Uranium Pyrophoricity Phenomena and Prediction (FAI/00-39) (original) (raw)
Related papers
2013
In 2012, a system was deployed to remove, transport, and interim store chemically reactive and highly radioactive sludge material from the Hanford Site's 105-K West Fuel Storage Basin that will be managed as spent/used nuclear fuel. The Knockout Pot (KOP) sludge in the 105-K West Basin was a legacy issue resulting from the spent nuclear fuel (SNF) washing process applied to 2200 metric tons of highly degraded fuel elements following long-term underwater storage. The washing process removed uranium metal and other non-uranium constituents that could pass through a screen with 0.25-inch openings; larger pieces are, by definition, SNF or fuel scrap. When originally retrieved, KOP sludge contained pieces of degraded uranium fuel ranging from 600 microns (μm) to 6350 μm mixed with inert material such as aluminum hydroxide, aluminum wire, and graphite in the same size range. In 2011, a system was developed, tested, successfully deployed and operated to pre-treat KOP sludge as part of ...
Surface science investigations on spent nuclear fuel model systems
Wissenschaftliche Berichte FZKA, 2009
Fig. 30 Raman spectra of a UO 2(+x) thin film on gold substrate (Au-3) before (A) and after (B) the solubility experiment in the phosphate electrolyte, pH 7.6. Fig. 31 Raman spectra from precipitate and reference spectra of hematite.
Corrosion of Spent Nuclear Fuel: The Long-Term Assessment
2004
Spent nuclear fuel (SNF), essentially UO 2 , accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO 2 in SNF is not stable under oxidizing conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term. This research program has been a broadly based effort to understand the long-term behavior of SNF and its alteration products in a geologic repository. Fortunately, our experiments and field studies have established that natural uraninite, UO 2+x , and its alteration products are excellent "natural analogues" for the study of the corrosion of UO 2 in SNF. This research program has addressed the following major issues: 1. What are the long-term corrosion products of natural UO 2+x , uraninite, under oxidizing and reducing conditions?
Understanding the Environment on the Surface of Spent Nuclear Fuel Interim Storage Containers
2013
A primary concern with dry storage of spent nuclear fuel is chloride-induced stress corrosion cracking, caused by deliquescence of salts deposited on the stainless steel canisters. However, limited access through the ventilated overpacks and high surface radiation fields impede direct examination of cask surfaces for CISCC, or sampling of surface deposits. Predictive models for CISCC must be able to predict the occurrence of a corrosive chemical environment (a chloride-rich brine formed by dust deliquescence) at specific locations (e.g. weld zones) on the canister surface. The presence of a deliquescent brine is controlled by the relative humidity (RH), which is a function of absolute humidity and cask surface temperature. This requires a thermal model that includes the canister and overpack design, canister-specific waste heat load, and passive cooling by ventilation. Brine compositions vary with initially-deposited salt assemblage, reactions with atmospheric gases, temperature, an...
Calculation of used nuclear fuel dissolution rates under anticipated Canadian waste vault conditions
Journal of Nuclear Materials, 1997
Dissolution rates of UO fuel are determined as a function of alpha and gamma dose rates. These room-temperature rates 2 are used to calculate the dissolution rates of used fuel at 1008C. Also, the alpha, beta and gamma dose rates in water in contact with the reference used fuel are calculated as a function of cooling time. These results are used to calculate used CANDU fuel dissolution rates as a function of time since emplacement in a defective copper container for the Canadian Nuclear Fuel Waste Management Program. It is shown that beta radiolysis of water is the main cause of oxidation of used CANDU fuel in a failed container and that the use of a corrosion model is required for ; 1000 a of emplacement in the waste vault. The results obtained here can be adopted to calculate used nuclear fuel dissolution rates for other waste management programs. q 1997 Elsevier Science B.V. h 4 93 7 w x 17 . However, any model of fuel dissolution within a 0022-3115r97r$17.00 q 1997 Elsevier Science B.V. All rights reserved. Ž . PII S 0 0 2 2 -3 1 1 5 9 7 0 0 2 7 2 -9
MRS Proceedings, 1997
ABSTRACTCalculation of used nuclear fuel dissolution rates in a geological disposal vault requires a knowledge of the redox conditions in the vault. For redox conditions less oxidizing than those causing UO2 oxidation to the U3O7 stage, a thermodynamically-based model is appropriate. For more oxidizing redox conditions a kinetic or an electrochemical model is needed to calculate these rates. The redox conditions in a disposal vault will be affected by the radiolysis of groundwater by the ionizing radiation associated with the fuel. Therefore, we have calculated the alpha-, beta- and gamma-dose rates in water in contact with the reference used fuel in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) as a function of cooling time. Also, we have determined dissolution rates of UO2 fuel as a function of alpha and gamma dose rates from our electrochemical measurements. These room-temperature rates are used to calculate the dissolution rates of used fuel at 100°C, the highest t...
Oxidation/Reduction Status of Water Pooled in a Penetrated Nuclear Waste Container
Nuclear power use is expected to expand in the future and result in hundreds of thousands of metric tons of spent nuclear fuel (SNF). One of the main concerns of nuclear energy use is SNF disposal. Storage in geological repositories is a reasonable solution for the accumulation of SNF. One of the key factors that determine the performance of the proposed geological repository at Yucca Mountain (YM), NV is the release of radionuclides from the engineered barrier system (EBS) by water transport. Over time, EBS, including nuclear waste containers, is expected to fail gradually due to general and localized corrosion. Physical and chemical disturbances in the environment of the repository will lead to different corrosion rates at different locations of the waste packages. Considering the inherent uncertainty of the failure sequence of a waste package, two main failure scenarios are expected: flow through model (penetrations are on the top and bottom of the waste package causing water to flow through it) and bathtub model (penetrations are on the top with the waste package filling with water). In this paper, we consider a bathtub category failed waste container and shed some light on chemical and physical processes that take place in the pooled water and their effects on radionuclide release. We consider two possibilities: temperature stratification of the pooled water versus mixing. Our calculations show that there will be temperature stratification of the pooled water in the lower half of the waste package, and mixing in the upper half. The effect of these situations on oxygen availability and consequently spent fuel alteration and waste container components corrosion is also considered.
Drying Results of K-Basin Fuel Element 6513U (Run 8)
1999
An N-Reactor outer fuel element which had been stored under water in the Hanford 100 Area K-West basin was subjected to a combination of low-and high-temperature vacuum drying treatments. These studies are part of a series of tests being conducted by Pacific Northwest National Laboratory on the drying behavior of spent nuclear fuel elements removed from both the K-West and K-East Basins. This report describes the fuel element, the test methodology, and the testing results. v Quality Assurance This work was conducted under the Quality Assurance Program, Pacific Northwest National Laboratory (PNNL) SNF-70-001, SNF Quality Assurance Program, as implemented by the PNNL SNF Characterization Project Operations Manual. This QA program has been evaluated and determined to effectively implement the requirements of DOE/RW-0333P, Quality Assurance Requirements and Description (QARD). Compliance with the QARD is mandatory for projects that generate data used to support the development of a permanent High-Level Nuclear Waste repository. Further, the U.S. Department of Energy has determined that the testing activities which generated the results documented in this report shall comply with the QARD.