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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

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

Strategy for the practical utilization of thorium fuel cycles

Description: There has been increasing interest in the utilization of thorium fuel cycles in nuclear power reactors for the past few years. This is due to a number of factors, the chief being the recent emphasis given to increasing the proliferation resistance of reactor fuel cycles and the thorium cycle characteristic that bred /sup 233/U can be denatured with /sup 238/U (further, a high radioactivity is associated with recycle /sup 233/U, which increases fuel diversion resistance). Another important factor influencing interest in thorium fuel cycles is the increasing cost of U/sub 3/O/sub 8/ ores leading to more emphasis being placed on obtaining higher fuel conversion ratios in thermal reactor systems, and the fact that thorium fuel cycles have higher fuel conversion ratios in thermal reactors than do uranium fuel cycles. Finally, there is increasing information which indicates that fast breeder reactors have significantly higher capital costs than do thermal reactors, such that there is an economic advantage in the long term to have combinations of fast breeder reactors and high-conversion thermal reactors operating together. Overall, it appears that the practical, early utilization of thorium fuel cycles in power reactors requires commercialization of HTGRs operating first on stowaway fuel cycles, followed by thorium fuel recycle. In the longer term, thorium utilization involves use of thorium blankets in fast breeder reactors, in combination with recycling the bred /sup 233/U to HTGRs (preferably), or to other thermal reactors.
Date: January 1, 1978
Creator: Kasten, P.R.
Partner: UNT Libraries Government Documents Department

Structure interaction due to thermal bowing of shrouds in steam generator of gas-cooled reactor

Description: The design of the gas-cooled reactor steam generators includes a tube bundle support plate system which restrains and supports the helical tubes in the steam generator. The support system consists of an array of radially oriented, perforated plates through which the helical tube coils are wound. These support plates have tabs on their edges which fit into vertical slots in the inner and outer shrouds. When the helical tube bundle and support plates are installed in the steam generator, they most likely cannot fit evenly between the inner and outer shrouds. This imperfection leads to different gaps between two extreme sides of the tube bundle and the shrouds. With different gaps through the tube bundle height, the helium flow experiences different cooling effects from the tube bundle. Hence, the temperature distribution in the shrouds will be non-uniform circumferentially since their surrounding helium flow temperatures are varied. These non-uniform temperatures in the shrouds result in the phenomenon of thermal bowing of shrouds.
Date: January 1, 1981
Creator: Woo, H.H.
Partner: UNT Libraries Government Documents Department

Nondestructive examination of 51 fuel and reflector elements from Fort St. Vrain Core Segment 1

Description: Fifty-one fuel and reflector elements irradiated in core segment 1 of the Fort St. Vrain High-Temperature Gas-Cooled Reactor (HTGR) were inspected dimensionally and visually in the Hot Service Facility at Fort St. Vrain in July 1979. Time- and volume-averaged graphite temperatures for the examined fuel elements ranged from approx. 400/sup 0/ to 750/sup 0/C. Fast neutron fluences varied from approx. 0.3 x 10/sup 25/ n/m/sup 2/ to 1.0 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/. Nearly all of the examined elements shrank in both axial and radial dimensions. The measured data were compared with strain and bow predictions obtained from SURVEY/STRESS, a computer code that employs viscoelastic beam theory to calculate stresses and deformations in HTGR fuel elements.
Date: December 1, 1980
Creator: Miller, C.M. & Saurwein, J.J.
Partner: UNT Libraries Government Documents Department

Size effect on the irradiation performance of coated fuel particles

Description: Outer coatings that were as near alike as possible were applied to two different sizes of inert TRISO particles that were larger than those commonly used to fuel HTGR reactors, and these particles were then irradiated in a test reactor to observe the influence of particle size on outer coating failures that resulted from irradiation-induced shrinkage of coatings onto the more stable SiC substrates over which they were applied. Outer coatings of plain pyrocarbon and of Si-alloyed pyrocarbon were used to make up two test pairs of particles with diameters of about 1050 ..mu..m and 1300 ..mu..m. For a fast-neutron fluence of 5.5 x 10/sup 25/ n/m/sup 2/ (E > 29fJ) at an irradiation temperature of 1125 K, failure was about twice as high in the larger 1300 ..mu..m particle of each test pair as in the smaller 1050 ..mu..m particle (16% versus 8%), with each of the coating types having roughly the same behavior.
Date: September 1, 1980
Creator: Bullock, R.E.
Partner: UNT Libraries Government Documents Department

