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Erosion/corrosion-induced pipe wall thinning in US Nuclear Power Plants

Description: Erosion/corrosion in single-phase piping systems was not clearly recognized as a potential safety issue before the pipe rupture incident at the Surry Power Station in December 1986. This incident reminded the nuclear industry and the regulators that neither the US Nuclear Regulatory Commission (NRC) nor Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code require utilities to monitor erosion/corrosion in the secondary systems of nuclear power plants. This report provides a brief review of the erosion/corrosion phenomenon and its major occurrence in nuclear power plants. In addition, efforts by the NRC, the industry, and the ASME Section XI Committee to address this issue are described. Finally, results of the survey and plant audits conducted by the NRC to assess the extent of erosion/corrosion-induced piping degradation and the status of program implementation regarding erosion/corrosion monitoring are discussed. This report will support a staff recommendation for an additional regulatory requirement concerning erosion/corrosion monitoring. 21 refs., 3 tabs.
Date: April 1, 1989
Creator: Wu, P.C.
Partner: UNT Libraries Government Documents Department

Cobalt-60 simulation of LOCA (loss of coolant accident) radiation effects

Description: The consequences of simulating nuclear reactor loss of coolant accident (LOCA) radiation effects with Cobalt-60 gamma ray irradiators have been investigated. Based on radiation induced damage in polymer base materials, it was demonstrated that electron/photon induced radiation damage could be related on the basis of average absorbed radiation dose. This result was used to estimate the relative effectiveness of the mixed beta/gamma LOCA and Cobalt-60 radiation environments to damage both bare and jacketed polymer base electrical insulation materials. From the results obtained, it is concluded that present simulation techniques are a conservative method for simulating LOCA radiation effects and that the practices have probably substantially overstressed both bare and jacketed materials during qualification testing. 9 refs., 8 figs., 5 tabs.
Date: July 1, 1989
Creator: Buckalew, W.H.
Partner: UNT Libraries Government Documents Department

Aging of nuclear station diesel generators: Evaluation of operating and expert experience: Workshop

Description: Pacific Northwest Laboratory (PNL) evaluated operational and expert experience pertaining to the aging degradation of diesel generators in nuclear service. The research, sponsored by the US Nuclear Regulatory Commission (NRC), identified and characterized the contribution of aging to emergency diesel generator failures. This report, Volume II, reports the results of an industry-wide workshop held on May 28 and 29, 1986, to discuss the technical issues associated with aging of nuclear service emergency diesel generators. The technical issues discussed most extensively were: man/machine interfaces, component interfaces, thermal gradients of startup and cooldown and the need for an accurate industry database for trend analysis of the diesel generator system.
Date: August 1, 1987
Creator: Hoopingarner, K.R. & Vause, J.W.
Partner: UNT Libraries Government Documents Department

Aging of nuclear station diesel generators: Evaluation of operating and expert experience: Phase 1, Study

Description: Pacific Northwest Laboratory evaluated operational and expert experience pertaining to the aging degradation of diesel generators in nuclear service. The research, sponsored by the US Nuclear Regulatory Commission (NRC), identified and characterized the contribution of aging to emergency diesel generator failures. This report, Volume I, reviews diesel-generator experience to identify the systems and components most subject to aging degradation and isolates the major causes of failure that may affect future operational readiness. Evaluations show that as plants age, the percent of aging-related failures increases and failure modes change. A compilation is presented of recommended corrective actions for the failures identified. This study also includes a review of current, relevant industry programs, research, and standards. Volume II reports the results of an industry-wide workshop held on May 28 and 29, 1986 to discuss the technical issues associated with aging of nuclear service emergency diesel generators.
Date: August 1, 1987
Creator: Hoopingarner, K.R.; Vause, J.W.; Dingee, D.A. & Nesbitt, J.F.
Partner: UNT Libraries Government Documents Department

Component Fragility Research Program: Phase 1, Demonstration tests: Volume 2, Appendices

