A new build in the Swedish nuclear power system would entail increased re-quirements for the proposed repository, which is adapted after the reactors of today. With a fast reactor, capable of burning nuclear waste, operating in parallel with the light water reactors, the increased requirements on the repository could be reduced. In this thesis, simulations of a light water reactor and a fast reactor have been performed by using the Monte Carlo code Serpent to investigate the changes in the fuel inventory. The light water reactor in the study is a boiling water reactor and the fast reactor of the type sodium-cooled fast reactor and they have been used for three different operation scenarios. By studying the fuel composition and the results from the simulations of the three scenarios conclusions can be drawn. Conclusions regarding the change of the fuel inventory and decay heat in Clab as well as the interim storage facility and in the repository. Depending on the operation alternative the changes dif-fered significantly and especially regarding the mass of burned actinides for different fuels in the fast reactor. The lowest increase of fuel assemblies was meet when using 50 years old fuel with 20MWd/kg U burnout and 2,0 % enrichment for start up of the fast reactor and 30 years old fuel assemblies with 50MWd/kg U burnout and 4,7 % enrichment for the further operation of the reactor. The increase of the number of fuel assemblies was 3174, which is equivalent to 641tons of heavy metal. Further this means an increase of the decay heat with 1,2MW. To decide whether or not it is possible to run a sodium-cooled fast reactor paral-lel with a light water reactor to compensate for the produced decay heat and fuel assemblies, further investigations concerning the deposit in the repository needs to be done.

<p>Generation IV (Gen IV) is a developing new generation of nuclear power reactors which is foreseen to bring about a safer and more sustainable production of nuclear power. A Swedish research program called GENIUS aims at developing the Gen IV technology, with emphasis on lead-cooled fast reactors. The present work is part of the GENIUS project, and deals with safeguards aspects for an envisioned future 100 MW Gen IV demonstration facility including storage and reprocessing plant. Also, the safeguards licensing aspects for the facilities have been investigated and results thereof are presented.</p><p>As a basis for the study, the changed usage and handling of nuclear fuel, as compared to that of today, have been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. Safeguards approaches have been considered for within and between each unit at the demonstration facility, with the main focus on system aspects rather than proposing safeguards instrumentation on a detailed level.</p><p>The proposed safeguards approach include the implementation of well-tried measures that are used at currently existing nuclear facilities as well as suggestions for new procedures. The former include, among others, regular inventory verifications, containment and surveillance measures as well as non-destructive and destructive measurements of nuclear materials. The traditional approaches may be improved and supplemented by modern techniques and approaches such as nuclear forensics, safeguards-by-design and improved on-line monitoring of streams of nuclear material. The safeguards approach for the demonstration facility should be outlined early in the licensing process, such that the facility units can be designed in a way that allows for implementation of adequate safeguards measures with minimal intrusion on the regular activities.</p><p>For operating a nuclear facility in Sweden, two separate permits are required. A license application for a new facility shall be handed both to the Swedish Radiation Safety Authority and to the environmental court, which in parallel prepare for decisions according to the Nuclear Activities Act and the Environmental Code, respectively. In terms of the Swedish legislation, there are no fundamental differences between Gen IV facilities and currently existing plants. However, comprehensive investigations and evaluations would be required in order to license new Gen IV facilities.</p>

Generation IV (Gen IV) is a developing new generation of nuclear power reactors which is foreseen to bring about a safer and more sustainable production of nuclear power. A Swedish research program called GENIUS aims at developing the Gen IV technology, with emphasis on lead-cooled fast reactors. The present work is part of the GENIUS project, and deals with safeguards aspects for an envisioned future 100 MW Gen IV demonstration facility including storage and reprocessing plant. Also, the safeguards licensing aspects for the facilities have been investigated and results thereof are presented. As a basis for the study, the changed usage and handling of nuclear fuel, as compared to that of today, have been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. Safeguards approaches have been considered for within and between each unit at the demonstration facility, with the main focus on system aspects rather than proposing safeguards instrumentation on a detailed level. The proposed safeguards approach include the implementation of well-tried measures that are used at currently existing nuclear facilities as well as suggestions for new procedures. The former include, among others, regular inventory verifications, containment and surveillance measures as well as non-destructive and destructive measurements of nuclear materials. The traditional approaches may be improved and supplemented by modern techniques and approaches such as nuclear forensics, safeguards-by-design and improved on-line monitoring of streams of nuclear material. The safeguards approach for the demonstration facility should be outlined early in the licensing process, such that the facility units can be designed in a way that allows for implementation of adequate safeguards measures with minimal intrusion on the regular activities. For operating a nuclear facility in Sweden, two separate permits are required. A license application for a new facility shall be handed both to the Swedish Radiation Safety Authority and to the environmental court, which in parallel prepare for decisions according to the Nuclear Activities Act and the Environmental Code, respectively. In terms of the Swedish legislation, there are no fundamental differences between Gen IV facilities and currently existing plants. However, comprehensive investigations and evaluations would be required in order to license new Gen IV facilities.

