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Summary of results from the Series 2 and Series 3 NNWSI [Nevada Nuclear Waste Storage Investigations] bare fuel dissolution tests

Description: The Nevada Nuclear Waste Storage Investigations (NNWSI) Project is studying dissolution and radionuclide release behavior of spent nuclear fuel in Nevada Test Site groundwater. Specimens were tested for multiple cycles in J-13 well water. The Series 2 tests were run in unsealed silica vessels under ambient hot cell air (25{sup 0}C) for five cycles for a total of 34 months. The Series 3 tests were run in sealed stainless steel vessels at 25{sup 0}C and 85{sup 0}C for three cycles for a total of 15 months. Selected summary results from Series 2 and Series 3 tests with bare fuel specimens are reported. Uranium concentrations in later test cycles ranged from 1 to 2 {mu}g/ml in the Series 2 Tests versus about 0.1 to 0.4 {mu}g/ml in Series 3 with the lowest concentrations occurring in the 85{sup 0}C tests. Preferential release of fission products Cs, I, Sr and Tc, and activation product C-14, was indicated relative to the actinides. Tc-99 and Cs-137 activities measured in solution after Cycle 1 increased linearly with time, with the rate of increase greater at 85{sup 0}C than at 25{sup 0}C. 8 refs., 8 figs., 3 tabs.
Date: November 1, 1987
Creator: Wilson, C. N.
Partner: UNT Libraries Government Documents Department

Results from NNWSI [Nevada Nuclear Waste Storage Investigations] Series 2 bare fuel dissolution tests

Description: The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Two bare spent fuel specimens plus the empty cladding hulls were tested in NNWSI J-13 well water in unsealed fused silica vessels under ambient hot cell air conditions (25{degree}C) in the currently reported tests. One of the specimens was prepared from a rod irradiated in the H. B. Robinson Unit 2 reactor and the other from a rod irradiated in the Turkey Point Unit 3 reactor. Results indicate that most radionuclides of interest fall into three groups for release modeling. The first group principally includes the actinides (U, Np, Pu, Am, and Cm), all of which reached solubility-limited concentrations that were orders of magnitude below those necessary to meet the NRC 10 CFR 60.113 release limits for any realistic water flux predicted for the Yucca Mountain repository site. The second group is nuclides of soluble elements such as Cs, Tc, and I, for which release rates do not appear to be solubility-limited and may depend on the dissolution rate of fuel. In later test cycles, {sup 137}Cs, {sup 90}Sr, {sup 99}Tc, and {sup 129}I were continuously released at rates between about 5 {times} 10{sup {minus}5} and 1 {times} 10{sup {minus}4} of inventory per year. The third group is radionuclides that may be transported in the vapor phase, of which {sup 14}C is of primary concern. Detailed test results are presented and discussed. 17 refs., 15 figs., 21 tabs.
Date: September 1, 1990
Creator: Wilson, C.N.
Partner: UNT Libraries Government Documents Department

Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests

Description: The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85{degree}C and 25{degree}C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs.
Date: June 1, 1990
Creator: Wilson, C.N.
Partner: UNT Libraries Government Documents Department

Fabrication of lithium ceramics by hot pressing

Description: Controlled density LiA10/sub 2/, Li/sub 2/Zr0/sub 3/, Li/sub 4/SiO/sub 4/ and Li/sub 2/O pellets were fabricated by hot pressing for irradiation testing as candidate tritium breeding materials. Pellet specifications, characterization data, and procedures for hot pressing, pellet grinding and halide removal are discussed.
Date: March 1, 1982
Creator: Wilson, C.N.
Partner: UNT Libraries Government Documents Department

Results from long-term dissolution tests using oxidized spent fuel

Description: Two semi-static dissolution tests using oxidized PWR spent fuel specimens are being conducted under ambient hot cell conditions in Nevada Test Site J-13 well water and unsealed fused silica vessels. The test specimens were oxidized at 250{degree}C in air to bulk oxygen-to-metal (O/M) values of 2.21 and 2.33. Following an initial 191-day test cycle, the specimens were restarted in fresh J-13 water for a second long-term test cycle. Results through the first 40 months of Cycle 2 are compared with results from similar tests at 25{degree}C and 85{degree}C using unoxidized spent fuel specimens. Increased concentrations of U, Am, Cm and NP were measured in 0.4-{mu}m filtered samples from the oxidized fuel tests compared to the unoxidized fuel tested at 25{degree}C; Pu concentrations were not affected by the fuel oxidation state. Most of the Am and Cm, and a portion of the Pu, measured in 0.4-{mu}m filtered samples was removed by 2-nm filtration. Fission product release results were normalized to specimen inventories and reported as fractional release. No attempt was made to normalize the data to surface area. Initial {sup 99}Tc release was greatly increased, and prolonged increases in the fractional release rates of {sup 99}Tc and {sup 129}I occurred as a result of fuel oxidation. Fractional release rates for {sup 137}Cs and {sup 90}Sr from oxidized fuel eventually decreased to levels similar to those observed with unoxidized fuel after equivalent testing times, suggesting that matrix dissolution rates normalized to fuel mass were not increased as a result of oxidation. 6 refs., 3 figs., 2 tabs.
Date: November 1, 1990
Creator: Wilson, C.N.
Partner: UNT Libraries Government Documents Department