Description: Appendices are presented which contain information concerning: details of controller and relay installation; resonance search transmissibility plots; time-history data from runs 17, 31, 46, and 56; and response spectra from runs 17, 31, 46, and 56. (JDB)
Date: August 1, 1987
Creator: Holman, G.S.; Chou, C.K.; Shipway, G.D. & Glozman, V.
Partner: UNT Libraries Government Documents Department

Beta and gamma dose calculations for PWR and BWR containments

Description: Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 /times/ 10/sup 8/ rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 /times/ 10/sup 8/ rad equipment qualification test region. 8 refs., 23 figs., 12 tabs.
Date: July 1, 1989
Creator: King, D.B.
Partner: UNT Libraries Government Documents Department

Component Fragility Research Program: Phase 1 component prioritization

Description: Current probabilistic risk assessment (PRA) methods for nuclear power plants utilize seismic ''fragilities'' - probabilities of failure conditioned on the severity of seismic input motion - that are based largely on limited test data and on engineering judgment. Under the NRC Component Fragility Research Program (CFRP), the Lawrence Livermore National Laboratory (LLNL) has developed and demonstrated procedures for using test data to derive probabilistic fragility descriptions for mechanical and electrical components. As part of its CFRP activities, LLNL systematically identified and categorized components influencing plant safety in order to identify ''candidate'' components for future NRC testing. Plant systems relevant to safety were first identified; within each system components were then ranked according to their importance to overall system function and their anticipated seismic capacity. Highest priority for future testing was assigned to those ''very important'' components having ''low'' seismic capacity. This report describes the LLNL prioritization effort, which also included application of ''high-level'' qualification data as an alternate means of developing probabilistic fragility descriptions for PRA applications.
Date: June 1, 1987
Creator: Holman, G.S. & Chou, C.K.
Partner: UNT Libraries Government Documents Department

Component Fragility Research Program: Phase 1, Demonstration tests: Volume 1, Summary report

Description: This report describes tests performed in Phase I of the NRC Component Fragility Research Program. The purpose of these tests was to demonstrate procedures for characterizing the seismic fragility of a selected component, investigating how various parameters affect fragility, and finally using test data to develop practical fragility descriptions suitable for application in probabilistic risk assessments. A three-column motor control center housing motor controllers of various types and sizes as well as relays of different types and manufacturers was subjected to seismic input motions up to 2.5g zero period acceleration. To investigate the effect of base flexibility on the structural behavior of the MCC and on the functional behavior of the electrical devices, multiple tests were performed on each of four mounting configurations: four bolts per column with top bracking, four bolts per column with no top brace, four bolts per column with internal diagonal bracking, and two bolts per column with no top or internal bracking. Device fragility was characterized by contact chatter correlated to local in-cabinet response at the device location. Seismic capacities were developed for each device on the basis of local input motion required to cause chatter; these results were then applied to develop probabilistic fragility curves for each type of device, including estimates of the ''high-confidence low probability of failure'' capacity of each.
Date: August 1, 1987
Creator: Holman, G.S.; Chou, C.K.; Shipway, G.D. & Glozman, V.
Partner: UNT Libraries Government Documents Department

Fuel performance annual report for 1986

Description: This annual report, the ninth in a series, provides a brief description of fuel performance during 1986 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related U.S. Nuclear Regulatory Commission evaluations are included. 550 refs., 12 figs., 31 tabs.
Date: March 1, 1988
Creator: Bailey, W.J. & Wu, S.
Partner: UNT Libraries Government Documents Department

Results of crack-arrest tests on two irradiated high-copper welds

Description: The objective of this study was to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288{degree}C to an average fluence of 1.9 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV). Evaluation of the results shows that the neutron-irradiation-induced crack-arrest toughness temperature shift is about the same as the Charpy V-notch impact temperature shift at the 41-J energy level. The shape of the lower-bound curves (for the range of test temperatures covered) did not seem to have been altered by irradiation compared to those of the ASME K{sub Ia} curve. 9 refs., 21 figs., 10 tabs.
Date: December 1, 1990
Creator: Iskander, S.K.; Corwin, W.R. & Nanstead, R.K. (Oak Ridge National Lab., TN (USA))
Partner: UNT Libraries Government Documents Department

