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in subjects: "Safety 220900* -- Nuclear Reactor Technology-- Reactor Safety"

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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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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

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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

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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

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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

Probabilistic risk assessment of HTGRs

Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the US Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed.

Date: August 1980

Creator: Fleming, K. N.; Houghton, W. J.; Hannaman, G. W. & Joksimovic, V.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Liquid metal fast breeder reactor safety: an overview including implications of alternate fuel cycles

Key issues concerning LMFBR safety are reviewed. A brief discussion of alternate fuel cycles for preventing diversion of fissionable materials is included.

Date: January 1, 1977

Creator: Fauske, H. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Physics of reactor safety. Volume II. Quarterly report, April-June 1980

The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

Date: August 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Inherent design features of the GCFR

This paper discusses several inherent design features of the GCFR that enhance its safety and presents analyses to demonstrate the degree of protection they provide. These features are a subset of a larger group of potential inherent features that form the third line of protection (LOP-3) for the GCFR. The function of LOP-3 is to demonstrate that the inherent response of the reactor system will limit core damage even if active cooling and shutdown systems in LOP-1 and LOP-2 fail. By providing t… more

Date: May 1, 1980

Creator: Medwid, W.; Breher, W.; Shenoy, A. & Elliott, R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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LMFBR system-wide transient analysis: the state of the art and US validation needs

This paper summarizes the computational capabilities in the area of liquid metal fast breeder reactor (LMFBR) system-wide transient analysis in the United States, identifies various numerical and physical approximations, the degree of empiricism, range of applicability, model verification and experimental needs for a wide class of protected transients, in particular, natural circulation shutdown heat removal for both loop- and pool-type plants.

Date: January 1, 1982

Creator: Khatib-Rahbar, M.; Guppy, J.G. & Cerbone, R.J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Self-mixing phenomenology in hypothetical core-disruptive accidents

Physical processes are investigated that lead to the thermal equilibration of a disrupted liquid metal fast breeder reactor (LMFBR) core following a hypothetical core-disruptive accident (HCDA). Their impact is assessed, particularly as relating to the SIMMER code. The turbulent structure in the core region is characterized and bounding estimates are derived of thermal equilibration (''self-mixing'') times. The implication of these results for LMFBR safety research is discussed briefly.

Date: December 1, 1980

Creator: Chapyak, E.J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Physics of reactor safety. Quarterly report, July-September 1980. Volume III

This Quarterly progress report summarizes work done during the months of July-September 1980 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Division of Reactor Safety Research of the US Nuclear Regulatory Commission. The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology… more

Date: November 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Large scale sodium test facility. [LMFBR]

The large scale sodium test facility in use at Sandia Laboratories was designed to be used for a wide range of sodium interaction experiments. Of initial concern is the interaction of hot sodium with concrete under conditions similar to accident conditions in any sodium cooled reactor. The tests to be run cover the cases of sodium spills and sprays on bare concrete and on cells with defective steel liners. The initial series of tests in progress at the facility involves rapidly dropping from 20… more

Date: January 1, 1977

Creator: King, D.L.; Smaardyk, J.E.; Sutherland, H.J. & Sallach, R.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Amplified-response-spectrum analysis of sodium-water reaction pressure waves. [LMFBR]

This report deals with a frequency spectrum evaluation of the SWAAM I predicted double rupture disc assembly operation pressure wave generated in the LLTR Series II A-2 test. It also evaluates the same wave predicted by the TRANSWRAP II code and the pressure wave actually measured upstream of the rupture disc assembly by the test instrumentation in Test A-2. The SWAAM I and TRANSWRAP II codes currently use the same analytical model to characterize the rupture disc until the disc strikes the kni… more

Date: October 28, 1981

Creator: Knittle, D.E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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General formulation of an HCDA bubble rising in a sodium pool and the effect of nonequilibrium on fuel transport. [LMFBR]

