A molecular dynamics study of the thermal properties of thorium oxide

There is growing interest in the exploitation of the thorium nuclear fuel cycle as an alternative to that of uranium. As part of a wider study of the suitability of thorium dioxide (thoria) as a nuclear fuel, we have used molecular dynamics to investigate the thermal expansion, oxygen diffusion, and heat capacity of pure thoria and uranium doped (1-10%) thoria between 1500K and 3600 K. Our results indicate that the thermal performance of the thoria matrix, even when doped with 10%U, is comparable to, and possibly better than, that of UO2

Multimegawatt Nuclear Reactor Design for Plasma Propulsion Systems

Peer Reviewedhttp://deepblue.lib.umich.edu/bitstream/2027.42/76042/1/AIAA-5456-353.pd

Modelling of fission gas release in UO2 doped fuel using transuranus code

The expected benefits from Cr-doped fuel are improved retention of fission gases within the pellets due to its large grain size. To demonstrate this, several experiments have been carried out by Halden reactor and Studsvik. These experiments are now being used to benchmark several fuel performance codes among them transuranus code. All this as part of a Coordinate Research Project (CRP) by IAEA named Testing and Simulation for Advanced Technology and Accident Tolerant Fuels (ATF-TS). This work is introducing a novel fission gas diffusivity model for doped fuel in transuranus code. It is observed the benefits of introducing this new model when comparing to the standard model already existing in transuranus. Nevertheless, more work needs to be carried out to fully understand all the phenomena involved in adding dopant in UO2 due to change of thermo mechanical properties

Nuclear Reactor Safeguarding with Neutrino Detection for MOX Loading Verification

The resurgence of interest in nuclear power around the world highlights the importance of effective methods to safeguard against nuclear proliferation. Many powerful safeguarding techniques have been developed and are currently employed, but new approaches are needed to address proliferation challenges from emerging advanced reactor designs and fuel cycles. Building on prior work that demonstrated monitoring of nuclear reactor operation using neutrino detectors, we develop and present a simple quantitative statistical test suitable for analysis of measured reactor neutrino data and demonstrate its efficacy in a semi-cooperative reactor monitoring scenario. In this approach, a moderate-sized neutrino detector is placed near the reactor site to help monitor possible MOX fuel diversion independent of inspection-based monitoring. We take advantage of differing time-dependent neutrino count rates during the operating cycle of a reactor core to monitor any deviations of measurements from expectations given a declared fuel composition. For a five-ton idealized detector placed 25m away from a hypothetical 3565 MWth reactor, the statistical test is capable of detecting the diversion of ~80kg plutonium at the 95% confidence level 90% of the time over a 540-day observation period.Comment: 35 pages, 24 figure

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part I: SCIANTIX

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as inter- related phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS

Dependence of transuranic content in spent fuel on fuel burnup

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007.Includes bibliographical references (p. 33).As the increasing demand for nuclear energy results in larger spent fuel volume, implementation of longer fuel cycles incorporating higher burnup are becoming common. Understanding the effect of higher burnup on the spent fuel composition and radioactive properties is essential to ensure that spent fuel receives proper cooling in storage before it is sent to a disposal site or proper treatment and reprocessing if its useful content is to be extracted prior to disposal. Using CASMO-4, a standard Westinghouse 4-loop pressurized water reactor model was created and simulated with a three batch fuel cycle. U-235 enrichment was adjusted to achieve fuel burnups of 30, 50, 70 and 100 MWD per kg of initial uranium. These burnups demanded reload enrichments of 3.15%, 4.63%, 6.26% and 9.01% U-235 w/o respectively. The resultant spent fuel transuranic isotopic compositions were then provided as input into ORIGEN to study the decay behavior of the spent fuel. It was found that when burnup increased from 30 MWD/kg to 100 MWD/kg, the activity more than doubled due to the decreased Pu-241 content and the increased Np-239 presence. More importantly, the activity per MWD significantly decreased despite absolute increases in unit mass. The net result is that the half-life of high burnup fuels is greatly increased in comparison to low burnup fuels for the first decade of life. Beginning from day 14 after shutdown and until 10 years later, the 100 MWD/kg fuel has a half-life of 129 days while the 30 MWD/kg spent fuel has a half life of 5 days. Previous work has suggested that different trends dominate decay behavior from years 10 to 100 years following discharge.by Drew A. Reese.S.B

Effects of fuel type on the safety characteristics of a sodium cooled fast reactor

A series of accident simulations were performed using INL's thermal hydraulics code RELAP5-3D to analyze steady-state and transient behavior of a sodium cooled fast reactor. The reactor chosen for this study was General Electric's S-PRISM, which is a 1,000 MWt pool-type sodium-cooled fast reactor, designed for either an Oxide or Metal fueled core. Once key core characteristics including power profiles, reactivity feedback coefficients and delayed neutron parameters were calculated, S-PRISM was redesigned for a Nitride fueled core to take advantage of the Nitride fuel's high thermal conductivity and melting temperature. Loss of flow, loss of heat sink, loss of power and inadvertent control rod withdrawal accidents were simulated for each core at beginning, middle and end of cycle to determine if one fuel type provides significant safety advantages over the others.Ph.D.Committee Chair: Seyed M. Ghiaasiaan; Committee Member: Bojan Petrovic; Committee Member: C. K. Wang; Committee Member: Hamid Garmestani; Committee Member: John Zino; Committee Member: Weston M. Stace