Data quality objectives for the reaction kinetics studies of K West fuel samples

The Data Quality Objectives (DQOs) were established for the reaction kinetics studies of the first group of fuel samples shipped from the K West Basin to the Hanford 327 Building hot cells for examinations. A Thermo-Gravimetric Analysis (TGA) system was selected for these measurements and associated hydrogen release and ignition temperature testing. These examinations are an extension of the conditioning testing of sibling samples described in WHC-SD-SNF-DQO-004, Data Quality Objectives for the Initial Fuel Conditioning Examinations

Analysis of sludge from Hanford K East Basin canisters

Sludge samples from the canisters in the Hanford K East Basin fuel storage pool have been retrieved and analyzed. Both chemical and physical properties have been determined. The results are to be used to determine the disposition of the bulk of the sludge and to assess the impact of residual sludge on dry storage of the associated intact metallic uranium fuel elements. This report is a summary and review of the data provided by various laboratories. Although raw chemistry data were originally reported on various bases (compositions for as-settled, centrifuged, or dry sludge) this report places all of the data on a common comparable basis. Data were evaluated for internal consistency and consistency with respect to the governing sample analysis plan. Conclusions applicable to sludge disposition and spent fuel storage are drawn where possible

Thermal Decomposition of Radiation-Damaged Polystyrene

The radiation-damaged polystyrene material (''polycube'') used in this study was synthesized by mixing a high-density polystyrene (''Dylene Fines No. 100'') with plutonium and uranium oxides. The polycubes were used on the Hanford Site in the 1960s for criticality studies to determine the hydrogen-to-fissile atom ratios for neutron moderation during processing of spent nuclear fuel. Upon completion of the studies, two methods were developed to reclaim the transuranic (TRU) oxides from the polymer matrix: (1) burning the polycubes in air at 873 K; and (2) heating the polycubes in the absence of oxygen and scrubbing the released monomer and other volatile organics using carbon tetrachloride. Neither of these methods was satisfactory in separating the TRU oxides from the polystyrene. Consequently, the remaining polycubes were sent to the Hanford Plutonium Finishing Plant (PFP) for storage. Over time, the high dose of alpha and gamma radiation has resulted in a polystyrene matrix that is highly cross-linked and hydrogen deficient and a stabilization process is being developed in support of Defense Nuclear Facility Safety Board Recommendation 94-1. Baseline processes involve thermal treatment to pyrolyze the polycubes in a furnace to decompose the polystyrene and separate out the TRU oxides. Thermal decomposition products from this degraded polystyrene matrix were characterized by Pacific Northwest National Laboratory to provide information for determining the environmental impact of the process and for optimizing the process parameters. A gas chromatography/mass spectrometry (GC/MS) system coupled to a horizontal tube furnace was used for the characterization studies. The decomposition studies were performed both in air and helium atmospheres at 773 K, the planned processing temperature. The volatile and semi-volatile organic products identified for the radiation-damaged polystyrene were different from those observed for virgin polystyrene. The differences were in the n umber of organic species generated and their concentrations

Data quality objectives for the initial K West fuel examinations

The Data Quality Objectives (DQOs) were established for the examinations of the first group of fuel samples shipped from the K West Basin to the Hanford 327 Building hot cells for examinations to support the Path Forward recommended to solve the safety and environmental concerns associated with the deteriorating fuel in the K Basins. A separate DQO will be prepared for each future shipment of samples to the hot cells. The material stored in the K West Basin must ultimately be removed from the basin and placed in a stable storage configuration until it can be disposed of in a repository. The condition of the fuel in the closed canisters is a major uncertainty for any of the proposed actions. The major question to answer is what are the conditions of the materials in the closed canisters? The data to be gathered during the canister opening, handling, transport, associated hot cell handling, and examinations will also support decisions related to the Path Forward primarily in areas of transportation and the Multi-Canister Overpack (MCO) design

Drying behavior of K-East canister sludge

A series of tests were conducted by Pacific Northwest National Laboratory to evaluate the drying behavior of sludge taken from the Hanford K-East Basin storage canisters. Some of the components of K-Basin sludge, such as oxides of uranium and its hydrates, could be associated with the spent nuclear fuel that will ultimately be loaded into Multi-Canister Overpacks (MCOs) and transferred to interim dry storage on the Hanford Site. The materials sealed in the MCOs must be compatible with the storage facility safety basis and the design accident analyses. Understanding the drying behavior of hydrates that may be formed by the reaction of uranium oxides (corrosion products) and water will help ensure these criteria are addressed. Drying measurements of sludge samples collected from K-East Basin canisters showed the water content (physically plus chemically bound) to range between 5 wt% and 75 wt%. Uranium oxide hydrates, the main source of gaseous products that can pressurize the MCOs during storage, constituted about 3 wt% to 15 wt% of the total water content of the initial weight. Most of the physically bound water was assumed to be released from the samples at ambient temperature when the system was pumped down to vacuum conditions of about 40 mTorr. The period for release of most free water in the K-East canister sludge was about 24 hours

Oxidation rate of K-Basin spent nuclear fuel in moist air

Experiments have been conducted by Pacific Northwest National Laboratory to determine the oxidation rate of damaged/corroded N-Reactor fuel material in moist air. Five SNF pieces (with regular geometrical shapes) sectioned from a damaged element stored in the K-West Basin were oxidized in flowing air containing moisture. The SNF oxidation behavior in moist air at a temperature of 198 C can best be fitted by parabolic oxidation kinetics. A linear rate equation gave the best fit to the oxidation data at 250 C and above. The results within the temperature range studied, therefore, show a transition from parabolic oxidation kinetics to linear oxidation kinetics. The transition temperature is somewhere between 198 C and 250 C. The tests at approximately 300 C gave results that were very different from the other tests at temperatures of 198 C, 250 C, and 349 C. The SNF sample weight change at this temperature showed erratic behavior. Visual examination indicated the sample fragmented into small pieces and powder as a result of rapid oxidation and hydration. Additional tests at temperatures close to 300 C (i.e., 300 {+-} 10 C) are recommended in order to fully understand the oxidation behavior of the damaged/corroded SNF samples in moist air at about 300 C