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Three-dimensional thermoelastic analysis of a Fort St. Vrain core support block

Description: A thermoelastic stress analysis of a graphite core support block in the Fort St. Vrain High-Temperature Gas-Cooled Reactor is described. The support block is subjected to thermal stresses caused by a loss of forced circulation accident of the reactor system. Two- and three-dimensional finite element models of the core support block are analyzed using the ADINAT and ADINA codes, and results are given that verify the integrity of this structural component under the given accident condition. 10 refs., 39 figs.
Date: September 1, 1981
Creator: Butler, T.A. & Anderson, C.A.
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

High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research. Quarterly progress report, January 1-March 31, 1980

Description: Work continued on development of the ORTAP, ORECA, and BLAST codes; and verification studies were continued on the ORECA, CORTAP, and BLAST codes. An improved steam turbine plant model (ORTURB) for use in ORTAP was developed and checked. Predictions from BLAST, CORTAP, and ORECA were compared with various transient test data from the Fort St. Vrain reactor.
Date: August 1, 1980
Creator: Ball, S.J.; Cleveland, J.C.; Conklin, J.C. & Harrington, R.M.
Partner: UNT Libraries Government Documents Department

Postirradiation examination of recycle test elements from the Peach Bottom Reactor

Description: The Recycle Test Elements were a series of tests of High-Temperature Gas-Cooled Reactor fuels irradiated in Core 2 of the Peach Bottom Unit 1 Reactor. They tested a wide variety of fissile and fertile fuel types of prime interest when the tests were designed. The fuel types included UO/sub 2/, UC/sub 2/, (2Th,U)O/sub 2/, (4Th,U)O/sub 2/, ThC/sub 2/, and ThO/sub 2/. The mixed thorium--uranium oxides and the pure thorium oxide were tested as Biso-coated particles only, while the others were tested as both Biso- and Triso-coated particles. The Biso coatings on the fissile kernels contained the fission products inadequately but on the fertile kernels they did so acceptably. The results from accelerated and real-time tests on the particle types agreed well.
Date: December 1, 1978
Creator: Tiegs, T.N. & Long, E.L. Jr.
Partner: UNT Libraries Government Documents Department

A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators

Description: This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.
Date: July 1, 1990
Creator: White, L.S.
Partner: UNT Libraries Government Documents Department

Thorium fuel cycle: a nuclear strategy and fuel recycle technology

Description: Use of thorium fuel cycles in thermal reactors appears to permit a moderate rate of introduction of fast breeder reactors into the nuclear economy and helps maintain a relatively low ratio of FBRs to thermal reactors in the future. To implement the benefits of thorium fuel cycles, however, will require fuel recycle research and development. Fuel recycle technology developed for uranium and plutonium cycles will be beneficial to thorium fuel cycle development; however, significant additional R and D is required to implement either the HEUTH or the DUTH cycles. The metal-clad reactors in general have relatively common generic technology development requirements, although there are significant differences between fast and thermal reactor fuel recycle needs. The thorium fuel recycle R and D requirements of HTGRs are more reactor-specific than of the other reactor types; however, some specific efforts will be required for all the different reactor types.
Date: January 1, 1978
Creator: Kasten, P.R.; Dahlberg, R.C. & Wymer, R.G.
Partner: UNT Libraries Government Documents Department

MELCOR simulation of long-term station blackout at Peach Bottom

Description: This paper presents the results from MELCOR (Version 1.8BC) calculations of the Long-Term Station Blackout Accident Sequence, with failure to depressurize the reactor vessel, at the Peach Bottom (BWR Mark I) plant, and presents comparisons with Source Term Code Package (STCP) calculations of the same sequence. This sequence assumes that batteries are available for six hours following loss of all power to the plant. Following battery failure, the reactor coolant system (RCS) inventory is boiled off through the relief valves by continued decay heat generation. This leads to core uncovery, heatup, clad oxidation, core degradation, relocation, and, eventually, vessel failure at high pressure. STCP has calculated the transient out to 13.5 hours after core uncovery. The results include the timing of key events, pressure and temperature response in the reactor vessel and containment, hydrogen production, and the release of source terms to the environment. 12 refs., 23 figs., 3 tabs.
Date: January 1, 1990
Creator: Madni, I.K.
Partner: UNT Libraries Government Documents Department

Fission product holdup in graphite. [HTGR]

