Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
Agenda and book of abstracts
Event organized and sponsored by the Sustainable Nuclear Energy Centre (SNEC)
The purpose of the gathering is:
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To provide to the industry an overview of the type of on-going MSc/PhD research projects.
To favor networking between the MSc/PhD students and the industry.
To favor the interactions between MSc/PhD students working on different but related topics.
Agenda
13:00 – 13:05
Introduction
13:05 – 14:45
Short students’ presentations
14:45 – 15:15
Coffee break
15:15 – 16:00
Students’ poster presentations
16:00 – 16:45
Presentation from Mats Ladeborn, responsible for nuclear power
development at Vattenfall,
Mats Ladeborn will present Vattenfall plans for new capacity. Mats
Ladeborn has worked within the nuclear industry since 1982 when he
joined the Swedish company Vattenfall. He is now Director for Nuclear
Development but has previously been Plant Manager for one unit at the
Ringhals Nuclear Power Plant and has also been acting Head of all
Vattenfall nuclear activities including eight companies and nine nuclear
reactors in Sweden and Germany. He is a former president in
FORATOM and he is president of the Swedish Atomic Forum. Mats has
a Master in technology management from Chalmers University of
Technology.
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
BSc category
BENCHMARK-COMPARISON OF THE CORE SIM NEUTRONIC
TOOL AND MCNP
Eirik Eide Pettersen
Guest scientist from Center of Nuclear Technologies, DTU, Denmark,
visiting the Division of Nuclear Engineering, Department of Nuclear Engineering, Chalmers
University of Technology
eirikep@nephy.chalmers.se
ABSTRACT
The CORE SIM neutronic tool is a two-group diffusion equation solver developed at the
Division of Nuclear Engineering at Chalmers University of Technology. One of the primary
features of the code is its ability to resolve reactor characteristics of both static and dynamic
systems. In previous work, steady-state calculations from CORE SIM have been compared
and validated with analytical solutions and industrial deterministic reactor modelling tools.
Now, the purpose of this project is to compare CORE SIM with the highly-trusted Monte
Carlo code MCNP and asses its accuracy in both static and dynamic benchmark scenarios.
Among the main advantages of CORE SIM are robust numerical algorithms, allowing CORE
SIM to calculate solutions to a wide range of reactor systems and geometries, in addition to
user-friendly operation accomplished through, for instance, the avoidance of input decks.
Combined with high portability, this makes CORE SIM a valuable tool for both educational
and research purposes, and while it can not rival the accuracy of commercial reactor
modelling tools, it provides, among other things, a clear alternative for preliminary
investigations. Therefore, it is of much interest to further explore the range of validity of the
CORE SIM neutronic tool, particularly for calculating dynamic systems. Furthermore, in
benchmarking CORE SIM's calculations of dynamic systems with a stochastic code, it is
hoped to develop a general methodology that can find application also in testing other
deterministic tools for calculations of dynamic systems. The benchmark reference, MCNP,
was chosen for its long track record and thorough testing. MCNP has been developed at the
Los Alamos National Laboratory since the 1950's, and is today widely regarded as an
industry-leading reactor modelling tool and a reliable reference in most reactor calculations.
A number of prominent issues arise when comparing a deterministic code like CORE SIM,
which relies on discretised energy groups and macroscopic cross sections for input, with the
continuous-energy Monte Carlo code MCNP, using materials and their densities and
temperatures for input. A three-step procedure is imagined to facilitate this comparison. First,
MCNP will be used to generate homogenised two-group macroscopic cross sections for a
given benchmark geometry. Second, the steady-state solution will be calculated by CORE
SIM using the MCNP-generated cross sections for input. Since the static solution also can be
found directly from MCNP, the accuracy of the steady-state results from CORE SIM can be
evaluated. The third step is required for comparing the calculated solutions of dynamic
systems. To do this, complete control over the macroscopic cross sections used by MCNP is
required in order to assure that identical cross sections are being compared. Thus, MCNP
must be forced to operate in a mode of reduced functionality where it uses only two energy
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
BSc category
groups and homogenised macroscopic cross sections for input. The methodology behind
accomplishing this is based on writing user-specified cross section library files (ACE files)
that can be read and used by MCNP. Naturally, a comparison must also be made between the
reduced and the full-featured MCNP to investigate and verify the accuracy of the reduced
version. Since the full-featured MCNP can not be used for solving dynamic systems, this
comparison can only be done for steady-state calculations.
