<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.

One of the future reactor technologies defined by the Generation-IV International Forum (GIF) is the Lead-Cooled Fast Reactor (LFR). An advantage with this reactor technology is that steam production is accomplished by means of heat exchangers located within the primary reactor vessel, which decreases costs and increases operational safety. However, a crack in a heat exchanger tube may create steam (void) into the coolant and this process has the potential to introduce reactivity changes, which may cause criticality issues. This fact motivates the development of a methodology to detect such voids. This thesis comprises theoretical investigations on a possible route to detect voids by studying changes of the neutron spectrum in a small LFR as a function of various types of in-core voids .The methodology includes a combination of fission chambers loaded with U-235 and Pu-242 operating in various positions. It is shown that such a combination results in information that can be made independent on reactor power, a feasible property in order to detect the relatively small spectral changes due to void. A sensitivity analysis of various combinations of detectors, fuel burnup and void has also been included in the investigation. The results show that the proposed methodology yields a reasonably large sensitivity to voids down to (1-2) % of the coolant volume. The results obtained so far point in the direction that the proposed methodology is an interesting subject for further studies.

This thesis is a follow up on "SPARC fast reactor design: Design of two passively metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control" by Tobias Lindström (2015). In this thesis the two reactors designed by Lindström in said thesis were evaluated. The goal was to determine the reactors ability to load follow as well as the burnup of the neutron absorber used in the passive control system. To be able to determine the dynamic behaviour of the reactors the reactivity feedbacks of the cores were modelled using Serpent, a Monte Carlo simulation software for 3D neutron transport calculations. These feedbacks were then implemented into a dynamic simulation of the core, primary and secondary circulation and steam generator. The secondary circulation and feedwater flow were used to regulate steam temperature and turbine power. The core was left at constant coolant flow and no control rods were used. The simulations showed that the reactor was able to load follow between 100 % and 40 % of rated power at a speed of 6 % per minute. It was also shown that the reactor could safely adjust its power between 100 % and 10 % of rated power suggesting that load following is possible below 40 % of rated power but at a lower speed. Finally the reactors were allowed compensate for the variations in a week of the Latvian wind power production in order to show one possible application of the reactor.

El Generation IV International Forum (GIF) [1] es un programa internacional dedicado a apoyar, coordinar y dirigir las iniciativas de investigación y desarrollo encaminados a implementar las soluciones tecnológicas que caracterizarán a la siguiente generación de reactores nucleares. Estos reactores se caracterizaran por una gestión más eficiente del combustible nuclear, un incremento en las exigencias de seguridad y una alta competitividad económica. Con tales objetivos, GIF propuso una serie de diseños potencialmente capaces de alcanzarlos. Estos diseños son tecnológicamente muy distintos a las plantas nucleares comerciales actuales al utilizar neutrones de espectro rápido y consecuentemente refrigeración por metales líquidos. Estos nuevos diseños requieren el desarrollo y validación de herramientas computacionales capaces de simular el comportamiento de la planta tanto en fase estacionaria como en transitoria y por tanto sean aplicables en los procesos de diseño y licitación de dichas plantas. El objetivo de esta tesis es el de adaptar los códigos computacionales actuales aplicados a la simulación de reactores refrigerados por agua a reactores rápidos refrigerados por metales líquidos, tales como el sodio o el plomo y el desarrollo de modelos capaces de simular de una manera consistente el comportamientos de los sistemas ante determinados eventos que constituyen la base de diseño de la planta Para ello se adaptaran dichos códigos a la fenomenología específica de estos reactores, se desarrollaran modelos termo-hidráulicos y neutrónicos tanto unidimensionales como tridimensionales de los diseños propuestos y se validarán los resultados para demostrar su aplicabilidad. El trabajo incluye la implementación de correlaciones específicas para habilitar los códigos para el cálculo de la condiciones termo-hidráulicas de los refrigerantes así como la adaptación de los esquemas de acoplamiento termo-hidráulico-neutrónicos existentes a esta nueva tecnología. / Lázaro Chueca, A. (2014). Development, assessment and application of computational tools for design safety analysis of liquid metal cooled fast breeder reactors [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/39353 / TESIS

In this thesis, the feasibility to use sodium cooled fast reactors loaded with MOX, metallic and nitride fuels for efficient transmutation of americium is investigated by performing transient analysis for cases with different americium contents in fuels, using safety parameters obtained with the SERPENT Monte Carlo code. It was then demonstrated that there is no solid limit for the Am introduction into oxide, metallic and nitride fuels that were loaded into sodium fast reactors. Instead, higher Am contents could be permitted if specific levels of power penalty were accepted. Transient analysis of a new Accelerator Driven System design with higher neutron source efficiency than the reference EFIT-400 design, was also performed. Based on simulation results, the suggested ADS design was proved to survive the full set of transients, preserving 130 K margin to cladding rupture during the most limiting transient. After comparing Am transmutation performances in SFRs and the suggested ADS, it can be concluded that: 1. Nitride fuel could provide the highest Am transmutation efficiency, when loaded into SFRs; 2. One SFR loaded with nitride fuel is sufficient to transmute Am inventory produced by more than 15 commercial LWRs within the same time period, which is three times higher than the supporting ratio reported for the suggested ADS; 3. The total fraction of ADS power in the power park is half of cases for critical reactors. / QC 20120201

