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Qualitative Reliability Issues for Solid and Liquid Wall Fusion Design

Description: This report is an initial effort to identify issues affecting reliability and availability of solid and liquid wall designs for magnetic fusion power plant designs. A qualitative approach has been used to identify the possible failure modes of major system components and their effects on the systems. A general set of design attributes known to affect the service reliability has been examined for the overview solid and liquid wall designs, and some specific features of good first wall design have been discussed and applied to these designs as well. The two generalized designs compare well in regard to these design attributes. The strengths and weaknesses of each design approach are seen in the comparison of specific features.
Date: January 1, 2001
Creator: Cadwallader, Lee Charles
Partner: UNT Libraries Government Documents Department

The Influence of the In-Situ Clad Staining on the Corrosion of Zircaloy in PWR Water Environment

Description: Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the diametrical direction under the external pressure of the water environment, but potentially grow later in life in the diametrical direction once the expanding fuel pellet contacts the cladding inner wall (Reference (b)). Finally, the Zircaloy cladding absorbs hydrogen as a by product of the corrosion reaction (Reference (c)). Once above the solubility limit in Zircaloy, the hydride precipitates as zirconium hydride (References (c) through (j)). Both hydrogen in solid solution and precipitated as Zirconium hydride cause a volume expansion of the Zircaloy metal (Reference (k)). Few studies are reported on that have investigated the influence that in-situ clad straining has on corrosion of Zircaloy. If Zircaloy corrosion rates are governed by diffusion of anions through a thin passivating boundary layer at the oxide-to-metal interface (References (l) through (n)), in-situ straining of the cladding could accelerate the corrosion process by prematurely breaking that passivating oxide boundary layer. References (o) through (q) investigated the influence that an applied tensile stress has on the corrosion resistance of Zircaloy. Knights and Perkins, Reference (o), reported that the applied tensile stress increased corrosion rates above a critical stress level in 400 C and 475 C steam, but not at lower temperatures nor in dry oxygen environments. This latter observation suggested that hydrogen either in the oxide or at the oxide-to-metal interface is involved in the observed stress effect. Kim et ...
Date: June 21, 2001
Creator: Kammenzind, B.F., Eklund, K.L. and Bajaj, R.
Partner: UNT Libraries Government Documents Department

Feasibility of Isotopic Measurements: Graphite Isotopic Ratio Method

Description: This report addresses the feasibility of the laboratory measurements of isotopic ratios for selected trace constituents in irradiated nuclear-grade graphite, based on the results of a proof-of-principal experiment completed at Pacific Northwest National Laboratory (PNNL) in 1994. The estimation of graphite fluence through measurement of isotopic ratio changes in the impurity elements in the nuclear-grade graphite is referred to as the Graphite Isotope Ratio Method (GIRM). Combined with reactor core and fuel information, GIRM measurements can be employed to estimate cumulative materials production in graphite moderated reactors. This report documents the laboratory procedures and results from the initial measurements of irradiated graphite samples. The irradiated graphite samples were obtained from the C Reactor (one of several production reactors at Hanford) and from the French G-2 Reactor located at Marcoule. Analysis of the irradiated graphite samples indicated that replicable measurements of isotope ratios could be obtained from the fluence sensitive elements of Ti, Ca, Sr, and Ba. While these impurity elements are present in the nuclear-grade graphite in very low concentrations, measurement precision was typically on the order of a few tenths of a percent to just over 1 percent. Replicability of the measurements was also very good with measured values differing by less than 0.5 percent. The overall results of this initial proof-of-principal experiment are sufficiently encouraging that a demonstration of GIRM on a reactor scale basis is planned for FY-95.
Date: April 30, 2001
Creator: Wood, Thomas W.; Gerlach, David C.; Reid, Bruce D. & Morgan, W. C.
Partner: UNT Libraries Government Documents Department

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages

Description: The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The ...
Date: February 27, 2001
Creator: Bernot, P.
Partner: UNT Libraries Government Documents Department

A Review of Proposed Upgrades to the High Flux Isotope Reactor and Potential Impacts to Reactor Vessel Integrity

