
using a damage tolerant approach. It is
usually believed that high frequency
oscillations of fluid temperatures during
turbulent mixing may be responsible for
crack arrest. In these conditions, large
stress gradients in thickness direction
are the cause for the reduction of the
stress intensity factor as the crack
grows. However, these observations are
typically derived from crack growth
analyses assuming that fluid or pipe
surface temperatures follow single
frequency sinusoidals.
The growth rates of stipulated surface
cracks are studied for diverse variations
of fluid temperatures, generated with an
improved spectral loading approach
recently developed by Jožef Stefan
Institute. The analyses use a rather
simple and linear one-dimensional model of
the pipe with numerically resolved
time-dependent temperatures and analytical
expressions for the linear elastic wall
thermal stresses varying only in the
radial direction. The linear elastic
fracture mechanics theory is then employed
to compute the time-dependent stress
intensity factors of the crack following
the method of weight functions for general
stress profiles. The uncertainties of
crack growth, which arise from the use of
comparatively short fluid temperature
histories to the expected fatigue life
time of months or years, are moreover
evaluated for diverse time lengths of the
fluid temperature histories using the
rainflow counting method and the Paris
law. The likelihood of crack arrest is
finally assessed for the diverse
variations of the fluid temperatures.
07.09.2016
11:30 Thermal Hydraulics II
Thermal
Hydraulics - 505
Simulation of the
Experiment PKL III H2.1 with the
TRACE5 Code
Jara
Turégano Lara1, Maria Lorduy2,
Sergio Gallardo2, Gumersindo
Verdú2
1Department of
Chemical and Nuclear Engineering,
Polytechnic University of Valencia, Camí
de Vera sn, 46022 Valencia, Spain
2Universidad
Politecnica de Valencia, Departamento de
Ingeniería Química y Nuclear, Camino de
Vera s/n, 46022 Valencia, Spain
sergalbe@iqn.upv.es
In the nuclear safety
it is especially important to know the
thermal hydraulic phenomena which take
place during an accident in a nuclear
power plant. However, it is impossible to
perform the experiments in a real scale in
order to obtain the data. For this reason,
there are facilities that replicate some
nuclear power plants scaled. Such is the
case of the PKL facility of AREVA placed
in Erlangen, Germany, which models the
most important components of the primary
and secondary side.
The purpose of this work is the analysis
of the results obtained with a PKL model
developed with TRACE5 Patch 4 and its
comparison with the experimental results.
PKL-III H2.1 experiment is based on a
Station Black Out (SBO) accidental
scenario with secondary and primary side
depressurization. This experiment consists
of 4 stages (conditioning, A, B and C
phases). The main thermal hydraulic
variables (pressure, flow, temperatures,
CET and PCT) have been studied.
Two models of PKL have been used (1-D vs
3-D vessel components) in order to compare
the behavior of the coolant from both of
them to the real vessel. The obtained
results show that the 3D model is able to
simulate the main thermal hydraulic
phenomena occurring during the
conditioning phase and the accidental
phase, among them the core uncovery and
the rise of the temperatures. The results
with both models have been compared with
the experimental results. It can be
concluded that, for this particular case,
the 3D-vessel component allows a proper
representation of the vessel collapsed
water levels as well as the Core Exit
Temperature (CET) and the Peak Cladding
Temperature (PCT). Moreover, the effect on
the steam generators and vessel emptying
have been studied when a finer
nodalization in the U-tubes and the vessel
is applied. As a general conclusion,
TRACE5 reproduces the main phenomena
observed in this test successfully.
