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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

Electrometallurgical treatment of metallic spent nuclear fuel stored at the Hanford Site

Description: The major component of the DOE spent nuclear fuel inventory is the metallic fuel stored at the Hanford site in the southeastern part of the state of Washington. Most of this fuel was discharged from the N-Reactor; a small part of the inventory is fuel from the early Hanford production reactors. The U.S. Department of Energy (DOE) plans to remove these fuels from the spent fuel storage pools in which they are presently stored, dry them, and place them in interim storage at a location at the Hanford site that is far removed from the Columbia River. It is not yet certain that these fuels will be acceptable for disposal in a mined geologic repository without further treatment, due to their potential pyrophoric character. A practical method for treatment of the Hanford metallic spent fuel, based on an electrorefining process, has been developed and has been demonstrated with unirradiated N-Reactor fuel and with simulated single-pass reactor (SPR) spent fuel. The process can be operated with any desired throughput rates; being a batch process, it is simply a matter of setting the size of the electrorefiner modules and the number of such modules. A single module, prototypic of a production-scale module, has been fabricated and testing is in progress at a throughput rate of 150 kg (heavy metal) per day. The envisioned production version would incorporate additional anode baskets and cathode tubes and provide a throughput rate of 333 kgHM/day. A system with four of these modules would permit treatment of Hanford metallic fuels at a rate of at least 250 metric tons per year.
Date: May 1, 1996
Creator: Laidler, J.J. & Gay, E.C.
Partner: UNT Libraries Government Documents Department

Strategy for phase 2 whole element furnace testing K West fuel

Description: A strategy was developed for the second phase of the whole element furnace testing of damaged fuel removed from the K West Basin. The Phase 2 testing can be divided into three groups covering oxidation of whole element in moist inert atmospheres, drying elements for post Cold Vacuum Drying staging tests, and drying additional K West elements to provide confirmation of the results from the first series of damaged K West fuel drying studies.
Date: March 13, 1998
Creator: Lawrence, L.A.
Partner: UNT Libraries Government Documents Department

Observations during the first K West fuel shipping campaign

Description: Three fuel elements were shipped to the 300 Area hotcells during the first characterization shipping campaign from K West Basin. This document summarizes observations made during this campaign including the gas, liquid, and sludge content of the observed canisters. Included in an appendix is a detailed evaluation of fuel element condition for each canister opened
Date: November 1, 1995
Creator: Makenas, B. J.
Partner: UNT Libraries Government Documents Department

KE Basin underwater visual fuel survey

Description: Results of an underwater video fuel survey in KE Basin using a high resolution camera system are presented. Quantitative and qualitative information on fuel degradation are given, and estimates of the total fraction of ruptured fuel elements are provided. Representative photographic illustrations showing the range of fuel conditions observed in the survey are included.
Date: February 1, 1995
Creator: Pitner, A.L.
Partner: UNT Libraries Government Documents Department

Radionuclide mass inventory, activity, decay heat, and dose rate parametric data for TRIGA spent nuclear fuels

Description: Parametric burnup calculations are performed to estimate radionuclide isotopic mass and activity concentrations for four different Training, Research, and Isotope General Atomics (TRIGA) nuclear reactor fuel element types: (1) Aluminum-clad standard, (2) Stainless Steel-clad standard, (3) High-enrichment Fuel Life Improvement Program (FLIP), and (4) Low-enrichment Fuel Life Improvement Program (FLIP-LEU-1). Parametric activity data are tabulated for 145 important radionuclides that can be used to generate gamma-ray emission source terms or provide mass quantity estimates as a function of decay time. Fuel element decay heats and dose rates are also presented parametrically as a function of burnup and decay time. Dose rates are given at the fuel element midplane for contact, 3.0-feet, and 3.0-meter detector locations in air. The data herein are estimates based on specially derived Beginning-of-Life (BOL) neutron cross sections using geometrically-explicit TRIGA reactor core models. The calculated parametric data should represent good estimates relative to actual values, although no experimental data were available for direct comparison and validation. However, because the cross sections were not updated as a function of burnup, the actinide concentrations may deviate from the actual values at the higher burnups.
Date: March 1, 1997
Creator: Sterbentz, J.W.
Partner: UNT Libraries Government Documents Department