Description: The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) was scheduled in October 2000 to implement design upgrades that include the enlargement of the HB-2 and HB-4 beam tubes. Higher dose rates and higher radiation embrittlement rates were predicted for the two beam-tube nozzles and surrounding vessel areas. ORNL had performed calculations for the upgraded design to show that vessel integrity would be maintained at acceptable levels. Pacific Northwest National Laboratory (PNNL) was requested by the U.S. Department of Energy Headquarters (DOE/HQ) to perform an independent peer review of the ORNL evaluations. PNNL concluded that the calculated probabilities of failure for the HFIR vessel during hydrostatic tests and for operational conditions as estimated by ORNL are an acceptable basis for selecting pressures and test intervals for hydrostatic tests and for justifying continued operation of the vessel. While there were some uncertainties in the embrittlement predictions, the ongoing efforts at ORNL to measure fluence levels at critical locations of the vessel wall and to test materials from surveillance capsules should be effective in dealing with embrittlement uncertainties. It was recommended that ORNL continue to update their fracture mechanics calculations to reflect methods and data from ongoing research for commercial nuclear power plants. Such programs should provide improved data for vessel fracture mechanics calculations.
Date: May 31, 2001
Creator: Simonen, Fredric A.
Partner: UNT Libraries Government Documents Department


Description: Pursuing verification of the nuclear data for actinides, we have made a run of experiments to determine reaction rates in facilities with different neutron spectra. The researches of the kind are particularly argent when going over from the transmutation physics studies to designing the transmutation reactors and developing their fuel cycle equipment. In this case, the nuclear data on the minor actinides (Np, Am, Cm) are notably interesting with the view to correct prediction of transmutation rates and t,o validation of hazardous nuclear and radiation environment for the external (off-reactor) fuel cycle. It is in the case of just those nuclides when the well-known ENDF/B6 and JENDL3.2 libraries give the most discrepant nuclear cross sections, thus necessitating the high- priority experimental tests.
Date: January 1, 2001
Creator: Titarenko, Y. E. (Yury E.); Batyaev, V. F. (Vyacheslav F.); Karpikhin, E. I. (Evgeny I.); Zhivun, V. M. (Valery M.); Koldobsky, A. B. (Aleksander B.); Mulambetov, R. D. (Ruslan D.) et al.
Partner: UNT Libraries Government Documents Department

Technique for the identification of dominant delayed neutron precursors.

Description: A technique for the identification of delayed neutron precursors has been developed based on the product of cumulative yield and probability of neutron emission. The motivation behind this work is to fix the decay constants of delayed neutrons to those of the dominant delayed neutron precursors. The desirability of identifying a single set of decay constants that would apply to all fissionable isotopes and be independent of the neutron energy spectrum has been addressed by several authors. The main advantages of a fixed-decay constant representation are simplifying the analysis of epithemal and fast reactors with multiple fissioning isotopes, and improving the fit to experimental data while preserving the inferred positive reactivity scale associated with the original six-group representation. It is well known that 27 1 delayed neutron precursors exist, but only a select number of those precursors contribute significantly to the decay of delayed neutron. Using data compiled by England and Rider, which lists fission yield and probability of neutron emission values for the 27 1 known delayed neutron precursors in 32 fissioning systems, thirteen precursors were identified that are consistently dominant for alI fissioning systems.
Date: January 1, 2001
Creator: Loaiza, D. J. (David J.) & Haskin, E. (Eric)
Partner: UNT Libraries Government Documents Department


Description: The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.
Date: February 8, 2001
Creator: Wilson, Michael L.
Partner: UNT Libraries Government Documents Department

Vertical Drop Of 21-Pwr Waste Package On Unyielding Surface

Description: The objective of this calculation is to determine the structural response of a 21-PWR (pressurized-water reactor) Waste Package (WP) subjected to the 2-m vertical drop on an unyielding surface at three different temperatures. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities in two different WP components. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP considered in this calculation, and all obtained results are valid for that design only.
Date: January 29, 2001
Creator: Mastilovic, S.; Scheider, A. & Bennett, S.M.
Partner: UNT Libraries Government Documents Department