07.09.2016
11:50 Thermal Hydraulics II
Thermal
Hydraulics - 522
Prediction of Wall
Condensation in the Presence of
Non-Condensable Gases through Various
Thermal-Hydraulic Codes
Erol
Bicer, Yeon-Joon Choo, Seong-Su Jeon,
Seung-Sin Kim, Yong-Hwy Kim, Soon-Joon
Hong
Future
and Challenge Technology Co., Ltd. (FNC
Tech.) , 46 Tapsil-ro, Giheung-gu,
Yongin-si, Gyeonggi-do, 17084, South Korea
ebicer@fnctech.com
Vapor film
condensation is an important topic in
various Light Water Reactor (LWR) nuclear
safety applications. In most cases,
condensation takes place in the presence
of non-condensable gases, such as the
condensation along containment walls in
the presence of air following a Loss of
Coolant Accident (LOCA). Recently, various
Passive Containment Cooling Systems (PCCS)
are under design and are expected to
operate with high non-condensable gas
content. Thus to determine whether the
current Lumped Parameter (LP), and/or
Multi-Dimensional simulation codes can
accurately simulate the condensation
phenomenon will be investigated. For this
purpose, four of COPAIN facility tests
(both forced and natural convection) were
simulated through MARS-KS, GOTHIC, CUPID,
and CFX codes. Early results indicate that
the one dimensional LP codes could not
provide accurate results (usually
underestimation) when the flow dynamics is
three dimensional. One main reason of the
inaccuracy is due to the empirical
correlations that are used in the
condensation models. Furthermore, the
velocity vector maps and temperature
contour maps are provided for further
investigation on heat flux along the wall.
Experimental data from COPAIN facility are
compared with the simulation results to
evaluate the applicability of wall
condensation model of each code.
07.09.2016
12:10 Thermal Hydraulics II
Thermal
Hydraulics - 507
On the discontinuity of
the dissipation rate associated with
the temperature variance at the
fluid-solid interface for cases with
conjugate heat transfer
Cedric
Flageul1, Sofiane
Benhamadouche2, Eric
Lamballais3, Dominique
Laurence4, Iztok Tiselj5
1Institut Jožef
Stefan, Jamova cesta 39, 1000 Ljubljana,
Slovenia
2EDF R&D, Fluid
Mechanics, Energy and Environment Dept., 6
Quai Wattier, 78401 Chatou, France
3Institute PPRIME,
Department of Fluid Flow, Heat Transfer
and Combustion, Université de Poitiers,
CNRS, ENSMA, Téléport 2 – Bd. Marie et
Pierre Curie, B.P. 30179, 86962
Futuroscope Chasseneuil Cedex, France
4The University of
Manchester, Materials Performance Centre,
School of Materials, PO Box 88, M60 1QD
Manchester, United Kingdom
5Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
cedric.flageul@ijs.si
Conjugate heat transfer
describes the thermal coupling between a
fluid and a solid. It is of prime
importance in industrial applications
where fluctuating thermal stresses are a
concern, e.g. in case of a severe
emergency cooling (PTS) or long-term
ageing of materials (T junctions). For
such complex applications, investigations
often rely on experiments, high Reynolds
RANS or wall-modelled LES. However,
experimental data on conjugate heat
transfer are scarce as walls in lab rigs
are often made of plexiglas and the
transported scalar studied is often a dye.
The development of RANS models for
conjugate heat transfer is relatively
recent (Craft et al., Journal of
turbulence, Vol. 11, 2010). In this paper,
we establish that the dissipation rate
associated with the temperature variance
is discontinuous at the fluid-solid
interface, in case of conjugate heat
transfer. There is currently no RANS model
for conjugate heat transfer that takes
into account this discontinuity.
06.09.2016
15:40 Nuclear Fusion and Plasma
Technology
Nuclear
Fusion and Plasma Technology - 912
Production of prompt and
delayed gamma rays in fusion reactors
Dijana
Makivič1, Igor Lengar2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Institut "Jožef
Stefan", Odsek za reaktorsko fiziko,
Jamova cesta 39, 1000 Ljubljana, Slovenia
dijana.mkv@gmail.com
Neutrons with an energy of
2.5 MeV originating from D-D plasma and
energy of 14.1 MeV from D-T plasma,
propagate through structural materials of
a fusion reactor and collide with nuclei
in the material. Prompt and delayed gamma
rays are created in the materials.
Measurement of the gamma ray spectra from
the plasma is an important diagnostic tool
for determining plasma properties such as
identification of fast ion species, their
tail temperature and relative
concentration. The measured spectra is
however disturbed by gamma rays from
structural materials. It is important to
evaluate the gamma rays which originate
from the neutron activation of materials,
in order that correction to the measured
gamma ray spectra from the plasma can be
made. Evaluation of delayed gamma rays is
important for determination of dose rates
and delayed heat after the shut down and
for ensuring safe maintenance of fusion
reactors.
Calculations of the
production of prompt gamma rays in a
fusion reactor were made using the Monte
Carlo N-Particle Transport Code (MCNP).
