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Hydrofluoric Acid Corrosion Testing on Unplated and Electroless Gold-Plated Samples

Description: The Molten Salt Reactor Experiment (MSRE) remediation requires that almost 40 kg of uranium hexafluoride (UF6) be converted to uranium oxide (UO). In the process of this conversion, six moles of hydrofluoric acid (HP) are produced for each mole of UF6 converted.
Date: August 1, 2000
Creator: Osborne, P.E.; Icenhour, A.S. & Del Cul, G.D.
Partner: UNT Libraries Government Documents Department

Scientific Analysis Cover Sheet for Radionuclide Screening

Description: The waste forms under consideration for disposal in the proposed repository at Yucca Mountain contain scores of radionuclides (Attachments V and VI). It would be impractical and highly inefficient to model all of these radionuclides in a total system performance assessment (TSPA). Thus, the purpose of this radionuclide screening analysis is to remove from further consideration (screen out) radionuclides that are unlikely to significantly contribute to radiation dose to the public from the proposed nuclear waste repository at Yucca Mountain. The remaining nuclides (those screened in) are recommended for consideration in TSPA modeling for license application. This analysis also covers radionuclides that are not screened in based on dose, but need to be included in TSPA modeling for other reasons. For example, U.S. Environmental Protection Agency (EPA) and U.S. Nuclear Regulatory Commission (NRC) regulations require consideration of the combined activity of Ra-226 and Ra-228 in groundwater (40 CFR 197.30, 10 CFR 63.331). Also, Cm-245, Pu-241, and U-235 decay indirectly to potentially important radionuclides, and are not identified by the screening analysis as important. The radionuclide screening analysis separately considers two different postclosure time periods: the 10,000-y regulatory period for the proposed repository at Yucca Mountain and the period after 10,000 y up to 1 million y after emplacement. The incremental effect of extending the screening for the regulatory period to 20,000 y is also addressed. Four release scenarios are considered: (1) the nominal scenario, which entails long-term degradation of disposal containers and waste forms, (2) a human-intrusion scenario, (3) an intrusive igneous event, and (4) an eruptive igneous event. Because the first three scenarios require groundwater transport, they are called groundwater scenarios below. The screening analysis considers the following waste forms: spent boiling water reactor (BWR) fuel, spent pressurized water reactor (PWR) fuel, U.S. Department of Energy (DOE) spent nuclear ...
Date: August 9, 2002
Creator: Ragan, G.
Partner: UNT Libraries Government Documents Department

Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

Description: Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers.
Date: August 21, 2000
Creator: O'Leary, P.M. & Pitts, M. L.
Partner: UNT Libraries Government Documents Department

Isotopic Bias and Uncertainty for Burnup Credit Applications

Description: The application of burnup credit requires calculating the isotopic inventory of the irradiated fuel. The depletion calculation simulates the burnup of the fuel under reactor operating conditions. The result of the depletion analysis is the predicted isotopic composition, which is ultimately input to a criticality analysis to determine the system multiplication factor (k{sub eff}). This paper demonstrates an approach for calculating the isotopic bias and uncertainty in k{sub eff} for commercial spent nuclear fuel burnup credit. This paper covers 74 different radiochemical assayed spent fuel samples from 22 different fuel assemblies that were irradiated in eight different pressurized water reactors (PWRs). The samples evaluated span an enrichment range of 2.556 wt% U-235 through 4.67 wt% U-235, and burnups from 6.92 GWd/MTU through 55.7 GWd/MTU.
Date: August 19, 2002
Creator: Scaglione, J.M.
Partner: UNT Libraries Government Documents Department

Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

Description: The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000).
Date: August 29, 2001
Creator: Wimmer, Larry B.
Partner: UNT Libraries Government Documents Department

Initial Cladding Condition

Description: The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M&O 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in evaluating the post-closure performance of the Monitored Geologic Repository ...
Date: August 22, 2000
Creator: Siegmann, E.
Partner: UNT Libraries Government Documents Department

Regenerative Heater Optimization for Steam Turbo-Generation Cycles of Generation IV Nuclear Power Plants with a Comparison of Two Concepts for the Westinghouse International Reactor Innovative and Secure (IRIS)

