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Survey of Remote Area Monitoring Systems at U.S. Light-Water-Cooled Power Reactors

Description: A study was made of the capabilities and operating practices, including calibration, of remote area monitoring (RAM) systems at light-water-cooled power reactors in the United States. The information was obtained by mail questionaire. Specific design capabilities, including range, readout and alarm features are documented along with the numbers and location of detectors, calibration and operational procedures. Comments of respondents regarding RAM systems are also included.
Date: April 1, 1982
Creator: Kathren, R. L. & Mileham, A. P.
Partner: UNT Libraries Government Documents Department
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HAZARDS SUMMARY REPORT FOR THE ARMY PACKAGE POWER REACTOR

Description: The APPR-I is described and the various hazards are reviewed. Because of the reactor's location near the nation's Capitol, containment is of the utmost importance. The maximum energy release in any possible accident is 7.4 million Btu's which is completely contained within a 7/8 inch thick steel cylindrical shell with hemispherical ends. The vapor container is 60 ft high and 32 ft in diameter and is lined on the inside with 2 ft of reinforced concrete which provides missile prote… more
Date: July 27, 1955
Partner: UNT Libraries Government Documents Department
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Power Histories for Fuel Codes

Description: Computations of power history effects on the pre-loss-of-coolant accident (LOCA) conditions of generic pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rods were performed at Pacific Northwest Laboratory using the U.S. Nuclear Regulatory Commission (NRC) code FRAPCON-2. Comparisons were made between cases where the fuel operated at a high ( 11 LOCA-limited") power throughout life (20,000 MWd/MTU) and those where the fuel was at a lower power for most of its burnup and r… more
Date: January 1, 1982
Creator: Gilbert, E. R.; Rausch, W. N. & Panisko, F. E.
Partner: UNT Libraries Government Documents Department
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Radiological Assessment of Steam Generator Removal and Replacement: Update and Revision

Description: A previous analysis of the radiological impact of removing and replacing corroded steam generators has been updated based on experience gained during steam generator repairs at Surry Unit 2. Some estimates of occupational doses involved in the operation have been revised but are not significantly different from the earlier estimates. Estimates of occupational doses and radioactive effluents for new tasks have been added. Health physics concerns that arose at Surry included the number of persons… more
Date: December 1, 1980
Creator: Hoenes, G. R.; Mueller, M. A. & McCormack, W. D.
Partner: UNT Libraries Government Documents Department
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Fuel Performance Annual Report for 1980

Description: This annual report, the third in a series, provides a brief description of fuel performance in conmercial nuclear power plants. Brief summaries of fuel surveillance programs and operating experience, fuel performance problems, and fuel design changes are provided. References to additional, more detailed, information and related NRC evaluation are included.
Date: December 1, 1981
Creator: Bailey, W. J.; Rising, K. H. & Tokar, M.
Partner: UNT Libraries Government Documents Department
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Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

Description: Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-st… more
Date: January 1, 1982
Creator: Wittenbrock, N. G.
Partner: UNT Libraries Government Documents Department
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Facilitation of Decommissioning Light Water Reactors

Description: Information on design features, special equipment, and construction methods useful in the facilitation of decommissioning light water reactors is presented in this report. A wide range of facilitation methods--from improved documentation to special decommissioning tools and techniques--is discussed. In addition, estimates of capital costs, cost savings, and radiation dose reduction associated with these facilitation methods are given.
Date: December 1, 1979
Creator: Moore, E. B., Jr.
Partner: UNT Libraries Government Documents Department
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An Independent Assessment of Evacuation Time Estimates for A Peak Population Scenario in the Emergency Planning Zone of the Seabrook Nuclear Power Station

Description: This study comprises two major tasks. First, it includes an independent assessment of the methods and assumptions used in calculating evacuation time estimates (ETEs) applicable to the general population for a peak population scenario in the emergency planning zone {EPZ) of the Seabrook Nuclear Power Station. This consists of a review and analysis of previous work by Public Service of New Hampshire {PSNH) and the Federal Emergency Management Agency (FEMA), as well as an independent calculation … more
Date: November 1, 1982
Creator: Moeller, M. P.; Urbanik, II, T.; Mclean, M. A. & Desrosiers, A. E.
Partner: UNT Libraries Government Documents Department
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Mechanistic Considerations Used in the Development of the PROFIT PCI Failure Model

Description: A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) Interactions (PC!) failure model for estimating the probability of failure in !ransient increases in power (PROFIT) was developed. PROFIT is based on 1) standard statistical methods applied to available PC! fuel failure data and 2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that … more
Date: May 1, 1980
Creator: Pankaskie, P. J.
Partner: UNT Libraries Government Documents Department
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REVIEW OF TRANSAMERICA DELAVAL INC. DIESEL GENERATOR OWNERS' GROUP ENGINE REQUALIFICATION PROGRAM