Severe accident testing of electrical penetration assemblies

Description: This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs.
Date: November 1, 1989
Creator: Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))
Partner: UNT Libraries Government Documents Department

Static load cycle testing of a low-aspect-ratio four-inch wall, TRG-type structure, TRG-5-4 (1. 0, 0. 56)

Description: This report is the second in a series of test reports that details the quasi-static cyclic testing of low height-to-length aspect ratio reinforced concrete structures. The test structures were designed according to the recommendations of a technical review group for the US Nuclear Regulatory Commission sponsored Seismic Category I Structures Program. The structure tested and reported here had 4-in.-thick shear and end walls, and the elastic deformation was dominated by shear. The background of the program and previous results are given for completeness. Details of the geometry, material property tests, construction history, ultrasonic testing, and modal testing to find the undamaged dynamic characteristics of the structures are given. Next, the static test procedure and results in terms of stiffness and load deformation behavior are given. Finally, results are shown relative to other known results, and conclusions are presented. 33 refs., 140 figs., 13 tabs.
Date: November 1, 1990
Creator: Farrar, C.R.; Bennett, J.G.; Dunwoody, W.E. (Los Alamos National Lab., NM (USA)) & Baker, W.E. (New Mexico Univ., Albuquerque, NM (USA))
Partner: UNT Libraries Government Documents Department

Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors

Description: Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs.
Date: November 1, 1989
Creator: Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A. et al.
Partner: UNT Libraries Government Documents Department

Source term calculations for assessing radiation dose to equipment

Description: This study examines results of analyses performed with the Source Term Code Package to develop updated source terms using NUREG-0956 methods. The updated source terms are to be used to assess the adequacy of current regulatory source terms used as the basis for equipment qualification. Time-dependent locational distributions of radionuclides within a containment following a severe accident have been developed. The Surry reactor has been selected in this study as representative of PWR containment designs. Similarly, the Peach Bottom reactor has been used to examine radionuclide distributions in boiling water reactors. The time-dependent inventory of each key radionuclide is provided in terms of its activity in curies. The data are to be used by Sandia National Laboratories to perform shielding analyses to estimate radiation dose to equipment in each containment design. See NUREG/CR-5175, Beta and Gamma Dose Calculations for PWR and BWR Containments.'' 6 refs., 11 tabs.
Date: July 1, 1989
Creator: Denning, R.S.; Freeman-Kelly, R.; Cybulskis, P. & Curtis, L.A.
Partner: UNT Libraries Government Documents Department

Estimation of fracture toughness of cast stainless steels during thermal aging in LWR systems

Description: A procedure and correlations are presented for predicting the change in fracture toughness of cast stainless steel components due to thermal aging during service in light water rectors (LWRs) at 280--330{degrees}C (535--625{degrees}F). The fracture toughness J-R curve and Charpy-impact energy of aged cast stainless steels are estimated from known mineral in formation. Fracture toughness of a specific cast stainless steel is estimated from the extent and kinetics of thermal embrittlement. The extent of thermal embrittlement is characterized by the room-temperature normalized'' Charpy-impact energy. A correlation for the extent of embrittlement at saturation,'' i.e., the minimum impact energy that would be achieved for the material after long-term aging, is given in terms of a material parameter, {Phi}, which is determined from the chemical composition. The fracture toughness J-R curve for the material is then obtained from correlations between room-temperature Charpy-impact energy and fracture toughness parameters. Fracture toughness as a function of time and temperature of reactor service is estimated from the kinetics of thermal embrittlement, which is determined from chemical composition. A common lower-bound'' J-R curve for cast stainless steels with unknown chemical composition is also defined for a given material specification, ferrite content, and temperature. Examples for estimating impact strength and fracture toughness of cast stainless steel components during reactor service are describes. 24 refs., 39 figs., 2 tabs.
Date: June 1, 1991
Creator: Chopra, O.K. (Argonne National Lab., IL (USA))
Partner: UNT Libraries Government Documents Department

Round-robin pretest analyses of a 1:6-scale reinforced concrete containment model subject to static internal pressurization

