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in subjects: "Reactor Components"

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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

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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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

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

open access

First wall/blanket/shield design and optimization system

First wall/blanket/shield design and optimization system (BSDOS) has been developed to provide a state-of-the-art design tool for fast accurate analysis. In addition, it has been designed to perform several other functions: (1) allowing comparison and evaluation studies for different concepts using the same data bases and ground rules, (2) permitting the use of any figure of merit in the evaluation studies, (3) optimizing the first wall/blanket/shield design parameters for any figure of merit u… more

Date: February 1, 1988

Creator: Gohar, Y.; Baker, C.; Attaya, H.; Cha, Y.; Majumdar, S. & Scandora, T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

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

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

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

open access

TCT hybrid preconceptual blanket design studies

The conceptual design of a tokamak fusion-fission (hybrid) reactor, which produces electric power and fissile material, has been performed in a cooperative effort between Princeton's Plasma Physics Laboratory (PPPL) and Battelle's Pacific Northwest Laboratories (PNL). PPPL, who had overall project lead responsibility, designed the fusion driver system. Its core consists of a tokamak plasma maintained in the two-component torus (TCT) mode by both D and T beams and having a single null poloidal d… more

Date: January 1, 1978

Creator: Aase, D. T.; Bampton, M. C. C.; Doherty, T. J.; Leonard, B. R.; McCann, R. A.; Newman, D. F. et al.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Irradiation experiment for lithium ceramics

Solid breeder materials are required in D-T fusion reactor blankets to convert fusion neutrons to tritium and thermal energy while providing some neutron shielding for the magnets. Lithium compounds such as Li/sub 2/O, LiAlO/sub 2/, Li/sub 4/SiO/sub 4/ and Li/sub 2/ZrO/sub 3/ have been proposed as solid breeder materials. Tritium is necessary to maintain the fuel supply for the D-T fusion reaction. No high burnup irradiation performance data is available for these materials in a neutron environ… more

Date: March 1, 1982

Creator: Hollenberg, G.W.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

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

open access

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

open access

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, December 1, 1979-February 29, 1980

Information is presented concerning bundle geometry with wrapped and bare rods; LMFBR outlet plenum flow mixing; and theoretical determination of local temperature fields in LMFBR fuel rod bundles.

Date: January 1, 1980

Creator: Todreas, N.E.; Golay, M.W. & Wolf, L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Machining of Uranium for Brookhaven Reactor

No Description Available.

Date: June 15, 1949

Creator: Kemmer, F. R.; Musgrave, T. P. & Fox, G. E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Acoustic measurements of the boiling stability tests on THORS sodium loop. [LMFBR]

Acoustic data of boiling stability tests on the THORS (Thermal-Hydraulic Out-of-Reactor Safety) facility were obtained using three sodium-immersible high temperature microphones. The data was analyzed in both the time and frequency domains and provides the following information: (1) the acoustic signal due to sodium boiling was clearly observed; (2) the signal level and the repetition rate of boiling pulses are directly proportional to the applied heat flux; (3) a typical boiling pulse consists… more

Date: January 1, 1977

Creator: Sheen, S.H.; Bobis, J.P. & Carey, W.M.

Item Type: Article

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

Structural materials for breeder reactor cores and coolant circuits

The structural components of principal interest in LMFBR cores and cooling circuits include the reactor vessel, primary and secondary piping, intermediate heat exchanger (IHX), and steam generator. Load-bearing components inside the vessel, among these the fuel cladding and duct, are also included. The operating conditions present in a fast-breeder nuclear reactor impose a number of requirements on the mechanical, physical, and neutronic properties of the materials used to construct these compo… more

Date: February 1, 1984

Creator: Diercks, D. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Analysis of fission product behavior in the Saclay Spitfire Loop Test SSL-1. [HTGR]

The behavior of the fission metal cesium and the fission gases krypton and xenon in the Saclay Spitfire Loop SSL-1 test has been compared to that predicted using General Atomic reference data and computer code models. This is the first in a series of analyses planned in order to provide quantitative validation of HTGR fission product design methods. In this analysis, the first attempt to rigorously verify fission product design methods, the FIPERQ code was used to model the diffusion of cesium … more

Date: February 1, 1978

Creator: Jensen, D. D.; Haire, M. J. & Ballagny, A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Design configuration of GCFR core assemblies

The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the … more

Date: May 1, 1980

Creator: LaBar, M.P.; Lee, G.E. & Meyer, R.J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Deposition and removal of radioactive isotopes from LMFBR components

The development of an analytical model to describe the production, transport and eventual removal of radioactive materials in the primary sodium of LMFBR's is a continuing Sodium Technology activity sponsored by the Department of Energy. This paper describes studies directed toward obtaining an understanding of the deposition from sodium of fuel cladding activated corrosion products onto stainless steel alloys and the effect of their diffusion into the base metal on the process required to deco… more

Date: January 1, 1980

Creator: Hill, E. F.; Lutton, J. M. & Maffei, H. P.

Item Type: Article

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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Once-through steam generator (OTSG) materials and water chemistry. [PWR]

Materials and water chemistry research results associated with the development of the Oconee-1 Reactor steam generator are presented. A summary of water chemistry data acquired during preoperational testing and power operation to date is also included. These data confirm the operational practicality of the nuclear once-through concept using volatile water treatment and high purity condensate demineralized feedwater.

Date: January 1, 1974

Creator: Pocock, F.J. & Levstek, D.F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

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

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Mirror hybrid reactor studies

The hybrid reactor studies are reviewed. The optimization of the point design and work on a reference design are described. The status of the nuclear analysis of fast spectrum blankets, systems studies for fissile fuel producing hybrid reactor, and the mechanical design of the machine are reviewed. (MHR)

Date: January 19, 1978

Creator: Bender, D.J.

Item Type: Article

Partner: UNT Libraries Government Documents Department