Description: Multicomponent time-dependent concentration diffusion and radioactive decay of isotopic species is an important aspect of fission product migration and release from fuel particles and fuel elements in a High Temperature Gas-Cooled Reactor (HTGR). After fission products escape from a fuel particle in an HTGR, it is still necessary for them to diffuse across the graphite web of a fuel block to a coolant hole before they can be entrained in the primary coolant. The time required for a given fission product species to diffuse across the graphite web has a direct influence on the time-dependent release associated with a significant increase in the power/flow ratio. The main purpose of the paper is to present the results of a study of the holdup time in graphite of Sr as a function of the diffusion constants. The study employs a newly-developed multicomponent time-dependent diffusion and decay code called DASH. Analysis methods for solving the type of problem discussed are well known, and some applications to fission product decay and diffusion in HTGRs have appeared in the literature. However, the methods employed are often subject to time step limitations, and the effects of decay are not adequately handled. The DASH code uses a one dimensional spatial discretization for the diffusion operator and an analytic matrix operator method to remove the time dependence. Comparisons of the solutions given by DASH with a number of analytic solutions have been made, and in all instances considered the agreement with analytical solutions is excellent and limited only by the inaccuracy inherent in the spatial discretization.
Date: January 1, 1978
Creator: Apperson, C. Jr.; Carruthers, L.M. & Anderson, C.A.
Partner: UNT Libraries Government Documents Department

Process options and projected mass flows for the HTGR refabrication scrap recovery system

Description: The two major uranium recovery processing options reviewed are (1) internal recovery of the scrap by the refabrication system and (2) transfer to and external recovery of the scrap by the head end of the reprocessing system. Each option was reviewed with respect to equipment requirements, preparatory processing, and material accountability. Because there may be a high cost factor on transfer of scrap fuel material to the reprocessing system for recovery, all of the scrap streams will be recycled internally within the refabrication system, with the exception of reject fuel elements, which will be transferred to the head end of the reprocessing system for uranium recovery. The refabrication facility will be fully remote; thus, simple recovery techniques were selected as the reference processes for scrap recovery. Crushing, burning, and leaching methods will be used to recover uranium from the HTGR refabrication scrap fuel forms, which include particles without silicon carbide coatings, particles with silicon carbide coatings, uncarbonized fuel rods, carbon furnace parts, perchloroethylene distillation bottoms, and analytical sample remnants. Mass flows through the reference scrap recovery system were calculated for the HTGR reference recycle facility operating with the highly enriched uranium fuel cycle. Output per day from the refabrication scrap recovery system is estimated to be 4.02 kg of /sup 2355/U and 10.85 kg of /sup 233/U. Maximum equipment capacities were determined, and future work will be directed toward the development and costing of the scrap recovery system chosen as reference.
Date: March 1, 1979
Creator: Tiegs, S. M.
Partner: UNT Libraries Government Documents Department

Radionuclide distributions and sorption behavior in the Susquehanna--Chesapeake Bay System

Description: Radionuclides released into the Susquehanna--Chesapeake System from the Three Mile Island, Peach Bottom, and Calvert Cliffs nuclear power plants are partitioned among dissolved, particulate, and biological phases and may thus exist in a number of physical and chemical forms. In this project, we have measured the dissolved and particulate distributions of fallout /sup 137/Cs; reactor-released /sup 137/Cs, /sup 134/Cs, /sup 65/Zn, /sup 60/Co, and /sup 58/Co; and naturally occurring /sup 7/Be and /sup 210/Pb in the lower Susquehanna River and Upper Chesapeake Bay. In addition, we chemically leached suspended particles and bottom sediments in the laboratory to determine radionuclide partitioning among different particulate-sorbing phases to complement the site-specific field data. This information has been used to document the important geochemical processes that affect the transport, sorption, distribution, and fate of reactor-released radionuclides (and by analogy, other trace contaminants) in this river-estuarine system. Knowledge of the mechanisms, kinetic factors, and processes that affect radionuclide distributions is crucial for predicting their biological availability, toxicity, chemical behavior, physical transport, and accumulation in aquatic systems. The results from this project provide the information necessary for developing accurate radionuclide-transport and biological-uptake models. 76 refs., 12 figs.
Date: January 1, 1989
Creator: Olsen, C.R.; Larsen, I.L.; Lowry, P.D.; McLean, R.I. & Domotor, S.L.
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

Evaluation of severe accident risks: Quantification of major input parameters: MAACS (MELCOR Accident Consequence Code System) input

