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in publisher: "U.S. Nuclear Regulatory Commission"

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NRC plan for cleanup operations at Three Mile Island Unit 2

This NRC Plan, which defines NRC's functional role in cleanup operations at Three Mile Island Unit 2 and outlines NRC's regulatory responsibilities in fulfilling this role, is the first revision to the initial plan issued in July 1980 (NUREG-0698). Since 1980, a number of policy developments have occurred which will have an impact on the course of cleanup operations. This revision reflects these developments in the area of NRC's review and approval process with regard to cleanup … more

Date: February 1, 1982

Creator: Lo, R. & Snyder, B.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Evaluation of station blackout accidents at nuclear power plants: Technical findings related to unresolved safety issue A-44: Final report

''Station Blackout,'' which is the complete loss of alternating current (AC) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on AC power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed inclu… more

Date: June 1, 1988

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Effects of water in film boiling over liquid metal melts

Liquid-liquid boiling experiments have been performed with H/sub 2/O and liquid metal melts in the 100-series test matrix (Runs 121, 126, 127) and the VE test matrix. Some of the pre-explosion unstable film boiling data as well as observations from the explosive series have been previously reported.

Date: January 1, 1986

Creator: Greene, G.A.; Finfrock, C. & Burson, S.B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

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: Article

Partner: UNT Libraries Government Documents Department

open access

Plant risk status information management system

The Plant Risk Status Information Management System (PRISIMS) is a PC program that presents information about a nuclear power plant's design, its operation, its technical specifications, and the results of the plant's probabilistic risk assessment (PRA) in a logically and easily accessible format. PRISIMS provides its user with unique information for integrating safety concerns into day-to-day operational decisions and/or long-range management planning.

Date: January 1, 1985

Creator: Campbell, D.J.; Ellison, B.C.; Glynn, J.C. & Flanagan, G.F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Understanding and managing the effects of battery charger and inverter aging

An aging assessment of battery chargers and inverters was conducted under the auspices of the NRC's Nuclear Plant Aging Research (NPAR) Program. The intentions of this program are to resolve issues related to the aging and service wear of equipment and systems at operating reactor facilities and to assess their impact on safety. Inverters and battery chargers are used in nuclear power plants to perform significant functions related to plant safety and availability. The specific impact of a batt… more

Date: January 1, 1992

Creator: Gunther, W. (Brookhaven National Lab., Upton, NY (United States)) & Aggarwal, S. (Nuclear Regulatory Commission, Washington, DC (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Inspection methods for safeguards systems at nuclear facilities

A project team at Lawrence Livermore National Laboratory has been developing inspection procedures and training materials for the NRC inspectors of safeguards systems at licensed nuclear facilities. This paper describes (1) procedures developed for inspecting for compliance with the Code of Federal Regulations, (2) training materials for safeguards inspectors on technical topics related to safeguards systems, such as computer surety, alarm systems, sampling techniques, and power supplies, and (… more

Date: October 16, 1981

Creator: Minichino, C. & Richard, E.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Seismic review of the R. E. Ginna Nuclear Power Plant as part of the Systematic Evaluation Program for the Nuclear Regulatory Commission

This paper is a progress report on work at the Lawrence Livermore National Laboratory (LLNL) to perform a limited seismic reassessment of the Robert E. Ginna Nuclear Power Plant. The reassessment is being done for the Nuclear Regulatory Commission (NRC) as part of the Systematic Evaluation Program. The reassessment focuses generally on the reactor coolant pressure boundary and on those systems and components necessary to shut down the reactor safely and to maintain it in a safe shutdown conditi… more

Date: May 27, 1980

Creator: Murray, R. C.; Nelson, T. A.; Ng, D. S.; Liaw, C. Y.; Levin, H. A. & Cheng, T. M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

The effect of cure conditions on the stability of cement waste forms after immersion in water

