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Review of zirconium-zircaloy pyrophoricity

Description: Massive zirconium metal scrap can be handled, shipped, and stored with no evidence of combustion or pyrophoricity hazards. Mechanically produced fine scrap such as shavings, turnings, or powders can burn but are not pyrophoric unless the particle diameter is less than 54 ..mu..m. Powders with particle diameters less than 54 ..mu..m can be both pyrophoric and explosive. Pyrophoric powders should be collected and stored underwater or under inert gas cover to reduce the flammability hazard. Opening sealed containers of zirconium stored underwater should be attempted with caution since hydrogen may be present. The factors that influence the ignition temperature have been explored in depth and recommendations are included for the safe handling, shipping, and storage of pyrophoric or flammable zirconium. 29 refs., 5 figs., 6 tabs.
Date: November 1, 1984
Creator: Cooper, T.D.
Partner: UNT Libraries Government Documents Department

Test procedure for cation exchange chromatography

Description: The purpose of this test plan is to demonstrate the synthesis of inorganic antimonate ion exchangers and compare their performance against the standard organic cation exchangers. Of particular interest is the degradation rate of both inorganic and organic cation exchangers. This degradation rate will be tracked by determining the ion exchange capacity and thermal stability as a function of time, radiation dose, and chemical reaction.
Date: August 24, 1994
Creator: Cooper, T. D.
Partner: UNT Libraries Government Documents Department

Technical documentation to support the evaluation of handling of plutonium metal

Description: In 1997, a can containing a plutonium metal ingot was opened. The sides of the inner storage can had collapsed. As the inner can was opened, an apparent flame appeared to issue from the opening. Based on the reaction and possible pressurization of the glovebox, a positive Unreviewed Safety Question (USQ) screening was issued. This document contains some of the technical documents to resolve the screening.
Date: August 31, 1999
Creator: COOPER, T.D.
Partner: UNT Libraries Government Documents Department

The chemistry of tributyl phosphate at elevated temperatures in the Plutonium Finishing Plant Process Vessels

Description: Potentially violent chemical reactions of the tributyl phosphate solvent used by the Plutonium Finishing Plant at the Hanford Site were investigated. There is a small probability that a significant quantity of this solvent could be accidental transferred to heated process vessels and react there with nitric acid or plutonium nitrate also present in the solvent extraction process. The results of laboratory studies of the reactions show that exothermic oxidation of tributyl phosphate by either nitric acid or actinide nitrates is slow at temperatures expected in the heated vessels. Less than four percent of the tributyl phosphate will be oxidized in these vented vessels at temperatures between 125{degrees}C and 250{degrees}C because the oxidant will be lost from the vessels by vaporization or decomposition before the tributyl phosphate can be extensively oxidized. The net amounts of heat generated by oxidation with concentrated nitric acid and with thorium nitrate (a stand-in for plutonium nitrate) were determined to be about -150 and -220 joules per gram of tributyl phosphate initially present, respectively. This is not enough heat to cause violent reactions in the vessels. Pyrolysis of the tributyl phosphate occurred in these mixtures at temperatures of 110{degrees}C to 270{degrees}C and produced mainly 1-butene gas, water, and pyrophosphoric acid. Butene gas generation is slow at expected process vessel temperatures, but the rate is faster at higher temperatures. At 252{degrees}C the rate of butene gas generated was 0.33 g butene/min/g of tributyl phosphate present. The measured heat absorbed by the pyrolysis reaction was 228 J/g of tributyl phosphate initially present (or 14.5 kcal/mole of tributyl phosphate). Release of flammable butene gas into process areas where it could ignite appears to be the most serious safety consideration for the Plutonium Finishing Plant.
Date: June 1, 1994
Creator: Barney, G. S. & Cooper, T. D.
Partner: UNT Libraries Government Documents Department

Spent nuclear fuel project surface area estimates for N-Reactor fuel in the K East basin