The MCNP6 code is currently capable to
compute the creation of new nuclides from
nuclear reactions with incident particles,
but does not create new nuclides as the
result of radioactive decay. In addition
the production of delayed gamma rays from
newly created unstable nuclides can be
calculated within different time bins. The
concentration of transmuted nuclides and
delayed gamma ray spectra for different
time bins were calculated with the MCNP6
code. Calculations were made for several
materials in a simple geometry. The same
calculation were made using the FISPACT
inventory code, with the capability to
calculate the activation and transmutation
induced by neutrons, without performing
the transport of particles. The comparison
was made between results obtained from
MCNP6 and from the FISPACT code in order
to make the evaluation of MCNP6 new
features and capabilities for production
of new nuclides and delayed gamma ray.
08.09.2016
09:10 Nuclear Fusion and Plasma
Technology
Nuclear
Fusion and Plasma Technology - 901
SOLPS-ITER Dashboard
Leon
Kos1, Ivan Lupell2,
Xavier Bonnin3
1University of
Ljubljana, Faculty of Mechanical
Engineering, LECAD Laboratory, Aškerčeva
cesta. 6, 1000 Ljubljana, Slovenia
2EUROfusion
Consortium, JET, Culham Science Centre,
OX14 3DB, Abingdon, United Kingdom
3ITER Organization,
Cadarache Centre, Building 519, 13108 St.
Paul lez Durance, France
leon.kos@lecad.fs.uni-lj.si
The design of the ITER
divertor and estimates of the required
fuelling throughput have relied for many
years on simulations performed by use of
the SOLPS plasma edge modelling tool. The
newly developed SOLPS GUI is a framework
tailored specifically for the SOLPS-ITER
code suite in a sense that code specifics
are built into the interface. Its design
allows users to extend functionality by
coupling custom widgets prepared for the
SOLPS GUI. These custom widgets are in
similar environments called actors as they
do act on some data, depending on input
received and then they pass results
further in a scientific workflow. Custom
widgets for SOLPS are operating in a
similar fashion in a way that they receive
and send the signals to other widgets for
further operation. In principle, no
programming is needed by users to create
their own “Dashboard” for analysing and
controlling the SOLPS simulations.
In contrast to scientific workflow engines
such as Kepler here we are more oriented
to look-and-feel experience than to create
a general purpose workflow engine. That’s
why the widgets in the SOLPS GUI are
designed to have “nice” input and output
presentation while we don’t care how
“nicely” wires are placed. “Wiring” is
usually taking significant space in other
workflow engines where actors are “small”
or have a unified size with separated or
neglected display output. The SOLPS GUI
uses the reverse approach with widgets
filling up the available Dashboard window
completely. There can be many widgets that
trigger part of the workflow, whereas
there are just play/pause/stop buttons
used in
Kepler. The SOLPS GUI signal/slot
philosophy provided by Qt framework is
similar to input/output ports in Kepler,
while the triggering is more explicit than
implicit. Users are therefore encouraged
to design their own Dashboard
by redesigning it to suit their needs. As
the dashboard is intended to be configured
with Qt designer this means that all
actions needs to be provided within the
widgets and connected by signals. “Wiring”
can be graphical too. The GUI is then
saved in XML files and compiled on-the-fly
at the GUI startup. Even when providing a
limited set of “custom” widgets, there can
exist many different dashboards for
running SOLPS simulations. They may differ
on the analysis, user’s preferences and
may be exchanged for reuse by others.
Presented graphical programming is not
just adding functionality easily but one
can simply/remove the unwanted custom
widgets and further extend it to other
simulation code suite.
08.09.2016
09:30 Nuclear Fusion and Plasma
Technology
Nuclear
Fusion and Plasma Technology - 902
Calculations to support
JET neutron yield calibration: Effects
of the neutron source anisotropy
Aljaž
Čufar1, Luka Snoj2,
Jet Contributors3
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Jožef Stefan
Institute, Reactor Physics Department,
Jamova cesta 39, 1000 Ljubljana, Slovenia
3EUROfusion
Consortium, JET, Culham Science Centre,
OX14 3DB, Abingdon, United Kingdom
aljaz.cufar@ijs.si
The calibration of
JET’s main neutron detectors, fission
chambers and activation system, is based
on the measurements of the detector
response to a calibration neutron source
placed in multiple positions inside the
vacuum vessel. The main requirement for
the calibration neutron source are well
known source characteristics such as the
source intensity and energy spectrum as
well as its anisotropy. The information
about these parameters is crucial if the
JET’s neutron detectors are to be
calibrated with the target accuracy of 10
%. These neutron source characteristics
are obtained through a combination of
Monte Carlo simulations and
characterisation measurements performed at
a neutronics laboratory.