Preliminary comparison of the thermal static flux calculated by MCNPX and by CORE SIM
(the latter using macroscopic cross sections generated with MCNPX) in a simple onedimensional two-region geometry. The considerably flatter flux profile from CORE SIM is
thought to originate from the simplifications made in deriving the two-group diffusion
equation.
Key Words: Reactor modelling, neutronics, CORE SIM, MCNP
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
MSc category
THE EVOLUTION OF REGUALATORY APPROACHES
IN NUCLEAR POWER OVERSIGHT
Björn Arkborn & Alexander Engström
Master students enrolled in the Nuclear Engineering master program
mi0arbj@student.chalmers.se & aleeng@student.chalmers.se
ABSTRACT
Due to the potential risk of nuclear power, regulatory agencies are used to assure that the
power plants live up to present nuclear safety regulation. Depending on cultural aspects,
traditions, economy and more, the approach of the regulatory agencies can vary with country
and time. Current globalization, such as for instance The European Union, raises demands for
regulatory conformity. This regulatory conformity will most likely affect the evolution of the
regulatory work and the approaches used for oversight. Following and trying to predict the
development of regulatory approaches will be of great importance, both for regulatory
agencies and nuclear power plant owners. The benefits from this knowledge are many, but
enhanced nuclear reactor safety is undoubtedly the largest.
The purpose of this project was to investigate and compare the different types of strategies
between regulatory agencies and nuclear power plants in Sweden, Finland, United States and
Canada. The idea was not to compare which strategy that is better, but rather to understand
why some are preferred and why some are not. Further, the idea was to get an understanding
of in which direction the regulatory work has changed, how it will change and what the
complications of that might be.
The analysis is based on a total of 18 qualitative interviews with experienced personnel from
the nuclear industry and the nuclear regulators together with extensive literature studies. The
interviews have been performed on site in Sweden, Finland, Canada and USA, lasting for
between 2-3 hours each. The interviews were recorded, transcribed and the information was
compiled and analyzed.
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
MSc category
The result from the study is not yet complete, as more analysis work still remains. Some
conclusions can still be made at this stage, the following result is clear:
•
Culture, tradition and accidents have had large impact of the regulatory models used in
all countries.
•
The regulatory approaches differ a lot between the analyzed countries and so do the
satisfactory levels. The result show however that the utilities have more or less the
same opinion of what an ideal authority – licensee relationship should be like.
•
The regulatory agency in Sweden is influenced by a number of international
organizations of which the following are the most influential:
o IAEA (International Atomic Energy Agency)
with its service IRRS (Integrated Regulatory Review Service)
o WENRA (Western European Nuclear Regulators' Association)
o CNRA (Committee on Nuclear Regulatory Activities)
•
Nuclear power oversight is undergoing harmonization. This harmonization will have
largest impact on countries with regulatory models that stands out in any way. With
Sweden having one of the least detailed nuclear regulations in the world,
harmonization will lead to a more detailed regulatory model.
•
Another factor that will affect the regulatory oversight and safety work in Sweden is
the release of the regulations that direct new build. This, together with harmonization
induced detail will most likely affect the present plants, and lead to need for
modernizations.
The project has been financially supported by Åforsk, EON and Vattenfall. Furthermore, EON
and Vattenfall have played a significant role in project feedback and in the acquiring of
experienced interviewees.