Description: A heatpipe-cooled fast reactor concept has been under development at Los Alamos National Laboratory for the past several years, to be used as a power source for nuclear electric propulsion (NEP) or as a planetary surface power system. The reactor core consists of an array of modules that are held together by a core lateral restraint system. Each module comprises a single heatpipe surrounded by 3-6 clad fuel pins. As part of the design development and performance assessment activities for these reactors, specialized methods and models have been developed to perform thermal and stress analyses of the core modules. The methods have been automated so that trade studies can be readily performed, looking at design options such as module size, heatpipe and clad thickness, use of sleeves to contain the fuel, material type, etc. This paper describes the methods and models that have been developed, and presents thermal and stress analysis results for a Mars surface power system and a NEP power source.
Date: January 1, 2001
Creator: Kapernick, R. J. (Richard J.) & Guffee, R. M. (Ray M.)
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

LTA Physics Design: Description of All MOX Pin LTA Design

Description: In this document issued according to Work Release 02. P. 99-lb the results of neutronics studies of <<100%Pu>> MOX LTA design are presented. The parametric studies of infinite MOX-UOX grids, MOX-UOX core fragments and of VVER-1000 core with 3 MOX LTAs are performed. The neutronics parameters of MOX fueled core have been performed for the chosen design MOX LTA using the Russian 3D code BIPR-7A and 2D code PERMAK-A with the constants prepared by the cell spectrum code TVS-M.
Date: September 28, 2001
Creator: Pavlovichev, A.M.
Partner: UNT Libraries Government Documents Department

Calculation of Quad-Cities Central Bundle Documented by the U.S. in FY98 Using Russian Computer Codes

Description: The report presents calculation results of isotopic composition of irradiated fuel performed for the Quad Cities-1 reactor bundle with UO{sub 2} and MOX fuel. The MCU-REA code was used for calculations. The code is developed in Kurchatov Institute, Russia. The MCU-REA results are compared with the experimental data and HELIOS code results.
Date: June 19, 2001
Creator: Pavlovichev, A.M.
Partner: UNT Libraries Government Documents Department

Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit

Description: This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.
Date: August 2, 2001
Creator: Wagner, J. C.
Partner: UNT Libraries Government Documents Department

Corrosion Tests of LWR Fuels - Nuclide Release

Description: Two BWR fuels [64 and 71 (MWd)/kgU], one of which contained 2% Gd, and two PWR fuels [30 and 45 (MWd)/kgU], are tested by dripping groundwater on the fuels under oxidizing and hydrologically unsaturated conditions for times ranging from 2.4 to 8.2 yr at 90 C. The {sup 99}Tc, {sup 129}I, {sup 137}Cs, {sup 97}Mo, and {sup 90}Sr releases are presented to show the effects of long reaction times and of gadolinium on nuclide release. This investigation showed that the five nuclides at long reaction times have similar fractional release rates and that the presence of 2% Gd reduced the {sup 99}Tc cumulative release fraction by about an order of magnitude over that of a fuel with a similar burnup.
Date: December 14, 2001
Creator: Finn, P.A.; Tsai, Y. & Cunnane, J.C.
Partner: UNT Libraries Government Documents Department


Description: Helium-cooled reactor technologies offer significant advantages in accomplishing the waste transmutation process. They are ideally suited for use with thermal, epithermal, or fast neutron energy spectra. They can provide a relatively hard thermal neutron spectrum for transmutation of fissionable materials such as Pu-239 using ceramic-coated transmutation fuel particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of transmutation with minimal or no intermediate reprocessing, (2) enhance passive decay heat removal via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly resistant to corrosion for long times. They also offer the possibility for the use of epithermal neutrons that can interact with transmutable materials more effectively because of the large atomic cross sections in this energy domain. A fast spectrum may be useful for deep burnup of certain minor actinides. For this application, helium is essentially transparent to neutrons, does not degrade neutron energies, and offers the hardest possible neutron energy environment. In this paper, we report results from recent work on materials transmutation balances, safety, value to a geological repository, and economic considerations.
Date: March 1, 2001
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

