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Westinghouse Reactor Protection System Unavailability, 1984--1995

Description: An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. Westinghouse commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.
Date: August 1, 1999
Creator: Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Rasmuson, D. & Marksberry, D.
Partner: UNT Libraries Government Documents Department

General Electric Reactor Protection System Unavailability, 1984--1995

Description: An analysis was performed of the safety-related performance of the reactor protection system (RPS) at U. S. General Electric commercial reactors during the period 1984 through 1995. RPS operational data were collected from the Nuclear Plant Reliability Data System and Licensee Event Reports. A risk-based analysis was performed on the data to estimate the observed unavailability of the RPS, based on a fault tree model of the system. Results were compared with existing unavailability estimates from Individual Plant Examinations and other reports.
Date: August 1, 1999
Creator: Eide, Steven Arvid; Calley, Michael Brennan; Gentillon, Cynthia Ann; Wierman, Thomas Edward; Hamzehee, H. & Rasmuson, D.
Partner: UNT Libraries Government Documents Department

The Graphite Isotope Ratio Method (GIRM): A Plutonium Production Verification Tool

Description: Over the lifetime of a production reactor, neutrons from the fission process not only convert U-238 into plutonium but also bring about changes in the elements of the reactor's core components. Components such as shielding, pressure vessels, coolant piping, control rods, structural supports, and, in the case of graphite moderated reactors, the solid graphite moderator are all affected. Because a reactor's total plutonium production is directly related to total neutron fluence, and, likewise, changes in the elements and isotopes of a reactor's core components are directly related to fluence; it was argued that measuring these changes could provide an accurate estimate of a reactor's total plutonium production. The U.S. Department of Energy funds a project at Pacific Northwest National Laboratory (PNNL) to develop this concept into a practical plutonium production verification tool for graphite moderated reactors. The following sections describe the GIRM project development process. The purpose of this document is to provide a simple, concise description of the graphite isotope ratio method (GIRM) for use as a verification tool in estimating a graphite-moderated reactor's total plutonium production. The description covers the theory behind the technique and how the method is actually applied.
Date: January 1, 1999
Creator: McNeece, JP; Reid, BD & Wood, TW
Partner: UNT Libraries Government Documents Department

LCEs for Naval Reactor Benchmark Calculations

Description: The purpose of this engineering calculation is to document the MCNP4B2LV evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (k{sub eff}) for various critical configurations. These LCE evaluations support the development and validation of the neutronics methodology used for criticality analyses involving Naval reactor spent nuclear fuel in a geologic repository.
Date: July 19, 1999
Creator: Anderson, W.J.
Partner: UNT Libraries Government Documents Department

AXIAL SOURCE PROFILE EFFECT ON WASTE PACKAGE TRANSPORTER SHIELDING

Description: The purpose of this scoping calculation is to support preliminary design of the Waste Package (WP) transporter radiation shield configuration. Spent Nuclear Fuel (SNF) is highly radioactive and site personnel must be protected during the period that the WPs are emplaced. Personnel protection is accomplished via a heavily shielded WP transporter that moves the waste from the surface to the emplacement drift. All previous WP transporter shielding calculations have assumed a Design Basis Fuel (DBF) in which the fuel burnup is uniform (e.g. Ref. 7.3, Ref. 7.4, and Ref. 7.12). In reality, SNF burnup varies significantly from one end of the fuel assembly to the other. Since source strengths are dependent upon fuel burnup, a model which varies the fuel burnup along the assembly axis will produce a more accurate depiction of the radiation field surrounding the WP transporter. The objective of this calculation is to determine the need for using the actual axial profile, as opposed to the uniform burnup assumption, in the WP transporter shield design. The scope of the calculation is as follows: (1) Determine the impact of axial source term variation on WP transporter contact dose rates. (2) Determine appropriate shielding modifications to account for expected dose rate peaking effects. Consistent with the previous subsurface shielding analyses, this calculation considers the bounding 21 Pressurized Water Reactor (PWR) WP only. The calculation will need to be revised and extended to Boiling Water Reactor (BWR) SNF upon selection of the WP design for the License Application (LA) and availability of the source terms from the WP Operations Group.
Date: June 30, 1999
Creator: Nielsen, A.
Partner: UNT Libraries Government Documents Department