Key Words: Oversight, Cultural characteristics, Regulations, Evolution
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
MSc category
STATISTICAL ANALYSIS OF PLANT DATA FOR REVISION OF
OPERATING RULES
MINIMIZING RISK OF PCI FAILURE IN BWRS
Alborz Azadrad
Master student enrolled in the Nuclear Engineering master program
work performed at Westinghouse Electric Sweden AB
azadrad@student.chalmers.se
ABSTRACT
The demand for production of clean electricity is increasing. Nuclear power plants produce
large amounts of non-polluting electric power. However, depending on daytime, weekday and
season this demand is not constant. In a nuclear reactor, control rods and circulation pumps
are used in order to regulate the energy production. Careless increases of power can lead to a
number of undesirable effects amongst which fuel rod cracks are the worst ones. This
phenomenon is known as Pellet-Cladding Interaction failures or PCI failures. These failures
occur because of thermal expansion and to a lesser degree fuel pellet burn-up. The extent of
this excessive contact pressure in the clad tube is so severe that it causes structural failure. A
known fact since the 70’s is that light water reactors are prone to PCI failures. In order to
reduce the PCI failures during normal operations of nuclear power plants, improvements of
nuclear fuel has been made as well as power maneuvering guidelines has been developed. All
of these remedies had one and the same purpose of lowering the Linear Heat Generation Rate
(LHGR) as it had a significant impact on the occurrence of PCI failures. This presentation
describes the study of a new methodology based on a combination of power maneuvering
guidelines, plant data from several plants, and ramp-up tests. In order to estimate the
utilization and operating range a comparison of the limits for current PCI threshold and
Thermal Mechanical Operating Limit (TMOL) were performed. Finally, a statistical
evaluation was made. The goal was two-fold, the primary to extend the clients’ (Westinghouse
Electric Sweden AB) knowledge of the PCI phenomenon, and the secondary an attempt to
establish an association between risk of PCI failure and corresponding power increase. During
the study computational tools such as Matlab, Excel, Linux, POLCA7, CM2 and STAV7 were
used. The results show that the ramp-up tests have a serious impact on the derived
probabilities of PCI failure while none of the utilities had violated any operating rules. There
are however margins for greater utilization. Other effects that were left out in this study were
burnup level, cladding defects, and pellet defects.
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
Figure showing irradiation effects of a fuel rod.
Key Words: PCI, Statistics, Nuclear, LHGR
MSc category
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
MSc category
DEVELOPMENT OF A FUEL PERFORMANCE CODE FOR THORIUMPLUTONIUM FUEL
WITH FOCUS ON THE RADIAL POWER PROFILE
Patrik Fredriksson
Master student enrolled in the Nuclear Engineering master program
work performed with Thor Energy, NO-0255 Oslo, Norway
at the Division of Nuclear Engineering, Department of Applied Physics,
Chalmers University of Technology, SE-412 96 Göteborg, Sweden
patrik@nephy.chalmers.se
ABSTRACT
Thorium-plutonium Mixed OXide fuel (Th-MOX) is considered for use as light water reactor
fuel. Both neutronic and material properties show some clear benefits over those of UOX and
U-MOX fuel, but for a new fuel type to be licensed for use in commercial reactors, its
behaviour must be possible to predict. For the thermal-mechanical behaviour, this is normally
done using a well validated fuel performance code, but given the scarce operation experience
with Th-MOX fuel, no such code is available today.
There is an ongoing work with developing a fuel performance code for prediction of
the thermal-mechanical behaviour of Th-MOX for light water reactors. The well-established
fuel performance code FRAPCON is modified by incorporation of new correlations for the
material properties of the thorium-plutonium mixed oxide, and by development of a new
subroutine for prediction of the radial power profiles and burnup profiles within the fuel
pellets.
