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Core seismic methods verification report

Description: This report presents the description and validation of the analytical methods for calculation of the seismic loads on an HTGR core and the core support structures. Analytical modeling, integration schemes, parameter assignment, parameter sensitivity, and correlation with test data are key topics which have been covered in detail. Much of the text concerns the description and the results of a series of scale model tests performed to obtain data for code correlation. A discussion of scaling laws, model properties, seismic excitation, instrumentation, and data reduction methods is also presented, including a section on the identification and calculation of statistical errors in the test data.
Date: December 1, 1979
Creator: Olsen, B.E.; Shatoff, H.D.; Rakowski, J.E.; Rickard, N.D.; Thompson, R.W.; Tow, D. et al.
Partner: UNT Libraries Government Documents Department

Investigation of stick-slip (chatter) phenomenon of HTGR thermal barrier attachment fixture sliding interfaces. Phase I test: Class A thermal barrier hardware and environment

Description: This test program was performed to investigate if significant chatter (stick-slip) would occur at the thermal barrier sliding surfaces. Given such conditions, cyclic loads could be induced in the thermal barrier attachment fixture and studs. A representative section of thermal barrier was tested with realistic HTGR temperature cycles in a high purity helium environment. No significant chatter was detected and there was no visible deterioration of the hardware after testing.
Date: July 1, 1979
Creator: Middleton, A.
Partner: UNT Libraries Government Documents Department

Fission-product retention in HTGR fuels

Description: Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed.
Date: January 1, 1982
Creator: Homan, F.J.; Kania, M.J. & Tiegs, T.N.
Partner: UNT Libraries Government Documents Department

Status of Prestressed Concrete Reactor Vessel (PCRV) experimental and analytical programs in the United States. [HTGR]

Description: During the past decade, the Oak Ridge National Laboratory has been engaged in a comprehensive program of experimental and analytical studies pertaining to the design and development of PCRVs. The program has been directed primarily toward the gas-cooled reactor since it has remained the only reactor concept currently utilizing PCRVs. However, interest has developed recently in potential applications to coal conversion systems. The purpose of the paper is to review the background and scope of the PCRV Research and Development Program and to summarize the status of the current studies.
Date: January 1, 1977
Creator: Callahan, J P & Dodge, W G
Partner: UNT Libraries Government Documents Department

Performance of HTGR fuel in HFIR capsule HT-33

Description: Irradiation capsule HT-33 was a cooperative effort between General Atomic Company (GA) and Oak Ridge National Laboratory (ORNL). In this capsule ThO/sub 2/ particles (fabricated by GA), low-enriched uranium particles, inert carbon particles, and various fuel rod matrices were tested under accelerated irradiation in the High-Flux Isotope Reactor. Visual examination showed good irradiation behavior for fuel rods with slug-injected matrices (using a pitch binder) and warm-molded matrices (using a thermosetting resin binder). Rod debonding improved somewhat with fuel rods that used GLCC H-451 ground graphite shim particles rather than Speer fluid coke shim particles. Measurements of permeability (by inert gas intrusion) of the pyrocarbon on the inert particles showed that the disorder created by the neutron flux did not increase the inert gas permeability. Metallographic examination of Triso-coated particles irradiated both with and without an outer pyrocarbon coating revealed that the outer coating is necessary to suppress SiC degradation at temperatures above approximately 1375/sup 0/C. The fission product behavior (determined by the electron microprobe) was similar in both low-enriched and high-enriched uranium particles made from weak-acid resins. Furthermore, fission product palladium caused severe SiC corrosion at time-averaged temperatures above 1400/sup 0/C.
Date: June 1, 1979
Creator: Tiegs, T.N. & Robbins, J.M.
Partner: UNT Libraries Government Documents Department

Structural model testing for prestressed concrete pressure vessels: a study of grouted vs nongrouted posttensioned prestressing tendon systems. [HTGR]