Irradiation performance of HTGR fuel in HFIR capsule HT-31

Description: The capsule was irradiated in the High Flux Isotope Reactor at ORNL to peak particle temperatures up to 1600/sup 0/C, fast neutron fluences (0.18 MeV) up to 9 x 10/sup 25/ n/m/sup 2/, and burnups up to 8.9% FIMA for ThO/sub 2/ particles. The oxygen release from plutonium fissions was less than calculated, possibly because of the solid solution of SrO and rare earth oxides in UO/sub 2/. Tentative results show that pyrocarbon permeability decreases with increasing fast neutron fluence. Fission products in sol-gel UO/sub 2/ particles containing natural uranium mostly behaved similarly to those in particles containing highly enriched uranium (HEU). Thus, much of the data base collected on HEU fuel can be applied to low-enriched fuel. Fission product palladium penetrated into the SiC on Triso-coated particles. Also the SiC coating provided some retention of /sup 110m/Ag. Irradiation above about 1200/sup 0/C without an outer pyrocarbon coating degraded the SiC coating on Triso-coated particles.
Date: May 1, 1979
Creator: Tiegs, T.N.; Robbins, J.M.; Hamner, R.L.; Montgomery, B.H.; Kania, M.J.; Lindemer, T.B. et al.
Partner: UNT Libraries Government Documents Department

Capsule HRB-15B postirradiation examination report

Description: Capsule HRB-15B design tested 184 thin graphite trays containing unbonded fuel particles to peak exposures of 6.6 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/ fast fluence, approx. 27% fissions per initial metal atom (FIMA) fissile burnup, and 6% FIMA fertile burnup at nominal time-averaged temperatures of 815 to 915/sup 0/C. The capsule tested a variety of low-enriched uranium (approx. 19.5% U-235) fissile particle types, including UC/sub 2/, UC/sub x/O/sub y/, UO/sub 2/, zirconium-buffered UO/sub 2/ (referred to in this report as UO/sub 2//sup *), and 1:1(Th,U)O/sub 2/ with both TRISO and silicon-BISO coatings. All fertile particles were ThO/sub 2/ with BISO, silicon-BISO, or TRISO coatings. The findings indicated that all TRISO particles retained virtually all of their fission product inventories, except small quantities of silver, at these irradiation temperatures, while some of the silicon-BISO particles released significant amounts of both silver and cesium. No kernel migration, pressure vessel, or outer pyrolytic carbon (OPyC) failures were observed in the fuel particles, which had total diameters of < 900 ..mu..m; however, the incidence of failed OPyC coatings was found to increase with particle size in the TRISO inert particles, which had diameters of 1000 to 1200 ..mu..m. UO/sub 2//sup */ particles exhibited no detrimental irradiation effects, but they contained pure carbon precipitates in the kernels after irradiation which were not observed in the undoped UO/sub 2/ particles. Postirradiation examination revealed no differences in the irradiation performance of three UC/sub x/O/sub y/ kernel types with varying oxygen/uranium ratios.
Date: June 1, 1981
Creator: Ketterer, J.W. & Bullock, R.E.
Partner: UNT Libraries Government Documents Department

Behavior of graphite during rapid depressurization

Description: Three grades of graphite used in the HTGR were subjected to depressurization rates of greater than 6500 psi/sec. All of the graphites survived the rapid depressurization with no outward signs of change. Examination of the graphites by NDE also indicated no effects of the depressurization on density or flaw distribution with the possible exception of one sample of grade 2020.
Date: January 1, 1981
Creator: Eatherly, W.P. & Beavan, L.A.
Partner: UNT Libraries Government Documents Department

CERL/ORNL research and development programs in support of prestressed concrete reactor vessel development