The new subroutine for the radial power profile is an extension of the old one used in
FRAPCON, modified to consider Th-MOX fuel instead of UOX and U-MOX fuel. Several
new isotopes were added in the subroutine, including Th232 and U233. Changes were also
made concerning the capture of epithermal neutrons in resonance regions using so called
shape functions, and the procedure of calculating the neutron flux. Instead of solving the
neutron flux analytically with a Bessel function, a numerical finite element method was
applied. This allowed the neutron flux to be dependent on inhomogeneities in the fuel.
Addition of the new isotopes necessitated the addition of new effective absorption and capture
cross sections, as well as modification of the old ones and new parameters for the shape
functions, which was obtained by employing a genetic algorithm (GA). A GA is a stochastic
optimization method useful for problems with many unknown parameters and which are
highly nonlinear. The GA found a set of cross sections and shape function parameters with the
help of a number of pre-generated radial power profile data cases, 61 and 94 for LWR and
HWR, obtained from simulations using a Monte Carlo code named Serpent. To validate the
new FRAPCON code and the new subroutine results were compared to data from a Th-MOX
test irradiation campaign which is currently ongoing in the Halden research reactor.
Results showed that the new subroutine could predict the radial power profile very
well for fuel pins with dimensions used in reactors today with a mean relative error of less
than 0.66% compared to the reference cases from Serpent. The addition of this subroutine in
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
MSc category
the new FRAPCON code improved the prediction of the centre temperature of a fuel pin used
in the Halden reactor, even if the change was very small.
Some more work needs to be done regarding the mechanical properties to correct for
effects exposed after the initial 60 days of irradiation. Regarding the radial power profile no
further work needs to be done or can be done without extending the problem to include more
advanced methods of solving the diffusion equation.
The centerline temperature for a Th-MOX fuel pin calculated by the new FRAPCON code
and measured during irradiation in the Halden research reactor.
Key Words: Th-MOX, Thorium, Power profile
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
PhD category
RADIATION EFFECTS IN ENGINEERING
MATERIALS AND ENGINEERS
Petty Cartemo
Nuclear Engineering, Applied Physics, Chalmers University of Technology
petty@nephy.chalmers.se
ABSTRACT
The PhD project partly aims at verifying suitable Generation IV reactor materials using the
Chalmers Pulsed Positron Beam. Several experiments were performed within GETMAT and
GENIUS as well as calibration studies which led to a deeper understanding of the
development and behavior of microscopic lattice defects under high irradiation. Other than in
present light water reactors, structure materials of the next generation of reactors have to deal
with higher temperatures and doses so that ageing phenomena such as embrittlement are
accelerated. To secure the performance of mayor reactor components it is of great importance
to have detailed knowledge on lattice properties and defect characteristics. The very sensitive
tool of Positron Annihilation Lifetime Spectroscopy (PALS) was chosen as part of the process
in finding suitable materials for Generation IV systems.
Positrons are strongly attracted to negative charges and annihilate instantaneously when
encountering free electrons such as found in vacancies and dislocations present within any
lattice structure. When a positron reaches the surface of a material it will penetrate it with
respect to kinetic energy before annihilation takes place. This time-dependent, very fast
process (ps), is measured in the lab and used for depth calibration and defect studies where the
latter used ion-irradiated model steel alloys to correlate radiation dose, temperature and defect
size (GETMAT, GENIUS).
Radiation damage is not only a key issue in terms of nuclear reactors and safety. If found in
living tissue it can have vital consequences, both negative and positive.
Here, a future collaboration with the Department of Radiophysics at Sahlgrenska, Göteborg
University is proposed. In case of a radiological accident involving the intake of radioactive
substances, whole body measurements are applied to quickly determine the amount, isotope
and in best case position of the specimen. Monte Carlo simulations can be used to validate
measurement results where the complex geometry of a human being (phantom) poses
difficulties on the task. A strongly simplified human body (IRINA phantom) will be defined in
terms of volume elements (voxel) where the resulting voxel phantom can be used in Monte
Carlo simulations omitting the need of difficult, software-specific geometry files. Being able
to model and validate whole body measurements is of importance for emergency preparedness
actions.