Description: The intent of this study is to discuss some of the many factors involved in the development of the design and layout of a steam turbo-generation unit as part of a modular Generation IV nuclear power plant. Of the many factors involved in the design and layout, this research will cover feed water system layout and optimization issues. The research is arranged in hopes that it can be generalized to any Generation IV system which uses a steam powered turbo-generation unit. The research is done using the ORCENT-II heat balance codes and the Salisbury methodology to be reviewed herein. The Salisbury methodology is used on an original cycle design by Famiani for the Westinghouse IRIS and the effects due to parameter variation are studied. The vital parameters of the Salisbury methodology are the incremental heater surface capital cost (S) in $/ft{sup 2}, the value of incremental power (I) in $/kW, and the overall heat transfer coefficient (U) in Btu/ft{sup 2}-degrees Fahrenheit-hr. Each is varied in order to determine the effects on the cycles overall heat rate, output, as well as, the heater surface areas. The effects of each are shown. Then the methodology is then used to compare the optimized original Famiani design consisting of seven regenerative feedwater heaters with an optimized new cycle concept, INRC8, containing four regenerative heaters. The results are shown. It can be seen that a trade between the complexity of the seven stage regenerative Famiani cycle and the simplicity of the INRC8 cycle can be made. It is desired that this methodology can be used to show the ability to evaluate modularity through the value of size a complexity of the system as well as the performance. It also shows the effectiveness of the Salisbury methodology in the optimization of regenerative cycles for such an ...
Date: August 1, 2002
Creator: Williams, W.C.
Partner: UNT Libraries Government Documents Department

Calculation Analysis of San Onofre Depletion MOX Fuel Experiment

Description: The report provides calculation results of isotopic composition of spent MOX fuel irradiated in Sun Onofre PWR reactor. The calculation was performed by means of the MCU/BURNUP Monte Carlo code. The code is developed in Kurchatov Institute, Russia. The predicted isotope contents are compared with the measured ones. A purpose of this work is a verification both the code and the model of experiment description. Predicted plutonium content exceeds the measured one approximately by 3%. It is arise mainly from error of {sup 239}Pu isotope. Isotopic contents of the main plutonium and uranium isotopes are predicted with satisfactory precision.
Date: August 31, 2001
Creator: Pavlovichev, AM
Partner: UNT Libraries Government Documents Department

The future of reactor neutrino experiments: A novel approach to measuring theta{sub 13}

Description: Results from non-accelerator neutrino oscillation experiments have provided evidence for the oscillation of massive neutrinos. The subdominant oscillation, the coupling of the electron neutrino flavor to the third mass eigenstate, has not been measured yet. The size of this coupling U{sub e3} and its corresponding mixing angle theta{sub 13} are critical for CP violation searches in the lepton sector and will define the future of accelerator neutrino physics. The current best limit on U{sub e3} comes from the CHOOZ reactor neutrino disappearance experiment. In this talk we review proposals for future measurements of theta-13 with reactor antineutrinos.
Date: August 24, 2003
Creator: Heeger, Karsten M.; Freedman, Stuart J. & Luk, Kam-Biu
Partner: UNT Libraries Government Documents Department

Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

Description: This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation--or neutrino oscillation--by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5% respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock mechanical and acoustic properties. The rock strength values ...
Date: August 1, 2004
Creator: Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven & Galic, Dom
Partner: UNT Libraries Government Documents Department

The Physics of transmutation systems : system capabilities and performances.

Description: This document is complementary to a document produced by Prof. Salvatores on ''The Physics of Transmutation in Critical or Subcritical Reactors and the Impact on the Fuel Cycle''. In that document, Salvatores describes the fundamental of transmutation, through basic physics properties and general parametric studies. In the present document we try to go one step further towards practical implementation (while recognizing that the practical issues such as technology development and demonstration, and economics, can only be mentioned in a very superficial manner). Section 1 briefly overviews the possible objectives of transmutation systems, and links these different objectives to possible technological paths. It also describes the overall constraints which have to be considered when developing and implementing transmutation systems. In section 2 we briefly overview the technological constraints which need to be accounted for when designing transmutation systems. In section 3 we attempt to provide a simplified classification of transmutation systems in order to clarify later comparisons. It compares heterogeneous and homogeneous recycle strategies, and single and multi-tier systems. Section 4 presents case analyses for assessing the transmutation performance of various individual systems, starting with LWR's (1. generic results; 2. multirecycle of plutonium; 3. an alternative: transmutation based on a Thorium fuel cycle), followed by Gas-Cooled Reactors (with an emphasis on the ''deep burn'' approach), and followed by Fast Reactors and Accelerator Driven systems (1. generic results; 2. homogeneous recycle of transuranics; 3. practical limit between Fast Reactors and Accelerator Driven Systems) Section 5 summarizes recent results on integrated system performances. It focuses first on interface effects between the two elements of a dual tier system, and then summarizes the major lessons learned from recent global physics studies.
Date: August 21, 2002
Creator: Finck, P. J.
Partner: UNT Libraries Government Documents Department

Transition to a nuclear/hydrogen energy system.

