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in subjects: "Reactor Safety"

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Assessment of light water reactor fuel damage during a reactivity initiated accident

This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalp… more

Date: January 1, 1980

Creator: MacDonald, P. E.; Seiffert, S. L.; Martinson, Z. R.; McCardell, R. K.; Owen, D. E. & Fukuda, S. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

De-entrainment on vertical elements in air droplet cross flow. [PWR]

De-entrainment phenomena on vertical elements in air-water droplet cross flow are generated using a horizontal array of water spray nozzles and a draft-induced wind tunnel. These conditions are used to obtain experimental values of the de-entrainment efficiency of isolated elements (25.4-, 63.5-, and 101.6-mm-diam cylinders and a 76.2-mm-square tube), and of an array of 101.6-mm-diam cylinders. A flow model is developed that extrapolates the de-entrainment efficiency of isolated elements throug… more

Date: January 1, 1980

Creator: Dallman, J.C. & Kirchner, W.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Severe accident core heatup transients in modular high temperature gas-cooled reactors without operating Reactor Cavity Cooling Systems

The ultimate decay heat removal system for the current Modular High Temperature Gas-Cooled reactors is a completely passive natural convection air cooling loop. This paper considers an extremely remote accident scenario, where even this passive system fails, and heat rejection is only via a layer of thermal insulation to the reactor silo structure and the surrounding soil. The results show that even in this case the peak fuel temperatures remain well within safe limits. However, vessel and conc… more

Date: January 1, 1988

Creator: Kroeger, P. G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

Date: May 1980

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

SIMMER as a safety analysis tool

SIMMER has been used for numerous applications in fast reactor safety, encompassing both accident and experiment analysis. Recent analyses of transition-phase behavior in potential core disruptive accidents have integrated SIMMER testing with the accident analysis. Results of both the accident analysis and the verification effort are presented as a comprehensive safety analysis program.

Date: January 1, 1982

Creator: Smith, L. L.; Bell, C. R.; Bohl, W. R.; Bott, T. F.; Dearing, J. F. & Luck, L. B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

COBRA-WC: a version of COBRA for single-phase multiassembly thermal hydraulic transient analysis. [LMFBR]

The objective of this report is to provide the user of the COBRA-WC (Whole Core) code a basic understanding of the code operation and capabilities. Included in this manual are the equations solved and the assumptions made in their derivations, a general description of the code capabilities, an explanation of the numerical algorithms used to solve the equations, and input instructions for using the code. Also, the auxiliary programs GEOM and SPECSET are described and input instructions for each … more

Date: July 1, 1980

Creator: George, T. L.; Basehore, K. L.; Wheeler, C. L.; Prather, W. A. & Masterson, R. E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

SACRD: a data base for fast reactor safety computer codes, operational procedures

SACRD (Safety Analysis Computerized Reactor Data) is a data base of nondesign-related information used in computer codes for fast reactor safety analyses. This document reports the procedures used in SACRD to help assure a reasonable level of integrity of the material contained in the data base. It also serves to document much of the computer software used with the data base.

Date: September 1, 1980

Creator: Forsberg, V. M.; Arwood, J. W.; Greene, N. M. & Raiford, G. B.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

HTGR safety philosophy

The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inher… more

Date: August 1, 1980

Creator: Joskimovic, V. & Fisher, C.R.

Item Type: Article

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

open access

Rapid quenching of molten lithium-aluminum jets in water under loss-of-control-rod-cooling conditions

A series of fifteen tests were performed to investigate the thermal interactions between molten LiAl control rod material and water under conditions prototypic of the loss-of-control-rod-cooling (LCRC) accident scenario. The experimental parameters such as melt mass, stream diameter, melt temperature and flowrate, water depth and water temperature were controlled or varied to agree with analytically determined conditions, thus insuring prototypicality of the experiments and applicability of the… more

Date: January 1, 1992

Creator: Greene, G. A.; Finfrock, C. C.; Schwarz, C. E.; Allison, D. K. & Hyder, M. L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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FLOWTRAN benchmarking with onset of flow instability data from 1988 Columbia University single-tube OFI experiment

Benchmarking FLOWTRAN, Version 16.2, with an Onset of Significant Voiding (OSV) criterion against measured Onset of Flow Instability (OFI) data from the 1988--89 Columbia University downflow tests has shown that FLOWTRAN with OSV is a conservative OFI predictor. Calculated limiting flow rates based on the Savannah River Site (SRS) OSV criterion were always higher than the measured flow rates at OFI. This work supplements recent FLOWTRAN benchmarking against 1963 downflow tests at Columbia Unive… more

Date: June 1, 1990

Creator: Chen, K.; Paul, P. K. & Barbour, K. L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Bibliography of reports on research sponsored by the NRC office of nuclear regulatory research, July--December 1977

A bibliography of 198 reports published by contractors of the NRC Office of Nuclear Regulatory Research during the period July through December 1977 is presented along with abstracts from the Nuclear Safety Information Center computer file. The bibliography has been sorted into the subject categories used by NRC to organize the research program. Within the subject categories, the reports are arranged first by contractor organization and then chronologically. A brief description of the NRC resea… more

Date: April 1, 1978

Creator: Buchanan, J.R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Effect of reactor coolant pumps following a small break in a pressurized water reactor

Small-break loss-of-coolant accidents were calculated to help determine whether to trip the reactor-coolant pumps early in the accident when the reactor scrams or to delay the pump trip (pump trip times ranged from 450 s to no trip at all). Four-in.-diam (approximate) cold-leg breaks in Babcock and Wilcox (B and W) and Westinghouse (W) pressurized-water reactors were investigated using the Transient Reactor Analysis Code, TRAC-PD2. The results indicated that for a 4-in.-diam cold-leg break the … more