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
Geant4 simulations predict reality by using the ICRP voxel phantom.
Key Words: radiation damage, positron beam, phantom, radiological accident
PhD category
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
PhD category
PLUTONIUM LOADING OF GANEX SOLVENTS WITH PHENYL
TRIFLUOROMETHYL SULFONE AND CYCLOHEXANONE AS
DILUENTS
Jenny Halleröd*, Christian Ekberg, Mark Foreman, Elin Löfström Engdahl and Emma
Aneheim
Nuclear Chemistry, Department of Chemical and Biological Engineering, Chalmers
University of Technology
hallerod@chalmers.se
ABSTRACT
Currently, the primary research focus at Chalmers University of Technology within the field
of partition and transmutation is in developing a Grouped ActiNide Extraction (GANEX)
separation process. The basic concept of partition and transmutation is to separate the
transuranic elements from the fission products in used nuclear fuel and then transmute them
using a fast neutron spectrum. The principle of the GANEX process uses two cycles; the first
to remove the uranium bulk is removed from the fuel dissolution liquor and the second (where
GANEX extraction actually occurs) to extract the transuranic elements and remaining
uranium together as a group thereby avoiding pure plutonium streams.
The GANEX process previously developed at Chalmers combines the two extractants 6,6’bis(5,6-dialkyl-[1,2,4-]triazin-3-yl)-2,2’-bipyridine (CyMe4-BTBP) and tri-butyl phosphate
(TBP) in cyclohexanone. However, cyclohexanone has a low flash point (44°C compared to
kerosene (37-65°C), for example) and it is somewhat soluble in the acidic aqueous phase. An
attempt has therefore been made to replace cyclohexanone with phenyl trifluoromethyl
sulfone (FS-13). FS-13 was developed for use in the UNiversal EXtraxtion (UNEX, extracts
elements through ion-exchange) process as an alternative to highly polar nitrobenzene. FS-13
was considered as an alternative to cyclohexanone in the GANEX process, due to its stability
for irradiation and acid, and its higher flash point (122 °C).
This study investigated the possibility of extracting plutonium into a GANEX solvent
containing 70 % FS-13, 30 % TBP and 10 mM CyMe4-BTBP. It was shown that the system
based on FS-13 could resist plutonium loading up to 40 g L-1, no metal precipitation was
observed. However, the separation of plutonium from americium increased at higher
plutonium concentrations due to competitive extraction between the metals.
Key Words: Solvent Extraction, GANEX, FS-13, Pu-loading, TBP and CyMe4BTBP
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
PhD category
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
PhD category
COMMERCIAL THORIUM FUEL MANUFACTURE AND
IRRADIATION: TESTING (Th,Pu)O2 AND (Th,U)O2 IN THE “SEVENTHIRTY” PROGRAM
Klara Insulander Björk
Chalmers University of Technology, Department of Applied Physics,
Division of Nuclear Engineering, SE-412 96 Göteborg, Sweden
Thor Energy, NO-0255 Oslo, Norway
klara.insulander@scatec.no
ABSTRACT
Thorium oxide based fuels are considered for use as nuclear fuel due to several reasons. The
vast thorium resources spread over the entire planet motivates its use from a resource
management perspective and it also has benefits from a non-proliferation standpoint, due to
the fact that no new plutonium or heavier actinides are generated during irradiation of
thorium. Furthermore, the neutronic properties of thorium fuels are beneficial in many
respects. A mixed oxide fuel containing thorium and plutonium dioxide makes an attractive
alternative to conventional MOX fuel due to better reactivity coefficients, and using thorium
as an additive to uranium based fuels offers advantages in terms of flatter power distributions
and correspondingly higher thermal margins.