The effects of global warming raise demands for reduced CO2 emissions, whereas at the same time the world’s need for energy increases. With the aim to resolve some of the difficulties facing today’s nuclear power, striving for safety, sustainability and waste minimization, a new generation of nuclear energy systems is being pursued: Generation IV. New reactor concepts and new nuclear facilities should be at least as resistant to diversion of nuclear material for weapons production, as were the previous ones. However, the emerging generation of nuclear power will give rise to new challenges to the international safeguards community, due to new and increased flows of nuclear material in the nuclear fuel cycle. Before a wide implementation of Generation IV nuclear power facilities takes place, there lies still an opportunity to formulate safeguards requirements for the next generation of nuclear energy systems. In this context, this thesis constitutes one contribution to the global efforts to make future nuclear energy systems increasingly resistant to nuclear material diversion attempts. This thesis comprises three papers, all of which concern safeguards and proliferation resistance in Generation IV nuclear energy systems and especially recycling facilities: In Paper I, proliferation resistances of three fuel cycles, comprising different reprocessing techniques, are investigated. The results highlight the importance of making group actinide extraction techniques commercial, due to the inherently less vulnerable isotopic and radiological properties of the materials in such processes. Paper II covers the schematic design and safeguards instrumentation of a Generation IV recycling facility. The identification of the safeguards needs of planned facilities can act as a guide towards the development of new instrumentation suitable for Generation IV nuclear energy systems. Finally, Paper III describes a mode of procedure for assessing proliferation resistance of a recycling facility for fast reactor fuel. The assessments may be used, as in this case, as an aid to maintain or increase the inherent proliferation resistance when performing facility design changes and upgrades.

The system code TRACE is the latest development of the U.S. Nuclear Regulatory Commission (US NRC). TRACE, developed for the analysis of operational conditions, transients and accidents of light water reactors (LWR), is a best-estimate code with two fluid, six equation models for mass, energy, and momentum conservation, and related closure models. Since TRACE is mainly applied to LWR specific issues, the validation process related to innovative nuclear systems (liquid metal cooled systems, systems operated with supercritical water, etc.) is very limited, almost not existing. In this work, essential contribution to the validation of TRACE related to lead and lead alloy cooled systems as well as systems operated with supercritical water is provided in a consistent and corporate way. In a first step, model discrepancies of the TRACE source code were removed. This inconsistencies caused the wrong prediction of the thermo physical properties of supercritical water and lead bismuth eutectic, and hence the incorrect prediction of heat transfer relevant characteristic numbers like Reynolds or Prandtl number. In addition to the correction of the models to predict these quantities, models describing the thermo physical properties of lead and Diphyl THT (synthetic heat transfer medium) were implemented. Several experiments and numerical benchmarks were used to validate the modified TRACE version. These experiments, mainly focused on wall-to-fluid heat transfer, revealed that not only the thermo physical properties are afflicted with inconsistencies but also the heat transfer models. The models for the heat transfer to liquid metals were enhanced in a way that the code can now distinguish between pipe and bundle flow by using the right correlation. The heat transfer to supercritical water was not existing in TRACE up to now. Completely new routines were implemented to overcome that issue. The comparison of the calculations to the experiments showed, on one hand, the necessity of these changes and, on the other hand, the success of the new implemented routines and functions. The predictions using the modified TRACE version were close to the experimental data. After validating the modified TRACE version, two design studies related to the Generation IV International Forum (GIF) were investigated. In the first one, a core of a lead-cooled fast reactor (LFR) was analyzed. To include the interaction between the thermal hydraulic and the neutron kinetic due to temperature and density changes, the TRACE code was coupled to the program system ERANOS2.1. The results gained with that coupled system are in accordance with theory and helped to identify sub-assemblies with the highest loads concerning fuel and cladding temperature. The second design which was investigated was the High Performance Light Water Reactor (HPLWR). Since the design of the HPLWR is not finalized, optimization of vital parameters (power, mass flow rate, etc.) are still ongoing. Since most of the parameters are affecting each other, an uncertainty and sensitivity analysis was performed. The uncertainty analysis showed the upper and lower boundaries of selected parameters, which are of importance from the safety point of view (e.g., fuel and cladding temperature, moderator temperature). The sensitivity study identified the most relevant parameters and their influence on the whole system.