Description: In December 1983, 13 nuclear utilities that own TDI diesel generators formally established an Owners• Group to address concerns regarding the reliability and operability of these engines. The Owners' Group program for engine requalification consisted of four major elements: 1) resolution of known problems with potentially generic implications, 2) a design review and quality revalidation (DR/QR) effort aimed at identifying and correcting potential problems with the important engine componen… more
Date: December 1, 1985
Creator: Berlinger, C. H.
Partner: UNT Libraries Government Documents Department
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Use of a Moving Heat Conductor Mesh to Perform Reflood Calculations with RELAP4/MOD6

Description: RELAP4 is a computer code which can be used for the transient therm~l hydraulic analysis of light water reactors and related systems .. Various versions of the RELAP4 code are widely used throughout the world for experimental system analysis, reactor design,and nuclear system safety studies. RELAP4/MOD6 includes many new analytical models which were developed primarily for the analysis of the reflood phase of a PWR loss-of-coolant accident (LOCA) transient. The key feature forming the basis for… more
Date: May 6, 1979
Creator: Fischer, S R; Ellis, L V & Chen, Y S
Partner: UNT Libraries Government Documents Department
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Technical Letter Report: Evaluation and Analysis of a Few International Periodic Safety Review Summary Reports

Description: At the request of the United States (U.S.) government, the International Atomic Energy Agency (IAEA) assembled a team of 20 senior safety experts to review the regulatory framework for the safety of operating nuclear power plants in the United States. This review focused on the effectiveness of the regulatory functions implemented by the NRC and on its commitment to nuclear safety and continuous improvement. One suggestion resulting from that review was that the U.S. Nuclear Regulatory Commissi… more
Date: December 17, 2013
Creator: Chopra, Omesh K.; Diercks, Dwight R.; Ma, David Chia-Chiun & Garud, Yogendra S.
Partner: UNT Libraries Government Documents Department
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In-Vessel Retention Technology Development and Use for Advanced PWR Designs in the USA and Korea

Description: In-Vessel Retention (IVR) of molten core debris by means of external reactor vessel flooding is a cornerstone of severe accident management for Westinghouse's AP600 (advanced passive light water reactor) design. The case for its effectiveness (made in previous work by the PI) has been thoroughly documented, reviewed as part of the licensing certification, and accepted by the US Nuclear Regulatory Commission. A successful IVR would terminate a severe accident, passively, with the core in a stabl… more
Date: May 18, 2004
Creator: Theofanous, T. G.; Oh, S. J. & Scobel, J. H.
Partner: UNT Libraries Government Documents Department
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Fuel Performance Annual Report for 1979

Description: This annual report, the second in a series, provides a brief description of fuel performance in commercial nuclear power plants. Brief summaries are given of fuel surveillance programs, fuel performance problems, and fuel design changes. References to additional, more detailed, information and related NRC evaluation are provided.
Date: January 1, 1981
Creator: Tokar, M.; Mailey, W. J. & Cunningham, M. E.
Partner: UNT Libraries Government Documents Department
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PNL Technical Review of Pressurized Thermal Shock Issues Supplement 1: Technical Critique of the NRC Near-Term Screening Criteria

Description: Pacific Northwest Laboratory (PNL) provided a technical critique of the draft report, NRC Staff Evaluation of Pressurized Thermal Shock, dated September 13, 1982. This report provided the basis for the NRC near-term regulatory position on pressurized thermal shock {PTS) and recommended a generic screening criteria for welds in the vessel beltline region. The PNL staff concluded that the screening criteria were adequate to meet the intent of the NRC safety goal and to retain past predictions of … more
Date: May 1, 1983
Creator: Pederson, L. T.; Apley, W. J.; Bian, S. H.; Pelto, P. J.; Simonen, E. P.; Simonen, F. A. et al.
Partner: UNT Libraries Government Documents Department
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LETTER REPORT SUMMARY RESULTS OF THE NRC TEAM INTERACTION SKILLS STUDY AT DIABLO CANYON POWER PLANT

Description: This report presents information to participants in the Team Interaction Skills study conducted at Diablo Canyon Power Plant from September to November 1989. A study was conducted to develop and assess measures of team interaction skills of nuclear power plant control room crews in simulated emergency conditions. Data were collected at a boiling water reactor (BWR) and pressurized water reactor (PWA) using three sets of rating scales; Behaviorally Anchored Rating Scales (BARS), Behavioral Frequ… more
Date: December 1, 1990
Creator: Hauth, J. T.; Toquam, J. L.; Bramwell, A. T. & Fleming, T. E.
Partner: UNT Libraries Government Documents Department
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EXPERIMENT OPERATIONS PLAN FOR A LOSS-OF-COOLANT ACCIDENT SIMULATION IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

