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in subjects: "Reactors"

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Hot helium flow test facility summary report

This report summarizes the results of a study conducted to assess the feasibility and cost of modifying an existing circulator test facility (CTF) at General Atomic Company (GA). The CTF originally was built to test the Delmarva Power and Light Co. steam-driven circulator. This circulator, as modified, could provide a source of hot, pressurized helium for high-temperature gas-cooled reactor (HTGR) and gas-cooled fast breeder reactor (GCFR) component testing. To achieve this purpose, a high-temp… more

Date: June 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department

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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

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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850/sup 0/C VHTR plant technical description

This report describes the conceptual design of an 842-MW(t) process heat very high temperature reactor (VHTR) plant having a core outlet temperature of 850/sup 0/C (1562/sup 0/F). The reactor is a variation of the high-temperature gas-cooled reactor (HTGR) power plant concept. The report includes a description of the nuclear heat source (NHS) and of the balance of reactor plant (BORP) requirements. The design of the associated chemical process plant is not covered in this report. The reactor de… more

Date: June 1, 1980

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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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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Nuclear power in the Soviet Bloc

The growth of Soviet Bloc nuclear power generation to the end of the century is evaluated on the basis of policy statements of objectives, past and current nuclear power plant construction, and trends in the potential for future construction. Central to this study is a detailed examination of individual reactor construction and site development that provides specific performance data not given elsewhere. A major commitment to nuclear power is abundantly clear and an expansion of ten times in nu… more

Date: March 1, 1982

Creator: Davey, W.G.

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

open access

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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Configuration Management Program Plan

Westinghouse Savannah River Company (WSRC) has established a configuration management (CM) plan to execute the SRS CM Policy and the requirements of the DOE Order 4700.1. The Reactor Restart Division (RRD) has developed its CM Plan under the SRS CM Program and is implementing it via the RRD CM Program Plan and the Integrated Action Plan. The purpose of the RRD CM program is to improve those processes which are essential to the safe and efficient operation of SRS production reactors. This docume… more

Date: September 25, 1991

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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Thorium fuel cycles for LWRs: fuel diversion assessments and recycle requirements

A number of fuel cycles have been proposed for evaluation in the nonproliferation alternative systems assessment program. Among these systems are light water reactors (LWR) operating on the thorium-uranium cycle or the plutonium-thorium cycle either inside or outside energy centers. These proposals support the President's nuclear power policy of accelerating research into alternative fuel cycles that do not permit direct access to materials usable for nuclear weapons but still retain the benefi… more

Date: January 1, 1978

Creator: Carter, W. L.; Rainey, R. H. & Johnson, D. R.

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

Evaluation of environmental dosimetry models for applicability to possible radioactive waste repository discharges

This report presents the results of a review of the available codes, for application to the National Waste Terminal Storage (NWTS) Program. Consideration was given to the types of radionuclides which may be of concern, the possible modes of release of these radionuclides and the various pathways by which members of the general public in the vicinity of a terminal storage facility may be exposed to the releases. Results reveal that the types and quantities of radioactive material requiring dispo… more

Date: September 1, 1977

Creator: Mauro, J. J.; Michlewicz, D. & Letizia, A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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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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An Evaluation of liquid metal leak detection methods for the Clinch River Breeder Reactor Plant

This report documents an independent review and evaluation of sodium leak detection methods described in the Clinch River Breeder Reactor Preliminary Safety Analysis Report. Only information in publicly available documents was used in making the assessments.

Date: December 1, 1977

Creator: Morris, C. J. & Doctor, S. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of i… more

Date: January 1, 1982

Creator: Wilson, W. B.; LaBauve, R. J. & England, T. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Design and Feasibility Study of a Pebble Bed Reactor-Steam Power Plant

The status of development studies being conducted on a pebble bed power reactor is outlined. The items discussed are fuel element manufacture, stability, and reprocessing, and component development. (D.L.C.)

Date: May 1, 1958

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Interaction of noble-metal fission products with pyrolytic silicon carbide

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain layers of pyrolytic carbon and silicon carbide, which act as a miniature pressure vessel and form the primary fission product barrier. Of the many fission products formed during irradiation, the noble metals are of particular interest because they interact significantly with the SiC layer and their concentrations are somewhat higher in the low-enriched uranium fuels currently under consideration. To study fission product-… more

Date: January 1, 1982

Creator: Lauf, R. J. & Braski, D. N.

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