Synergistic Failure of BWR Internals

Description: Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components.
Date: October 1, 1999
Creator: Ware, Arthur Gates & Chang, T-Y
Partner: UNT Libraries Government Documents Department

Review of ENDF/B-VI Fission-Product Cross Section

Description: In response to concerns raised in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 93-2, the U.S. Department of Energy (DOE) developed a comprehensive program to help assure that the DOE maintain and enhance its capability to predict the criticality of systems throughout the complex. Tasks developed to implement the response to DNFSB recommendation 93-2 included Critical Experiments, Criticality Benchmarks, Training, Analytical Methods, and Nuclear Data. The Nuclear Data Task consists of a program of differential measurements at the Oak Ridge Electron Linear Accelerator (ORELA), precise fitting of the differential data with the generalized least-squares fitting code SAMMY to represent the data with resonance parameters using the Reich-Moore formalism along with covariance (uncertainty) information, and the development of complete evaluations for selected nuclides for inclusion in the Evaluated Nuclear Data File (ENDFB). The current ENDF/B library was developed for fast and thermal fission reactors and fusion reactors. Criticality safety practitioners recognize that many situations around the DOE complex are characterized by neutron spectra in the intermediate-energy region, as opposed to the high-energy region for fast reactors and fusion systems and the low-energy region for thermal reactors. Consequently, the Nuclear Data Task focuses primarily on the intermediate-energy region so that upgrades to existing evaluated data will remove deficiencies in the current ENDF/B evaluations. The ORELA allows high-resolution measurements in the intermediate-energy region and the SAMMY fitting code provides high quality resonance parameters in the resolved and unresolved energy range using the sophisticated Reich-Moore (RM) formalism for superior representation of the data in the intermediate energy region. In addition, the SAMMY fitting procedure provides covariance information for the resonance parameters that can be used in subsequent analyses to assess the uncertainty in calculated results and provide a better interpretation of criticality safety margins. Thus, the thrust of the Nuclear Data Task is ...
Date: January 1, 1999
Creator: Wright, R.Q.
Partner: UNT Libraries Government Documents Department

Probaability of Criticality for MOX SNF

Description: The purpose of this calculation is to provide a conservative (upper bound) estimate of the probability of criticality for mixed oxide (MOX) spent nuclear fuel (SNF) of the Westinghouse pressurized water reactor (PWR) design that has been proposed for use. with the Plutonium Disposition Program (Ref. 1, p. 2). This calculation uses a Monte Carlo technique similar to that used for ordinary commercial SNF (Ref. 2, Sections 2 and 5.2). Several scenarios, covering a range of parameters, are evaluated for criticality. Parameters specifying the loss of fission products and iron oxide from the waste package are particularly important. This calculation is associated with disposal of MOX SNF.
Date: September 28, 1999
Creator: Gottlieb, P.
Partner: UNT Libraries Government Documents Department

Simulating the Effect on Criticality of Simultaneous Matrix Degradation and Assembly Collapse for the 21 PWR Waste Package

Description: The purpose of this calculation is to evaluate the effects of fission products loss on the reactivity of commercial pressurized water reactor (PWR) spent nuclear fuel (SNF) in 21 PWR waste packages (WPs) in the event of simultaneous fuel matrix degradation and assembly collapse.
Date: September 23, 1999
Creator: Alsaed, A. A.
Partner: UNT Libraries Government Documents Department

Dose Calculations for the Codsiposal WP of HLW Glass and the Shippingport LWBR SNF

Description: The purpose of this calculation is to determine the surface dose rates of a codisposal waste package (WP) containing an intact seed assembly of the Shippingport light-water breeder reactor (LWBR) spent nuclear fuel (SNF) and the Savannah River Site (SRS) high-level waste (HLW) in glass form. The Shippingport LWBR SNF is loaded in a Department of Energy (DOE) standardized 18-in. canister. The canister is surrounded by five 4.5-m-long Hanford pour canisters containing the HLW glass. Gamma dose rate calculation for the WP containing only the HLW glass is also performed. The results will provide information about the contribution of DOE SNF to the total dose rate on the WP surfaces.
Date: November 5, 1999
Creator: Radulescu, G.
Partner: UNT Libraries Government Documents Department

Volumes, Masses, and Surface Areas for Shippingport LWBR Spent Nuclear Fuel in a DOE SNF Canister