Description: In support of the evolution of PCRV designs being developed both in the UK and USA, research and developments programmers are being conducted at the CEGB Central Electricity Research Laboratories (CERL) and the Oak Ridge National Laboratory (ORNL) respectively. In the UK, recent work has focused on elevated temperature effects on concrete properties and instrument systems for PCRVs. The concrete development program at ORNL consists of generic studies designed to provide technical support for ongoing prestressed concrete reactor vessel-related activities, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Recent activities have been related to the development of properties for high-strength concrete mix designs for the PCRV of a 2240 MW(t) HTGR-SC/C lead plant project, and the development of PCRV model testing techniques.
Date: January 1, 1984
Creator: Hornby, I.W. & Naus, D.J.
Partner: UNT Libraries Government Documents Department

Analytical methods for determining the inelastic response of prestressed concrete reactor vessels and vessel closures. [HTGR]

Description: Numerical prediction of the limiting capacity of thick prestressed concrete structures such as multicavity PCRVs and PCRV closure plugs requires a constitutive model that is capable of accurately predicting the nonlinear response of concrete when subjected to states of high triaxial compressive stress. Four nonlinear concrete constitutive models have veen investigated at ORNL: the models contained in the 1976 and 1977 versions of the ADINA finite element code and the two endochronic models proposed by Z.P. Bazant. The paper presents comparisons between two of these constitutive models and published stress-strain data for various loading conditions and also comparisons of results obtained from experimental tests and finite element analyses of some typical concrete structures.
Date: January 1, 1978
Creator: Dodge, W.G. & Fanning, D.N.
Partner: UNT Libraries Government Documents Department

Operation and postirradiation examination of ORR capsule OF-2: accelerated testing of HTGR fuel

Description: Irradiation capsule OF-2 was a test of High-Temperature Gas-Cooled Reactor fuel types under accelerated irradiation conditions in the Oak Ridge Research Reactor. The results showed good irradiation performance of Triso-coated weak-acid-resin fissile particles and Biso-coated fertile particles. These particles had been coated by a fritted gas distributor in the 0.13-m-diam furnace. Fast-neutron damage (E > 0.18 MeV) and matrix-particle interaction caused the outer pyrocarbon coating on the Triso-coated particles to fail. Such failure depended on the optical anisotropy, density, and open porosity of the outer pyrocarbon coating, as well as on the coke yield of the matrix. Irradiation of specimens with values outside prescribed limits for these properties increased the failure rate of their outer pyrocarbon coating. Good irradiation performance was observed for weak-acid-resin particles with conversions in the range from 15 to 75% UC/sub 2/.
Date: March 1, 1979
Creator: Tiegs, T.N. & Thoms, K.R.
Partner: UNT Libraries Government Documents Department

Safety aspects of forced flow cooldown transients in Modular High Temperature Gas-Cooled Reactors

Description: During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs), the main Heat Transport System (HTS) and the Shutdown Cooling System n removed by the passive Reactor (SCS) are assumed to have failed. Decay heat is the Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This report used the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits.
Date: May 1, 1993
Creator: Kroger, P.G. (Brookhaven National Lab., Upton, NY (United States))
Partner: UNT Libraries Government Documents Department

Investigations of postulated accident sequences for the Fort St. Vrain HTGR

Description: The systems analysis capability of the ORNL HTGR Safety analysis research program includes a family of computer codes: an overall plant NSSS simulation (ORTAP), and detailed component codes for investigating core neutronic accidents (CORTAP), shutdown emergency-cooling accidents via a 3-dimensional core model (ORECA), and once-through steam generator transients (BLAST). The component codes can either be run independently or in the overall NSSS code. Verification efforts have consisted primarily of using existing Fort St. Vrain reactor dynamics data to compare against code predictions. Comparisons of core thermal conditions made for reactor scrams from power levels between 30 and 50% showed good agreement. An optimization program was used to rationalize the difference between the predicted and measured refueling region outlet temperatures, and, in general, excellent agreement was attained by adjustment of models and parameters within their uncertainty ranges. However, more work is required to establish a unique and valid set of models.
Date: January 1, 1978
Creator: Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Hatta, M. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

1170-MW(t) HTGR-PS/C plant application study report: tar sands oil recovery application

