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in subjects: "22 General Studies Of Nuclear Reactors"

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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

Nuclear facility decommissioning and site remedial actions. Volume 6. A selected bibliography

This bibliography of 683 references with abstracts on the subject of nuclear facility decommissioning, uranium mill tailings management, and site remedial actions is the sixth in a series of annual reports prepared for the US Department of Energy's Remedial Action Programs. Foreign as well as domestic literature of all types - technical reports, progress reports, journal articles, conference papers, symposium proceedings, theses, books, patents, legislation, and research project descriptions - … more

Date: September 1, 1985

Creator: Owen, P. T.; Michelson, D. C. & Knox, N. P.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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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

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Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

This report highlights Brookhaven National Laboratory's activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclea… more

Date: January 1, 1991

Creator: Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L. & Kuper, J.B.H.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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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

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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

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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

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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

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Safety rod latch inspection

During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small button'' in the latch mechanism had broken off of the lock plunger'' and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and met… more

Date: February 1, 1992

Creator: Leader, D.R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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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

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Analysis of the results of the Midland PRA

This paper presents the results of a limited review of the Midland PRA (MPRA), aimed at facilitating regulatory analyses and at providing insights into safety related plant failures. In particular, stress was laid on the root causes of accident sequences, particularly their failure modes, viz., hardware, human, maintenance, test, and repair. Because this information was difficult to extract in this detailed and sophisticated PRA (and in certain others) a special algorithm was developed to displ… more

Date: January 1, 1986

Creator: Bozoki, G. & Teichmann, T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Second Nuclear Era

The Institute for Energy Analysis with support from The Andrew W. Mellon Foundation has studied the decline of the present nuclear era in the United States and the characteristics of a Second Nuclear Era which might be instrumental in restoring nuclear power to an appropriate place in the energy options of our country. The study has determined that reactors operating today are much safer than they were at the time of the TMI accident. A number of concepts for a supersafe reactor were reviewed a… more

Date: March 1, 1984

Creator: Weinberg, Alvin M.; Spiewak, Irving; Barkenbus, Jack N.; Livingston, Robert S. & Phung, Doan L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Use of principal components analysis and three-dimensional atmospheric-transport models for reactor-consequence evaluation

This work explores the use of principal components analysis coupled to three-dimensional atmospheric transport and dispersion models for evaluating the environmental consequences of reactor accidents. This permits the inclusion of meteorological data from multiple sites and the effects of topography in the consequence evaluation; features not normally included in such analyses. The technique identifies prevailing regional wind patterns and their frequencies for use in the transport and dispersi… more

Date: January 1, 1982

Creator: Gudiksen, P. H.; Walton, J. J.; Alpert, D. J. & Johnson, J. D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Testing program for determining the mechanical properties of concrete to temperatures of 621/sup 0/C

Concrete temperatures in a Liquid Metal Fast Breeder Reactor (LMFBR) in excess of normal code limits can result from postulated large sodium spills in equipment cells. Elevated temperature concrete property data which may have application for providing a basis for the design and evaluation of such postulated accident conditions is limited. Data thus needed to be developed commensurate with LMFBR plant applications for critical physical and mechanical concrete properties under prototypic thermal… more

Date: January 1, 1980

Creator: Oland, C.B.; Naus, D.J. & Robinson, G.C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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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

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Characterization of solids in the Three Mile Island Unit 2 Reactor defueling water: Addendum

Shortly after ORNL/TM-10362 was issued, it was discovered that a series of 31 figures had been inadvertently omitted. These figures, which consist of scanning electron microscope (SEM) photographs and energy-dispersive X-ray fluorescence scans, provide significant information about the results obtained in the tests performed with water sample W3. This Addendum includes these figures. Details of and comments on the SEM photographs may be found in ORNL/TM-10362.

Date: March 1, 1988

Creator: Campbell, D.O.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Development of a computer program for reactor core thermal hydraulics: finite differencing in a 3-D staggered mesh suitable for irregular boundary applications

A formulation utilizing a finite difference scheme for the direct numerical simulation of time varying two-phase flow problems in three space dimensions with irregular boundary geometry is presented. The formulation of such problems introduces a complicated set of coupled non-linear partial differential equations for the conservation of mass, momentum, energy, appropriate constitutive equations, and equations of state coupling the two phases. Discretization of the partial differential equations… more

Date: unknown

Creator: Sha, W.T.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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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

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR]

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-… more

Date: February 1, 1978

Creator: Benedetti, R. L.; Lords, L. V. & Kiser, D. M.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Source-jerk analysis using a semi-explicit inverse kinetic technique

A method is proposed for measuring the effective reproduction factor, k, in subcritical systems. The method uses the transient response of a subcritical system to the sudden removal of an extraneous neutron source (i.e., a source jerk). The response is analyzed using an inverse kinetic technique that least-squares fits the exact analytical solution corresponding to a source-jerk transient as derived from the point-reactor model. It has been found that the technique can provide an accurate means… more

Date: January 1, 1985

Creator: Spriggs, G.D. & Pederson, R.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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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

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Integrity of PWR pressure vessels during overcooling accidents

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance… more

Date: January 1, 1982

Creator: Cheverton, R.D.; Iskander, S.K. & Whitman, G.D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Computer program SCAP-BR for gamma-ray streaming through multi-legged ducts

A computer program, SCAP-BR, has been developed at Burns and Roe for the gamma-ray streaming analysis through multi-legged ducts. SCAP-BR is a modified version of the single scattering code, SCAP, incorporating capabilities of handling multiple scattering and volumetric source geometries. It utilizes the point kernel integration method to calculate both the line-of-sight and scattered gamma dose rates by employing the ray tracing technique through complex shield geometries. The multiple scatter… more

Date: December 8, 1977

Creator: Byoun, T. Y.; Babel, P. J. & Dajani, A. T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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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