With respect to nitrides being considered as potential fast reactor fuels, research is conducted on the out-of-pile thermophysical properties, sintering and fabrication processes, gas migration mechanisms, self-diffusion and point defect behaviour of actinide nitrides, their surrogate materials, and the inert matrix material ZrN . The experimental research, carried out in the framework of qualifying fuel for the European Lead Cooled Training Reactor (ELECTRA), shows that sintered ZrN and (Dy,Zr)N pellet densities are influenced by the oxygen concentration in the material. The effect is confirmed in sintered (Pu,Zr)N pellets. Oxygen concentration also plays a role in the thermophysical properties of inert matrix nitride fuels, but does not have an impact on the electrical properties of these materials. With the fuel fabrication methods applied here, clean nitride powders can be synthesized. However, the subsequent fabrication phases, including milling and solid solution formation, increases the impurity levels significantly. Research of equal importance is performed on materials free of fabrication-induced impurities, whose properties are studied by employing first-principles methods. ZrN, UN and (U,Zr)N are studied, whereas the results from ZrN are expected to be applicable for actinide nitrides as a first approximation. The migration of noble gases in ZrN, on the atomic scale, confirms the experimentally observed tendency for noble gases with higher atomic number to be retained in the fuel matrix, while the majority of He is released to the fuel pin. Materials modelling implies that self-diffusion of nitrogen and metal atoms in inert matrix nitride fuels is accelerated under irradiation, since noble gas retention reduces migration barriers which govern self-diffusion. Unlike Kr and Xe, He has the capacity to be released into the fuel matrix, after having been trapped in a vacancy. The results are expected to aid in providing an explanation to the macroscopic diffusion phenomena in nitride fuels, as the diffusion behaviour of noble gases is sparsely studied. In addition, a study on the miscibility of ZrN and UN in a narrow composition range suggests solubility, based on the negative mixing energies. The results obtained from research on inert matrix nitride fuel underline several beneficial properties which are desirable in a fast reactor fuel. The relevance of these results is analyzed and contextualized in the thesis, from the perspective of current research and development in the field. / <p>QC 20130611</p>

The lead-cooled fast reactor (LFR) is one of the concepts of the Generation IV reactorsystems. There are some issues that have to be solved before a research orcommercial LFR can be built. The objective of this thesis was to identify these keyissues and analyse them by studying results from previous research: choice of fuel,corrosion on structural materials and corrosion/erosion on pumps.The major fuel candidates for the LFR are MOX fuel (Mixed OXide fuel), metallic fuel,nitride fuel and carbide fuel. Nitride fuel has desirable properties but its production ismore difficult than for MOX fuel.Most of today’s commercial steels are not corrosion resistant at higher temperaturesbut they could possibly be used for an LFR test demonstrator with an operatingtemperature lower than 450 ºC. A new type of steel called oxide dispersionstrengthened (ODS) steel and a new ceramic material MAXTHAL both showpromising corrosion resistance even at higher temperatures.By controlling the oxygen concentration a protective oxide film is produced. Flowingliquid coolant causes erosion and wears down the oxide film. Pumps are exposed tocoolant velocities of 10-15 m/s causing both erosion and corrosion. There is nosolution today, but MAXTHAL shows promising results in tests with liquid lead of lowvelocity. There are also other issues unsolved, such as irradiation damage onstructural materials, thus more research is needed.Economic and political aspects were not covered in this study. This thesis work wasperformed at Vattenfall Research and Development AB.

The Monte Carlo code Serpent has been used to model the material flows andisotope compositions for a lead cooled fast reactor. The demonstration sized trainingreactor ELECTRA was chosen for the investigation, and different fuel cycle scenarioswere studied. The scenarios differed in operation length (3 months, 1 year or 5years) and recycling technique (single and double PUREX or GANEX). The simulations gave detailed information on the changes of the isotope composition,activity and decay heat. The analysis of the generated waste also showed that thechoice of recycling method had great impact on the final storage time of the wastefrom the reprocessing. Performing double GANEX recycling, as compared to singlePUREX, reduced the storage time by a factor of about 3500. The results can be used for future work related to even more detailed studies ofmaterial flows and for designing an appropriate safeguards system.

A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner Reactor (SABR). SABR is a fusion-fission hybrid reactor that combines the leading sodium cooled fast reactor technology with the leading tokamak plasma technology based on ITER physics. Two general fuel cycles were considered for the SABR system. The first fuel cycle is one in which all of the transuranics from light water reactors are burned in SABR. The second fuel cycle is a minor actinide burning fuel cycle in which all of the minor actinides and some of the plutonium produced in light water reactors are burned in SABR, with the excess plutonium being set aside for starting up fast reactors in the future. The minor actinide burning fuel cycle is being considered in European Scenario Studies. The fuel cycles were evaluated on the basis of TRU/MA transmutation rate, power profile, accumulated radiation damage, and decay heat to the repository. Each of the fuel cycles are compared against each other, and the minor actinide burning fuel cycles are compared against the EFIT transmutation system, and a low conversion ratio fast reactor.