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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

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

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

Radiation degradation in EPICOR-2 ion exchange resins

Description: The Low-Level Waste Data base Development -- EPICOR-II Resin/Liner Investigation Program funded by the US Nuclear Regulatory Commission is investigating chemical and physical conditions for organic ion exchange resins contained in several EPICOR-II prefilters. Those prefilters were used during cleanup of contaminated water from the Three Mile Island Nuclear Power Station after the March 1979 accident. The work was performed by EG G Idaho, Inc. at the Idaho Engineering Laboratory. This is the final report of this task and summarizes results and analyses of three samplings of ion exchange resins from prefilters PF-8 and -20. Results are compared with baseline data from tests performed on unirradiated resins supplied by Epicor, Inc. to determine the extent of degradation due to the high internal radiation dose received by the organic resins. Results also are compared with those of other researchers. 18 refs., 23 figs., 7 tabs.
Date: September 1, 1990
Creator: McConnell, J.W. Jr.; Johnson, D.A. & Sanders, R.D. Sr.
Partner: UNT Libraries Government Documents Department

Recommendations for the shallow-crack fracture toughness testing task within the HSST (Heavy-Section Steel Technology) Program

Description: Recommendations for Heavy-Section Steel Technology Program's investigation into the influence of crack depth on the fracture toughness of a steel prototypic of those in a reactor pressure vessel are included in this report. The motivation for this investigation lies in the fact that probabilistic fracture mechanics evaluations show that shallow flaws play a dominant role in the likelihood of vessel failure, and shallow-flaw specimens have exhibited an elevated toughness compared with conventional deep-notch fracture toughness specimens. Accordingly, the actual margin of safety of vessels may be greater than that predicted using existing deep-notch fracture-toughness results. The primary goal of the shallow-crack project is to investigate the influence of crack depth on fracture toughness under conditions prototypic of a reactor vessel. A limited data base of fracture toughness values will be assembled using a beam specimen of prototypic reactor vessel material and with a depth of 100 mm (4 in.). This will permit comparison of fracture-toughness data from deep-cracked and shallow-crack specimens, and this will be done for several test temperatures. Fracture-toughness data will be expressed in terms of the stress-intensity factor and crack-tip-opening displacement. Results of this investigation are expected to improve the understanding of shallow-flaw behavior in pressure vessels, thereby providing more realistic information for application to the pressurized-thermal shock issues. 33 refs., 17 figs.
Date: September 1, 1990
Creator: Theiss, T.J. (Oak Ridge National Lab., TN (USA))
Partner: UNT Libraries Government Documents Department

The Seismic Category I Structures Program results for FY 1987

Description: The accomplishments of the Seismic Category I Structures Program for FY 1987 are summarized. These accomplishments include the quasi-static load cycle testing of large shear wall elements, an extensive analysis of previous data to determine if equivalent linear analytical models can predict the response of damaged shear wall structures, and code committee activities. In addition, previous testing and results that led to the FY 1987 program plan are discussed and all previous data relating to shear wall stiffness are summarized. Because separate reports have already summarized the experimental and analytical work in FY 1987, this report will briefly highlight this work and the appropriate reports will be references for a more detailed discussion. 12 refs., 23 figs., 18 tabs.
Date: October 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

Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station

Description: This study presents the results of a comparison of a previous decommissioning cost study by Pacific Northwest Laboratory (PNL) and a recent decommissioning cost study of TLG Engineering, Inc., for the same commercial nuclear power reactor station. The purpose of this comparative analysis on the same plant is to determine the reasons why subsequent estimates for similar plants by others were significantly higher in cost and external occupational radiation exposure (ORE) than the PNL study. The primary purpose of the original study by PNL (NUREG/CR-0672) was to provide information on the available technology, the safety considerations, and the probable costs and ORE for the decommissioning of a large boiling water reactor (BWR) power station at the end of its operating life. This information was intended for use as background data and bases in the modification of existing regulations and in the development of new regulations pertaining to decommissioning activities. It was also intended for use by utilities in planning for the decommissioning of their nuclear power stations. The TLG study, initiated in 1987 and completed in 1989, was for the same plant, Washington Public Supply System's Unit 2 (WNP-2), that PNL used as its reference plant in its 1980 decommissioning study. Areas of agreement and disagreement are identified, and reasons for the areas of disagreement are discussed. 31 refs., 3 figs., 22 tabs.
Date: December 1, 1990
Creator: Konzek, G.J. & Smith, R.I. (Pacific Northwest Lab., Richland, WA (USA))
Partner: UNT Libraries Government Documents Department

Depressurization as an accident management strategy to minimize the consequences of direct containment heating

Description: Probabilistic Risk Assessments (PRAs) have identified severe accidents for nuclear power plants that have the potential to cause failure of the containment through direct containment heating (DCH). Prevention of DCH or mitigation of its effects may be possible using accident management strategies that intentionally depressurize the reactor coolant system (RCS). The effectiveness of intentional depressurization during a station blackout TMLB' sequence was evaluated considering the phenomenological behavior, hardware performance, and operational performance. Phenomenological behavior was calculated using the SCDAP/RELAP5 severe accident analysis code. Two strategies to mitigate DCH by depressurization of the RCS were considered. One strategy, called early depressurization, assumed that the reactor head vent and pressurizer power-operated relief valves (PORVs) were latched open at steam generator dryout. The second strategy, called late depression, assumed that the head vent and PORVs were latched open at a core exit temperature of {approximately}922 K (1200{degree}F). Depressurization of the RCS to a low value that may mitigate DCH was predicted prior to reactor pressure vessel breach for both early and late depressurization. The strategy of late depressurization is preferred over early depressurization because there are greater opportunities to recover plant functions prior to core damage and because failure uncertainties are lessened. 22 refs., 38 figs., 6 tabs.
Date: October 1, 1990
Creator: Hanson, D.J.; Golden, D.W.; Chambers, R.; Miller, J.D.; Hallbert, B.P. & Dobbe, C.A. (EG and G Idaho, Inc., Idaho Falls, ID (USA))
Partner: UNT Libraries Government Documents Department

Low-level radioactive waste disposal facility closure

Description: Part I of this report describes and evaluates potential impacts associated with changes in environmental conditions on a low-level radioactive waste disposal site over a long period of time. Ecological processes are discussed and baselines are established consistent with their potential for causing a significant impact to low-level radioactive waste facility. A variety of factors that might disrupt or act on long-term predictions are evaluated including biological, chemical, and physical phenomena of both natural and anthropogenic origin. These factors are then applied to six existing, yet very different, low-level radioactive waste sites. A summary and recommendations for future site characterization and monitoring activities is given for application to potential and existing sites. Part II of this report contains guidance on the design and implementation of a performance monitoring program for low-level radioactive waste disposal facilities. A monitoring programs is described that will assess whether engineered barriers surrounding the waste are effectively isolating the waste and will continue to isolate the waste by remaining structurally stable. Monitoring techniques and instruments are discussed relative to their ability to measure (a) parameters directly related to water movement though engineered barriers, (b) parameters directly related to the structural stability of engineered barriers, and (c) parameters that characterize external or internal conditions that may cause physical changes leading to enhanced water movement or compromises in stability. Data interpretation leading to decisions concerning facility closure is discussed. 120 refs., 12 figs., 17 tabs.
Date: November 1, 1990
Creator: White, G.J.; Ferns, T.W.; Otis, M.D.; Marts, S.T.; DeHaan, M.S.; Schwaller, R.G. et al.
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

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

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