Date: January 1, 1982

Creator: Elliott, J.L.; Lime, J.F. & Willcutt, G.J.E. Jr.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Characterization of radioactive ion exchange media waste generated at Three Mile Island

The March 1979 accident at General Public Utilities Nuclear Corporation (GPUNC) Three Mile Island Nuclear Power Station Unit 2 (TMI-2), resulted in the transfer of more than 1100 m/sup 3/ of contaminated water to the auxiliary and fuel handling building. The principal sources of the water were the makeup and letdown purification system and the containment building sump. The contaminated water was processed through an ion exchange system designated as EPICOR II. The EPICOR-II System is a three-s… more

Date: October 1, 1981

Creator: Runion, T. C.; Holzworth, R. E.; Ogle, R. E.; Burton, H. M. & Bixby, W. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Passive safety and the advanced liquid metal reactors

Advanced Liquid Metal Reactors being developed today in the USA are designed to make maximum use of passive safety features. Much of the LMR safety work at Argonne National Laboratory is concerned with demonstrating, both theoretically and experimentally, the effectiveness of the passive safety features. The characteristics that contribute to passive safety are discussed, with particular emphasis on decay heat removal systems, together with examples of Argonne's theoretical and experimental pro… more

Date: January 1, 1988

Creator: Hill, D.J.; Pedersen, D.R. & Marchaterre, J.F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

PNL technical review of pressurized thermal-shock issues. [PWR]

Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the r… more

Date: July 1, 1982

Creator: Pedersen, L. T.; Apley, W. J.; Bian, S. H.; Defferding, L. J.; Morgenstern, M. H.; Pelto, P. J. et al.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Fission product release from high gap-inventory LWR fuel under LOCA conditions

Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200/sup 0/C. The basis for the model was tes… more

Date: January 1, 1980

Creator: Lorenz, R. A.; Collins, J. L.; Osborne, M. F. & Malinauskas, A. P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Comparison of TRAC and RELAP5 reactor system calculations for a DEGB LOCA in K-14. 1

A comparison of TRAC and RELAP5 predictions of steady-state and DEGB LOCA results (FI phase) for K-14.1 has been made. Both codes had been previously benchmarked against 1985 L Reactor AC Flow data and were under configuration control. The purpose of the code-to-code comparison is to provide insight on the transient uncertainty in TRAC plenum and tank bottom plenum pressures. The comparisons focus on LOCA results between 0.5 and 2.0 s, which is the primary period of interest for Flow Instabilit… more

Date: September 1, 1990

Creator: Griggs, D.P. (Westinghouse Savannah River Co., Aiken, SC (United States)) & Liebmann, M.L. (Wais and Associates (United States))

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Seismic hazard evaluation for the high-flux isotope reactor (HFIR) Oak Ridge National Laboratory, Oak Ridge, Tennessee

This study investigates the probabilistic hazard of earthquake-induced ground shaking at the HFIR facility, Oak Ridge, Tennessee. These results will be used to calculate plant response and potential effects in a Probabilistic Risk Assessment (PRA). For this purpose, several guidelines apply to this work. First, both the frequency of exceedance and the uncertainty in frequency of exceedance of various ground motion levels must be represented. These are required by the PRA so that the frequency a… more

Date: September 1, 1991

Creator: McGuire, R.K. & Toro, G.R. (Risk Engineering, Inc., Golden, CO (United States))

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Natural convection phenomena in a nuclear power plant during a postulated TMLB' accident

After the TMI (Three Mile Island) accident, there has been significant interest in analyzing and understanding the phenomena that may occur in a PWR (Pressurized Water Reactor) accident which may lead to partial or total core meltdown and degradation. Natural convection is one of the important phenomena. In the present paper the results of two numerical simulations of (1) four-loop PWR and (2) three-loop PWR are presented. The simulations were performed with the COMMIX(2) computer code. Our ana… more

Date: January 1, 1987

Creator: Domanus, H. M.; Schmitt, R. C.; Sha, W. T.; Shah, V. L. & Han, J. T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Integral fast reactor safety features

The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on… more

Date: January 1, 1988

Creator: Cahalan, J. E.; Kramer, J. M.; Marchaterre, J. F.; Mueller, C. J.; Pedersen, D. R.; Sevy, R. H. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Revised GCFR safety program plan

This paper presents a summary of the recently revised gas-cooled fast breeder reactor (GCFR) safety program plan. The activities under this plan are organized to support six lines of protection (LOPs) for protection of the public from postulated GCFR accidents. Each LOP provides an independent, sequential, quantifiable risk barrier between the public and the radiological hazards associated with postulated GCFR accidents. To implement a quantitative risk-based approach in identifying the importa… more

Date: May 1, 1980

Creator: Kelley, A. P.; Boyack, B. E. & Torri, A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Probability based load criteria for the design of nuclear structures: a critical review of the state-of-the-art

Structures for nuclear power plant facilities must be designed to withstand safely and effectively all kinds of loads and load combinations that may be expected to occur during their lifetime. The traditional methods of structural design attempt to account for the inevitable variability in the loads, material strengths, in-service environments, and fabrication process, etc., through the use of safety factor, allowable stresses or load and resistance factors. These approaches may result in an un… more

Date: April 1, 1981

Creator: Shinozuka, M.; Ellingwood, B.R.; Wang, P.C.; Meyer, C.; Wen, Y.K.; Kao, S. et al.

Item Type: Report

Partner: UNT Libraries Government Documents Department