Description: Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs.
Date: December 1, 1990
Creator: Sprung, J.L.; Jow, H-N (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)) & Helton, J.C. (Arizona State Univ., Tempe, AZ (USA))
Partner: UNT Libraries Government Documents Department

New approches for high temperature gas cooled reactors (HTGRs)

Description: Several approaches are being evaluated in the US HTR Program to explore designs which might improve the commercial viability of nuclear power. The general approach is to reduce the power level of the reactor and increase ability to use passive methods for removing afterheat energy following extreme accidents. One approach most fully discussed in this paper is represented by modular HTRs for which the unit size and design are constrained such that extreme accidents do not result in significant release of radioactivity from the reactor circuit. Through such an approach, it should be possible to minimize the amount of nuclear grade components required in the balance-of-plant and achieve an economic system. Attaining such performance should provide low investment risk to the owner.
Date: January 1, 1984
Creator: Kasten, P.R.; Cleveland, J.C. & Bowers, H.I.
Partner: UNT Libraries Government Documents Department

Technical Division quarterly progress report, April 1--June 30, 1978

Description: Fuel cycle research and development: results are presented on fluidized-bed calcination and on post-treatment of commercial wastes; study was done on the use of microwave energy in processing wastes and on the use of bidentate compounds for separation of actinides from commercial power reactor reprocessing waste. Work on the krypton-85 storage development program, including the results of rubidium corrosion tests, is reported. In the HTGR fuel reprocessing section, the results of x-ray and Auger spectroscopy analysis of CO oxidation catalyst are reported. Special materials production: the long-term management of high-level ICPP wastes is reported: development of a calcine pelletizing pilot plant, actinide removal, actinide extraction by DHDECMP, and calcined solids retrieval and handling. Design work was completed for the fluorinel pilot-plant upgrade. Other development results reported are on the progress of the Rover plant, and on flowsheet development for electrolytic and second-cycle waste, for Fluorinel waste, and for Tank WM-183. Other results reported include: assistance to the Waste Calcining Facility and to the New Waste Calcining Facility, methods for the monitoring of gaseous effluents, and a mathematical model to describe chloride buildup in the waste calcining scrubbing solution. Other projects supporting energy developments: results are reported on nuclear materials safety, the installation and operation of a geothermal fluidized-bed dryer, the in-plant source-term measurement at the Turkey Point station, burnup methods for fast breeder reactor fuels, absolute thermal fission yields, analytical support to light-water breeder reactor developments, cerium analysis of actinide removal project solutions, a spark source mass spectrometric computer program, and on environmental iodine species behavior.
Date: December 1, 1978
Creator: Plung, D.L. (ed.)
Partner: UNT Libraries Government Documents Department

Applications of high-strength concrete to the development of the prestressed concrete reactor vessel (PCRV) design for an HTGR-SC/C plant

Description: The PCRV research and development program at ORNL consists of generic studies to provide technical support for ongoing PCRV-related studies, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Recent activities under this program have concentrated on the development of high-strength concrete mix designs for the PCRV of a 2240 MW(t) HTGR-SC/C plant, and the testing of models to both evaluate the behavior of high-strength concretes (plain and fibrous) and to develop model testing techniques. A test program to develop and evaluate high-strength (greater than or equal to 63.4 MPa) concretes utilizing materials from four sources which are in close proximity to potential sites for an HTGR plant is currently under way. The program consists of three phases. Phase I involves an evaluation of the cement, fly ash, admixtures and aggregate materials relative to their capability to produce concretes having the desired strength properties. Phase II is concerned with the evaluation of the effects of elevated temperatures (less than or equal to 316/sup 0/C) on the strength properties of mixes selected for detailed evaluation. Phase III involves a determination of the creep characteristics and thermal properties of the selected mixes. An overview of each of these phases is presented as well as results obtained to date under Phase I which is approximately 75% completed.
Date: January 1, 1984
Creator: Naus, D.J.
Partner: UNT Libraries Government Documents Department

Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P. O. 1236909 Final report

Description: The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel.
Date: December 1, 1978
Creator: Borgonovi, G.M.; McCartin, T.J. & Miller, C.L.
Partner: UNT Libraries Government Documents Department

Recent developments in graphite. [Use in HTGR and aerospace]

