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The purpose of this research was to identify fission product signatures that could be used to uniquely identify a specific spent fuel assembly in order to improve international safeguards. This capability would help prevent and deter potential diversion of spent fuel for a nuclear weapons program. The power history experienced by a fuel assembly is distinct and could serve as the basis of a method for unique identification. Using fission product concentrations to characterize the assembly power history would limit the ability of a proliferator to deceive the identification method. As part of the work completed, the TransLat lattice physics code was successfully benchmarked for fuel depletion. By developing analytical models for potential monitor isotopes an understanding was built of how specific isotope characteristics affect the production and destruction mechanisms that determine fission product concentration. With this knowledge potential monitor isotopes were selected and tested for concentration differences as a result of power history variations. Signature ratios were found to have significant concentration differences as a result of power history variations while maintaining a constant final burnup. A conceptual method for implementation of a fission product identification system was proposed in conclusion.

The purpose of this research was to identify fission product signatures that could be used to uniquely identify a specific spent fuel assembly in order to improve international safeguards. This capability would help prevent and deter potential diversion of spent fuel for a nuclear weapons program. The power history experienced by a fuel assembly is distinct and could serve as the basis of a method for unique identification. Using fission product concentrations to characterize the assembly power history would limit the ability of a proliferator to deceive the identification method. As part of the work completed, the TransLat lattice physics code was successfully benchmarked for fuel depletion. By developing analytical models for potential monitor isotopes an understanding was built of how specific isotope characteristics affect the production and destruction mechanisms that determine fission product concentration. With this knowledge potential monitor isotopes were selected and tested for concentration differences as a result of power history variations. Signature ratios were found to have significant concentration differences as a result of power history variations while maintaining a constant final burnup. A conceptual method for implementation of a fission product identification system was proposed in conclusion.

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Made available in DSpace on 2014-10-09T12:37:15Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:09:08Z (GMT). No. of bitstreams: 1 04785.pdf: 3458454 bytes, checksum: 560fdc27a291126a48fceeb5a4a5137a (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

This document gives an overview of the proposed MSc study. The main goal of the study is to model the cases listed in the code benchmark study of the International Atomic Energy Agency CRP–6 fuel performance study (Verfondern & Lee, 2005). The platform that will be employed is the GETTER code (Keshaw & van der Merwe, 2006). GETTER was used at PBMR for the release calculations of metallic and some non–metallic long–lived fission products. GETTER calculates the transport of fission products from their point of fission to release from the fuel surface taking into account gas precursors and activation products. Results show that for certain experiments the codes correspond very well with the experimental data whilst in others there are orders of magnitude differences. It can be seen that very similar behaviour is observed in all codes. Improvements are needed in updating the strontium diffusion coefficient and in understanding, on a deeper level, the transport of silver in TRISO particles and how it deviates from simple diffusion models. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.

This document gives an overview of the proposed MSc study. The main goal of the study is to model the cases listed in the code benchmark study of the International Atomic Energy Agency CRP–6 fuel performance study (Verfondern & Lee, 2005). The platform that will be employed is the GETTER code (Keshaw & van der Merwe, 2006). GETTER was used at PBMR for the release calculations of metallic and some non–metallic long–lived fission products. GETTER calculates the transport of fission products from their point of fission to release from the fuel surface taking into account gas precursors and activation products. Results show that for certain experiments the codes correspond very well with the experimental data whilst in others there are orders of magnitude differences. It can be seen that very similar behaviour is observed in all codes. Improvements are needed in updating the strontium diffusion coefficient and in understanding, on a deeper level, the transport of silver in TRISO particles and how it deviates from simple diffusion models. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.

abstract: Fission products in nuclear fuel pellets can affect fuel performance as they change the fuel chemistry and structure. The behavior of the fission products and their release mechanisms are important to the operation of a power reactor. Research has shown that fission product release can occur through grain boundary (GB) at low burnups. Early fission gas release models, which assumed spherical grains with no effect of GB diffusion, did not capture the early stage of the release behavior well. In order to understand the phenomenon at low burnup and how it leads to the later release mechanism, a microstructurally explicit model is needed. This dissertation conducted finite element simulations of the transport behavior using 3-D microstructurally explicit models. It looks into the effects of GB character, with emphases on conditions that can lead to enhanced effective diffusion. Moreover, the relationship between temperature and fission product transport is coupled to reflect the high temperature environment. The modeling work began with 3-D microstructure reconstruction for three uranium oxide samples with different oxygen stoichiometry: UO2.00 UO2.06 and UO2.14. The 3-D models were created based on the real microstructure of depleted UO2 samples characterized by Electron Backscattering Diffraction (EBSD) combined with serial sectioning. Mathematical equations on fission gas diffusion and heat conduction were studied and derived to simulate the fission gas transport under GB effect. Verification models showed that 2-D elements can be used to model GBs to reduce the number of elements. The effect of each variable, including fuel stoichiometry, temperature, GB diffusion, triple junction diffusion and GB thermal resistance, is verified, and they are coupled in multi-physics simulations to study the transport of fission gas at different radial location of a fuel pellet. It was demonstrated that the microstructural model can be used to incorporate the effect of different physics to study fission gas transport. The results suggested that the GB effect is the most significant at the edge of fuel pellet where the temperature is the lowest. In the high temperature region, the increase in bulk diffusivity due to excess oxygen diminished the effect of GB diffusion. / Dissertation/Thesis / Doctoral Dissertation Materials Science and Engineering 2014

京都大学 / 0048 / 新制・課程博士 / 博士(エネルギー科学) / 甲第21192号 / エネ博第366号 / 新制||エネ||72(附属図書館) / 京都大学大学院エネルギー科学研究科エネルギー基礎科学専攻 / (主査)教授 野平 俊之, 教授 萩原 理加, 教授 佐川 尚 / 学位規則第4条第1項該当 / Doctor of Energy Science / Kyoto University / DFAM

It is important to have a method of experimentally calculating fission product yields. Statistical calculations and simulations produce very large uncertainties. Experimental calculations, depending on the methods used, tend to produce lower uncertainties. This work set up a method to calculate fission product yields using gamma ray spectroscopy. In order to produce a method that was theoretically sound, a simulation was set up using OrigenArp to calculate theoretical concentrations of fission products from the irradiation of natural uranium. From these concentrations, the fission product yields were calculated to verify that they would agree with expected values. Moving forward in the work, the total flux at the point of irradiation, in the pneumatic transfer system, was calculated and determined to be 3.9070E+11 ± 6.9570E+10 n/cm^2/s at 100 kW. Once the flux was calculated, the method for calculating fission product yields was implemented and yields were calculated for 10 fission products. The yields calculated were in very good agreement (within 10.04%) with expected values taken from the ENDF-349 library. This method has strong potential in nuclear forensics as it can provide a means for developing a library of experimentally-determined fission product yields, as well as rapid post-nuclear detonation analysis. / text