Description: Analyses of a 1:6-scale reinforced concrete containment model that will be tested to failure at Sandia National Laboratories in the spring of 1987 were conducted by the following organizations in the United States and Europe: Sandia National Laboratories (USA), Argonne National Laboratory (USA), Electric Power Research Institute (USA), Commissariat a L'Energie Atomique (France), HM Nuclear Installations Inspectorate (UK), Comitato Nazionale per la ricerca e per lo sviluppo dell'Energia Nucleare e delle Energie Alternative (Italy), UK Atomic Energy Authority, Safety and Reliability Directorate (UK), Gesellschaft fuer Reaktorsicherheit (FRG), Brookhaven National Laboratory (USA), and Central Electricity Generating Board (UK). Each organization was supplied with a standard information package, which included construction drawings and actual material properties for most of the materials used in the model. Each organization worked independently using their own analytical methods. This report includes descriptions of the various analytical approaches and pretest predictions submitted by each organization. Significant milestones that occur with increasing pressure, such as damage to the concrete (cracking and crushing) and yielding of the steel components, and the failure pressure (capacity) and failure mechanism are described. Analytical predictions for pressure histories of strain in the liner and rebar and displacements are compared at locations where experimental results will be available after the test. Thus, these predictions can be compared to one another and to experimental results after the test.
Date: May 1, 1987
Creator: Clauss, D.B. (ed.)
Partner: UNT Libraries Government Documents Department

Models for estimation of service life of concrete barriers in low-level radioactive waste disposal

Description: Concrete barriers will be used as intimate parts of systems for isolation of low level radioactive wastes subsequent to disposal. This work reviews mathematical models for estimating the degradation rate of concrete in typical service environments. The models considered cover sulfate attack, reinforcement corrosion, calcium hydroxide leaching, carbonation, freeze/thaw, and cracking. Additionally, fluid flow, mass transport, and geochemical properties of concrete are briefly reviewed. Example calculations included illustrate the types of predictions expected of the models. 79 refs., 24 figs., 6 tabs.
Date: September 1, 1990
Creator: Walton, J.C.; Plansky, L.E. & Smith, R.W. (EG and G Idaho, Inc., Idaho Falls, ID (USA))
Partner: UNT Libraries Government Documents Department

Results from the Nuclear Plant Aging Research Program: Their use in inspection activities

Description: The US NCR's Nuclear Plant Aging Research (NPAR) Program has determined the susceptibility to aging of components and systems, and the potential for aging to impact plant safety and availability. The NPAR Program also identified methods for detecting and mitigating aging in components. This report describes the NPAR results which can enhance NRC inspection activities. Recommendations are provided for communicating pertinent information to NRC inspectors. These recommendations are based on a detailed assessment of the NRC's Inspection Program, and feedback from resident and regional inspectors as described within. Examples of NPAR report summaries and aging inspection guides for components and systems are included. 13 refs., 3 figs., 4 tabs.
Date: September 1, 1990
Creator: Gunther, W. & Taylor, J. (Brookhaven National Lab., Upton, NY (USA))
Partner: UNT Libraries Government Documents Department

Steam generator group project: Task 13 final report: Nondestructive examination validation

Description: The Steam Generator Group Project (SGGP) was a multi-task effort using the retired-from-service Surry 2A pressurized water reactor steam generator as a test bed to investigate the reliability and effectiveness of in-service nondestructive eddy current (EC) inspection equipment and procedures. The information developed provided the technical basis for recommendations for improved in- service inspection and tube plugging criteria of steam generators. This report describes the results and analysis from Task 13--NDE Validation. The primary objective of Task 13 was to validate the EC inspection to detect and size tube defects. Additional objectives were to assess the nature and severity of tube degradation from all regions of the generator and to measure the remaining integrity of degraded specimens by burst testing. More than 550 specimens were removed from the generator and included in the validation studies. The bases for selecting the specimens and the methods and procedures used for specimen removal from the generator are reported. Results from metallurgical examinations of these specimens are presented and discussed. These examinations include visual inspection of all specimens to locate and identify tube degradation, metallographic examination of selected specimens to establish defect severity and burst testing of selected specimens to establish the remaining integrity of service-degraded tubes. Statistical analysis of the combined metallurgical and EC data to determine the probability of detection (POD) and sizing accuracy are reported along with a discussion of the factors which influenced the EC results. Finally, listings of the metallurgical and corresponding EC data bases are given. 12 refs., 141 figs., 24 tabs.
Date: August 1, 1988
Creator: Bradley, E.R.; Doctor, P.G.; Ferris, R.H. & Buchanan, J.A.
Partner: UNT Libraries Government Documents Department