Natural convection heat transfer within horizontal spent nuclear fuel assemblies

Description: Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.
Date: December 1, 1995
Creator: Canaan, R.E.
Partner: UNT Libraries Government Documents Department

Radiation doses in granite around emplacement holes in the Spent Fuel Test - Climax. Final report

Description: Final comparisons are made between measured and calculated radiation doses around the holes in which the spent fuel was emplaced in the Spent Fuel Test - Climax. Neutron doses were found to be negligible compared with gamma doses. Good agreement was found between the doses predicted by Monte Carlo calculations and those measured by short-exposure thermoluminescence dosimetry. Poor agreement was found between the calculational results and doses measured by exposure of LiF optical-absorption-type dosimeters for long periods, probably because of an inability to accurately correct for fade resulting from elevated temperature exposure over several months. The maximum dose to the rock occurred at the walls of the emplacement holes, and amounted to 1.6 MGy (1.6 x 10{sup 8} rad) in granite for the emplacement period of nearly 3 years. It is recommended that dose evaluations for future high-level nuclear waste storage facilities also be performed by combining calculations and dosimetry. Passive dosimetry techniques, if used, should involve short exposures, so that laboratory calibrations can be performed with duplicate time, temperature, dose rate, and dose parameters. An attractive alternative would be to use active ionization chambers, inserted only periodically. These could be calibrated under appropriate temperature and pressure conditions, and could be read directly. 23 references, 7 figures, 8 tables.
Date: July 26, 1984
Creator: Van Konynenburg, R.A.
Partner: UNT Libraries Government Documents Department

Investigation of Burnup Credit Issues in BWR Fuel

Description: Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel.
Date: September 20, 1999
Creator: Broadhead, B.L. & DeHart, M.D.
Partner: UNT Libraries Government Documents Department

Remotely operable modular shear system for shearing nuclear reactor spent fuel conceptual design. Final summary report

Description: A unique feature of the shear system is the requirement for a removable master tool module and tool submodule assembly. Each tooling submodule is designed specifically to meet the compaction, gagging, and shearing requirements for specific spent fuel assemblies. These submodules are interchangeable within a common master tool module housing. The cross section of the spent fuel assemblies range from a 4.575-in. hexagonal shrouded to a 8.75-in. square grid-unshrouded. A number of interrelated initial design problems had to be solved: a hydraulic force system arrangement with minimum spacing between cylinders; readily removable couplings between the force system and associated tooling submodule; couplings with maximum stiffness and minimum geometry; a roller bearing system for each tool submodule and hydraulic stem assembly; and long life tool operation under high loads, wear, high temperaure, and corrosive conditions. It was established that the cylinder arrangement should consist of three 200-ton, 10-in. dia tandem cylinders, and one 100-ton, 10-in. dia standard cylinder at 3000 psi operating pressure. The shear would be operated by two - 200 ton tandem cylinders in a vertical straddle mode about the horizontal center line of the tooling arrangement. The compactor would be operated by one - 200 ton tandem cylinder at the center line of the fuel assembly and tooling arrangement. The gag would be operated by one - 100 ton standard cylinder at the center line of the compacted fuel assembly and tooling arrangement.
Date: June 16, 1978
Creator: Buckingham, D.
Partner: UNT Libraries Government Documents Department

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

Description: The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.
Date: March 1, 2011
Creator: Morrell, Douglas
Partner: UNT Libraries Government Documents Department

Drying results of K-Basin fuel element 0309M (Run 3)

Description: An N-Reactor outer fuel element that had been stored underwater 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. The drying test series was designed to test fuel elements that ranged from intact to severely damaged. The fuel element discussed in this report was removed from K-West canister 0309M during the second fuel selection campaign, conducted in 1996, and has remained in wet storage in the Postirradiation Testing Laboratory (PTL, 327 Building) since that time. The fuel element was broken in two pieces, with a relatively clean fracture, and the larger piece was tested. A gray/white coating was observed. This was the first test of a damaged fuel element in the furnace. K-West canisters can hold up to seven complete fuel assemblies, but, for purposes of this report, the element tested here is designated as Element 0309M. Element 0309M was subjected to drying processes based on those proposed under the Integrated Process Strategy, which included a hot drying step.
Date: July 1, 1998
Creator: Oliver, B.M.; Klinger, G.S.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J. & Ritter, G.A.
Partner: UNT Libraries Government Documents Department

