83 Matching Results

Search Results

Advanced search parameters have been applied.
in subjects: "Bwr Type Reactors"  
Results: 1 - 24 next
open access

Fatigue and environmentally assisted cracking in light water reactors

Description: Fatigue and stress corrosion cracking (SCC) for low-alloy steel used in piping and in steam generator and reactor pressure vessels have been investigated. Fatigue data were obtained on medium-sulfur-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor water, and in air. Analytical studies focused on the behavior of carbon steels in boiling water reactor (BWR) environments. Crack-growth rates of composite fracture-mechanics specime… more
Date: March 1, 1992
Creator: Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

Description: The effects of aging on Residual Heat Removal systems in Boiling Water Reactors have been studied as part of the Nuclear Plant Aging Research Program. The aging phenomena has been characterized by analyzing operating experience from various national data bases. In addition, actual plant data was obtained to supplement and validate the data base findings.
Date: January 1, 1992
Creator: Lofaro, R. J. & Aggarwal, S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Evaluation of aging degradation of structural components

Description: Irradiation embrittlement of the neutron shield tank (NST) A212 Grade B steel from the Shippingport reactor, as well as thermal embrittlement of CF-8 cast stainless steel components from the Shippingport and KRB reactors, has been characterized. Increases in Charpy transition temperature (CTT), yield stress, and hardness of the NST material in the low-temperature low-flux environment are consistent with the test reactor data for irradiations at < 232{degrees}C. The shift in CTT is not as severe… more
Date: March 1, 1992
Creator: Chopra, O. K. & Shack, W. J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Assessment of engineering plant analyzer with Peach Bottom 2 stability tests

Description: Engineering Plant Analyzer (EPA) has been developed to simulate plant transients for Boiling Water Reactor (BWR). Recently, this code has been used to simulate LaSalle-2 instability event which was initiated by a failure in the feed water heater. The simulation was performed for the scram conditions and for the postulated failure in the scram. In order to assess the capability of the EPA to simulate oscillatory flows as observed in the LaSalle event, EPA has been benchmarked with the available … more
Date: January 1, 1992
Creator: Rohatgi, U.S.; Mallen, A.N.; Cheng, H.S. & Wulff, W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Oxidation of spent fuel in air at 175 degree to 195 degree C

Description: Oxidation tests in dry air were conducted on four LWR spent fuels at 175{degrees} and 195{degrees}C to determine the effect of the fuel characteristics on the oxidation state likely to exist at the time leaching occurs in a potential repository. Weight changes were measured and samples were examined by XRD, ceramography, TEM, and TGA. Despite local variations in the grain boundary susceptibility to oxidation, all four fuels progressed toward an apparent endpoint at an oxygen-to-metal (O/M) rati… more
Date: April 1, 1992
Creator: Einziger, R. E.; Buchanan, H. C.; Thomas, L. E. (Pacific Northwest Lab., Richland, WA (United States)) & Stout, R. B. (Lawrence Livermore National Lab., CA (United States))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Status of the MELTSPREAD-1 Computer Code for the Analysis of Transient Spreading of Core Debris Melts

Description: A transient, one dimensional, finite difference computer code (MELTSPREAD-1) has been developed to predict spreading behavior of high temperature melts flowing over concrete and/or steel surfaces submerged in water, or without the effects of water if the surface is initially dry. This paper provides a summary overview of models and correlations currently implemented in the code, code validation activities completed thus far, LWR spreading-related safety issues for which the code has been applie… more
Date: January 1, 1992
Creator: Farmer, M. T.; Sienicki, J. J.; Spencer, B. W. & Chu, C. C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Characterization of core debris/concrete interactions for the Advanced Neutron Source

Description: This report provides the results of a recent study conducted to explore the molten core/concrete interaction (MCCI) issue for the Advanced Neutron Source (ANS). The need for such a study arises from the potential threats to reactor system integrity posed by MCCI. These threats include direct attack of the concrete basemat of the containment; generation and release of large quantities of gas that can pressurize the containment; the combustion threat of these gases; and the potential generation, … more
Date: February 1, 1992
Creator: Hyman, C. R. & Taleyarkhan, R. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Melt Coolability Modeling and Comparison to MACE Test Results