Derivation of a waste package source term for NNWSI from the results of laboratory experiments

Description: Results are performed for the dissolution of Turkey Point pressurized water reactor (PWR) spent fuel in J-13 well water at ambient hot cell temperatures. These results are compared with those previously obtained on Turkey Point fuel in deionized water, on H.B. Robinson PWR fuel in J-13 water, and by other workers using various fuels in dilute bicarbonate groundwaters. A model is presented that represents the conditions under which maximum dissolution of spent fuel could occur in a repository sited at Yucca Mountain, Nevada. Using an experimentally determined upper limit of 5 mg/l for uranium solubility in J-13 water, a fractional release rate of 6.4 x 10{sup -8} per year is obtained by assuming that all water entering the repository carries away the maximum amount of uranium. 14 refs., 3 figs., 3 tabs.
Date: September 1, 1985
Creator: Oversby, V. M. & Wilson, C N.
Partner: UNT Libraries Government Documents Department

Experimental study of the dissolution of spent fuel at 85{sup 0} in natural ground water

Description: Semi-static dissolution tests using pressurized water reactor spent fuel rod segments and NNWSI reference J-13 well water in sealed stainless steel vessels at 85{sup 0}C are being conducted in support of the Waste Package Task of the NNWSI Project. Test specimens include: bare fuel plus the empty cladding hulls, fuel rod segments with artificially induced cladding defects and water-tight end caps, and undefected fuel rod segments with water-tight end caps. The test conditions approximate those expected in the proposed NNWSI Project repository when the waste package has cooled sufficiently to allow water to enter a breached container and contact the fuel rods, some of which may exhibit various degrees of cladding failure. Periodic solution samples (unfiltered and filtered) were analyzed for most radionuclides for which cumulative release limits are listed by the US Environmental Protection Agency. Results from the first six-month cycle of the 85{sup 0}C tests are presented and are compared with results from the first cycle of a previous test series run at 25{sup 0}C in fused silica test vessels.
Date: December 1, 1986
Creator: Wilson, C. N. & Shaw, H. F.
Partner: UNT Libraries Government Documents Department

Spent fuel cladding containment credit tests

Description: Preliminary tests are being conducted to evaluate the effectiveness of defected cladding as a barrier to radionuclide release from spent fuel rods stored in a geological repository. The tests are being conducted at the Hanford Engineering Development Laboratory Waste Package Task of the Nevada Nuclear Waste Storage Investigations (NNWSI) tuff repository project. In these tests, spent PWR fuel rod specimens with various artificially induced cladding defects are leach tested in a test matrix which also includes both bare fuel specimens (unclad) and undefected spent fuel rod specimens. Artificial cladding defects are made by laser drilling and sawing to give defect areas in the 10{sup 4} to 10{sup 6} {mu}m{sup 2} range. Periodic samples are taken of the leach solution and fused quartz rods contained in the test vessels. Results for the first 180 days of testing are presented. 5 references, 3 figures, 2 tables.
Date: February 1, 1984
Creator: Wilson, C.N. & Oversby, V.M.
Partner: UNT Libraries Government Documents Department

Spent fuel dissolution studies FY 1991 to 1994

Description: Dissolution and transport as a result of groundwater flow are generally accepted as the primary mechanisms by which radionuclides from spent fuel placed in a geologic repository could be released to the biosphere. To help provide a source term for performance assessment calculations, dissolution studies on spent fuel and unirradiated uranium oxides have been conducted over the past few years at Pacific Northwest National Laboratory (PNNL) in support of the Yucca Mountain Site Characterization Project. This report describes work for fiscal years 1991 through 1994. The objectives of these studies and the associated conclusions, which were based on the limited number of tests conducted so far, are described in the following subsections.
Date: December 1, 1995
Creator: Gray, W.J. & Wilson, C.N.
Partner: UNT Libraries Government Documents Department

Melter system technology testing for Hanford Site low-level tankwaste vitrification

Description: Following revisions to the Tri-Party Agreement for Hanford Site cleanup, which specified vitrification for Complete melter feasibility and system operability immobilization of the low-level waste (LLW) tests, select reference melter(s), and establish reference derived from retrieval and pretreatment of the radioactive LLW glass formulation that meets complete systems defense wastes stored in 177 underground tanks, commercial requirements (June 1996). Available melter technologies were tested during 1994 to 1995 as part of a multiphase program to select reference Submit conceptual design and initiate definitive design technologies for the new LLW vitrification mission.
Date: May 3, 1996
Creator: Wilson, C. N.
Partner: UNT Libraries Government Documents Department