Description: The purpose of this calculation is to estimate volumes, masses, and surface areas associated with (a) an empty Department of Energy (DOE) 18-inch diameter, 15-ft long spent nuclear fuel (SNF) canister, (b) an empty DOE 24-inch diameter, 15-ft long SNF canister, (c) Shippingport Light Water Breeder Reactor (LWBR) SNF, and (d) the internal basket structure for the 18-in. canister that has been designed specifically to accommodate Seed fuel from the Shippingport LWBR. Estimates of volumes, masses, and surface areas are needed as input to structural, thermal, geochemical, nuclear criticality, and radiation shielding calculations to ensure the viability of the proposed disposal configuration.
Date: October 22, 1999
Creator: Davis, J.W.
Partner: UNT Libraries Government Documents Department

Chemical and Radiochemical Constituents in Water from Wells in the Vicinity of the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, Idaho, 1996

Description: The U.S. Geological Survey, in response to a request from the U.S. Department of Energy's Pittsburgh Naval Reactors Office, Idaho Branch Office (IBO), samples water from 13 wells during 1996 as part of a long-term project to monitor water quality to the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility (NRF), Idaho National Engineering and Environmental Laboratory, Idaho. The IBO requires information about the mobility of radionuclide- and chemical-waste constituents in the Snake River Plain aquifer. Waste-constituent mobility is determined principally by (1) the rate and direction of ground-water flow; (2) the locations, quantities, and methods of waste disposal; (3) waste-constituents chemistry; and (4) the geochemical processes taking place in the aquifer. The purpose of the data-collection program is to provide IBO with water-chemistry data to evaluate the effect of NRF activities on the water quality of the Snake River Plain aquifer. Water samples were analyzed for naturally occurring constituents and man-made contaminants.
Date: October 1, 1999
Creator: Knobel, L. L.; Bartholomay, R. C.; Tucker, B. J. & Williams, L. M.
Partner: UNT Libraries Government Documents Department

Development of small, fast reactor core designs using lead-based coolant.

Description: A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations.
Date: June 11, 1999
Creator: Cahalan, J. E.; Hill, R. N.; Khalil, H. S. & Wade, D. C.
Partner: UNT Libraries Government Documents Department

Age-Related Degradation of Nuclear Power Plant Structures and Components

Description: This paper summarizes and highlights the results of the initial phase of a research project on the assessment of aged and degraded structures and components important to the safe operation of nuclear power plants (NPPs). A review of age-related degradation of structures and passive components at NPPs was performed. Instances of age-related degradation have been collected and reviewed. Data were collected from plant generated documents such as Licensing Event Reports, NRC generic communications, NUREGs and industry reports. Applicable cases of degradation occurrences were reviewed and then entered into a computerized database. The results obtained from the review of degradation occurrences are summarized and discussed. Various trending analyses were performed to identify which structures and components are most affected, whether degradation occurrences are worsening, and what was the most common aging mechanisms. The paper also discusses potential aging issues and degradation-susceptible structures and passive components which would have the greatest impact on plant risk.
Date: March 29, 1999
Creator: Braverman, J.; Chang, T.-Y.; Chokshi, N.; Hofmayer, C.; Morante, R. & Shteyngart, S.
Partner: UNT Libraries Government Documents Department

IWTS metal-water reaction rate evaluation (Fauske and Associates report 99-26)

Description: The report presents a thermal stability analysis of partially metallic particulate in two IWTS components, the knock out pot and settlers. Particulate in the knock out pot is thermally stable for combinations of average particle size and metal mass fraction which appear realistic. Particulate in the settlers is thermally stable when a realistic account of particle reactions over time, metal fraction, and size distribution is considered.
Date: July 29, 1999
Creator: DUNCAN, D.R.
Partner: UNT Libraries Government Documents Department

