<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 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.

With the initiative taken by the SUNRISE project (Sustainable Nuclear Energy Research in Sweden) to construct a Lead-cooled research reactor, this thesis intends to extend the knowledge within nuclear fuel development. By using integral iterative modelling and simulating techniques that mimic real-world phenomena, novel fuel materials like uranium nitride are assessed for future validation.  The work deals with the fuel performance analysis of the SUNRISE LFR, employing the TRANSURANUS fuel performance code. This code contains a collection of model parameters that simulate the thermo-mechanical behaviour of the fuel cladding system on an engineering scale of the reactor core. A comparative study is performed for UO$_2$ and UN fuels using the same input data such as fuel geometry. In addition, predefined information relating to the neutronics analysis for the reactor was used as input to the TRANSURANUS code along with literature reviews to select the accurate models on the reactor, fuel, and its behaviour. Furthermore, a sensitivity study is carried out to assess the models and parameters affected by more significant uncertainty.  The uncertainty analysis of the UN fuel's swelling models is performed using the Dakota toolkit. The sampling of input data using the Dakota software coupled with the nuclear simulation program TRANSURANUS produced partial rank correlation coefficients significant to the modelling. However, since the assessed models displayed the same correlation coefficients, the results conclude that a deeper understanding of the theoretical swelling model might be required. / I samverkan med initiativet av SUNRISEprojektet (Sustainable Nuclear Energy Research inSweden) som syftar att bygga en blykyld forskningsreaktor, avser denna avhandling att utökakunskapen inom kärnbränsleutveckling. Med användning av integral iterativ modellering ochsimuleringstekniker som efterliknar verkliga fenomen bedöms nya bränslematerial somuranmononitrid för framtida validering. Arbetet behandlar analysen av bränsleprestanda för SUNRISE LFR, med användning avTRANSURANUS bränsleprestandakod. Denna kod innehåller en samling modellparametrarsom simulerar det termomekaniska beteendet hos bränslebetäckningssystemet i en tekniskskala för reaktorkärnan. En jämförande studie utförs för UO2 och UN-bränslen med sammaingångsdata som t.ex bränslegeometrin. Dessutom användes fördefinierad information om denneutroniska analysen för reaktorn som ingångsdata till TRANSURANUSkoden tillsammans medgranskning av litteratur för att välja lämpliga modeller för reaktorn, bränslet och dess beteende.Därtill genomfördes en känslighetsstudie för att bedöma de modeller och parametrar sompåverkas av mer betydande osäkerhet. Osäkerhetsanalysen av UN-bränslets svällningsmodeller utförs med hjälp av Dakota-verktyget.Samlingen av indata med Dakota-programmet i kombination medkärnkraftssimuleringsprogrammet TRANSURANUS gav korrelationskoefficienter för partiell rangviktiga för modelleringen. Eftersom de utvärderade modellerna visade sammakorrelationskoefficienter, tyder slutsatsen på att en djupare förståelse av den teoretiskasvällningsmodellen krävs