Description: Spent N-reactor fuel will be moved from wet to dry storage at Hanford Washington. The majority ofthis fuel exists as intact fuel assemblies, however, small amounts ofscrap will be included. Varying amounts of uranium metal are exposed in these fuel assemblies, depending upon the amount of mechanical damage sustained by the zircaloy cladding. The total exposed uranium surface area in each storage pool is estimated through the release of radioisotopes to the storage pools. The exposed uranium surface area of individual fuel assemblies in the K-East basin were estimated through the results of a camera survey. The exposed uranium surface area of scrap is estimated from the known particle side range and an estimated log normal particle size distribution. This document uses the radioisotope release calculations, the estimated scrap surface area, and the carnera survey results to estimate the ``worst case`` amount of surface area that could exist in a given ``MCO`` container containing 4 levels of fuel assemblies and one scrap basket. The total exposed uranium metal surface area for this ``worst case`` was 120,000 cm{sup 2}.
Date: September 30, 1996
Creator: Cooper, T.D.
Partner: UNT Libraries Government Documents Department

Spent nuclear fuel project detonation phenomena of hydrogen/oxygen in spent fuel containers

Description: Movement of Spent N Reactor fuels from the Hanford K Basins near the Columbia River to Dry interim storage facility on the Hanford plateau will require repackaging the fuel in the basins into multi-canister overpacks (MCOs), drying of the fuel, transporting the contained fuel, hot conditioning, and finally interim storage. Each of these functions will be accomplished while the fuel is contained in the MCOs by several mechanisms. The principal source of hydrogenand oxygen within the MCOs is residual water from the vacuum drying and hot conditioning operations. This document assesses the detonation phenomena of hydrogen and oxygen in the spent fuel containers. Several process scenarios have been identified that could generate detonation pressures that exceed the nominal 10 atmosphere design limit ofthe MCOS. Only 42 grams of radiolized water are required to establish this condition.
Date: September 30, 1996
Creator: Cooper, T. D.
Partner: UNT Libraries Government Documents Department

Spent Nuclear Fuel project estimate of volatile fission products release from multi-canister overpacks

Description: Spent N-Reactor fuel will be moved from wet pool storage to dry storage at Hanford Washington. This fuel will be sequentially loaded into a Multiple Container Overpack (MCO), moved to the cold vacuum drying station, drained, cold vacuum dried, shipped to the Canister Storage Building (CSB), staged for up to 2 years,hot vacuum dried at 300 degrees C, hot conditioned at 150 degrees C, and finally, sealed and stored for up to 75 years in the CSB.During each proposed process step, the volatile radioactive fission products released to the atmosphere were estimated.Tritium is the only volatile fission product released insignificant amounts during each process step. For an accident scenario involving interior MCO temperature of 600 degrees C for up to 8 hours, it was estimated that many volatile fission products are released.
Date: August 1996
Creator: Cooper, T. D.
Partner: UNT Libraries Government Documents Department

SNFP detonation phenomena of hydrogen/oxygen in spent fuel containers

Description: Movement of spent nuclear fuels from the Hanford K Basins near the Columbia River to dry interim storage facility on the Hanford plateau will require repackaging the fuel in the basin into multi-canister overpacks (MCOs), drying of the fuel, transporting the contained fuel, hot conditioning, and finally interim storage. Each of these functions will be accomplished while the fuel is contained in the MCOs. Hydrogen and oxygen can be generated within the MCOs by several mechanisms. The principal source of hydrogen and oxygen within the MCOs is residual water from the vacuum drying and hot conditioning operations. This document assesses the detonation phenomena of hydrogen and oxygen in the spent fuel containers. Several process scenarios have been identified that could generate detonation pressures that exceed the nominal 10 atmosphere design limit of the MCOs. Only 42 grams of radiolized water are required to establish this condition.
Date: May 30, 1996
Creator: Cooper, T.D.
Partner: UNT Libraries Government Documents Department

Spent Nuclear Fuel Project (SNFP) gas generation from N-Fuel in multi-canister overpacks

Description: During the conversion from wet pool storage for spent nuclear fuel at Hanford, gases will be generated from both radiolysis and chemical reactions. The gas generation phenomenon needs to be understood as it applies to safety and design issues,specifically over pressurization of sealed storage containers,and detonation/deflagration of flammable gases. This study provides an initial basis to predict the implications of gas generation on the proposed functional processes for spent nuclear fuel conversion from wet to dry storage. These projections are based upon examination of the history of fuel manufacture at Hanford, irradiation in the reactors, corrosion during wet pool storage, available fuel characterization data and available information from literature. Gas generation via radiolysis and metal corrosion are addressed. The study examines gas generation, the boundary conditions for low medium and high levels of sludge in SNF storage/processing containers. The functional areas examined include: flooded and drained Multi-Canister Overpacks, cold vacuum drying, shipping and staging and long term storage.
Date: August 1, 1996
Creator: Cooper, T.D.
Partner: UNT Libraries Government Documents Department