This report which improved the formulation of the previous reports is designed to investigate the effect of the interfacial nonequilibrium mass transfer and the radiative heat transfer on the amount of the fuel vapor condensed before the bubble reaches to the cover-gas region. Consideration is given to a fuel dominated bubble which is assumed to have just penetrated into the sodium pool in a spherical form subsequent to an Hypothetical Core Disruptive Accident (HCDA). The two-phase bubble mixtu… more

Date: January 1, 1980

Creator: Kocamustafaogullari, G. & Chan, S.H.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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WSRC Reactor Tank Inspection Program (RTIP) status report

Westinghouse Savannah River Company (WSRC) recently completed the initial phase of nondestructive inspections of the Savannah River Site's (SRS) reactor tanks. This program required almost three years to be conceptualized, fabricated, and tested. An additional 20 months were required to complete the NDE inspection of the P, K and L reactor tanks. The overall cost of the program to date is approximately $25 MM. This status report will address: (1) A brief review of the RTIP program and the const… more

Date: January 1, 1992

Creator: Loibl, M. W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Substitute safety rods: Physics design and NTG calibration

Under certain assumed accident conditions, an SRS reactor may loose most of its bulk moderator while maintaining flow to fuel assemblies. If this occurs immediately after operation at power, components normally dependent on convective heat transfer to the moderator will heat up with the possibility of melting that component. One component at risk is the safety rod. Tests have shown that the current cadmium safety rod, which contains aluminum as well as cadmium, can fail at temperatures only sli… more

Date: July 1, 1991

Creator: Baumann, N.P.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Critical time step estimation for three-dimensional explicit impact analysis

A method to obtain a bound on the critical time step of impact problems that employ explicit integration and penalty type contact elements is presented. The method uses a Gerschgorin bound obtained from a partially assembled stiffness matrix for the nodes that belong to elements involved in the contact region. A numerical example is presented to illustrate the method's effectiveness. 6 refs., 4 figs., 1 tab.

Date: January 1, 1989

Creator: Kulak, R.F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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A simple model for radial expansion reactivity in LMRs (liquid metal reactors)

Presented in this report is a simple analytical model developed for evaluating the radial expansion reactivity in small modular liquid metal reactors (LMRs). The present model is based on a non-leakage representation of the effective neutron multiplication factor. The resultant analytical expression for the radial expansion reactivity is simple and can be used directly in a system code for safety analyses. Applications of the present model to PRISM and SAFR resulted in a good agreement with the… more

Date: January 1, 1988

Creator: Cheng, H.S. & Van Tuyle, G.J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Designation of facility usage categories for Hanford Site facilities

This report summarizes the Hanford Site methodology used to ensure facility compliance with the natural phenomena design criteria set forth in the US Department of Energy Orders and guidance. The current Hanford Site methodology for Usage Category designation is based on an engineered feature's safety function and on the feature's assigned Safety Class. At the Hanford Site, Safety Class assignments are deterministic in nature and are based on teh consequences of failure, without regard to the l… more

Date: October 1, 1991

Creator: Woodrich, D. D.; Ellingson, D. R.; Scott, M. A. & Schade, A. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Meteorological evaluation of multiple reactor contamination probabilities for a Hanford Nuclear Energy Center

The conceptual Hanford energy center is composed of nuclear power plants, hence the name Hanford Nuclear Energy Center (HNEC). Previous topical reports have covered a variety of subjects related to the HNEC including: electric power transmission, fuel cycle, and heat disposal. This report discusses the probability that a radiation release from a single reactor in the HNEC would contaminate other facilities in the center. The risks, in terms of reliability of generation, of this potential contam… more

Date: March 1, 1978

Creator: Ramsdell, J.V. & Diebel, D.I.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Compressible analysis of inlet plenum pressure rise due to sodium boiling in fuel subassemblies during pump coastdown of an LMFBR

The effect of sodium compressibility and steel elasticity on the rise in inlet plenum pressure occurring during boiling in a loss-of-flow accident in an LMFBR has been investigated using the require consideration in accident analysis. The pressure rise is less for pool than for loop designs. 3 refs., 1 fig., 9 tabs.

Date: May 1, 1980

Creator: Kalimullah & Hummel, H.H.

Item Type: Report

Partner: UNT Libraries Government Documents Department