When an anisotropic neutron source is put
into a complex geometry, such as the
tokamak’s vacuum vessel, various
complications can arise. Small changes in
the position or orientation of the source
can significantly affect the detector
response, depending on the anisotropy
profile and detector’s response function.
Monte Carlo simulations are an important
tool in the preparatory phase to the
experimental in-situ calibration as it is
important to understand the difficulties
and possible sources of errors before the
calibration experiment is performed.
For the calibration of JET’s neutron
detectors to neutrons with energy of 14
MeV a compact accelerator based DT neutron
generator will be used as a calibration
source. Such a generator is inherently
anisotropic while additional anisotropy is
introduced as a result of the materials
surrounding the area where neutrons are
produced. Small changes in the position or
orientation of the source, when positioned
inside the tokamak, can lead to
significant change of the detector
responses. This influences the accuracy of
the calibration process. The investigation
of the effects of uncertainties in the
generator’s position and orientation will
be presented along with their effects on
the accuracy of the calibration.
08.09.2016
09:50 Nuclear Fusion and Plasma
Technology
Nuclear
Fusion and Plasma Technology - 903
Fast online MPC for ITER
plasma current and shape control
Samo
Gerkšič
Institut
Jožef Stefan, Jamova cesta 39, 1000
Ljubljana, Slovenia
samo.gerksic@ijs.si
In a tokamak reactor,
the Plasma Current and Shape Controller
(PCSC) is the component of Plasma Magnetic
Control (PMC) that commands the voltages
applied to the poloidal field coils, to
control the coil currents and the plasma
parameters, such as the plasma shape,
current, and position. The PCSC acts on
the system pre-stabilised by the Vertical
Stabilisation controller. The task of PMC
is to maintain the prescribed plasma shape
and distance from the plasma facing
components, in presence of disturbances,
e.g. H-L transitions or ELMs, subject to
changes of local dynamics in different
operating points.
Model Predictive Control (MPC) is
considered as the most important technique
in advanced process control technique in
the process industry. It has gained wide
industrial acceptance by facilitating a
systematic approach to control of
large-scale multivariable systems, with
efficient handling of constraints on
process variables and by enabling plant
optimisation. These advantages are
considered beneficial for PCSC, and
potentially also for other control systems
of a tokamak. The main obstacle to using
MPC for control of such processes is the
restriction of the most relevant MPC
methods to processes with relatively slow
dynamics due to the relatively long
achievable sampling times, because
time-consuming on-line optimization
problems are being repeatedly solved at
each sample time of the CSC control loop
for determining control actions.
In this work we explore the practical
feasibility of using MPC for PCSC in the
ITER tokamak, employing recently developed
fast on-line quadratic programming (QP)
optimization methods with complexity
reduction techniques. A survey of the
available QP methods suitable for the
on-line solution of MPC optimization
problems is given, with emphasis on
first-order methods, which have been
recently considered as the most promising
candidates for fast online MPC control.
The prototype MPC controller [1] is based
on the control scheme of [2]. Using a
modification of the QP solver QPgen [3], a
five-fold speed-up compared to the
state-of-the-art commercial solver CPLEX
was achieved, with peak computation times
less than 10 ms on a computer with a
four-core Intel processor running
real-time Linux. This is already
considered sufficiently fast for the 100
ms sample time estimated to be suitable
for the ITER CSC control loop.
[1] S. Gerkšič, G. De Tommasi, "Model
predictive control of plasma current and
shape for ITER", 28th Symposium on Fusion
Technology (SOFT 2014), San Sebastián,
Spain
[2] G. Ambrosino et al., IEEE Trans.