LWR spent fuel characteristics relevant to performance as a wasteform in a potential tuff repository

Description: A testing program has been initiated to determine the probable condition of spent fuel during the post-containment period under NNWSI site specific conditions, and to determine relevant radionuclide release rates for spent fuel. The current testing program is focused on three subject areas: (1) spent fuel leaching/dissolution behavior, (2) spent fuel oxidation, and (3) cladding corrosion. Results are presented. 3 refs.
Date: June 1, 1985
Creator: Wilson, C.N.; Einziger, R.E.; Woodley, R.E. & Oversby, V.M.
Partner: UNT Libraries Government Documents Department

Vectra GSI, Inc. low-level waste melter testing Phase 1 test report

Description: A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Vectra GSI, Inc. was one of seven vendors selected for Phase 1 of the melter demonstration tests using simulated LLW that were completed during fiscal year 1995. The attached report prepared by Vectra GSI, Inc. describes results of melter testing using slurry feed and dried feeds. Results of feed drying and prereaction tests using a fluid bed calciner and rotary dryer also are described.
Date: February 21, 1996
Creator: Stegen, G.E. & Wilson, C.N.
Partner: UNT Libraries Government Documents Department

Evaluation of melter technologies for vitrification of Hanford site low-level tank waste - phase 1 testing summary report

Description: Following negotiation of the fourth amendment to the Tri- Party Agreement for Hanford Site cleanup, commercially available melter technologies were tested during 1994 and 1995 for vitrification of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of the radioactive defense wastes stored in 177 underground tanks. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high-sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes also were tested. The technologies and Phase 1 testing results were evaluated and a preliminary technology down-selection completed. This report describes the Phase 1 LLW melter vendor testing and the tested technologies, and summarizes the testing results and the preliminary technology recommendations.
Date: June 27, 1996
Creator: Wilson, C. N.
Partner: UNT Libraries Government Documents Department

Vitrification technology for Hanford Site tank waste

Description: The US Department of Energy`s (DOE) Hanford Site has an inventory of 217,000 m{sup 3} of nuclear waste stored in 177 underground tanks. The DOE, the US Environmental Protection Agency, and the Washington State Department of Ecology have agreed that most of the Hanford Site tank waste will be immobilized by vitrification before final disposal. This will be accomplished by separating the tank waste into high- and low-level fractions. Capabilities for high-capacity vitrification are being assessed and developed for each waste fraction. This paper provides an overview of the program for selecting preferred high-level waste melter and feed processing technologies for use in Hanford Site tank waste processing.
Date: April 1, 1995
Creator: Weber, E.T.; Calmus, R.B. & Wilson, C.N.
Partner: UNT Libraries Government Documents Department

Low-Level waste phase 1 melter testing off gas and mass balance evaluation

Description: Commercially available melter technologies were tested during 1994-95 as part of a multiphase program to test candidate technologies for vitrification of the low-level waste (LLW) stream to be derived from retrieval and pretreatment of Hanford Site tank wastes. Seven vendors were selected for Phase 1 testing to demonstrate vitrification of a high sodium content liquid LLW simulant. The tested melter technologies included four Joule-heated melters, a carbon electrode melter, a combustion melter, and a plasma melter. Various dry and slurry melter feed preparation processes were also tested. Various feed material samples, product glass samples, and process offgas streams were characterized to provide data for evaluation of process decontamination factors and material mass balances for each vitrification technology. This report describes the melter mass balance evaluations and results for six of the Phase 1 LLW melter vendor demonstration tests.
Date: June 28, 1996
Creator: Wilson, C.N.
Partner: UNT Libraries Government Documents Department

Melter technology evaluation for vitrification of Hanford Site low-level waste

Description: The current plan at the Hanford Site, in accordance with the Tri-Party Agreement among Washington State, the US Environmental Protection Agency, and the US Department of Energy, is to convert the low-level tank waste fraction into a silicate glass. The low-level waste will be composed primarily of sodium nitrate and nitrite salts concentrated in a highly alkaline aqueous solution. The capability to process up to 200 metric tons/day off glass will be established to produce an estimated 210,000 m{sup 3} for onsite disposal. A program to test and evaluate high-capacity melter technologies is in progress. Testing performed by seven different industrial sources using Joule heating, combustion, plasma, and carbon arc melters is described.
Date: April 1995
Creator: Wilson, C. N.; Burgard, K. C.; Weber, E. T. & Brown, N. R.
Partner: UNT Libraries Government Documents Department