Spent Nuclear Fuel (SNF) surface area estimates for N Reactor fuel in the K-East Basin

Description: Spent N-reactor fuel will be moved from wet to dry storage at Hanford Washington. The majority of this fuel exists as intact fuel assemblies, however, small amounts of scrap will be included. Varying amounts of uranium metal are exposed in these fuel assemblies, depending upon the amount of mechanical damage sustained by the zircaloy cladding. The total exposed uranium surface area in each storage pool is estimated through the release of radioisotopes to the storage pools. The exposed uranium surface area of individual fuel assemblies in the K-East basin were estimated through the results of a camera survey.This document uses the radioisotope release calculations and the camera survey results to estimate the `worst case` amount of surface area that could exist in a given `MCO` container containing 4 levels of fuel assemblies and one scrap basket. The total exposed uranium metal surface area for this `worst case` was 127,233 cm{sup 2}.
Date: August 1, 1996
Creator: Cooper, T.D.
Partner: UNT Libraries Government Documents Department

Redox Bias in Loss on Ignition Moisture Measurement for Relatively Pure Plutonium-Bearing Oxide Materials

Description: This paper evaluates potential analytical bias in application of the Loss on Ignition (LOI) technique for moisture measurement to relatively pure (plutonium assay of 80 wt.% or higher) oxides containing uranium that have been stabilized according to stabilization and storage standard DOE-STD-3013-2000 (STD- 3013). An immediate application is to Rocky Flats (RF) materials derived from high-grade metal hydriding separations subsequently treated by multiple calcination cycles. Specifically evaluated are weight changes due to oxidation/reduction of multivalent impurity oxides that could mask true moisture equivalent content measurement. Process knowledge and characterization of materials representing complex-wide materials to be stabilized and packaged according to STD-3013, and particularly for the immediate RF target stream, indicate that oxides of uranium, iron and gallium are the only potential multivalent constituents expected to be present above 0.5 wt.%. The evaluation show s that of these constituents, with few exceptions, only uranium oxides can be present at a sufficient level to produce weight gain biases significant with respect to the LOI stability test. In general, these formerly high-value, high-actinide content materials are reliably identifiable by process knowledge and measurement. Significant bias also requires that UO2 components remain largely unoxidized after calcination and are largely converted to U3O8 during LOI testing at only slightly higher temperatures. Based on well-established literature, it is judged unlikely that this set of conditions will be realized in practice. We conclude that it is very likely that LOI weight gain bias will be small for the immediate target RF oxide materials containing greater than 80 wt.% plutonium plus a much smaller uranium content. Recommended tests are in progress to confirm these expectations and to provide a more authoritative basis for bounding LOI oxidation/reduction biases. LOI bias evaluation is more difficult for lower purity materials and for fuel-type uranium-plutonium oxides. However, even in these cases ...
Date: February 26, 2002
Creator: Eller, P. G.; Stakebake, J. L. & Cooper, T. D.
Partner: UNT Libraries Government Documents Department

Redox bias in loss of ignition moisture measurement for relatively pure plutonium-bearing oxide materials.

