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in subjects: "Reactor Accidents"

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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

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

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

open access

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

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

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

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

LOFT system structural response during subcooled blowdown

The Loss-of-Fluid Test (LOFT) facility is a highly instrumented, pressurized water reactor test system designed to be representative of large pressurized water reactors (LPWRs) for the simulation of loss-of-coolant accidents (LOCAs). Detailed structural analysis and appropriate instrumentation (accelerometers and strain gages) on the LOFT system provided information for evaluation of the structural response of the LOFT facility for loss-of-coolant experiment (LOCE) induced loads. In general, th… more

Date: January 1, 1978

Creator: Martinell, J. S.

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

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

Mechanical strength and stability of lithium aluminate

Pacific Northwest Laboratory (PNL) investigated the strength and resistance to thermal shock of lithium aluminate annular pellets. The room temperature, axial compressive fracture strength of pellets made at Westinghouse Advanced Energy Systems (WAES) varied from 80 to 133 ksi. The strength at 430{degrees}C (806{degrees}F) was to 30 to 40% lower. The strength at 900{degrees}C (1652{degrees}F) showed a wide variation with one measurement near 90 ksi. These strength values are consistent with oth… more

Date: June 1, 1992

Creator: Brimhall, J.L.

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

open access

Quarterly Progress Report on Fission Product Behavior in LWRs for the Period October-December 1977

Analysis of release data obtained during High Burnup Fuel Test 10 (HBU-10) has been completed. In this test the fuel rod segment was ruptured by internal pressurization at 900/sup 0/C, at which time the temperature was rapidly increased to 1200/sup 0/C and maintained at this temperature for 10 min. Approximately 0.061% of the total cesium inventory in the rod segment was released; this was accompanied by the release of about 1.69% of the total /sup 85/Kr inventory. Moreover, about 0.022% of the… more

Date: February 1, 1978

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

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Radiation protection: an analysis of thyroid blocking. [Effectiveness of KI in reducing radioactive uptake following potential reactor accident]

An analysis was performed to provide guidance to policymakers concerning the effectiveness of potassium iodide (KI) as a thyroid blocking agent in potential reactor accident situations, the distance to which (or area within which) it should be distributed, and its relative effectiveness compared to other available protective measures. The analysis was performed using the Reactor Safety Study (WASH-1400) consequence model. Four categories of accidents were addressed: gap activity release acciden… more

Date: January 1, 1980

Creator: Aldrich, D.C. & Blond, R.M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Some Scoping Experiments for a Space Reactor

Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failur… more

Date: July 7, 1983

Creator: Alexander, C. A. & Ogden, J. S.

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