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in creator/contributor: "U.S. Nuclear Regulatory Commission"

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Reviews of ASME Section 11 pump and valve relief requests: Post Generic Letter 89-04

This paper presents a discussion of ASME Section 11 Pump and Valve Inservice Testing relief request reviews by the NRC and their contractors. Topics that will be discussed include the scope of USNRC reviews in Technical Evaluation Reports (TERs) (and Safety Evaluation, SEs); including the basis for granting relief requests, the status of relief requests in IST Program updates, and the Generic Letter 89-04 approval process; and the level of technical detail required in submitted programs. This p… more

Date: January 1, 1992

Creator: DiBiasio, A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Physical fitness training reference manual for security force personnel at fuel cycle facilities possessing formula quantities of special nuclear materials

The recommendations contained throughout this NUREG are being provided to the Nuclear Regulatory Commission (NRC) as a reference manual which can be used by licensee management as they develop a program plan for the safe participation of guards, Tactical Response Team members (TRTs), and all other armed response personnel in physical fitness training and in physical performance standards testing. The information provided in this NUREG will help licensees to determine if guards, TRTs, and other … more

Date: September 1, 1991

Creator: Arzino, P. A.; Caplan, C. S. & Goold, R. E. (California State Univ., Hayward, CA (United States). Foundation)

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Fuel performance annual report for 1989

This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 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 US Nuclear Regulatory Commission evaluations are included.

Date: June 1, 1992

Creator: Bailey, W.J.; Berting, F.M. (Pacific Northwest Lab., Richland, WA (United States)) & Wu, S. (Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology)

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Health effects models for nuclear power plant accident consequence analysis: Low LET radiation

This report describes dose-response models intended to be used in estimating the radiological health effects of nuclear power plant accidents. Models of early and continuing effects, cancers and thyroid nodules, and genetic effects are provided. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes -- are considered. In addition, model… more

Date: January 1, 1990

Creator: Evans, J. S.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Nuclear plant-aging research on reactor protection systems

This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show th… more

Date: January 1, 1988

Creator: Meyer, L.C.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Current nuclear industry practices with regard to the integration of surveillances

Commercial nuclear industry practices regarding the integration of technical specification (TS) surveillance and maintenance activities are identified in this report. A questionnaire was developed and used to obtain current industry practices from NRC Regional personnel and INEL engineers with utility experience. Some of these practices indicate that the scheduling of TS surveillance and maintenance items could be more effectively coordinated. Also, must utilities do not formally consider risk … more

Date: April 1, 1992

Creator: Stewart, M. & Smith, C.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Validation issues in aging risk evaluations

Validation issues in aging risk evaluations are examined by considering how variabilities and uncertainties due to sparse component aging data, modeling assumptions, and risk quantification approaches may affect aging risk evaluation results and inferences. Sensitivity studies using a NUREG-1150 PWR evaluated the effect of component aging data uncertainties and variations in test and maintenance frequencies on aging prioritizations. Preliminary results indicate that while individual component r… more

Date: January 1, 1992

Creator: Hassan, M.; Samanta, P. (Brookhaven National Lab., Upton, NY (United States)) & Vesely, W. (Science Applications International Corp., San Diego, CA (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Reviewing the impact of advanced control room technology

Progress to date on assessing the nature of the expected changes in human performance and risk associated with the introduction of digital control, instrumentation, and display systems is presented. Expected changes include the shift toward more supervisory tasks, development of intervention strategies, and reallocation of function between human and machine. Results are reported in terms of the scope of new technology, human performance issues, and crews experience with digital control systems … more

Date: January 1, 1992

Creator: Wilhelmsen, C. A.; Gertman, D. I.; Ostrom, L. T.; Nelson, W. R.; Galyean, W. J. & Byers, J. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Aging assessment of BWR control rod drive systems

This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to… more

Date: January 1, 1991

Creator: Greene, R.H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Work on surveillance dosimetry and benchmarks used for US power reactors and for irradiation experiments

No Description Available.

Date: January 1, 1991

Creator: Kam, F. B. K. & Stallmann, F. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Transformer failure and common-mode loss of instrument power at Nine Mile Point Unit 2 on August 13, 1991

On August 13, 1991, at Nine Mile Point Unit 2 nuclear power plant, located near Scriba, New York, on Lake Ontario, the main transformer experienced an internal failure that resulted in degraded voltage which caused the simultaneous loss of five uninterruptible power supplies, which in turn caused the loss of several nonsafety systems, including reactor control rod position indication, some reactor power and water indication, control room annunciators, the plant communications system, the plant … more

Date: October 1, 1991

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Identification and assessment of containment and release management strategies for a BWR Mark I containment