Titanium oxide cesium getters for low O/M FBR fuel pins

Description: Fission product cesium may contribute to cladding strain in low oxygen-to-metal ratio (O/M) FBR fuel pins through localized reaction with fuel or UO/sub 2/ blanket pellets. Titanium oxide pellets were laboratory irradiation tested as candidate cesium getters for FBR fuel pins. Results indicate satisfactory performance.
Date: August 15, 1979
Creator: Wilson, C. N.; Gibby, R. L. & Weber, E. T.
Partner: UNT Libraries Government Documents Department

Transportable vitrification system demonstration on mixed waste. Revision 1

Description: The Transportable Vitrification System (TVS) is a large scale, fully integrated, vitrification system for the treatment of low-level and mixed wastes in the form of sludges, soils, incinerator ash, and many other waste streams. It was demonstrated on surrogate waste at Clemson University and at the Oak Ridge Reservation (ORR) prior to treating actual mixed waste. Treatment of a combination of dried B and C Pond sludge and CNF sludge was successfully demonstrated at ORR in 1997. The demonstration produced 7,616 kg of glass from 7,328 kg of mixed wastes with a 60% reduction in volume. Glass formulations for the wastes treated were developed using a combination of laboratory crucible studies with the actual wastes and small melter studies at Clemson with both surrogate and actual wastes. Initial characterization of the B and C Pond sludge had not shown the presence of carbon or fluoride, which required a modified glass formulation be developed to maintain proper glass redox and viscosity. The CNF sludge challenges the glass formulations due to high levels of phosphate and iron. The demonstration was delayed several times by permitting problems, a glass leak, and electrical problems. The demonstration showed that the two wastes could be successfully vitrified, although the design glass production rate was not achieved. The glass produced met the Universal Treatment Standards and the emissions from the TVS were well within the allowable permit limits.
Date: April 22, 1998
Creator: Zamecnik, J.R.; Whitehouse, J.C.; Wilson, C.N. & Van Ryn, F.R.
Partner: UNT Libraries Government Documents Department

Transportable Vitrification System Demonstration on Mixed Waste

Description: This paper describes preliminary results from the first demonstration of the Transportable Vitrification System (TVS) on actual mixed waste. The TVS is a fully integrated, transportable system for the treatment of mixed and low-level radioactive wastes. The demonstration was conducted at Oak Ridge`s East Tennessee Technology Park (ETTP), formerly known as the K-25 site. The purpose of the demonstration was to show that mixed wastes could be vitrified safely on a `field` scale using joule-heated melter technology and obtain information on system performance, waste form durability, air emissions, and costs.
Date: January 1, 1998
Creator: Zamecnik, J.R.; Whitehouse, J.C.; Wilson, C.N. & Van Ryn, F.R.
Partner: UNT Libraries Government Documents Department

In-reactor performance of methods to control fuel-cladding chemical interaction. [LMFBR]

Description: Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 ..mu..m wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction (FCCI), oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: (1) control of total oxygen inventory and chemical activity (..delta.. anti GO/sub 2/) by use of low oxygen-to-metal ratio (O/M) fuel; and (2) incorporation of a material within the fuel pin to provide in-situ control of oxygen activity (..delta.. anti GO/sub 2/) and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins.
Date: January 1, 1979
Creator: Weber, E. T.; Gibby, R. L.; Wilson, C. N.; Lawrence, L. A. & Adamson, M. G.
Partner: UNT Libraries Government Documents Department

Spent fuel waste form characteristics: Grain and fragment size statistical dependence for dissolution response

Description: The Yucca Mountain Project of the US Department of Energy is investigating the suitability of the unsaturated zone at Yucca Mountain, NV, for a high-level nuclear waste repository. All of the nuclear waste will be enclosed in a container package. Most of the nuclear waste will be in the form of fractured UO{sub 2} spent fuel pellets in Zircaloy-clad rods from electric power reactors. If failure of both the container and its enclosed clad rods occurs, then the fragments of the fractured UO{sub 2} spent fuel will be exposed to their surroundings. Even though the surroundings are an unsaturated zone, a possibility of water transport exists, and consequently, UO{sub 2} spent fuel dissolution may occur. A repository requirement imposes a limit on the nuclide release per year during a 10,000 year period; thus the short term dissolution response from fragmented fuel pellet surfaces in any given year must be understood. This requirement necessitates that both experimental and analytical activities be directed toward predicting the relatively short term dissolution response of UO{sub 2} spent fuel. The short term dissolution response involves gap nuclides, grain boundary nuclides, and grain volume nuclides. Analytical expressions are developed that describe the combined geometrical influences of grain boundary nuclides and grain volume nuclides on the dissolution rate of spent fuel. 7 refs., 1 fig.
Date: April 1, 1991
Creator: Stout, R.B.; Leider, H.; Weed, H.; Nguyen, S.; McKenzie, W.; Prussin, S. et al.
Partner: UNT Libraries Government Documents Department