Description: The paper explores the motivation for the transition to a nuclear/hydrogen system. For such a transition to be successful the technologies employed must be able to generate enough hydrogen to displace a significant fraction of the petroleum fuels used in the transportation and process heat sectors. This hydrogen must be generated in a manner that is compatible with the environment and independent of foreign fuels. Nuclear energy, along with contributions from wind, solar, and geothermal resources meet the criteria of environmental compatibility and resource independence. However, nuclear energy is the only one of these sources that has a high enough energy density to generate copious quantities of hydrogen. The status of the relevant nuclear and hydrogen technologies are discussed and how they are coupled to bring about a transition to a nuclear/hydrogen system. Should the world adopt such a system then the growth rate of nuclear energy would greatly accelerate. With an accelerated growth for nuclear energy the uranium resources would be depleted in a few decades with the once through fuel cycle currently in use. It is pointed out that deployment of fast breeder reactors would become important in the nearer term.
Date: August 13, 2002
Creator: Walters, L.; Wade, D. & Lewis, D.
Partner: UNT Libraries Government Documents Department

Calculation of Doses Due to Accidentally Released Plutonium From An LMFBR

Description: Experimental data and analytical models that should be considered in assessing the transport properties of plutonium aerosols following a hypothetical reactor accident have been examined. Behaviors of released airborne materials within the reactor containment systems, as well as in the atmosphere near the reactor site boundaries, have been semiquantitatively predicted from experimental data and analytical models. The fundamental chemistry of plutonium as it may be applied in biological systems has been used to prepare models related to the intake and metabolism of plutonium dioxide, the fuel material of interest. Attempts have been made to calculate the possible doses from plutonium aerosols for a typical analyzed release in order to evaluate the magnitude of the internal exposure hazards that might exist in the vicinity of the reactor after a hypothetical LMFBR (Liquid-Metal Fast Breeder Reactor) accident. Intake of plutonium (using data for {sup 239}Pu as an example) and its distribution in the body were treated parametrically without regard to the details of transport pathways in the environment. To the extent possible, dose-response data and models have been reviewed, and an assessment of their adequacy has been made so that recommended or preferred practices could be developed.
Date: August 7, 2001
Creator: Fish, B.R.
Partner: UNT Libraries Government Documents Department

RELAP5 / MOD3.2 analysis of INSC standard problem INSCSP - R7 : void fraction distribution over RBMK fuel channel height for experiments performed in the ENTEK BM test facility.

Description: The RELAP5/MOD3.2 computer program has been used to analyze a series of tests investigating void fraction distribution over height in RBMK fuel channels performed in Facility BM at the ENTEK. This is RBMK Standard Problem 7 in Joint Project 6, which is the investigation of Computer Code Validation for Transient Analysis of RBMK and VVER Reactors, between the United States and Russian Minatom International Nuclear Safety Centers. The experiment facility and data, RELAP5 nodalization, and results are shown for all tests. Agreement between RELAP5 and the experiment data is reasonable.
Date: August 22, 2002
Creator: Garner, P. L.
Partner: UNT Libraries Government Documents Department

Development and use of GREET 1.6 fuel-cycle model for transportation fuels and vehicle technologies.

Description: Since 1995, with funds from the U.S. Department of Energy's (DOE's) Office of Transportation Technologies (OTT), Argonne National Laboratory has been developing the Greenhouse gases, Regulated Emissions, and Energy use in Transportation (GREET) model. The model is intended to serve as an analytical tool for use by researchers and practitioners in estimating fuel-cycle energy use and emissions associated with alternative transportation fuels and advanced vehicle technologies. Argonne released the first version of the GREET mode--GREET 1.0--in June 1996. Since then, it has released a series of GREET versions with revisions, updates, and upgrades. In February 2000, the latest public version of the model--GREET 1.5a--was posted on Argonne's Transportation Technology Research and Development Center (TTRDC) Web site (www.transportation.anl.gov/ttrdc/greet).
Date: August 23, 2001
Creator: Wang, M. Q.
Partner: UNT Libraries Government Documents Department

ORNL Nuclear Safety Research and Development Program Bimonthly Report for July-August 1968