This report identifies and assesses accident management strategies which could be important for preventing containment failure and/or mitigating the release of fission products during a severe accident in a BWR plant with a Mark 1 type of containment. Based on information available from probabilistic risk assessments and other existing severe accident research, and using simplified containment and release event trees, the report identifies the challenges a Mark 1 containment could face during t… more

Date: September 1, 1991

Creator: Lin, C. C. & Lehner, J. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Evaluation of Inservice Inspection Examinatiions

In order to evaluate the effectiveness of Section 11, Division 1, Rules for Inservice Inspection of Nuclear Power Plant Components,'' of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code examinations, 26 inservice inspection (ISI) summary reports from 24 facilities were reviewed. It was found that these ASME Code examinations and tests are instrumental in revealing indications and defects in welds and plant components. In addition, this study uncovered that fac… more

Date: May 1990

Creator: Aldrich, D. A. & Cook, J. F.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Severe accident testing of electrical penetration assemblies

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. … more

Date: November 1, 1989

Creator: Clauss, D.B. (Sandia National Labs., Albuquerque, NM (USA))

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Steam generator secondary pH during a steam generator tube rupture

The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL t… more

Date: December 1, 1991

Creator: Adams, J. P. & Peterson, E. S.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Representing cognitive activities and errors in HRA trees

A graphic representation method is presented herein for adapting an existing technology--human reliability analysis (HRA) event trees, used to support event sequence logic structures and calculations--to include a representation of the underlying cognitive activity and corresponding errors associated with human performance. The analyst is presented with three potential means of representing human activity: the NUREG/CR-1278 HRA event tree approach; the skill-, rule- and knowledge-based paradigm… more

Date: January 1, 1992

Creator: Gertman, D.I.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Human factors design review guidelines for advanced nuclear control room technologies

Advanced control rooms (ACRs) for future nuclear power plants are being designed utilizing computer-based technologies. The US Nuclear Regulatory Commission reviews the human engineering aspects of such control rooms to ensure that they are designed to good human factors engineering principles and that operator performance and reliability are appropriately supported in order to protect public health and safety. This paper describes the rationale, general approach, and initial development of an … more

Date: January 1, 1991

Creator: O'Hara, J.; Brown, W. (Brookhaven National Lab., Upton, NY (United States)); Granda, T. & Baker, C. (Carlow Associates, Inc., Fairfax, VA (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fatigue and environmentally assisted cracking in light water reactors

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: Article

Partner: UNT Libraries Government Documents Department

open access

PRISM reactor system design and analysis of postulated unscrammed events

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of … more

Date: January 1, 1991

Creator: Van Tuyle, G.J.; Slovik, G.C. (Brookhaven National Lab., Upton, NY (United States)); Rosztoczy, Z. & Lane, J. (Nuclear Regulatory Commission, Washington, DC (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Rcs Pressure Under Reduced Inventory Conditions Following a Loss of Residual Heat Removal

The thermal-hydraulic response of a closed-reactor coolant system to loss of residual heat removal (RHR) cooling is investigated. The processes examined include: core coolant boiling and steam generator reflux condensation, pressure increase on the primary side, heat transfer mechanisms on the steam generator primary and secondary sides, and effects of noncondensible gas on heat transfer processes.

Date: January 1, 1992

Creator: Palmrose, D. E.; Hughes, E. D. & Johnsen, G. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Standard Review Plan Update and Development Program

This implementing procedures document (IPD) was prepared for use in implementing tasks under the standard review plan update and development program (SRP-UDP). The IPD provides comprehensive guidance and detailed procedures for SRP-UDP tasks. The IPD is mandatory for contractors performing work for the SRP-UDP. It is guidance for the staff. At the completion of the SRP-UDP, the IPD will be revised (to remove the UDP aspects) and will replace NRR Office Letter No. 800 as long-term maintenance pr… more

Date: May 1, 1992

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Correlation of analysis with high level vibration test results for primary coolant piping

Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was … more

Date: January 1, 1992

Creator: Park, Y.J.; Hofmayer, C.H. (Brookhaven National Lab., Upton, NY (United States)) & Costello, J.F. (Nuclear Regulatory Commission, Washington, DC (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

The use of NPAR (Nuclear Plant Aging Research) results in plant inspection activities

The US NRC's Nuclear Plant Aging Research (NPAR) Program is a hardware oriented research program which has produced a large data base of equipment and system operating, maintenance, and testing information. Equipment and systems which have a propensity for age related degradation are identified, and methods for detecting and mitigating aging effects have been evaluated. As plants age, it becomes increasingly important that NRC inspectors be cognizant of plant aging phenomena. This paper describ… more

Date: January 1, 1989

Creator: Gunther, W. & Taylor, J.

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