The thermomechanical properties of thorium dioxide are also very beneficial in many
respects, compared to those of UOX fuel. A higher thermal conductivity, lower thermal
expansion and heat capacity and lower fission product diffusion rates are all properties which
should result in good in-reactor fuel performance, ultimately resulting in larger operation
margins. Operation experience with thorium fuels is however very limited and several
phenomena that occur with burnup have never been subject to detailed experimental study. In
order to assess these phenomena and to quantify the thermomechanical behaviour of thorium
based fuel, the Norwegian company Thor Energy is currently undertaking an irradiation
experiment in the Halden Research Reactor in Norway.
Thorium based fuels are being tested in this experiment with the aim of producing the data
necessary for licensing of these fuels in the today’s light water reactors. The fuel types
currently under irradiation are thorium oxide fuel with plutonium as the fissile component,
and uranium fuel with thorium as an additive for enhancement of thermo-mechanical and
neutronic fuel properties. Fuel temperatures, rod pressures and dimensional changes are
monitored on-line for quantification of thermomechanical behaviour and fission gas release.
The data gathered during this program will serve the safety licensing case for subsequent lead
test thorium fuel rods (LTR) and/or lead test assemblies (LTA) in a commercial reactor.
Furthermore, the understanding of thorium oxide fuel performance gained from this test
campaign will enable the development of a fuel performance code for predicting thorium fuel
behavior during commercial operation. Since an important benefit of thorium based fuel is the
potential for high burnup, an important objective of the irradiation campaign is to measure
Further information about the project can be found at
http://www.thorenergy.no/no/Topmenu/Projects/Irradiation.aspx
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
PhD category
high burnup data. Preliminary irradiation results show benefits in terms of lower fuel
temperatures, mainly caused by improved thermal conductivity of the thorium fuels.
In parallel with the irradiation, a manufacture procedure for thorium-plutonium mixed oxide
fuel is developed with the aim to manufacture industrially relevant high-quality fuel pellets
for the next phase of the irradiation campaign. The thorium-uranium oxide fuels currently
under irradiation have been manufactured with equipment normally used for conventional
uranium fuel manufacture. Pre-studies for the manufacture of thorium-plutonium mixed oxide
fuel indicate that fuel of high quality can be manufactured without great difficulty.
The indications so far of the Seven-Thirty program is that from a technical standpoint,
thorium based oxide fuel can be manufactured and used in commercially operating light water
reactors with some clear benefits.
The IFA-730 irradiation rig, showing the six fuel pins currently under irradiation.
Key Words: Thorium, experiment, irradiation, Halden test reactor
Further information about the project can be found at
http://www.thorenergy.no/no/Topmenu/Projects/Irradiation.aspx
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
PhD category
MULTIPHYSICS SIMULATIONS OF NUCLEAR REACTORS –
MODELING AND IMPLEMENTATION FOR FINE-MESH
SIMULATIONS
Klas Jareteg
Department of Applied Physics
Division of Nuclear Engineering
Chalmers University of Technology
SE-412 96 Gothenburg, Sweden
klas.jareteg@chalmers.se
ABSTRACT
The behavior of a Light Water Reactor (LWR) core is determined by a number of coupled
physical fields and phenomena. The neutron distribution determines the amount and rate of
energy released, leading to heating of the solid fuel. The temperature distribution in the fuel is
further coupled to the flow of liquid or vapor water and the temperature dependent neutron
reaction probabilities in the fuel. The fluid flow of the coolant not only determines the
conjugate heat transfer from the fuel pins, but also the density dependent reaction
probabilities for the neutrons in the water.
To accurately simulate the reactor core, all mentioned aspects must be modeled, including the
couplings between the neutron distribution, the heat transfer and the fluid flow. In current
applied industrial methodologies, the coupled dependencies are treated in a coarse manner,
often by applying simplified relations or an a posteriori scheme. In the presented work, the
goal is to perform the coupling between the thermal-hydraulics (fluid flow and heat transfer)
and the neutronics in a direct manner, determining the interdependencies with a higher
resolution. The implemented framework aims at fine-mesh simulations for a single or a few
nuclear fuel assemblies.