We investigated the effects of curing conditions on the stability of cement-solidified ion-exchange resins after immersion in water. The test specimens consisted of partially depleted mixed-bed bead resins solidified in one of three vendor-supplied Portland I cement formulations, in a reference cement formulation, or in a gypsum-based binder formulation. We cured samples prepared using each formulation in sealed containers for periods of 7, 14, or 28 days as well as in air or with an accelerate… more

Date: January 1, 1988

Creator: Siskind, B.; Adams, J. W.; Clinton, J. H.; Piciulo, P. L. & McDaniel, K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 2. Evaluation of seismic designs: a review of seismic design requirements for Nuclear Power Plant Piping

This document reports the position and recommendations of the NRC Piping Review Committee, Task Group on Seismic Design. The Task Group considered overlapping conservation in the various steps of seismic design, the effects of using two levels of earthquake as a design criterion, and current industry practices. Issues such as damping values, spectra modification, multiple response spectra methods, nozzle and support design, design margins, inelastic piping response, and the use of snubbers are … more

Date: April 1, 1985

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Thermohydraulic Modeling and Simulation of Breeder Reactors

This paper deals with the modeling and simulation of system-wide transients in LMFBRs. Unprotected events (i.e., the presumption of failure of the plant protection system) leading to core-melt are not considered in this paper. The existing computational capabilities in the area of protected transients in the US are noted. Various physical and numerical approximations that are made in these codes are discussed. Finally, the future direction in the area of model verification and improvements is d… more

Date: January 1, 1982

Creator: Agrawal, A. K.; Khatib-Rahbar, M.; Curtis, R. T.; Hetrick, D. L. & Girijashankar, P. V.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Safety research programs sponsored by Office of Nuclear Regulatory Research: Progress report, January 1--March 31, 1989

This progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Regulatory Applications, Division of Engineering, Division of Safety Issue Resolution, and Division of Systems Research of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research following the reorganization in July 1988. The previous reports have covered the period October 1, 1976 through December 31, 1988.

Date: August 1, 1989

Creator: Weiss, A. J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Report to the Nuclear Regulatory Commission from the staff panel on the Commission's determination of an Extraordinary Nuclear Occurrence (ENO)

The Panel finds that the first criterion, pertaining to whether the accident caused a discharge of radioactive material or levels of radiation offsite as defined in 10 CFR 140.84, has not been met. It further finds that there is presently insufficient information to support any definitive finding as to whether or not the second criterion, relating to damage to persons or property offsite as defined in 10 CFR 140.85, has been met. Since the Panel has not found that both criteria have been met, i… more

Date: 1980~

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

A strategy for minimizing common mode human error in executing critical functions and tasks

Human error in execution of critical functions and tasks can be costly. The Three Mile Island and the Chernobyl Accidents are examples of results from human error in the nuclear industry. There are similar errors that could no doubt be cited from other industries. This paper discusses a strategy to minimize common mode human error in the execution of critical functions and tasks. The strategy consists of the use of human redundancy, and also diversity in human cognitive behavior: skill-, rule-,… more

Date: January 1, 1992

Creator: Beltracchi, L. (Nuclear Regulatory Commission, Washington, DC (United States)) & Lindsay, R.W. (Argonne National Lab., IL (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Instrument accuracy in reactor vessel inventory tracking systems

Instrumentation needs for detection of inadequate core cooling. Studies of the Three Mile Island accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. Industry studies by plant owners and reactor vendors supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC as well as to provide more complete information for operator control of safety injection flow to minimize … more

Date: January 1, 1986

Creator: Anderson, J. L.; Anderson, R. L.; Morelock, T. C.; Hauang, T. L. & Phillips, L. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

1982 engineering conference on reliability for the electrical power industry

Emergency onsite ac power systems at nuclear power plants are a major concern in plant risk assessments because of the relatively large frequency of loss of offsite power and the dependence of most other safety systems on ac power. Detailed reviews of onsite ac power system designs and reviews of experience with diesel generators at US nuclear power plants form the basis of system reliability analyses that show significant improvements in reliability can be obtained at moderate cost for some pl… more

Date: January 1, 1982

Creator: Campbell, D. J.; Arendt, J. S.; Battle, R. E. & Baranowsky, P. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds

The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant… more

Date: April 1, 2008

Creator: Schuster, G. J.; Simonen, F. A. & Doctor, S. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Methods and findings of a systems interaction study of a Westinghouse PWR

This paper describes the methods and findings of a systems interaction study of a Westinghouse PWR. BNL conducted the study as a methods application that was performed to support the resolution of Unresolved Safety Issue A-17 on Systems Interactions. The method calls for a fault tree model of the plant to be developed in stages, corresponding to successively increasing levels of scope and detail. A functional model is developed first, resolved only to sufficient detail to reflect support system… more

Date: January 1, 1985

Creator: Youngblood, R.; Hanan, N.; Fitzpatrick, R.; Xue, D.; Bozoki, G.; Fresco, A. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Verification and update of BNL mixed waste survey

This report briefly describes attempts to verify the results of a previous survey on the amount of mixed wastes generated at various facilities during 1985. The original survey indicated some lack of understanding of current EPA regulations. This telephone survey verification indicated a better understanding of these regulations in recent months. Changes in EPA regulations and the addition of new compounds to the list of hazardous wastes are causing problems for organizations trying to comply. … more

Date: January 1, 1987

Creator: Bowerman, B.S. & Siskind, B.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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DHCVIM: A direct heating containment vessel interactions module: Applications to Sandia National Laboratory Surtsey experiments

Direct containment heating is the mechanism of severe nuclear reactor accident containment loading which results from transfer of thermal and chemical energy from high temperature, finely divided, molten core material to the containment atmosphere. The Direct Heating Containment Vessel Interactions Module, DHCVIM, has been developed at BNL to mechanistically model the mechanisms of containment loading resulting from the direct heating accident sequence. The calculational procedure is being used… more

Date: January 1, 1987

Creator: Ginsberg, T. & Tutu, N.K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Qualification Testing Evaluation (QTE) program for safety-related equipment

The nuclear power industry is required to demonstrate that certain safety-related equipment is ''qualified'' and will function even in the event of a severe reactor accident. Demonstration of qualification by testing is the preferred approach. International interest in equipment qualification, and its recognition as being paramount to safety, is rapidly increasing, with most major supplier-countries developing sophisticated qualification testing facilities. An aspect of the demonstration of qua… more

Date: January 1, 1980

Creator: Bonzon, L. L.; Gillen, K. T.; Clough, R. L.; Salazar, E. A.; Buckalew, W. H.; Thome, F. V. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Update of Part 61 Impacts Analysis Methodology. Methodology report. Volume 1

Under contract to the US Nuclear Regulatory Commission, the Envirosphere Company has expanded and updated the impacts analysis methodology used during the development of the 10 CFR Part 61 rule to allow improved consideration of the costs and impacts of treatment and disposal of low-level waste that is close to or exceeds Class C concentrations. The modifications described in this report principally include: (1) an update of the low-level radioactive waste source term, (2) consideration of addi… more

Date: January 1, 1986

Creator: Oztunali, O.I. & Roles, G.W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Effect of reactor conditions on MSIV (main steam isolation valves)-ATWS power level

In a boiling water reactor (BWR) when there is closure of the main steam isolation valves (MSIVs), the energy generated in the core will be transferred to the pressure suppression pool (PSP) via steam flows out of the relief valves. The pool has limited capacity as a heat sink and hence, if there is no reactor trip (an ATWS event), there is the possibility that the pool temperature may rise beyond acceptable limits. The present study was undertaken to determine how the initial reactor condition… more

Date: January 1, 1987

Creator: Diamond, D. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Initial pipe break analyses for advanced LMR (liquid metal reactor) concepts using MINET

In support of an initial NRC review of DOE sponsored advanced liquid metal reactors (LMRs), BNL has performed some very conservative calculations of postulated primary loop pipe breaks using the MINET Code. This report briefly describes the results obtained from these calculations. 5 refs., 2 figs.

Date: June 1, 1987

Creator: Van Tuyle, G. J.; Chan, B. C. & Slovik, G. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department