The scientific endeavors that took place at Hanford Engineer Works, beginning in World War II and continuing thereafter, are often overlooked in the literature on the Manhattan Project, the Atomic Energy Commission, and in regional histories. To historians of science, Hanford is described as an industrial facility that illustrates the perceived differences between academic scientists on the one hand and industrial scientists and engineers on the other. To historians of the West such as Gerald Nash, Richard White, and Patricia Limerick, Hanford has functioned as an example of the West's transformation during in World War II, the role of science in this transformation, and the recurring impacts of industrialization on the western landscape. This thesis describes the establishment and gradual expansion of a multi-disciplinary research program at Hanford whose purpose was to assess and manage the biological and environmental effects of plutonium production. By drawing attention to biological research, an area in which Hanford scientists gained distinction by the mid 1950s, this study explains the relative obscurity of Hanford's scientific research in relation to the prominent, physics-dominated national laboratories of the Atomic Energy Commission. By the mid 1960s, with growing public concern over radiation exposure and changes in the government's funding patterns for science, Hanford's ecologically relevant research provided a recognizable and valuable identity for the newly independent, regionally-based research laboratory. With funding shifts favoring the biological and environmental sciences in the latter half of the twentieth-century, Hanford scientists were well prepared to take advantage of expanding opportunities to carve out a permanent niche on the border of American science. / Graduation date: 2003

The MURE code is based on the coupling of a Monte Carlo static code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to analyse two different questions, concerning the mono-recycling of Am in present French Pressurized Water Reactor, and the conversion of high enriched uranium (HEU) used in the Miniature Neutron Source Reactor in Ghana into low enriched uranium (LEU) due to proliferation resistance issues. In both cases, a detailed comparison is made on burnup and the induced radiotoxicity of waste or spent fuel. The UOX fuel assembly, as in the open cycle system, was designed to reach a burn-up of 46GWd/T and 68GWd/T. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of Plutonium and addition of depleted Uranium to reach burn-ups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. Spent UOX fuel, after 30 years of cooling in the repository required higher concentration of Pu to be reprocessed into a MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has high radiotoxic level, high mid-term residual heat and a precursor for other long lived isotope. An innovative strategy consists of reprocessing not only the plutonium from the UOX spent fuel but also the americium isotopes which dominate the radiotoxicity of present waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all the Am. The main objective is to propose a "waiting strategy" for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOXAm (MOX and Americium isotopes) fuel was fabricated to see the effect of americium in MOX fuel on the burn-up, neutronic behavior and on radiotoxicity. The MOXAm fuel showed relatively good indicators both on burnup and on radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared to the open cycle. All fuel types understudy in the PWR cycle showed good safety inherent feature with the exception of the some MOXAm assemblies which have a positive void coefficient in specific configurations, which could not be consistent with safety features. The core lifetimes of the current operating 90.2% HEU UAl fuel and the proposed 12.5% LEU UOX fuel of the MNSR were investigated using MURE code. Even though LEU core has a longer core life due to its higher core loading and low rate of uranium consumption, the LEU core will have it first beryllium top up to compensate for reactivity at earlier time than the HEU core. The HEU and LEU cores of the MNSR exhibited similar neutron fluxes in irradiation channels, negative feedback of temperature and void coefficients, but the LEU is more radiotoxic after fission product decay due to higher actinides presence at the end of its core lifetime.

Les phénomènes de diffusion se produisant lors du frittage des oxydes mixtes d'uranium et deplutonium (MOX) dépendent du potentiel d'oxygène de l'atmosphère du four, qui détermine lanature et la concentration des défauts ponctuels dans le matériau. Les travaux de thèse ont porté surune meilleure compréhension de l'influence du potentiel d'oxygène sur la densification, la formationde la solution solide et l'interdiffusion U-Pu lors du frittage des combustibles MOX. Pour cela, unlarge domaine de potentiel d'oxygène a été étudié, entre -600 et -100 kJ.mol-1 à 1700°C, afin demettre en évidence les différents mécanismes diffusionnels et leur impact sur la microstructurelorsqu'on s'éloigne de la composition stoechiométrique i.e. lorsque la concentration en défautsaugmente.Les études ont montré que plus le potentiel d'oxygène augmente, plus la densification du mélange70 % UO2+x + 30 % PuO2 s'effectue à basse température. Lors du chauffage, les oxydes de départ(UO2+x et PuO2-x) densifient dans un premier temps puis la solution solide se forme à une températureplus élevée d'environ 200°C. La solution solide apparaît à plus basse température quand le potentield'oxygène augmente, avec une cinétique de formation plus rapide. L'étude de l'interdiffusion U-Puindique qu'un traitement thermique avec un potentiel d'oxygène supérieur à -150 kJ.mol-1 à 1700°Cpermet d'obtenir un coefficient d'interdiffusion supérieur d'un à deux ordres de grandeur à ceuxobtenus entre -550 et -350 kJ.mol-1 à 1700°C et conduit donc à une homogénéisation U-Pu accrue.Cette étude permet de donner des recommandations sur le choix de l'atmosphère et de proposer uncycle de frittage optimisé en fonction de l'application ou de la caractéristique souhaitée.