Description: This report summarizes a study to apply an 1170-MW(t) high-temperature gas-cooled reactor - process steam/cogeneration (HTGR-PS/C) to tar sands oil recovery and upgrading. The raw product recovered from the sands is a heavy, sour bitumen; upgrading, which involves coking and hydrodesulfurization, produces a synthetic crude (refinable by current technology) and petroleum coke. Steam and electric power are required for the recovery and upgrading process. Proposed and commercial plants would purchase electric power from local utilities and obtain from boilers fired with coal and with by-product fuels produced by the upgrading. This study shows that an HTGR-PS/C represents a more economical source of steam and electric power.
Date: May 1, 1981
Creator: Rao, R. & McMain, Jr., A. T.
Partner: UNT Libraries Government Documents Department

1170-MW(t) HTGR-PS/C plant application study report: SRC-II process application

Description: The solvent refined coal (SRC-II) process is an advanced process being developed by Gulf Mineral Resources Ltd. (a Gulf Oil Corporation subsidiary) to produce a clean, non-polluting liquid fuel from high-sulfur bituminous coals. The SRC-II commercial plant will process about 24,300 tonnes (26,800 tons) of feed coal per stream day, producing primarily fuel oil plus secondary fuel gases. This summary report describes the integration of a high-temperature gas-cooled reactor operating in a process steam/cogeneration mode (HTGR-PS/C) to provide the energy requirements for the SRC-II process. The HTGR-PS/C plant was developed by General Atomic Company (GA) specifically for industries which require energy in the form of both steam and electricity. General Atomic has developed an 1170-MW(t) HTGR-PS/C design which is particularly well suited to industrial applications and is expected to have excellent cost benefits over other sources of energy.
Date: May 1, 1981
Creator: Rao, R. & McMain, A. T., Jr.
Partner: UNT Libraries Government Documents Department

Krypton absorption in liquid CO/sub 2/ (KALC): effects of the minor components N/sub 2/, CO, and Xe

Description: Results are presented for the fourth major campaign for quantifying krypton removal from simulated High-Temperature Gas-Cooled Reactor reprocessing off-gas by the Krypton Absorption in Liquid CO/sub 2/ (KALC) process. This process utilizes the high solubility of krypton in liquid CO/sub 2/. Mass transfer experiments for the absorption, fractionation, and stripping operations of the KALC process indicate that the addition of N/sub 2/ and CO do not alter the mass transfer characteristics exhibited by O/sub 2/ and krypton in the basic CO/sub 2/--O/sub 2/--Kr system. Decontamination factors for xenon in the absorber and stripper were several orders of magnitude less than those for krypton under similar conditions. Indications are that the fate of xenon is controlled by the heat input to the stripper reboiler. Experiments on the solubility of O/sub 2/ and CO indicate that CO is more soluble than O/sub 2/ at temperatures below -21/sup 0/C.
Date: February 1, 1979
Creator: Gilliam, T. M.; Fowler, V. L. & Inman, D. J.
Partner: UNT Libraries Government Documents Department

MEU/Th fuel cycle optimization for the Lead Plant

Description: The reference equilibrium cycle fuel composition for the Lead Plant was specified previously by a C/Th ratio of 850 and a fuel rod diameter of 1.17 cm, which is optimal for non-recycle operation and close to optimal for recycle of bred U-233. Subsequent work has emphasized the importance of full recycle of all discharged uranium to maintain the competitive advantage of the MEU/Th cycle. Cycles with full recycle optimize at higher thorium loadings and larger rod diameters. This is an additional benefit for core design and reduces fabrication problems. New optimization studies based on full recycle lead to an equilibrium cycle composition characterized by a C/Th ratio of 600 and a rod diameter of 1.35 cm. The average packing fraction of fuel particles in the rod is 0.43. The C/Th ratio for the initial core is 350, which can also be accommodated with the 1.35 cm rod diameter. Mass flow data for 30 year operation and fuel cycle cost data have been obtained for this cycle and for several other thorium loadings.
Date: December 1, 1978
Creator: Merrill, M.H. & Lane, R.K.
Partner: UNT Libraries Government Documents Department