System design description for the whole element furnace testing system

Description: This document provides a detailed description of the Hanford Spent Nuclear Fuel (SNF) Whole Element Furnace Testing System located in the Postirradiation Testing Laboratory G-Cell (327 Building). Equipment specifications, system schematics, general operating modes, maintenance and calibration requirements, and other supporting information are provided in this document. This system was developed for performing cold vacuum drying and hot vacuum drying testing of whole N-Reactor fuel elements, which were sampled from the 105-K East and K West Basins. The proposed drying processes are intended to allow dry storage of the SNF for long periods of time. The furnace testing system is used to evaluate these processes by simulating drying sequences with a single fuel element and measuring key system parameters such as internal pressures, temperatures, moisture levels, and off-gas composition.
Date: May 1, 1998
Creator: Ritter, G.A.; Marschman, S.C.; MacFarlan, P.J. & King, D.A.
Partner: UNT Libraries Government Documents Department

Strategy for analysis of coatings and subsurface sludge recovered during hot cell examinations of N Reactor elements from Hanford K Basins

Description: Subsurface sludge and/or surface coating material have been collected for four N Reactor fuel elements from K West Basin and one element from K East Basin. It is proposed that examinations of the fairly small volumes of recovered material proceed in order to identify the constituents and their potential impacts on fuel and sludge processing. Lists of potential examination techniques and material allocations are given in this report.
Date: September 22, 1997
Creator: Makenas, B.J.
Partner: UNT Libraries Government Documents Department

Planning and supervision of reactor defueling using discrete event techniques

Description: New fuel handling and conditioning activities for the defueling of the Experimental Breeder Reactor II are being performed at Argonne National Laboratory. Research is being conducted to investigate the use of discrete event simulation, analysis, and optimization techniques to plan, supervise, and perform these activities in such a way that productivity can be improved. The central idea is to characterize this defueling operation as a collection of interconnected serving cells, and then apply operational research techniques to identify appropriate planning schedules for given scenarios. In addition, a supervisory system is being developed to provide personnel with on-line information on the progress of fueling tasks and to suggest courses of action to accommodate changing operational conditions. This paper provides an introduction to the research in progress at ANL. In particular, it briefly describes the fuel handling configuration for reactor defueling at ANL, presenting the flow of material from the reactor grid to the interim storage location, and the expected contributions of this work. As an example of the studies being conducted for planning and supervision of fuel handling activities at ANL, an application of discrete event simulation techniques to evaluate different fuel cask transfer strategies is given at the end of the paper.
Date: December 31, 1995
Creator: Garcia, H.E.; Imel, G.R. & Houshyar, A.
Partner: UNT Libraries Government Documents Department

Radiological consequences of a hypothetical disruption of a maximally loaded FFTF fuel cask

Description: Radiological consequences at the site boundary were estimated for non-mechanistic disruption of an Interim Storage Cask (ISC) loaded with 7 assemblies at the maximum available burnup. The hypothetical disruption consisted of a crushing/shearing of the Core Component Container (CCC) along with all 7 assemblies and the creation of a large escape path out of the cask. The resulting site boundary dose of 1.6 mSv is far below the 250 mSv risk guidelines for highly unlikely events.
Date: September 22, 1995
Creator: Scott, P.A.
Partner: UNT Libraries Government Documents Department

Hanford`s spent nuclear fuel retrieval: an agressive agenda

Description: Starting December 1997, spent nuclear fuel that has been stored in the K Reactor Fuel Storage Basins will be retrieved over a two year period and repackaged for long term dry storage. The aging and sometimes corroding fuel elements will be recovered and processed using log handled tools and teleoperated manipulator technology. The U.S. Department of Energy (DOE) is committed to this urgent schedule because of the environmental threats to the groundwater and nearby the Columbia River.
Date: December 6, 1996
Creator: Shen, E.J., Westinghouse Hanford
Partner: UNT Libraries Government Documents Department