Age-related degradation of Westinghouse 480-volt circuit breakers

Description: An aging assessment of Westinghouse DS-series low-voltage air circuit breakers was performed as part of the Nuclear Plant Aging Research (NPAR) program. The objectives of this study are to characterize age-related degradation within the breaker assembly and to identify maintenance practices to mitigate their effect. Since this study has been promulgated by the failures of the reactor trip breakers at the McGuire Nuclear Station in July 1987, results relating to the welds in the breaker pole lever welds are also discussed. The design and operation of DS-206 and DS-416 breakers were reviewed. Failure data from various national data bases were analyzed to identify the predominant failure modes, causes, and mechanisms. Additional operating experiences from one nuclear station and two industrial breaker-service companies were obtained to develop aging trends of various subcomponents. The responses of the utilities to the NRC Bulletin 88-01, which discusses the center pole lever welds, were analyzed to assess the final resolution of failures of welds in the reactor trips. Maintenance recommendations, made by the manufacturer to mitigate age-related degradation were reviewed, and recommendations for improving the monitoring of age-related degradation are discussed. As described in Volume 2 of this NUREG, the results from a test program to assess degradation in breaker parts through mechanical cycling are also included. The testing has characterized the cracking of center-pole lever welds, identified monitoring techniques to determine aging in breakers, and provided information to augment existing maintenance programs. Recommendations to improve breaker reliability using effective maintenance, testing, and inspection programs are suggested. 13 refs., 21 figs., 8 tabs.
Date: July 1, 1990
Creator: Subudhi, M.; Shier, W. & MacDougall, E. (Brookhaven National Lab., Upton, NY (USA))
Partner: UNT Libraries Government Documents Department

Influence of fluence rate on radiation-induced mechanical property changes in reactor pressure vessel steels

Description: This report describes a set of experiments undertaken using a 2 MW test reactor, the UBR, to qualify the significance of fluence rate to the extent of embrittlement produced in reactor pressure vessel steels at their service temperature. The test materials included two reference plates (A 302-B, A 533-B steel) and two submerged arc weld deposits (Linde 80, Linde 0091 welding fluxes). Charpy-V (C{sub v}), tension and 0.5T-CT compact specimens were employed for notch ductility, strength and fracture toughness (J-R curve) determinations, respectively. Target fluence rates were 8 {times} 10{sup 10}, 6 {times} 10{sup 11} and 9 {times} 10{sup 12} n/cm{sup 2} {minus}s{sup {minus}1}. Specimen fluences ranged from 0.5 to 3.8 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV. The data describe a fluence-rate effect which may extend to power reactor surveillance as well as test reactor facilities now in use. The dependence of embrittlement sensitivity on fluence rate appears to differ for plate and weld deposit materials. Relatively good agreement in fluence-rate effects definition was observed among the three test methods. 52 figs., 4 tabs.
Date: March 1, 1990
Creator: Hawthorne, J.R. & Hiser, A.L. (Materials Engineering Associates, Inc., Lanham, MD (USA))
Partner: UNT Libraries Government Documents Department

Technology, safety and costs of decommissioning a reference boiling water reactor power station: Technical support for decommissioning matters related to preparation of the final decommissioning rule