Description: Nongrouted tendons are predominantly used in this country as the prestressing system for prestressed concrete pressure vessels (PCPVs) because they are more easily surveyed to detect reductions in prestressing level and distress such as results from corrosion. Grouted tendon systems, however, offer advantages which may make them cost-effective for PCPV applications. Literature was reviewed to (1) provide insight on the behavior of grouted tendon system, (2) establish performance histories for structures utilizing grouted tendons, (3) examine corrosion protection procedures for prestressing tendons, (4) identify arguments for and against using grouted tendons, and (5) aid in the development of the experimental investigation. The experimental investigation was divided into four phases: (1) grouted-nongrouted tendon behavior, (2) evaluation of selected new material systems, (3) bench-scale corrosion studies, and (4) preliminary evaluation of acoustic emission techniques for monitoring grouted tendons in PCPVs. The groutability of large tendon systems was also investigated.
Date: April 1, 1979
Creator: Naus, D.J.
Partner: UNT Libraries Government Documents Department

One-fifth-scale fuel element column shake test. Final report. [HTGR]

Description: A series of seismic tests were performed on a one-fifth-scale HTGR fuel element column. The column was subjected to static deflections, free vibrations, and uniaxial and biaxial horizontal and vertical sinusoidal and earthquake time history excitations. Main parameters measured were column displacements and element impact forces, dowel forces, and rocking angles.
Date: September 1, 1979
Creator: Fischer, J.; Rakowski, J.E. & Olsen, B.E.
Partner: UNT Libraries Government Documents Department

Numerical determination of inert gas permeability parameters of High-Temperature Gas-Cooled Reactor (HTGR) fuel-particles

Description: The high temperature diffusion of inert gases through an outer layer of dense carbon into the ThO/sub 2/ core of biso-coated High-Temperature Gas-Cooled Reactor (HTGR) fuel-particles is studied numerically. A mathematical model of diffusion through a dense spherical shell into a spherical core is used to numerically calculate the theoretical gas content of the core. This theoretical calculation, in tandem with an optimizing computer code and experimental data, is used to determine the diffusion coefficient of the shell and the porosities of the shell and inner core. The activation energy is also determined for use in an Arrhenius relationship between the diffusion coefficients and absolute temperature.
Date: November 1, 1979
Creator: Tolliver, J.S.
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, January 1, 1979-March 31, 1979

Description: This report presents the results of work performed from January 1, 1979 through March 31, 1979 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. Work covered in this report includes the activities associated with the creep-rupture testing of the test materials for the purpose of verifying the stresses selected for the screening creep test program, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment.
Date: July 19, 1979
Partner: UNT Libraries Government Documents Department

Postirradiation thermal analysis of capsule P13T. [HTGR]

Description: In determining fuel rod temperature histories for the P13T capsule, a technique which combined measured temperature and dimensional data, TAC-2D computer modeling, and a calculational procedure was employed. TAC-2D models were constructed for each of the capsule's four fuel bodies and temperature matching runs were made at five time points of the irradiation history. The agreement between TAC-calculated and measured temperatures was good; at all times the TAC-calculated temperatures were within 20/sup 0/C of the Chromel-Alumel (C/A) measurements and 40/sup 0/C of the corrected tungsten-rhenium (W/Re) temperatures. Thermocouple decalibration was treated in detail and corrected temperatures for all W/Re thermocouples were calculated over the irradiation period.
Date: December 1, 1978
Creator: Ketterer, J.W.
Partner: UNT Libraries Government Documents Department

HTGR core model response to simultaneous horizontal and vertical excitations

Description: An experimental program was undertaken to investigate the effects of simultaneous horizontal and vertical excitation on the response of the HTGR core. The tests were conducted with block array models of the core excited with both fixed frequency and sweeping frequency harmonic forcing functions. The effects on both free standing block arrays and on block arrays preloaded in the vertical direction were investigated. The results of the tests as well as their importance as regards to the full core response, are presented.
Date: January 1, 1978
Creator: Bezler, P. & Curreri, J.R.
Partner: UNT Libraries Government Documents Department