Description: Pressurized water reactor loss-of-coolant accident phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship between the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. This document contains both experiment proposal and ass… more
Date: April 1, 1981
Creator: Russcher, G. E.; Cannon, L. W.; Goodman, R. L.; Hesson, G. M.; King, L. L.; McDuffie, P. N. et al.
Partner: UNT Libraries Government Documents Department
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Efficiency Studies with Gamma Ray Portion of Specialized Reactor-Shield Monte Carlo Program 18-0

Description: Application studies were made with Specialized Reactor-Shield Monte Carlo Program 18-0 to determine the efficiency and feasibility of calculating energy deposition due to primary core gamma rays throughout the XNJ140E-1 reactor-shield assembly. Monte Carlo results are presented in tabular form for all geometrical regions used to describe the shield. Described here is a means of obtaining adequate and valid heating rates in about 47 hours on the IBM-704 digital computer. Comparison of Monte Carl… more
Date: August 1, 1961
Creator: Capo, M. A.
Partner: UNT Libraries Government Documents Department
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An Analysis of Evacuation Time Estimates Around 52 Nuclear Power Plant Sites Analysis and Evaluation

Description: On November 29, 1979, the NRC sent a letter to 52 nuclear power plants requesting evacuation time estimates for 10 sectors within a 10-mile radius of each plant. The requirements for these evacuation times are contained in NUREG-0654, Rev. 1, and include such factors as population density, weather conditions, warning time, response time and confirmation time. Fifty responses were received. The analysis of these findings are presented for review.
Date: May 1, 1981
Creator: Urbanik, II, T. & Desrosiers, A. E.
Partner: UNT Libraries Government Documents Department
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PRELIMINARY ANALYSIS OF SUPERCHARGED NUCLEAR RAMJET PROPULSION SYSTEM

Description: A preliminary analysis has been made to compare the performance.of a ceramic ramjet reactor powerplant, such as described in General Electric Report No. XDC-56-5-81, with that of such a powerplant supercharged by a metallic-vapor-cycle compressor jet. Performance at sea level, Mach 2.5 is ·reported for the vapor-cycle compressor jet alone, for the ramjet alone, and for the compressor jet - ramjet combination. Results indicate that adding the compressor-jet as a supercharger for the ramjet provi… more
Date: May 17, 1957
Creator: Boppart, J. A.
Partner: UNT Libraries Government Documents Department
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Pre-Phase 1 Aging Assessment of the BWR Isolation Condenser System

Description: The isolation condenser system (ICS) is part of the emergency core cooling system in five U.S. boiling-water reactors. In the event that the reactor pressure vessel becomes isolated from the main condenser, the ICS removes decay heat from the reactor. The ICS is important to reactor safety because it is relied on to help mitigate core damage during a loss-of-coolant accident. In support of the U.S. Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, staff from the Pacif… more
Date: August 1, 1995
Creator: Orton, R. D.
Partner: UNT Libraries Government Documents Department
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Pre-Phase 1 Aging Assessment of the BWR and PWR Accumulators

Description: Accumulators are important components used in many systems at commercial boiling water reactors (BWRs) and pressurized water reactors in the United States. The accumulators are vessels attached to fluid systems to provide 1) a limited backup source of stored fluid energy for hydraulic/pneumatic mechanical equipment, 2) a damping effect on pressure pulses in fluid systems, and 3) a volume of fluid to be injected passively into a fluid system. Accumulators contain a gas that is compressed or expa… more
Date: August 1, 1995
Creator: Buckely, G. D.
Partner: UNT Libraries Government Documents Department
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Application of Linear Propagation of Errors to Fuel Rod Temperature and Stored Energy Calculations

Description: Linear propagatlon of errors evaluates modeling uncertainty by approximating a function of interest by first-order Taylor's series expansions and then approximating the variance of the function by the variance of the linear approximation. This report discusses uncertainty analysis for different nuclear fuel rod designs, the process of model validation, and the effect of cracked pellet fuel models upon temperabre uncertainty. Using a postulated power history, the uncertainty for the predicted th… more
Date: October 1, 1980
Creator: Cunningham, M. E.; Olsen, A. R.; Lanning, D. D. & Willford, R. E.
Partner: UNT Libraries Government Documents Department
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