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Comparison of predicted and measured fission product behavior in the Fort St. Vrain HTGR during the first three cycles of operation

Description: Fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors which is consistent with plateout probe measurements.
Date: October 1, 1985
Creator: Hanson, D.L.; Jovanovic, V. & Burnette, R.D.
Partner: UNT Libraries Government Documents Department

Continuous improvement of the MHTGR safety and competitive performance

Description: An increase in reactor module power from 350 to 450 MW(t) would markedly improve the economics of the Modular High Temperature Gas-Cooled Reactor (MHTGR). The higher power level was recommended as the result of an in-depth cost reduction study undertaken to compete with the declining price of fossil fuel. The safety assessment confirms that the high level of safety, which relies on inherent characteristics and passive features, is maintained at the elevated power level. Preliminary systems, nuclear, and safety performance results are discussed for the recommended 450 MW(t) design. Optimization of plant parameters and design modifications accommodated the operation of the steam generator and circulator at the higher power level. Events in which forced cooling is lost, designated as conduction cooldowns are described in detail. For the depressurized conduction cooldown, without full helium inventory, peak fuel temperatures are significantly lowered. A more negative temperature coefficient of reactivity was achieved while maintaining an adequate fuel cycle and reactivity control. Continual improvement of the MHTGR delivers competitive performance without relinquishing the high safety margins demanded of the next generation of power plants.
Date: May 1, 1992
Creator: Eichenberg, T.W.; Etzel, K.T.; Mascaro, L.L. & Rucker, R.A.
Partner: UNT Libraries Government Documents Department

Thermal-stress analysis of a Fort St. Vrain core-support block under accident conditions

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.
Date: January 1, 1982
Creator: Carruthers, L.M.; Butler, T.A. & Anderson, C.A.
Partner: UNT Libraries Government Documents Department

Development of a pneumatic transfer system for HTGR recycle fuel particles

Description: In support of the High-Temperature Gas-Cooled Reactor (HTGR) Fuel Refabrication Development Program, an experimental pneumatic transfer system was constructed to determine the feasibility of pneumatically conveying pyrocarbon-coated fuel particles of Triso and Biso designs. Tests were conducted with these particles in each of their nonpyrophoric forms to determine pressure drops, particle velocities, and gas flow requirements during pneumatic transfer as well as to evaluate particle wear and breakage. Results indicated that the material can be pneumatically conveyed at low pressures without excessive damage to the particles or their coatings.
Date: February 1, 1978
Creator: Mack, J.E. & Johnson, D.R.
Partner: UNT Libraries Government Documents Department

Thorium assessment study quarterly progress report, third quarter fiscal 1977

Description: The objective of the program described is to contribute to the ongoing assessment of the potential role of thorium fuel cycles for alleviating safeguards concerns. Scenarios include (1) no fuel recycle permitted, (2) fuel recycle permitted only in secure regions (''energy parks'') with denatured (chemically non-separable) fuels only outside these regions, and (3) no limits on fuel recycle. A further objective is to provide nuclear mass balance data on HTGRs required by ERDA contractors for comparative cost-benefit studies.
Date: September 30, 1977
Creator: Spiewak, I.; Bartine, D. E.; Burns, T. J.; Cleveland, J. C.; Thomas, W. E. & White, J. R.
Partner: UNT Libraries Government Documents Department

Needs for development in nondestructive testing for advanced reactor systems

Description: The needs for development of nondestructive testing (NDT) techniques and equipment were surveyed and analyzed relative to problem areas for the Liquid-Metal Fast Breeder Reactor, the Molten-Salt Breeder Reactor, and the Advanced Gas-Cooled Reactor. The paper first discusses the developmental needs that are broad-based requirements in nondestrutive testing, and the respective methods applicable, in general, to all components and reactor systems. Next, the requirements of generic materials and components that are common to all advanced reactor systems are examined. Generally, nondestructive techniques should be improved to provide better reliability and quantitativeness, improved flaw characterization, and more efficient data processing. Specific recommendations relative to such methods as ultrasonics, eddy currents, acoustic emission, radiography, etc., are made. NDT needs common to all reactors include those related to materials properties and degradation, welds, fuels, piping, steam generators, etc. The scope of applicability ranges from initial design and material development stages through process control and manufacturing inspection to in-service examination.
Date: January 1, 1978
Creator: McClung, R.W.
Partner: UNT Libraries Government Documents Department

Experiments with a lime slurry in a stirred tank for the fixation of carbon-14-contaminated CO/sub 2/ from simulated HTGR fuel reprocessing off-gas