Studies of Flexible MOX/LEU Fuel Cycles

Description: This project was a collaborative effort involving researchers from Oak Ridge National Laboratory and North Carolina State University as well as Texas A and M University. The background, briefly, is that the US is planning to use some of its excess weapons Plutonium (Pu) to make mixed-oxide (MOX) fuel for existing light-water reactors (LWRs). Considerable effort has already gone into designing fuel assemblies and core loading patterns for the transition from full-uranium cores to partial-MOX and full-MOX cores. However, these designs have assumed that any time a reactor needs MOX assemblies, these assemblies will be supplied. In reality there are many possible scenarios under which this supply could be disrupted. It therefore seems prudent to verify that a reactor-based Pu-disposition program could tolerate such interruptions in an acceptable manner. Such verification was the overall aim of this project. The task assigned to the Texas A and M team was to use the HELIOS code to develop libraries of two-group homogenized cross sections for the various assembly designs that might be used in a Westinghouse Pressurized Water Reactor (PWR) that is burning weapons-grade MOX fuel. The NCSU team used these cross sections to develop optimized loading patterns under several assumed scenarios. Their results are documented in a companion report.
Date: March 1, 1999
Creator: Adams, M.L. & Alonso-Vargas, G.
Partner: UNT Libraries Government Documents Department

ASME code and ratcheting in piping components. Final technical report

Description: The main objective of this research is to develop an analysis program which can accurately simulate ratcheting in piping components subjected to seismic or other cyclic loads. Ratcheting is defined as the accumulation of deformation in structures and materials with cycles. This phenomenon has been demonstrated to cause failure to piping components (known as ratcheting-fatigue failure) and is yet to be understood clearly. The design and analysis methods in the ASME Boiler and Pressure Vessel Code for ratcheting of piping components are not well accepted by the practicing engineering community. This research project attempts to understand the ratcheting-fatigue failure mechanisms and improve analysis methods for ratcheting predictions. In the first step a state-of-the-art testing facility is developed for quasi-static cyclic and seismic testing of straight and elbow piping components. A systematic testing program to study ratcheting is developed. Some tests have already been performed an d the rest will be completed by summer'99. Significant progress has been made in the area of constitutive modeling. A number of sophisticated constitutive models have been evaluated in terms of their simulations for a broad class of ratcheting responses. From the knowledge gained from this evaluation study two improved models are developed. These models are demonstrated to have promise in simulating ratcheting responses in piping components. Hence, implementation of these improved models in widely used finite element programs, ANSYS and/or ABAQUS, is in progress. Upon achieving improved finite element programs for simulation of ratcheting, the ASME Code provisions for ratcheting of piping components will be reviewed and more rational methods will be suggested. Also, simplified analysis methods will be developed for operability studies of piping components and systems. Some of the future works will be performed under the auspices of the Center for Nuclear Power Plant Structures, Equipment and Piping. Proposals for future funding ...
Date: May 14, 1999
Creator: Hassan, T. & Matzen, V.C.
Partner: UNT Libraries Government Documents Department

Effects of Fuel Ethanol Use on Fuel-Cycle Energy and Greenhouse Gas Emissions

Description: We estimated the effects on per-vehicle-mile fuel-cycle petroleum use, greenhouse gas (GHG) emissions, and energy use of using ethanol blended with gasoline in a mid-size passenger car, compared with the effects of using gasoline in the same car. Our analysis includes petroleum use, energy use, and emissions associated with chemicals manufacturing, farming of corn and biomass, ethanol production, and ethanol combustion for ethanol; and petroleum use, energy use, and emissions associated with petroleum recovery, petroleum refining, and gasoline combustion for gasoline. For corn-based ethanol, the key factors in determining energy and emissions impacts include energy and chemical usage intensity of corn farming, energy intensity of the ethanol plant, and the method used to estimate energy and emissions credits for co-products of corn ethanol. The key factors in determining the impacts of cellulosic ethanol are energy and chemical usage intensity of biomass farming, ethanol yield per dry ton of biomass, and electricity credits in cellulosic ethanol plants. The results of our fuel-cycle analysis for fuel ethanol are listed below. Note that, in the first half of this summary, the reductions cited are per-vehicle-mile traveled using the specified ethanol/gasoline blend instead of conventional (not reformulated) gasoline. The second half of the summary presents estimated changes per gallon of ethanol used in ethanol blends. GHG emissions are global warming potential (GWP)-weighted, carbon dioxide (CO2)-equivalent emissions of CO2, methane (CH4), and nitrous oxide (N2O).
Date: February 8, 1999
Creator: Saricks, C.; Santini, D. & Wang, M.
Partner: UNT Libraries Government Documents Department