Description: The accomplishments during the months of July and August in the research and development program under way at ORNL as part of the U.S. Atomic Energy Commission's Nuclear Safety Program are summarized, Included in this report are work on various chemical reactions, as well as the release, characterization, and transport of fission products in containment systems under various accident conditions and on problems associated with the removal of these fission products from gas streams. Although most of this work is in general support of water-cooled power reactor technology, including LOFT and CSE programs, the work reflects the current safety problems, such as measurements of the prompt fuel element failure phenomena and the efficacy of containment spray and pool-suppression systems for fission-product removal. Several projects are also conducted in support of the high-temperature gas-cooled reactor (HTGR). Other major projects include fuel-transport safety investigations, a series of discussion papers on various aspects of water-reactor technology, antiseismic design of nuclear facilities, and studies of primary piping and steel, pressure-vessel technology. Experimental work relative to pressure-vessel technology includes investigations of the attachment of nozzles to shells and the implementation of joint AEX-PVFX programs on heavy-section steel technology and nuclear piping, pumps, and valves. Several of the projects are directly related to another major undertaking; namely, the AEC's standards program, which entails development of engineering safeguards and the establishment of codes and standards for government-owned or -sponsored reactor facilities. Another task, CHORD-S, is concerned with the establishment of computer programs for the evaluation of reactor design data, The recent activities of the NSIC and the Nuclear Safety journal in behalf of the nuclear community are also discussed.
Date: August 17, 2001
Creator: Cottrell, W.B.
Partner: UNT Libraries Government Documents Department

ORNL Nuclear Safety Research and Development Program Bimonthly Report for September-October 1968

Description: The accomplishments during the months of September and October in the research and development program under way at ORNL as part of the U.S. Atomic Energy Commission's Nuclear Safety Program are summarized. Included in this report are work on various chemical reactions, as well as the release, characterization, and transport of fission products in containment systems under various accident conditions and on problems associated with the removal of these fission products from gas streams. Although most of this work is in general support of water-cooled power reactor technology, including LOFT and CSE programs, the work reflects the current safety problems, such as measurements of the prompt fuel element failure phenomena and the efficacy of containment spray and pool-suppression systems for fission-product removal. Several projects are also conducted in support of the high-temperature gas-cooled reactor (HTGR). Other major projects include fuel-transport safety investigations, a series cf discussion papers on various aspects of water-reactor technology, antiseismic design of nuclear facilities, and studies of primary piping and steel pressure-vessel technology. Experimental work relative to pressure-vessel technology includes investigations of the attachment of nozzles to shells and the implementation of joint AEC-PVRC programs on heavy-section steel technology and nuclear piping, pumps, and valves. Several of the projects are directly related to another major undertaking; namely, the AEC's standards program, which entails development of engineering safeguards and the establishment of codes and standards for government-owned or -sponsored reactor facilities. Another task, CHORD-S, is concerned with the establishment of computer programs for the evaluation of reactor design data. The recent activities of the NSIC and the Nuclear Safety journal in behalf of the nuclear community are also discussed.
Date: August 17, 2001
Creator: Cottrell, W.B.
Partner: UNT Libraries Government Documents Department

Nuclear Energy Research Initiative. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants. Annual Report

Description: The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-informed approach for the design and regulation of nuclear power plants. This approach will include the development and.lor confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRs) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go farther by focusing on the design of new plants.
Date: August 1, 2000
Creator: Ritterbusch, S.E.
Partner: UNT Libraries Government Documents Department

Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

Description: OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.
Date: August 1, 2000
Partner: UNT Libraries Government Documents Department

Energy Return on Investment from Recycling Nuclear Fuel

Description: This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.
Date: August 17, 2011
Partner: UNT Libraries Government Documents Department

CRITICALITY SAFETY CONTROL OF LEGACY FUEL FOUND AT 105-K WEST FUEL STORAGE BASIN

Description: In August 2004, two sealed canisters containing spent nuclear fuel were opened for processing at the Hanford Site's K West fuel storage basin. The fuel was to be processed through cleaning and sorting stations, repackaged into special baskets, placed into a cask, and removed from the basin for further processing and eventual dry storage. The canisters were expected to contain fuel from the old Hanford C Reactor, a graphite-moderated reactor fueled by very low-enriched uranium metal. The expected fuel type was an aluminum-clad slug about eight inches in length and with a weight of about eight pounds. Instead of the expected fuel, the two canisters contained several pieces of thin tubes, some with wire wraps. The material was placed into unsealed canisters for storage and to await further evaluation. Videotapes and still photographs of the items were examined in consultation with available retired Hanford employees. It was determined that the items had a fair probability of being cut-up pieces of fuel rods from the retired Hanford Plutonium Recycle Test Reactor (PRTR). Because the items had been safely handled several times, it was apparent that a criticality safety hazard did not exist when handling the material by itself, but it was necessary to determine if a hazard existed when combining the material with other known types of spent nuclear fuel. Because the PRTR operated more than 40 years ago, investigators had to rely on a combination of researching archived documents, and utilizing common-sense estimates coupled with bounding assumptions, to determine that the fuel items could be handled safely with other spent nuclear fuel in the storage basin. As older DOE facilities across the nation are shut down and cleaned out, the potential for more discoveries of this nature is increasing. As in this case, it is likely that only incomplete records ...
Date: August 19, 2005
Creator: Jensen, M. A.
Partner: UNT Libraries Government Documents Department