In the current tool, developed within the project, the neutron distribution is determined by a
multi-group discrete ordinates method. This method estimates the energy as well as the spatial
and angular dependencies of the neutron population within the system. The thermal-hydraulic
problem is solved by a CFD formulation of the single-phase fluid problem, including implicit
conjugate heat transfer between the coolant and the fuel. Such a formulation allows highresolution profiles of the fuel temperature as well as the water temperature and density
between the pins. The neutronics and thermal-hydraulics are coupled directly on the finest
level of the applied computational grid, avoiding any coarsening or loss of resolution in the
coupled parameters. An example of the result of a calculation on a quarter of a 15-by-15 fuel
pin lattice is displayed in Figure 1. The tool permits determining a resolved temperature
profile in the coolant, which implicitly depends on the heat transfer in the fuel and the neutron
distribution, as a result of the above mentioned coupling mechanisms and as captured by the
implemented models.
Further information about the project can be found at
http://klas.nephy.chalmers.se
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
PhD category
Figure 1. Moderator (water) temperature distribution in a slice of a 15-by-15 fuel assembly, with
3 horizontal planes presented. Figure not to scale. (Jareteg et al, 2014)
Due to the severe computational load of the high-resolution, fine-mesh calculations, high
performance computations (HPC) are applied in the implemented code. This includes full
parallelization of the solver, which allows the problem to be decomposed and solved
concurrently on a large number of CPUs. Other HPC aspects in the work include utilization of
modern sparse matrix solvers, use of fast computational languages (primarily C++) and an
efficient and easily extendable implementation of all models, based on the open source library
OpenFOAM®.
The modeled and implemented multiphysics simulation framework allows for high resolution
estimates of local quantities within LWR fuel assemblies, and thus gives novel calculation
capabilities for safety parameters such as the local fuel temperature. An extension to Boiling
Water Reactor conditions is currently under development, and will result in axial and
horizontal void distributions and the influence of such heterogeneities on the coupled
problem.
Key Words: nuclear reactor multiphysics, deterministic reactor modelling, fine-mesh
simulations, coupled neutronics/thermal-hydraulics
References
Jareteg et al. (2014). Coupled fine-mesh neutronics and thermal-hydraulics - modeling and implementation for
PWR fuel assemblies. submitted to Special Issue ”LWR Multipyhiscs” in Annals of Nuclear Energy.
Further information about the project can be found at
http://klas.nephy.chalmers.se
Gathering MSc/PhD students – industry
Chalmers University of Technology, Gothenburg, Sweden, June 10, 2014
PhD category
DILUENT AND SOLVENT EFFECTS IN BTBP –BASED
EXTRACTION SYSTEMS
Elin Löfström-Engdahl
Nuclear Chemistry, Department of Chemical and Biological Engineering, Chalmers
University of Technology
Elinlo@chalmers.se
ABSTRACT
Used nuclear fuel taken directly from a reactor is radiotoxic for mankind and its environment
for a long time. One of the major contributions to the long time radiotoxicity is the presence
of the so called actinides. If the actinides could be transmuted into less radiotoxic nuclides the
strain of the final storage would decrease, both according to storage time and volume
efficiency. However, this transmutation demands a partitioning of the actinides from the rest
of the used fuel. This separation can be achieved by solvent extraction.
A solvent extraction system utilizes the principle that oil and water do not mix when they are
in contact. By adding specifically designed extraction molecules to the organic phase, chosen
elements, such as the actinides, are transferred, extracted, into the organic phase. At the same
time, the rest of the used fuel is left in the aqueous part of the system.
My work has focused on solvent extraction systems based on a special class of extracting
molecules, so called BTBPs. The BTBPs extracts trivalent actinides, but it has earlier been
showed that the extraction is affected by the diluent used. Such diluent effects have been
investigated during this work.