Description: Overall, the HTGR graphite situation is in excellent shape. In both of the critical requirements, fuel blocks and support structures, adequate graphites are at hand and improved grades are sufficiently far along in truncation. In the aerospace field, GraphNOL N3M permits vehicle performance with confidence in trajectories unobtainable with any other existing material. For fusion energy applications, no other graphite can simultaneously withstand both extreme thermal shock and neutron damage. Hence, the material promises to create new markets as well as to offer a better candidate material for existing applications.
Date: January 1, 1983
Creator: Cunningham, J.E.
Partner: UNT Libraries Government Documents Department

Postirradiation examination and evaluation of Peach Bottom fuel test elements FTE-14 and FTE-15

Description: Peach Bottom fuel test elements FTE-14 and FTE-15 were companion nonaccelerated tests of fuel rods and fuel particles representative of the Large High-Temperature Gas-Cooled Reactor (LHTGR). The purpose of the tests was to broaden the data base of H-327 graphite and various fuel types; specifically, UO/sub 2/, UC/sub 2/, weak acid resin UC/sub x//O/sub y/, and several fertile fuel types were tested. The irradiation reached peak fuel temperatures of 1600/sup 0/C volume- and time-averaged temperatures of 1300/sup 0/C, and fast fluence exposures up to 2 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/. Experimental results were compared with predictions based on accelerated irradiation tests, postirradiation heating, and other Peach Bottom test elements to validate HTGR design codes. The nuclear design predictions were modified by measurements which allowed the verification of thermal design calculations and thermocouple readings.
Date: February 1, 1979
Creator: Holzgraf, J.F.; McCord, F.; Miller, C.M.; Norman, B.L.; Saurwein, J.J. & Wallroth, C.F.
Partner: UNT Libraries Government Documents Department

Uranium and thorium loadings determined by chemical and nondestructive methods in HTGR fuel rods for the Fort St. Vrain Early Validation Irradiation Experiment

Description: The Fort St. Vrain Early Validation Irradiation Experiment is an irradiation test of reference and of improved High-Temperature Gas-Cooled Reactor fuels in the Fort St. Vrain Reactor. The irradiation test includes fuel rods fabricated at ORNL on an engineering scale fuel rod molding machine. Fuel rods were nondestructively assayed for /sup 235/U content by a technique based on the detection of prompt-fission neutrons induced by thermal-neutron interrogation and were later chemically assayed by using the modified Davies Gray potentiometric titration method. The chemical analysis of the thorium content was determined by a volumetric titration method. The chemical assay method for uranium was evaluated and the results from the as-molded fuel rods agree with those from: (1) large samples of Triso-coated fissile particles, (2) physical mixtures of the three particle types, and (3) standard solutions to within 0.05%. Standard fuel rods were fabricated in order to evaluate and calibrate the nondestructive assay device. The agreement of the results from calibration methods was within 0.6%. The precision of the nondestructive assay device was established as approximately 0.6% by repeated measurements of standard rods. The precision was comparable to that estimated by Poisson statistics. A relative difference of 0.77 to 1.5% was found between the nondestructive and chemical determinations on the reactor grade fuel rods.
Date: January 1, 1979
Creator: Angelini, P. & Rushton, J.E.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor safety studies for the Division of Accident Evaluation quarterly progress report, January 1-March 31, 1985

Description: Modeling, code development, and analyses of the modular High-Temperature Gas-Cooled Reactor (HTGR) continued with work on the side-by-side design. Fission-product release and transport experiments were completed. A description and assessment report on Oak Ridge National Laboratory HTGR safety codes was issued.
Date: October 1, 1985
Creator: Ball, S.J.; Cleveland, J.C.; Harrington, R.M.; Weber, C.F. & Wilson, J.H.
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

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

Instrumentation and Controls Division biennial progress report, September 1, 1976--September 1, 1978

Description: Progress is summarized in the following research and development areas: electronic circuits;instruments; radiation monitoring; process systems and instrumentation; thermometry; instrumentation for engineering experiments and test loops; HTGR fuel recycle development; reactor measurements and analysis; automatic control and data acquisition; electronic engineering support for research facilities; miscellaneous engineering services, studies, and developments; maintenance; and environmental science studies.
Date: November 1, 1978
Creator: Sadowski, G.S. (ed.)
Partner: UNT Libraries Government Documents Department

Modular high-temperature gas-cooled reactor core heatup accident simulations

Description: The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs.
Date: January 1, 1989
Creator: Ball, S.J. & Conklin, J.C.
Partner: UNT Libraries Government Documents Department