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Thermal-hydraulic modeling of porous bed reactors

Description: Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 ..mu..m in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs.
Date: January 1, 1987
Creator: Araj, K.J. & Nourbakhsh, H.P.
Partner: UNT Libraries Government Documents Department

Annular core for Modular High-Temperature Gas-Cooled Reactor (MHTGR)

Description: The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93-m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.
Date: August 1, 1987
Creator: Turner, R.F.; Baxter, A.M.; Stansfield, O.M. & Vollman, R.E.
Partner: UNT Libraries Government Documents Department

Measurement and modelling of postirradiation fission product release from HTGR fuel particles under accident conditions

Description: A study was performed to provide a description of the release of fission products from failed fuel particles during a core heatup event in an HTGR. The need for this study was established in the Accident Initiation and Progression Analysis program. The release of fission products was measured from laser-failed BISO ThO/sub 2/, TRISO UC/sub 2/, and weak acid resin (WAR) particles over a range of burnups. The burnups were 0.25, 1.4 and 15.7% FIMA for ThO/sub 2/ particles, 23.5 and 74% FIMA for UC/sub 2/ particles, and 60% FIMA for WAR particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium. Two types of experiments were performed: isothermal and temperature rise experiments. The range of the temperatures was from 1200/sup 0/ to 2300/sup 0/C. In the temperature rise experiments, the heating rates were between 50/sup 0/ and 450/sup 0/C/h.
Date: December 1, 1978
Creator: Myers, B.F. & Morrissey, R.E.
Partner: UNT Libraries Government Documents Department

Postirradiation examination of capsule GF-4. [HTGR]

Description: The GF-4 capsule test was irradiated in the SILOE reactor at Grenoble, France between April 8, 1975 and July 26, 1976. High-enriched uranium (HEU) UC/sub 2/ and weak acid resin (WAR) UC/sub x/O/sub y/ fissile and ThO/sub 2/ fertile particles were tested. Postirradiation examination of cured-in-place fuel rods showed no fuel rod/graphite element interaction. In addition, all rods exhibited adequate structural integrity. Irradiation-induced dimensional changes for rods containing all TRISO-coated fuel were consistent with model predictions; however, rods containing BISO-coated fuel exhibited greater volumetric contractions than predicted.
Date: October 1, 1980
Creator: Kovacs, W.J. & Sedlak, B.J.
Partner: UNT Libraries Government Documents Department

Irradiation performance of HTGR fuel in HFIR experiment HRB-13

Description: Irradiation capsule HRB-13 tested High-Temperature Gas-Cooled Reactor (HTGR) fuel under accelerated conditions in the High Flux Isotope Reactor (HFIR) at ORNL. The ORNL part of the capsule was designed to provide definitive results on how variously misshapen kernels affect the irradiation performance of weak-acid-resin (WAR)-derived fissile fuel particles. Two batches of WAR fissile fuel particles were Triso-coated and shape-separated into four different fractions according to their deviation from spericity, which ranged from 9.6 to 29.7%. The fissile particles were irradiated for 7721 h. Heavy-metal burnups ranged from 80 to 82.5% FIMA (fraction of initial heavy-metal atoms). Fast neutron fluences (>0.18 MeV) ranged from 4.9 x 10/sup 25/ neutrons/m/sup 2/ to 8.5 x 10/sup 25/ neutrons/m/sup 2/. Postirradiation examination showed that the two batches of fissile particles contained chlorine, presumably introduced during deposition of the SiC coating.
Date: March 1, 1982
Creator: Tiegs, T.N.
Partner: UNT Libraries Government Documents Department

Nondestructive assay of green HTGR fuel rods

Description: This report describes the nondestructive (NDA) work done at Los Alamos during 1979 and 1980 as part of the New Brunswick Laboratory-sponsored evaluation of NDA of the uranium content of fabricated fuel rods for high-temperature gas-cooled reactors (HTGR). The methods used (delayed neutron and passive gamma ray) are concisely described, and the results are summarized and compared in graphical and tabular form. The results indicate that, with the use of proper physical standards, accuracies within about 1 percent should be achievable by NDA procedures.
Date: May 1, 1981
Creator: Barschall, H.H.; Meier, M.M. & Parker, J.L.
Partner: UNT Libraries Government Documents Department

Postirradiation examination and evaluation of Fort St. Vrain fuel element 1-0743

Description: Fort St. Vrain (FSV) fuel element 1-0743 was irradiated in core location 17.04.F.06 from July 3, 1976 until February 1, 1979. The element experienced an average fast neutron exposure of about 0.95 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/, a time-and-volume-averaged fuel temperature in the vicinity of 680/sup 0/C, fissile and fertile particle burnups of approximately 6.2% and 0.3%, respectively, and a total burnup of 12,210 MWd/tonne. The postirradiation examination revealed that the element was in excellent condition. No cracks were observed on any of the element surfaces. The structural integrity of the fuel rods was good. No evidence of mechanical interaction between the fuel rods and fuel body was observed. Calculated irradiation parameters obtained with HTGR design codes were compared with measured data. Radial and axial power distributions, irradiation temperatures, neutron fluences, and fuel burnups were in good agreement with measurements. Calculated fuel rod strains were about a factor of three greater than were observed.
Date: May 1, 1981
Creator: Saurwein, J.J.; Miller, C.M. & Young, C.A.
Partner: UNT Libraries Government Documents Department

HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release]

Description: The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.
Date: September 1, 1977
Partner: UNT Libraries Government Documents Department

HTGR Generic Technology Program Fuels and Core Development Program quarterly progress report for the period ending May 31, 1978. [Graphite and fuel irradiation; fission product release]

Description: The work reported includes the design, analysis, and testing of the reactor core and its components comprising the fuel elements, hexagonal reflector elements, plenum elements, neutron sources, control rods, and reserve shutdown material. Also included are studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.
Date: June 1, 1978
Partner: UNT Libraries Government Documents Department