Description: An important question in the assessment of severe accidents in light water nuclear reactors is the ability of water to quench a molten corium-concrete interaction and thereby terminate the accident progression. As part of the Melt Attack and Coolability Experiment (MACE) Program, phenomenological models of the corium quenching process are under development. The modeling approach considers both bulk cooldown and crust-limited heat transfer regimes, as well as criteria for the pool thermal hydrau… more
Date: January 1, 1992
Creator: Farmer, M. T.; Sienicki, J. J. & Spencer, B. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Aerosols released during large-scale integral MCCI tests in the ACE Program

Description: As part of the internationally sponsored Advanced Containment Experiments (ACE) program, seven large-scale experiments on molten core concrete interactions (MCCIs) have been performed at Argonne National Laboratory. One of the objectives of these experiments is to collect and characterize all the aerosols released from the MCCIs. Aerosols released from experiments using four types of concrete (siliceous, limestone/common sand, serpentine, and limestone/limestone) and a range of metal oxidation … more
Date: January 1, 1992
Creator: Fink, J. K.; Thompson, D. H.; Spencer, B. W. & Sehgal, B. R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

Description: In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would b… more
Date: February 1, 1992
Creator: Arcieri, W.C. & Hanson, D.J. (EG and G Idaho, Inc., Idaho Falls, ID (United States))
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Results of Aerosol Code Comparisons With Releases From ACE MCCI Tests

Description: Results of aerosol release calculations by six groups from six countries are compared with the releases from ACE MCCI Test L6. The codes used for these calculations included: SOLGASMIX-PV, SOLGASMIX Reactor 1986, CORCON.UW, VANESA 1.01, and CORCON mod2.04/VANESA 1.01. Calculations were performed with the standard VANESA 1.01 code and with modifications to the VANESA code such as the inclusion of various zirconium-silica chemical reactions. Comparisons of results from these calculations were mad… more
Date: January 1, 1992
Creator: Fink, J. K.; Corradini, M.; Hidaka, A.; Hontanon, E.; Mignanelli, M. A.; Schroedl, E. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Phase I Aging Assessment of the Boiling-Water Reactor (BWR) Standby Liquid Control System

Description: Pacific Northwest Laboratory conducted a Phase I aging assessment of the standby liquid control (SLC) system used in boiling-water reactors. The study was based on detailed reviews of SLC system component and operating experience information obtained from the Nuclear Plant Reliability Database System, the Nuclear Document System, Licensee Event Reports, and other databases. Sources dealing with sodium pentaborate, borates, boric acid, and the effects of environment and corrosion in the SLC syst… more
Date: October 1, 1992
Creator: Orton, R. D.; Johnson, A. B.; Buckley, G. D. & Larson, L. L.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

A first look at LOCAs in the SBWR using RELAP5/MOD3

Description: The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering Laboratory has developed an input model of the SBWR for the RELAP5/MOD3 thermal-hydraulic safety analysis code. Preliminary … more
Date: January 1, 1992
Creator: Ghan, L.S.; Shaw, R.A. & Kullberg, C.M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Risk-based evaluation of Allowed Outage Times (AOTs) considering risk of shutdown

Description: When safety systems fail during power operation, Technical Specifications (TS) usually limit the repair within Allowed Outage Time (AOT). If the repair cannot be completed within the AOT, or no AOT is allowed, the plant is required to be shut down for the repair. However, if the capability to remove decay heat is degraded, shutting down the plant with the need to operate the affected decay-heat removal systems may impose a substantial risk compared to continued power operation over a usual repa… more
Date: January 1, 1992
Creator: Mankamo, T. (Avaplan Oy, Espoo (Finland)); Kim, I.S. & Samanta, P.K. (Brookhaven National Lab., Upton, NY (United States))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Strategies for denaturing the weapons-grade plutonium stockpile

Description: In the next few years, approximately 50 metric tons of weapons-grade plutonium and 150 metric tons of highly-enriched uranium (HEU) may be removed from nuclear weapons in the US and declared excess. These materials represent a significant energy resource that could substantially contribute to our national energy requirements. HEU can be used as fuel in naval reactors, or diluted with depleted uranium for use as fuel in commercial reactors. This paper proposes to use the weapons-grade plutonium … more
Date: October 1, 1992
Creator: Buckner, M. R. & Parks, P. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

Description: The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO[sub 2] fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontam… more
Date: October 1, 1992
Creator: Bond, W. D.; Mailen, J. C. & Michaels, G. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