Plasma Science, 37(7), 2009, 1324-1331
[3] Giselsson P., Improving Fast Dual
Ascent for MPC - Part II: The Embedded
Case, arXiv (2014)
08.09.2016
10:30 Radiation and Environmental
Protection
Radiation
and Environment Protection - 1001
MetroERM - Metrology for
Radiological Early Warning Networks in
Europe
Denis
Glavič Cindro1, Toni Petrovič1,
Matjaž Vencelj1, Benjamin
Zorko2
1Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
2Institut "Jožef
Stefan", Jamova cesta 39, 1000 Ljubljana,
Slovenia
denis.cindro@ijs.si
In an event of a major
radiological emergency, the early and
reliable knowledge of radioactivity
concentrations in air, and subsequently
the assessment of contamination levels of
farmland and of dose rate levels in urban
areas are of key importance in organizing
sound countermeasures for the protection
of the general public from the dangers
arising both from direct external
radiation and from intake of radioactivity
by ingestion or inhalation of contaminated
food or air.
Therefore in 2014, a 3-year EMRP joint
research project Metrology for
radiological early warning networks in
Europe (MetroERM) has been launched. The
aim of this project is to develop methods
for the harmonization of reported values
on both dose rate and airborne
radioactivity concentrations so that data
related to the same trans-boundary event
measured by different networks using
different detectors are directly
comparable. This will allow consistent
data collation and evaluation and
consecutively reliable conclusions could
be drawn by the responsible authorities.
In addition, within this project the
development of new measurement techniques
based on novel spectrometry systems with
state of the art detection materials, such
as LaBr3, CdZnTe or CeBr3 is in progress
with the aim to allow both the calculation
of dose rates and the calculation of
contamination levels including
nuclide-specific information at the same
time.
Another intention of this project is to
improve the capacity of the early warning
networks by the development of new methods
and systems for rapid ad-hoc radioactive
air concentration measurements to
efficiently supplement global early
warning data with accurate information on
airborne radionuclide content. With the
development of new modular air sampling
systems which can be easily transported to
locations for the detection of airborne
radioactive particulate, the information
content provided by early warning networks
in real time will be considerably
increased.
At JSI a novel portable aerosol sampling
device which provides real time airborne
radioactive particulate monitoring was
developed. It incorporates a 1 inch CeBr3
scintillation detector with ~4% FWHM
energy resolution at 662 keV positioned
centrally within a concertinaed filter
assembly. An improved air pump with stable
high flow rates, up to 200 m3/h, enables
low level airborne radioactivity
detection. To perform gamma spectrometry,
a fully digital signal processing unit
with a 4k-channel MCA was developed
in-house. The temperature drift of the
detection system is compensated on the
software side by a microcontroller-based
system connected to a digital thermometer
that is in good thermal contact with the
detector. The same microcontroller unit is
used to handle the user interface, via a 5
inch color touch screen, and handles all
inputs/outputs (I/O), including 3G network
communications. It enables a prompt and
continuous online detection and data
evaluation from remote stations, as well
as remote control of the unit settings and
functions. The system is incorporated in a
heavy-duty portable case which can be
easily transported to different
measurement locations.
The project MetroERM and the JSI
contribution within this project will be
presented and discussed.
08.09.2016
10:50 Radiation and Environmental
Protection
Radiation
and Environment Protection - 1002
New Ceramic Waste Forms
for High Level Radioactive Wastes
Neslihan
Yanikömer1, Sinan Asal2,
Sevilay Haciyakupoglu2, Sema
Erentürk2
1Istanbul Gelisim
University, Faculty of Engineering and
Architecture , 34315, Avcilar, Istanbul,
Turkey
2Istanbul Technical
University, Energy Institute, 34469
Maslak, Istanbul, Turkey
nyanikomer@gelisim.edu.tr
High level radioactive
wastes can be stored in geological
repositories after immobilizing in glass,
ceramics, glass-ceramics and glass
composite materials. Although waste
volumes can be reduced and products having
high chemical and mechanical durabilities
can be obtained with the waste
vitrification method, the radioactive
waste capacity in glass matrix is limited
between 20-35% (wt). Later, the waste
capacity of the composites increased to
50-70% (wt) with the development of
ceramic materials and at the same time
materials having high resistance can also
be obtained.
Purpose of this study is to prepare of
different ceramic matrices having high
chemical and mechanical durability, high
waste capacity, a processing temperature
as low as possible and having shielding
feature against radiation from the
radioactive waste in their matrices.