Description: This paper evaluates potential analytical bias in application of the Loss on Ignition (LOI) technique for moisture measurement to relatively pure (plutonium assay of 80 wt.% or higher) oxides containing uranium that have been stabilized according to stabilization and storage standard DOE-STD-3013-2000 (STD-3013). An immediate application is to Rocky Flats (RF) materials derived from highgrade metal hydriding separations subsequently treated by multiple calcination cycles. Specifically evaluated are weight changes due to oxidatiodreduction of multivalent impurity oxides that could mask true moisture equivalent content measurement. Process knowledge and characterization of materials representing complex-wide materials to be stabilized and packaged according to STD-3013, and particularly for the immediate RF target stream, indicate that oxides of uranium, iron and gallium are the only potential multivalent constituents expected to be present above 0.5 wt.%. The evaluation shows that of these constituents, with few exceptions, only uranium oxides can be present at a sufficient level to produce weight gain biases significant with respect to the LO1 stability test. In general, these formerly high-value, high-actinide content materials are reliably identifiable by process knowledge and measurement. Si&icant bias also requires that UO1 components remain largely unoxidized after calcination and are largely converted to U30s clsning LO1 testing at only slightly higher temperatures. Based on wellestablished literature, it is judged unlikely that this set of conditions will be realized in practice. We conclude that it is very likely that LO1 weight gain bias will be small for the immediate target RF oxide materials containing greater than 80 wt.% plutonium plus a much smaller uranium content. Recommended tests are in progress to confum these expectations and to provide a more authoritative basis for bounding LO1 oxidatiodreduction biases. LO1 bias evaluation is more difficult for lower purity materials and for fuel-type uranium-plutonium oxides. However, even in these cases testing may ...
Date: January 1, 2001
Creator: Eller, P. G. (Phillip Gary); Stakebake, J. L. (Jerry L.) & Cooper, T. D. (Thruman D.)
Partner: UNT Libraries Government Documents Department

Spent nuclear fuel project recommended reaction rate constants for corrosion of N-Reactor fuel

Description: The US Department of Energy (DOE) established the Spent Nuclear Fuel Project (SNF Project) to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored in the Hanford Site`s K Basins. The SNF Project has been tasked by the DOE with moving the spent N-Reactor fuel from wet storage to contained dry storage in order to reduce operating costs and environmental hazards. The chemical reactivity of the fuel must be understood at each process step and during long-term dry storage. Normally, the first step would be to measure the N-fuel reactivity before attempting thermal-hydraulic transfer calculations; however, because of the accelerated project schedule, the initial modeling was performed using literature values for uranium reactivity. These literature values were typically found for unirradiated, uncorroded metal. It was fully recognized from the beginning that irradiation and corrosion effects could cause N-fuel to exhibit quite different reactivities than those commonly found in the literature. Even for unirradiated, uncorroded uranium metal, many independent variables affect uranium metal reactivity resulting in a wide scatter of data. Despite this wide reactivity range, it is necessary to choose a defensible model and estimate the reactivity range of the N-fuel until actual reactivity can be established by characterization activities. McGillivray, Ritchie, and Condon developed data and/or models that apply for certain samples over limited temperature ranges and/or reaction conditions (McGillivray 1994, Ritchie 1981 and 1986, and Condon 1983). These models are based upon small data sets and have relatively large correlation coefficients.
Date: June 15, 1998
Creator: Cooper, T.D.
Partner: UNT Libraries Government Documents Department

Technology status in support of refined technical baseline for the Spent Nuclear Fuel project. Revision 1

Description: The Spent Nuclear Fuel Project (SNFP) has undertaken technology acquisition activities focused on supporting the technical basis for the removal of the N Reactor fuel from the K Basins to an interim storage facility. The purpose of these technology acquisition activities has been to identify technology issues impacting design or safety approval, to establish the strategy for obtaining the necessary information through either existing project activities, or the assignment of new work. A set of specific path options has been identified for each major action proposed for placing the N Reactor fuel into a ``stabilized`` form for interim storage as part of this refined technical basis. This report summarizes the status of technology information acquisition as it relates to key decisions impacting the selection of specific path options. The following specific categories were chosen to characterize and partition the technology information status: hydride issues and ignition, corrosion, hydrogen generation, drying and conditioning, thermal performance, criticality and materials accountability, canister/fuel particulate behavior, and MCO integrity. This report represents a preliminary assessment of the technology information supporting the SNFP. As our understanding of the N Reactor fuel performance develops the technology information supporting the SNFP will be updated and documented in later revisions to this report. Revision 1 represents the incorporation of peer review comments into the original document. The substantive evolution in our understanding of the technical status for the SNFP (except section 3) since July 1995 have not been incorporated into this revision.
Date: October 20, 1995
Creator: Puigh, R.J.; Toffer, H.; Heard, F.J.; Irvin, J.J. & Cooper, T.D.
Partner: UNT Libraries Government Documents Department