Description: The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In … more
Date: January 1, 1992
Creator: Khan, H.J.; Cheng, H.S. & Rohatgi, U.S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Radionuclide characterization at US commercial light-water reactors for decommissioning assessment: Distributions, inventories, and waste disposal considerations

Description: A continuing research program, conducted by Pacific Northwest Laboratory for the US Nuclear Regulatory Commission, characterizing radionuclide concentrations associated with US light-water reactors has been conducted for more than a decade. The research initially focused upon sampling and analytical measurements for the purpose of establishing radionuclide distributions and inventories for decommissioning assessment, since very little empirical data existed. The initial phase of the research pr… more
Date: September 1, 1992
Creator: Abel, K. H.; Robertson, D. E. & Thomas, C. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Stress Corrosion Cracking Susceptibility of Irradiated Type 304 Stainless Steels

Description: Slow-strain-rate tensile tests and microstructural analysis by Auger electron spectroscopy were conducted on specimens of high- and commercial-purity (HP and CP) heats of Type 304 stainless steel obtained from neutron absorber tubes and a control blade sheath after irradiation up to 2.5 [times] 10[sup 2l] n[center dot]cm[sup [minus]2] (E > l MeV) in boiling water reactors (BWRs). The susceptibility of the HP absorber tubes to intergranular stress corrosion cracking (IGSCC) was higher than that … more
Date: August 1, 1992
Creator: Chung, H. M.; Ruther, W. E.; Sanecki, J. E. & Kassner, T. F.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Stress Corrosion Cracking Susceptibility of Irradiated Type 304 Stainless Steels

Description: Slow-strain-rate tensile tests and microstructural analysis by Auger electron spectroscopy were conducted on specimens of high- and commercial-purity (HP and CP) heats of Type 304 stainless steel obtained from neutron absorber tubes and a control blade sheath after irradiation up to 2.5 {times} 10{sup 2l} n{center_dot}cm{sup {minus}2} (E > l MeV) in boiling water reactors (BWRs). The susceptibility of the HP absorber tubes to intergranular stress corrosion cracking (IGSCC) was higher than that … more
Date: August 1, 1992
Creator: Chung, H. M.; Ruther, W. E.; Sanecki, J. E. & Kassner, T. F.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Prediction of BWR performance under the influence of Isolation Condenser-using RAMONA-4 code

Description: The purpose of the Boiling Water Reactor (BWR) Isolation Condenser (IC) is to passively control the reactor pressure by removing heat from the system. This type of control is expected to reduce the frequency of opening and closing of the Safety Relief Valves (SRV). A comparative analysis is done for a BWR operating with and without the influence of an IC under Main Steam Isolation Valve (MSIV) closure. A regular BWR, with forced flow and high thermal power, has been considered for analysis. In … more
Date: December 31, 1992
Creator: Khan, H. J.; Cheng, H. S. & Rohatgi, U. S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Generic component failure data base

Description: This report discusses comprehensive component generic failure data base which has been developed for light water reactor probabilistic risk assessments. The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) was used to generate component failure rates. Using this approach, most of the failure rates are based on actual plant data rather then existing estimates.
Date: December 31, 1992
Creator: Eide, S. A. & Calley, M. B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

A first look at LOCAs in the SBWR using RELAP5/MOD3

Description: The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering Laboratory has developed an input model of the SBWR for the RELAP5/MOD3 thermal-hydraulic safety analysis code. Preliminary … more
Date: December 31, 1992
Creator: Ghan, L. S.; Shaw, R. A. & Kullberg, C. M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Oxidation of spent fuel in air at 175{degree} to 195{degree}C

Description: Oxidation tests in dry air were conducted on four LWR spent fuels at 175{degrees} and 195{degrees}C to determine the effect of the fuel characteristics on the oxidation state likely to exist at the time leaching occurs in a potential repository. Weight changes were measured and samples were examined by XRD, ceramography, TEM, and TGA. Despite local variations in the grain boundary susceptibility to oxidation, all four fuels progressed toward an apparent endpoint at an oxygen-to-metal (O/M) rati… more
Date: April 1, 1992
Creator: Einziger, R. E.; Buchanan, H. C.; Thomas, L. E. & Stout, R. B.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
Back to Top of Screen