Immobilization of some important fission
products (137Cs and 90Sr) was performed in
these ceramics prepared in proper
composition. Then, the suitability of the
main matrices and matrices containing
wastes to the long-term underground
storage conditions was investigated with
detailed testing.
Keywords: Nuclear waste; Ceramics; Cs-137;
Sr-90; Chemical durability; Leaching
08.09.2016
11:10 Radiation and Environmental
Protection
Radiation
and Environment Protection - 1003
Dual track approach to
strategy and planning for high level
waste and spent fuel deep geological
disposal
Tomaž
Žagar, Leon Kegel, Matej Rupret
ARAO
– Agencija za radioaktivne odpadke,
Celovška cesta 182, 1000 Ljubljana,
Slovenia
tomaz.zagar@gov.si
The paper will give
overview of Slovenian national plan and
strategy for managing radioactive waste
and spent fuel with the focus on the high
level waste and spent fuel management via
dual track approach to deep geological
disposal as the final solution for spent
fuel and high level waste. Dual track
approach is a novel and modern approach to
treat in parallel international/regional
deep geological repository development and
national disposal program development.
Slovenia has a very small nuclear program:
it owns one nuclear power plant in
co-ownership with Croatia in 50:50 share
located in Krško, Slovenia (Krško NPP). In
addition to operating nuclear power plan
there is also one research reactor (TRIGA)
and central interim storage facility for
radioactive waste from small producers,
both near capital Ljubljana.
International and national regulatory and
legal frameworks require a national
programme for managing radioactive waste
and spent fuel. The main goal of these
programmes is to ensure safe and efficient
management of radioactive waste and spent
fuel. Experience shows that the route to
an operational deep geological disposal
facility is long and burdened with
uncertainties, even for large nuclear
programs. For countries with small or very
small nuclear programs the financial and
human resources required for the
construction and operation of a geological
disposal facility are significant, this is
why idea of regional and international
cooperation regarding radioactive waste
management is not new and has its roots in
the previous century. However, due to
uncertainties over implementation of
regional/multinational repository, the
national repository option must be kept
open in national programme. Both plans can
be implemented in parallel in national
programmes in so called dual track
approach.
08.09.2016
11:30 Panel
discussion
Panel
discussion on
Challenges in Education, Training and
Knowledge Management
Sustainable development of
the humankind in the future will, among
others, require access to sufficient,
environmentally acceptable and affordable
energy sources. Development of abundant
and affordable low carbon energy sources
might well represent one of the most
important and complex challenges that
humankind will have to adequately solve
within a few decades.
The sheer complexity of this
challenge calls for new knowledge,
excellently educated and motivated
individuals and intensive knowledge
management activities within all nuclear
stakeholders.
Moderator
Prof. Michel Giot,
Professor Emeritus, Universitécatholique de
Louvain
• Introductory statement by
the moderator
• Introduction of panelists and
introductory statements by panelists
• Moderated discussion:
• EU Commissionneworientationsofsupport
• Costsandbenefitsofnetworking: SWOT
analysis
• Codesandinnovation (how to match).
Panelists:
• Dr. Franck Wastin, Head of Unit
Knowledge for Nuclear Safety, Security and
Safeguards, European Commision, DG Joint
Research Centre
• Dr. Roger Garbil, Scientific officer,
European Commission, DG RTD, Unit Nuclear
fission
• Mr. Robert Stakenborghs, General
Manager, ILD Evisive, Baton Rouge, LA,
USA, Chair of Executive Committee of the
ASME Nuclear Engineering Division
• Dr. Asif Arastu, Technical Consultant,
Unisont Engineering, San Francisco, CA
(retired from Bechtel Power Corporation);
USA, Past Chair of Executive Committee of
the ASME Nuclear Engineering Division
• Mr. Clayton Smith, Director & Senior
Fellow, Technical Services, Fluor,
Greenville, SC, USA, Vice-Chair ASME Board
of Nuclear Codes and Standards, Member BPV
III Standards Committee and Committee on
Nuclear Certification
• Prof. Leon Cizelj, Head, Reactor
Engineering Division, Jožef Stefan
Institute, Slovenia, and President of the
European Nuclear Education Network (ENEN)