A detailed reactor core modeling of the LOVIISA-2 PWR and FORSMARK-3BWR was performed in the Serpent 2 Continuous Energy Monte-Carlocode.Both models of the reactors were completed but the approximations ofthe atomic densities of nuclides present in the core differedsignificantly.In the LOVIISA-2 PWR, the predicted atomic density for the nuclidesapproximated by Chebyshev Rational Approximation method (CRAM)coincided with the corrected atomic density simulated by the Serpent2 program. In the case of FORSMARK-3 BWR, the atomic density fromCRAM poorly approximated the data returned by the simulation inSerpent 2. Due to boiling of the moderator in the core of FORSMARK-3,the model seemed to encounter problems of fission density, whichyielded unusable results.The results based on the models of the reactor cores are significantto the FANCSEE Nuclear fuel cycle simulator, which will be used as adataset for the nuclear fuel cycle burnup in the reactors. / FANCSEE

Fundamental understanding of the behaviour of nuclear fuel has been of great importance. Enhancing this knowledge not only by means of experimental observations, but also via multi-scale modelling is of current interest. The overall goal of this thesis is to understand the impact of atomic interactions on the nuclear fuel material properties. Two major topics are tackled in this thesis. The first topic deals with non-stoichiometry in uranium dioxide (UO2) to be addressed by empirical potential (EP) studies. The second fundamental question to be answered is the effect of the atomic fraction of americium (Am), neptunium (Np) containing uranium (U) and plutonium (Pu) mixed oxide (MOX) on the material properties.UO2 has been the reference fuel for the current fleet of nuclear reactors (Gen-II and Gen-III); it is also considered today by the Gen-IV International Forum for the first cores of the future generation of nuclear reactors on the roadmap towards minor actinide (MA) based fuel technology. The physical properties of UO2 highly depend on material stoichiometry. In particular, oxidation towards hyper stoichiometric UO2 – UO2+x – might be encountered at various stages of the nuclear fuel cycle if oxidative conditions are met; the impact of physical property changes upon stoichiometry should therefore be properly assessed to ensure safe and reliable operations. These physical properties are intimately linked to the arrangement of atomic defects in the crystalline structure. The first paper evaluates the evolution of defect concentration with environment parameters – oxygen partial pressure and temperature by means of a point defect model, with reaction energies being derived from EP based atomic scale simulations. Ultimately, results from the point defect model are discussed, and compared to experimental measurements of stoichiometry dependence on oxygen partial pressure and temperature. Such investigations will allow for future discussions about the solubility of different fission products and dopants in the UO2 matrix at EP level.While the first paper answers the central question regarding the dominating defects in non-stoichiometry in UO2, the focus of the second paper was on the EP prediction of the material properties, notably the lattice parameter of Am, Np containing U and Pu MOX as a function of atomic fractions.The configurational space of a complex U1-y-y’-y’’PuyAmy’Npy’’O2 system, was assessed via Metropolis-Monte Carlo techniques. From the predicted configuration, the relaxed lattice parameter of Am, Np bearing MOX fuel was investigated and compared with available literature data. As a result, a linear behaviour of the lattice parameter as a function of Am, Np content was observed, as expected for an ideal solid solution. These results will allow to support and increase current knowledge on Gen-IV fuel properties, such as melting temperature, for which preliminary results are presented in this thesis, and possibly thermal conductivity in the future. / Doctorat en Sciences de l'ingénieur et technologie / info:eu-repo/semantics/nonPublished

科学研究費補助金 研究種目:一般研究(C) 課題番号:60580177 研究代表者:飯田 孝夫 研究期間:1985-1986年度

As nuclear energy grows in the United States and around the world, the expansion of the nuclear fuel cycle is inevitable. All currently deployed commercial reprocessing plants are based on the Plutonium - Uranium Extraction (PUREX) process. However, this process is not implemented in the U.S. for a variety of reasons, one being that it is considered by some as a proliferation risk. The 2001 Nuclear Energy Policy report recommended that the U.S. "develop reprocessing and treatment technologies that are cleaner, more efficient, less waste-intensive, and more proliferation-resistant." The Uranium Extraction (UREX+) reprocessing technique has been developed to reach these goals. However, in order for UREX+ to be considered for commercial implementation, a safeguards approach is needed to show that a commercially sized UREX+ facility can be safeguarded to current international standards. A detailed safeguards approach for a UREX+1a reprocessing facility has been developed. The approach includes the use of nuclear material accountancy (MA), containment and surveillance (C/S) and solution monitoring (SM). Facility information was developed for a hypothesized UREX+1a plant with a throughput of 1000 Metric Tons Heavy Metal (MTHM) per year. Safeguard goals and safeguard measures to be implemented were established. Diversion and acquisition pathways were considered; however, the analysis focuses mainly on diversion paths. The detection systems used in the design have the ability to provide near real-time measurement of special fissionable material in feed, process and product streams. Advanced front-end techniques for the quantification of fissile material in spent nuclear fuel were also considered. The economic and operator costs of these systems were not considered. The analysis shows that the implementation of these techniques result in significant improvements in the ability of the safeguards system to achieve the objective of timely detection of the diversion of a significant quantity of nuclear material from the UREX+1a reprocessing facility and to provide deterrence against such diversion by early detection.