Methods for very high temperature design

Description: Design rules and procedures for high-temperature, gas-cooled reactor components are being formulated as an ASME Boiler and Pressure Vessel Code Case. A draft of the Case, patterned after Code Case N-47, and limited to Inconel 617 and temperatures of 982/degree/C (1800/degree/F) or less, will be completed in 1989 for consideration by relevant Code committees. The purpose of this paper is to provide a synopsis of the significant differences between the draft Case and N-47, and to provide more complete accounts of the development of allowable stress and stress rupture values and the development of isochronous stress vs strain curves, in both of which Oak Ridge National Laboratory (ORNL) played a principal role. The isochronous curves, which represent average behavior for many heats of Inconel 617, were based in part on a unified constitutive model developed at ORNL. Details are also provided of this model of inelastic deformation behavior, which does not distinguish between rate-dependent plasticity and time-dependent creep, along with comparisons between calculated and observed results of tests conducted on a typical heat of Inconel 617 by the General Electric Company for the Department of Energy. 4 refs., 15 figs., 1 tab.
Date: January 1, 1989
Creator: Blass, J.J.; Corum, J.M. & Chang, S.J.
Partner: UNT Libraries Government Documents Department

Reactor-safety research programs. Quarterly report, October-December 1982. Volume 4

Description: Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized-water-reactor steam-generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models being developed to provide better digital codes to compute the bahavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.
Date: April 1, 1983
Creator: Edler, S.K. (ed.)
Partner: UNT Libraries Government Documents Department

In-pile steam oxidation of model HTGR fuel elements

Description: Model HTGR fuel elements were exposed to various concentrations of steam while being irradiated under several sets of temperature conditions in the Oak Ridge Research Reactor. In one test, catalysis by iron impurities in the graphite casing of the fuel element caused a highly localized attack on the graphite by the steam; this resulted in the formation of deep pits in the casing. Furthermore, the iron impurities were sufficiently mobile to cause pitting attack on the pyrolytic carbon coatings of the fuel particles as well. The presence of steam induced a rapid increase in the release of gaseous fission products. However, the cessation of steam ingress in the primary system resulted in a pronounced, but correspondingly smaller, reduction in the level of gaseous release. The incidence of fuel failure was greater than anticipated; however, even though the coatings of greater than 30% of the fuel had failed, the release of fission products beyond the fuel element itself was largely confined to iodine and the noble gases. A novel mode of fuel failure was observed under the rather severe conditions of the tests; this involved the attack of the pyrolytic carbon coatings on intact particles by uncoated fragments of uranium fuel kernel material from failed particles.
Date: March 1, 1979
Creator: Freid, S.H.; de Nordwall, H.J. & Malinauskas, A.P.
Partner: UNT Libraries Government Documents Department

Interpretation of bend strength increase of graphite by the couple-stress theory. [HTGR]

Description: This paper presents a continued evaluation of the applicability of the couple-stress constitutive theory to graphite. The evaluation is performed by examining four-point bend and uniaxial tensile data of various sized cylindrical and square specimens for three grades of graphites. These data are superficially inconsistent and, usually, at variance with the predictions of classical theories. Nevertheless, this evaluation finds that they can be consistently interpreted by the couple-stress theory. This is compatible with results of an initial evaluation that considered one size of cylindrical specimen for H-451 graphite.
Date: May 1, 1981
Creator: Tang, P.Y.
Partner: UNT Libraries Government Documents Department

Relationship between carburization and zero-applied-stress creep dilation in Alloy 800H and Hastelloy X. [HTGR]

Description: Typical HTGR candidate alloys can carburize when exposed to simulated service environments. The carbon concentration gradients so formed give rise to internal stresses which could cause dilation. Studies performed with Hastelloy X and Alloy 800H showed that dilations of up to almost 1% can occur at 1000/sup 0/C when carbon pickup is high. Dilation was normally observed only when the carbon increase was >1000 ..mu..g/cm/sup 2/ and ceased when the diffusing carbon reached the center of the specimen.
Date: January 1, 1981
Creator: Inouye, H. & Rittenhouse, P.L.
Partner: UNT Libraries Government Documents Department