Fuel Summary Report: Shippingport Light Water Breeder Reactor

Description: The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.
Date: January 1, 1999
Creator: Illum, D.B.; Olson, G.L. & McCardell, R.K.
Partner: UNT Libraries Government Documents Department

Elevated-Temperature Mechanical Properties of an Advanced Type 316 Stainless Steel

Description: Type 316FR stainless steel is a candidate material for the Japanese Demonstration Fast Breeder Reactor Plant to be built in Japan early in the next century. Like type 316L(N), it is a low-carbon grade of stainless steel with a more closely specified nitrogen content and chemistry optimized to enhance elevated-temperature performance. Early in 1994, under sponsorship of The Japan Atomic Power Company, work was initiated at Oak Ridge National Laboratory (ORNL) aimed at obtaining an elevated-temperature mechanical-properties database on a single heat of this material. The product form was 50-mm plate manufactured by the Nippon Steel Corporation. Data include results from long-term creep-rupture tests conducted at temperatures of 500 to 600 C with test times up to nearly 40,000 h, continuous-cycle strain-controlled fatigue test results over the same temperature range, limited creep-fatigue data at 550 and 600 C, and tensile test properties from room temperature to 650 C. The ORNL data were compared with data obtained from several different heats and product forms of this material obtained at Japanese laboratories. The data were also compared with results from predictive equations developed for this material and with data available for type 316 and type 316L(N) stainless steel.
Date: August 1, 1999
Creator: Brinkman, C.R.
Partner: UNT Libraries Government Documents Department

Comparison among five hydrodynamic codes with a diverging-converging nozzle experiment

Description: A realistic open-cycle gas-core nuclear rocket simulation model must be capable of a self-consistent nozzle calculation in conjunction with coupled radiation and neutron transport in three spatial dimensions. As part of the development effort for such a model, five hydrodynamic codes were used to compare with a converging-diverging nozzle experiment. The codes used in the comparison are CHAD, FLUENT, KIVA2, RAMPANT, and VNAP2. Solution accuracy as a function of mesh size is important because, in the near term, a practical three-dimensional simulation model will require rather coarse zoning across the nozzle throat. In the study, four different grids were considered. (1) coarse, radially uniform grid, (2) coarse, radially nonuniform grid, (3) fine, radially uniform grid, and (4) fine, radially nonuniform grid. The study involves code verification, not prediction. In other words, the authors know the solution they want to match, so they can change methods and/or modify an algorithm to best match this class of problem. In this context, it was necessary to use the higher-order methods in both FLUENT and RAMPANT. In addition, KIVA2 required a modification that allows significantly more accurate solutions for a converging-diverging nozzle. From a predictive point of view, code accuracy with no tuning is an important result. The most accurate codes on a coarse grid, CHAD and VNAP2, did not require any tuning. Their main comparison among the codes was the radial dependence of the Mach number across the nozzle throat. All five codes yielded a very similar solution with fine, radially uniform and radially nonuniform grids. However, the codes yielded significantly different solutions with coarse, radially uniform and radially nonuniform grids. For all the codes, radially nonuniform zoning across the throat significantly increased solution accuracy with a coarse mesh. None of the codes agrees in detail with the weak shock located downstream of ...
Date: September 1, 1999
Creator: Thode, L. E.; Cline, M. C.; DeVolder, B. G.; Sahota, M. S. & Zerkle, D. K.
Partner: UNT Libraries Government Documents Department

Preliminary results from Charpy impact testing of irradiated JPDR weld metal and commissioning of a facility for machining of irradiated materials

Description: Forty two full-size Charpy specimens were machined from eight trepans that originated from the Japan Power Demonstration Reactor (JPDR). They were also successfully tested and the preliminary results are presented in this report. The trends appear to be reasonable with respect to the location of the specimens with regards to whether they originated from the beltline or the core regions of the vessel, and also whether they were from the inside or outside regions of the vessel wall. A short synopsis regarding commissioning of the facility to machine irradiated materials is also provided.
Date: September 1, 1999
Creator: Iskander, S.K.; Hutton, J.T.; Creech, L.E.; Nanstad, R.K.; Manneschmidt, E.T.; Rosseel, T.M. et al.
Partner: UNT Libraries Government Documents Department

Synergistic failure of BWR internals

Description: Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components.
Date: October 25, 1999
Creator: Ware, A. G. & Chang, T. Y.
Partner: UNT Libraries Government Documents Department