Gathering MSc/PhD students industry. Chalmers University of Technology, Gothenburg, Sweden, June 10, Agenda and book of abstracts

Transcription

1 Agenda and book of abstracts Event organized and sponsored by the Sustainable Nuclear Energy Centre (SNEC)

2 The purpose of the gathering is: 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.

3 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.

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5 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 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

6 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

7 MSc category THE EVOLUTION OF REGUALATORY APPROACHES IN NUCLEAR POWER OVERSIGHT Björn Arkborn & Alexander Engström Master students enrolled in the Nuclear Engineering master program & 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.

8 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

9 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 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.

10 MSc category Figure showing irradiation effects of a fuel rod. Key Words: PCI, Statistics, Nuclear, LHGR

11 MSc category DEVELOPMENT OF A FUEL PERFORMANCE CODE FOR THORIUM- PLUTONIUM 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 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

12 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

13 PhD category RADIATION EFFECTS IN ENGINEERING MATERIALS AND ENGINEERS Petty Cartemo Nuclear Engineering, Applied Physics, Chalmers University of Technology 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.

14 PhD category Geant4 simulations predict reality by using the ICRP voxel phantom. Key Words: radiation damage, positron beam, phantom, radiological accident

15 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 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 (CyMe 4 -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 CyMe 4 -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 CyMe 4 - BTBP

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17 PhD category COMMERCIAL THORIUM FUEL MANUFACTURE AND IRRADIATION: TESTING (Th,Pu)O 2 AND (Th,U)O 2 IN THE SEVEN- THIRTY PROGRAM Klara Insulander Björk Chalmers University of Technology, Department of Applied Physics, Division of Nuclear Engineering, SE Göteborg, Sweden Thor Energy, NO-0255 Oslo, Norway klara.insulander@scatec.no Further information about the project can be found at 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

18 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

19 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 Gothenburg, Sweden 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

20 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

21 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 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.