Description: Preparation of the final Decommissioning Rule by the Nuclear Regulatory Commission (NRC) staff has been assisted by Pacific Northwest Laboratory (PNL) staff familiar with decommissioning matters. These efforts have included updating previous cost estimates developed during the series of studies of conceptually decommissioning reference licensed nuclear facilities for inclusion in the Final Generic Environmental Impact Statement (FGEIS) on decommissioning; documenting the cost updates; evaluating the cost and dose impacts of post-TMI-2 backfits on decommissioning; developing a revised scaling formula for estimating decommissioning costs for reactor plants different in size from the reference boiling water reactor (BWR) described in the earlier study; and defining a formula for adjusting current cost estimates to reflect future escalation in labor, materials, and waste disposal costs. This report presents the results of recent PNL studies to provide supporting information in three areas concerning decommissioning of the reference BWR: updating the previous cost estimates to January 1986 dollars; assessing the cost and dose impacts of post-TMI-2 backfits; and developing a scaling formula for plants different in size than the reference plant and an escalation formula for adjusting current cost estimates for future escalation.
Date: July 1, 1988
Creator: Konzek, G.J. & Smith, R.I.
Partner: UNT Libraries Government Documents Department

Correlation of irradiation-induced transition temperature increases from C sub v and K sub Jc /K sub Ic data

Description: Reactor pressure vessel (RPV) surveillance capsules contain Charpy-V (C{sub v}) specimens, but many do not contain fracture toughness specimens; accordingly, the radiation-induced shift (increase) in the brittle-to-ductile transition region ({Delta}T) is based upon the {Delta}T determined from notch ductility (C{sub v}) tests. Since the ASME K{sub Ic} and K{sub IR} reference fracture toughness curves are shifted by the {Delta}T from C{sub v}, assurance that this {Delta}T does not underestimate {Delta}T associated with the actual irradiated fracture toughness is required to provide confidence that safety margins do not fall below assumed levels. To assess this behavior, comparisons of {Delta}T's defined by elastic-plastic fracture toughness and C{sub v} tests have been made using data from RPV base and weld metals in which irradiations were made under test reactor conditions. Using as-measure'' fracture toughness values (K{sub Jc}), average comparisons between {Delta}T(C{sub v}) and {Delta}T(K{sub Jc}) are: (a) All data: {Delta}T(K{sub Jc} 100 MPa{radical}{bar m}) = {Delta}T(C{sub v} 41 J) +10{degree}C; (b) Plates only: {Delta}T(K{sub Jc} 100 MPa{radical}{bar m}) = {Delta}T(C{sub v} 41 J) +15{degree}C; and (c) Welds only: {Delta}T(K{sub Jc} 100 MPa{radical}{bar m}) = {Delta}T(C{sub v} 41 J) {minus}1{degree}C. Fluence rate is found to have no significant effect on the relationship between {Delta}T(C{sub v}) and {Delta}T(K{sub Jc}). 12 refs., 12 figs., 5 tabs.
Date: March 1, 1990
Creator: Hiser, A.L. (Materials Engineering Associates, Inc., Lanham, MD (USA))
Partner: UNT Libraries Government Documents Department

Aging evaluation of class 1E batteries: Seismic testing

Description: This report presents the results of a seismic testing program on naturally aged class 1E batteries obtained from a nuclear plant. The testing program is a Phase 2 activity resulting from a Phase 1 aging evaluation of class 1E batteries in safety systems of nuclear power plants, performed previously as a part of the US Nuclear Regulatory Commission's Nuclear Plant Aging Research Program and reported in NUREG/CR-4457. The primary purpose of the program was to evaluate the seismic ruggedness of naturally aged batteries to determine if aged batteries could have adequate electrical capacity, as determined by tests recommended by IEEE Standards, and yet have inadequate seismic ruggedness to provide needed electrical power during and after a safe shutdown earthquake (SSE) event. A secondary purpose of the program was to evaluate selected advanced surveillance methods to determine if they were likely to be more sensitive to the aging degradation that reduces seismic ruggedness. The program used twelve batteries naturally aged to about 14 years of age in a nuclear facility and tested them at four different seismic levels representative of the levels of possible earthquakes specified for nuclear plants in the United States. Seismic testing of the batteries did not cause any loss of electrical capacity. 19 refs., 29 figs., 7 tabs.
Date: August 1, 1990
Creator: Edson, J.L. (EG and G Idaho, Inc., Idaho Falls, ID (USA))
Partner: UNT Libraries Government Documents Department