Core Benchmarks Descriptions

Description: Actual regulations while designing of new fuel cycles for nuclear power installations comprise a calculational justification to be performed by certified computer codes. It guarantees that obtained calculational results will be within the limits of declared uncertainties that are indicated in a certificate issued by Gosatomnadzor of Russian Federation (GAN) and concerning a corresponding computer code. A formal justification of declared uncertainties is the comparison of calculational results obtained by a commercial code with the results of experiments or of calculational tests that are calculated with an uncertainty defined by certified precision codes of MCU type or of other one. The actual level of international cooperation provides an enlarging of the bank of experimental and calculational benchmarks acceptable for a certification of commercial codes that are being used for a design of fuel loadings with MOX fuel. In particular, the work is practically finished on the forming of calculational benchmarks list for a certification of code TVS-M as applied to MOX fuel assembly calculations. The results on these activities are presented.
Date: May 24, 2001
Creator: Pavlovichev, A.M.
Partner: UNT Libraries Government Documents Department

OBEST: The Object-Based Event Scenario Tree Methodology

Description: Event tree analysis and Monte Carlo-based discrete event simulation have been used in risk assessment studies for many years. This report details how features of these two methods can be combined with concepts from object-oriented analysis to develop a new risk assessment methodology with some of the best features of each. The resultant Object-Based Event Scenarios Tree (OBEST) methodology enables an analyst to rapidly construct realistic models for scenarios for which an a priori discovery of event ordering is either cumbersome or impossible (especially those that exhibit inconsistent or variable event ordering, which are difficult to represent in an event tree analysis). Each scenario produced by OBEST is automatically associated with a likelihood estimate because probabilistic branching is integral to the object model definition. The OBEST method uses a recursive algorithm to solve the object model and identify all possible scenarios and their associated probabilities. Since scenario likelihoods are developed directly by the solution algorithm, they need not be computed by statistical inference based on Monte Carlo observations (as required by some discrete event simulation methods). Thus, OBEST is not only much more computationally efficient than these simulation methods, but it also discovers scenarios that have extremely low probabilities as a natural analytical result--scenarios that would likely be missed by a Monte Carlo-based method. This report documents the OBEST methodology, the demonstration software that implements it, and provides example OBEST models for several different application domains, including interactions among failing interdependent infrastructure systems, circuit analysis for fire risk evaluation in nuclear power plants, and aviation safety studies.
Date: March 1, 2001
Partner: UNT Libraries Government Documents Department


Description: The US Department of Energy (US DOE), under the US government's International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators, are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper.
Date: January 7, 2001
Partner: UNT Libraries Government Documents Department

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

Description: The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.
Date: June 1, 2001
Creator: Ellis, RJ
Partner: UNT Libraries Government Documents Department

Degradation and Failure Characteristics of NPP Containment Protective Coating Systems

Description: Nuclear power plants (NPPs) must ensure that the emergency core cooling system (ECCS) or safety-related containment spray system (CSS) remains capable of performing its design safety function throughout the life of the plant. This requires ensuring that long-term core cooling can be maintained following a postulated loss-of-coolant accident (LOCA). Adequate safety operation can be impaired if the protective coatings which have been applied to the concrete and steel structures within the primary containment fail, producing transportable debris which could then accumulate on BWR ECCS suction strainers or PWR ECCS sump debris screens located within the containment. This document will present the data collected during the investigation of coating specimens from plants.
Date: April 10, 2001
Creator: Sindelar, R.L.
Partner: UNT Libraries Government Documents Department

An Empirical Approach to Bounding the Axial Reactivity Effects of PWR Spent Nuclear Fuel

Description: One of the significant issues yet to be resolved for using burnup credit (BUC) for spent nuclear fuel (SNF) is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters (such as local power, fuel temperature, moderator temperature, burnable poison rod history, and soluble boron concentration) affect the isotopic inventory of fuel that is depleted in a pressurized water reactor (PWR). However, obtaining the detailed operating histories needed to model all PWR fuel assemblies to which BUC would be applied is an onerous and costly task. Simplifications therefore have been suggested that could lead to using ''bounding'' depletion parameters that could be broadly applied to different fuel assemblies. This paper presents a method for determining a set of bounding depletion parameters for use in criticality analyses for SNF.
Date: April 4, 2001
Creator: O'Leary, P. M. & Scaglione, J. M.
Partner: UNT Libraries Government Documents Department

Phenomena and Parameters Important to Burnup Credit

Description: Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given.
Date: January 10, 2001
Creator: Parks, C. V.
Partner: UNT Libraries Government Documents Department