Irradiation performance of HTGR fuel rods in HFIR experiments HT-26 and -27

Description: Capsules HT-26 and -27 were irradiated in the target region of the High Flux Isotope Reactor. The purpose was to study matrix-particle interaction in High-Temperature Gas-Cooled Reactor fuel rods before and after irradiation. The first objective was to fabricate fuel with a failed fuel fraction less than 0.1 percent. This was accomplished for packed Al/sub 2/O/sub 3/ and in-block carbonization techniques, using annealed Biso-coated fertile particles and either annealed or unannealed Triso-coated fissile particles. The second objective, to irradiate fuel rods and determine the failed fuel fraction (less than 1 percent for acceptable results), was not completely accomplished. In HT-26 the capsule evidently leaked, and the fuel rods suffered attack by steam. In HT-27 the fuel rods only in the high-temperature region suffered attack by zirconium (used as an oxygen getter in the capsule). These problems resulted in failed particles and consequently made it impossible to determine failure fractions due to matrix-particle interaction. Results from the other fuel rods in HT-27 showed that particle failure fractions of less than 1 percent can be obtained from fuel rods carbonized in-tube and irradiated at temperatures up to 900/sup 0/C fuel surface temperature.
Date: August 1, 1976
Creator: Tiegs, T. N.; Caputo, A. J.; Long, E. L. Jr. & Montgomery, B. H.
Partner: UNT Libraries Government Documents Department

Fatigue tests of dowel-socket systems. [HTGR]

Description: A test program was conducted to determine the fatigue behavior of LHTGR fuel element dowel/socket systems. Two dowel/socket systems, namely, a four-dowel system and a five-dowel system, were tested to failure under shear loads applied through a fatigue test apparatus to simulate repetitive loading during a seismic event.
Date: June 15, 1976
Creator: Chiang, D. D.
Partner: UNT Libraries Government Documents Department

Thermocouple evaluation model and evaluation of chromel--alumel thermocouples for High-Temperature Gas-Cooled Reactor applications

Description: Factors affecting the performance and reliability of thermocouples for temperature measurements in High-Temperature Gas-Cooled Reactors are investigated. A model of an inhomogeneous thermocouple, associated experimental technique, and a method of predicting measurement errors are described. Error drifts for Type K materials are predicted and compared with published stability measurements. 60 references.
Date: March 1, 1977
Creator: Washburn, B. W.
Partner: UNT Libraries Government Documents Department

The effects of specimen geometry and size on the fracture toughness of nuclear graphites

Description: In a joint Oak Ridge National Laboratory (ORNL)/Japan Atomic Energy Research Institute (JAERI) study, various fracture toughness techniques were applied to Toyo Tanso grade IG-110 graphite to establish if specimen geometry influences on fracture toughness. The test geometries investigated were: compact tension (CT), disc compact tension (DCT), short rod (SR), chevron-notched short-red (CNSR), cylindrical bend specimen (BS), and centrally slotted disc (CSD). Specimen geometries which allow slow crack propagation, such as the CNSR and CT, yielded higher fracture toughness values than those where fracture is very rapid, e.g., the CSD. In a further ORNL study, the CNSR specimen geometry was selected to investigate the effect of specimen size on fracture toughness. Three specimen sizes and three grades of graphite were examined: Great Lakes Carbon grade H-451, Stackpole grade 2020, and Toyo Tanso grade IG-110. Grade H-451 was the toughest graphite, while Stackpole 2020 was the least tough. Fracture toughness increased with increasing specimen size for all graphites tested. This result was attributed to rising R-curve behavior. 13 refs., 8 figs., 3 tabs.
Date: January 1, 1991
Creator: Romanoski, G.R. & Burchell, T.D.
Partner: UNT Libraries Government Documents Department

Irradiation performance of HTGR fuel rods in HFIR experiments HRB-11 and -12

Description: Capsules HRB-11 and -12 were irradiated in support of development of weak-acid-resin-derived recycle fuel for the high-enriched uranium (HEU) fuel cycle for the HTGR. Fissil fuel particles with initial oxygen-to-metal ratios between 1.0 and 1.7 performed acceptably to full burnup for HEU fuel. Particles with ratios below 1.0 showed excessive chemical interaction between rare earth fission products and the SiC layer.
Date: June 1, 1980
Creator: Homan, F.J.; Tiegs, T.N.; Kania, M.J.; Long, E.L. Jr.; Thoms, K.R.; Robbins, J.M. et al.
Partner: UNT Libraries Government Documents Department

Effect of steam oxidation on strength, elastic modulus, and strain at fracture of graphite 2020. [HTGR]

Description: Stackpole Carbon's graphite grade 2020, a candidate material for support posts in GA HTGRs, was oxidized in steam-helium mixtures to 25 wt % burnoff. The effects of the oxidation on ultimate strength, elastic modulus, and strain at fracture are reported. A large number of nonoxidized and oxidized tensile specimens were tested to failure. Average losses of 6.7 wt % in strength and 7.4 wt % in elastic modulus at 1% burnoff were obtained. No significant change in strain at fracture was observed for burnoffs up to 5%. The elastic modulus and ultimate tensile strength of the oxidized samples correlated with bulk density.
Date: December 1, 1977
Creator: Velasquez, C.; Hightower, G. & Burnette, R.
Partner: UNT Libraries Government Documents Department