The system code TRACE is the latest development of the U.S. Nuclear Regulatory Commission (US NRC). TRACE, developed for the analysis of operational conditions, transients and accidents of light water reactors (LWR), is a best-estimate code with two fluid, six equation models for mass, energy, and momentum conservation, and related closure models. Since TRACE is mainly applied to LWR specific issues, the validation process related to innovative nuclear systems (liquid metal cooled systems, systems operated with supercritical water, etc.) is very limited, almost not existing. In this work, essential contribution to the validation of TRACE related to lead and lead alloy cooled systems as well as systems operated with supercritical water is provided in a consistent and corporate way. In a first step, model discrepancies of the TRACE source code were removed. This inconsistencies caused the wrong prediction of the thermo physical properties of supercritical water and lead bismuth eutectic, and hence the incorrect prediction of heat transfer relevant characteristic numbers like Reynolds or Prandtl number. In addition to the correction of the models to predict these quantities, models describing the thermo physical properties of lead and Diphyl THT (synthetic heat transfer medium) were implemented. Several experiments and numerical benchmarks were used to validate the modified TRACE version. These experiments, mainly focused on wall-to-fluid heat transfer, revealed that not only the thermo physical properties are afflicted with inconsistencies but also the heat transfer models. The models for the heat transfer to liquid metals were enhanced in a way that the code can now distinguish between pipe and bundle flow by using the right correlation. The heat transfer to supercritical water was not existing in TRACE up to now. Completely new routines were implemented to overcome that issue. The comparison of the calculations to the experiments showed, on one hand, the necessity of these changes and, on the other hand, the success of the new implemented routines and functions. The predictions using the modified TRACE version were close to the experimental data. After validating the modified TRACE version, two design studies related to the Generation IV International Forum (GIF) were investigated. In the first one, a core of a lead-cooled fast reactor (LFR) was analyzed. To include the interaction between the thermal hydraulic and the neutron kinetic due to temperature and density changes, the TRACE code was coupled to the program system ERANOS2.1. The results gained with that coupled system are in accordance with theory and helped to identify sub-assemblies with the highest loads concerning fuel and cladding temperature. The second design which was investigated was the High Performance Light Water Reactor (HPLWR). Since the design of the HPLWR is not finalized, optimization of vital parameters (power, mass flow rate, etc.) are still ongoing. Since most of the parameters are affecting each other, an uncertainty and sensitivity analysis was performed. The uncertainty analysis showed the upper and lower boundaries of selected parameters, which are of importance from the safety point of view (e.g., fuel and cladding temperature, moderator temperature). The sensitivity study identified the most relevant parameters and their influence on the whole system.

<p>The use of hydrogen is predicted to increase substantially in the future, both as chemical feedstock and also as energy carrier for transportation. The annual world production of hydrogen amounts to some 50 million tonnes and the majority is produced using fossil fuels like natural gas, coal and naphtha. High temperature nuclear reactors (HTRs) represent a novel way to produce hydrogen at large scale with high efficiency and less carbon footprint. The aim of this master thesis has been to evaluate the feasibility of HTRs for hydrogen production by analyzing both the reactor concept and its potential to be used in certain hydrogen niche markets. The work covers the production, storage, distribution and use of hydrogen as a fuel for vehicles and aviation and as chemical feedstock for the oil refining and ammonia production industry.</p><p>The study indicates that HTRs may be suitable for hydrogen production under certain conditions. However, the use of hydrogen as an energy carrier necessitates a widespread hydrogen infrastructure (e.g. pipe-lines, refuelling stations and large scale storage), which is associated with major energy losses. Both mentioned industries could benefit from nuclear-based hydrogen with less infrastructural changes, but the potential market is by far smaller than if hydrogen is used as an energy carrier. A maximum of about 60 HTRs of 600 MWth worldwide has been estimated for the ammonia production industry. The Swedish refineries are likely too small to utilize the HTR but in the larger refineries HTR might be applicable.</p>

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 oC. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX. / UOIT

A large set of experimental data, obtained in Russia, was analyzed and a new heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those for proposed SuperCritical Water-cooled nuclear Reactor (SCWR) concepts. Thus, this new correlation, for forced convective heat transfer in the normal heat-transfer regime, can be used for preliminary heat-transfer calculations in SCWR fuel channels. It has demonstrated a good fit for Heat Transfer Coefficient (HTC) values (±25%) and for wall temperature calculations (±15) for the analyzed dataset. This correlation can be used for supercritical water heat exchangers linked to indirectcycle concepts and the co-generation of hydrogen, for future comparisons with other independent datasets, with bundle data, as the reference case, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids. / UOIT

The use of hydrogen is predicted to increase substantially in the future, both as chemical feedstock and also as energy carrier for transportation. The annual world production of hydrogen amounts to some 50 million tonnes and the majority is produced using fossil fuels like natural gas, coal and naphtha. High temperature nuclear reactors (HTRs) represent a novel way to produce hydrogen at large scale with high efficiency and less carbon footprint. The aim of this master thesis has been to evaluate the feasibility of HTRs for hydrogen production by analyzing both the reactor concept and its potential to be used in certain hydrogen niche markets. The work covers the production, storage, distribution and use of hydrogen as a fuel for vehicles and aviation and as chemical feedstock for the oil refining and ammonia production industry. The study indicates that HTRs may be suitable for hydrogen production under certain conditions. However, the use of hydrogen as an energy carrier necessitates a widespread hydrogen infrastructure (e.g. pipe-lines, refuelling stations and large scale storage), which is associated with major energy losses. Both mentioned industries could benefit from nuclear-based hydrogen with less infrastructural changes, but the potential market is by far smaller than if hydrogen is used as an energy carrier. A maximum of about 60 HTRs of 600 MWth worldwide has been estimated for the ammonia production industry. The Swedish refineries are likely too small to utilize the HTR but in the larger refineries HTR might be applicable.