US/FRG umbrella agreement for cooperation in GCR development. Fuel, fission products, and graphite subprogram. Quarterly status report, January 1, 1982-March 31, 1982

Description: Technical highlights of the period were: Mr. David L. Hanson began a one-year assignment from GA to KFA as a technical specialist working in the area of fission product transport validation; benchmark calculations for verification of fission product release codes were exchanged between GA and HRB (PWS FP-2); fuel capsule R2-K13 was shut down during the entire period while the Studsvik reactor was undergoing licensing tests; a Specialists Meeting on SiC characterization was organized for June 1982 (PWS FD-10); the graphite capsules HTK-5 and HTK-6 were fabricated and HTK-5 started up at ORNL (PWS GD-1); and the graphite creep irradiation experiment in Petten designed for no direct readout was started up (PWS GD-4).
Date: April 1, 1982
Creator: Turner, R.F.
Partner: UNT Libraries Government Documents Department

HTGR fuel element structural design considerations

Description: The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development.
Date: September 1, 1986
Creator: Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R. & Yu, H.
Partner: UNT Libraries Government Documents Department

Postirradiation examination of capsule P13Q. [HTGR]

Description: Capsule P13Q was the sixth in a series of irradiation tests conducted under the HTGR Fuels and Core Development Program. It was the first accelerated irradiation test of large-diameter graphite-fuel bodies irradiated to peak LHTGR fast fluences. The primary purpose of the test was to evaluate the irradiation performance of the integral bodies and cured-in-place fuel rods. One TRISO UC/sub 2/ and two BISO ThO/sub 2/ coated particle batches were used in the fuel rods. The postirradiation examination revealed that the performance of the H-451 graphite bodies and fuel rods irradiated to a peak fluence of 9.5 x 10/sup 25/ n/m/sup 2/ (E greater than 29 fJ)/sub HTGR/ and to an average peak fuel rod temperature of 1175/sup 0/C was acceptable. A range of fuel rod variables was tested and none were detrimental to the integrity of the rods. The coated fuel particles behaved in a manner predicted by previous irradiation data.
Date: September 1, 1977
Creator: Young, C.A. & Scott, C.B.
Partner: UNT Libraries Government Documents Department

Corrosion-induced changes in pore-size distributions of fuel-matrix material

Description: In order to understand the mechanism of metallic fission-product adsorption and desorption as well as diffusion in graphitic materials, a detailed knowledge of the material microstructure is essential. Different types of grahitic matrix material used or to be used in fuel elements of the German HTR Program were measured at ORNL in cooperation with the Hahn-Meitner-Institut Berlin. Actual measurements of fission product diffusion and adsorption/desorption were performed at HMI Berlin.
Date: January 1, 1981
Creator: Krautwasser, P. & Eatherly, W.P.
Partner: UNT Libraries Government Documents Department

HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1981

Description: This document reports the technical accomplishments on the HTGR Fuel Technology Program at General Atomic during the first half of FY-81. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was initiated during this period was the preparation of input data for a long-range technology program plan.
Date: May 1, 1981
Partner: UNT Libraries Government Documents Department

Particle fuel bed tests

Description: Gas-cooled reactors, using packed beds of small diameter coated fuel particles have been proposed for compact, high-power systems. The particulate fuel used in the tests was 800 microns in diameter, consisting of a thoria kernel coated with 200 microns of pyrocarbon. Typically, the bed of fuel particles was contained in a ceramic cylinder with porous metallic frits at each end. A dc voltage was applied to the metallic frits and the resulting electric current heated the bed. Heat was removed by passing coolant (helium or hydrogen) through the bed. Candidate frit materials, rhenium, nickel, zirconium carbide, and zirconium oxide were unaffected, while tungsten and tungsten-rhenium lost weight and strength. Zirconium-carbide particles were tested at 2000 K in H/sub 2/ for 12 hours with no visible reaction or weight loss.
Date: January 1, 1985
Creator: Horn, F.L.; Powell, J.R. & Savino, J.M.
Partner: UNT Libraries Government Documents Department