Description: The fixation of CO/sub 2/ with a lime slurry in a stirred tank reactor appears to be feasible. The rate of reaction is fast, and virtually complete removal of CO/sub 2/ can be attained. At a gas residence time of <1 min, the decontamination factor (DF) is >100 in a single stage reactor for CO/sub 2/ concentrations ranging from 5 to 100%. It has been determined that two-stage contacting sequences which result in a cumulative DF > 10/sup 4/ are feasible. The reaction rate is constant up to 90% utilization of the lime and then rapidly decreases, as does the pH for the remainder of the reaction. The reaction appears to be liquid-phase-controlled, and the overall gas-side mass transfer coefficient (K/sub G/..cap alpha..) increases with impeller speed and gas flow rate, ranging from 0.4 x 10/sup -6/ to 6 x 10/sup -6/ g-moles of CO/sub 2/ per (cm/sup 3/-sec-atm). The reaction rate data are also correlated by a model of mass transfer accompanied by a fast pseudo first-order chemical reaction from which good agreement of calculated and predicted interfacial area is obtained. It was noted that temperature (21 to 46/sup 0/C) and lime concentration (0.5 to 1.5 M) had very little effect on mass transfer rate and DF. The settling rate of the CaCO/sub 3/ product increased significantly with impeller speed and temperature and decreased with gas flow rate. Scale-up calculations indicate that reasonably sized equipment would provide adequate removal of CO/sub 2/ for a full-scale reprocessing plant.
Date: March 1, 1978
Creator: Holladay, D. W.
Partner: UNT Libraries Government Documents Department

Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

Description: The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low (121 to 210{degrees}C (250--410{degrees}F)) compared to those for commercial light-water reactors (LWRs) ({approximately}288{degrees}C (550{degrees}F)). The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 {times}10{sup 18} neutron/cm{sup 2} (0.68 {times} 10{sup 18} neutron/cm{sup 2} (>1 MeV)) at temperatures of 288, 204, 163, and 121{degrees}C (550, 400, 325, and 250{degrees}F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs.
Date: January 1, 1991
Creator: Haggag, F.M.; Nanstad, R.K. (Oak Ridge National Lab., TN (United States)) & Byrne, S.T. (ABB/Combustion Engineering, Inc., Windsor, CT (United States))
Partner: UNT Libraries Government Documents Department

General Atomic reprocessing pilot plant: description and results of initial testing

Description: In June 1976 General Atomic completed the construction of a reprocessing head-end cold pilot plant. In the year since then, each system within the head end has been used for experiments which have qualified the designs. This report describes the equipment in the plant and summarizes the results of the initial phase of reprocessing testing.
Date: December 1, 1977
Partner: UNT Libraries Government Documents Department

Effect of steam oxidation on the strength and elastic modulus of graphite H-451

Description: Graphite grade H-451, the reference material used as fuel and reflector elements in General Atomic large high-temperature gas-cooled reactors, was oxidized to 20% burnoff by weight in steam-helium mixtures, and the effects of oxidation on ultimate tensile strength (UTS) and elastic modulus (E) were studied. Large numbers of small cylindrical samples cored from selected locations in a log of preproduction H-451 graphite were oxidized at 1073 and 1273 K (1472/sup 0/ and 1832/sup 0/F) in helium containing 3% water vapor and 5% H/sub 2/. The UTS and E of oxidized specimens were measured and the results compared with those for nonoxidized control specimens. The average rate of change of UTS and E with burnoff at 1273 K (1832/sup 0/F) ( up to 20% burnoff) averaged 3.6% and 5.2%, respectively. Statistical analysis (comparison of population means) of the low burnoff strength data showed no significant differences between the nonoxidized control and oxidized specimens at burnoffs of up to approximately 2%.
Date: December 1, 1977
Creator: Velasquez, C.; Johnson, W.; Hightower, G. & Burnette, R.
Partner: UNT Libraries Government Documents Department

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

Description: 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 graphite and release to the coolant stream. The comparisons showed that the cesium profile shape in the graphite web and the partition coefficient between fuel rod matrix material and fuel element graphite were correctly modeled, although the overall release was significantly underpredicted. Uncertainties in the source term (fissile particle failure fraction) and total release to the coolant precluded an accurate appraisal of the validity of FIPERQ. However, several recommendations are presented to improve the applicability of future in-pile test data for the validation of fission metal release codes. The half-life dependence of fission gas release during irradiation was found to be in good agreement with the model used in the reference design materials, providing assurance that this aspect of the fission gas release predictions is properly modeled.
Date: February 1, 1978
Creator: Jensen, D.D.; Haire, M.J. & Ballagny, A.
Partner: UNT Libraries Government Documents Department

Assessment of grade H-451 graphite for replaceable fuel and reflector elements in HTGR

Description: Experimental data for grade H-451 graphite are presented and assessed to support licensing of H-451 graphite for use as fuel element blocks in a high-temperature gas-cooled reactor (HTGR). Additional data from the literature on graphite grades similar to grade H-451 are presented to supplement the H-451 data. Evaluation programs at General Atomic Company (GA) covering characterization, irradiation, and oxidation studies, along with studies carried out in the Great Lakes Carbon Corporation H-451 development program, are reported.
Date: December 1, 1977
Creator: Engle, G.B.
Partner: UNT Libraries Government Documents Department

HTGR fuel recycle program. Quarterly progress report for the period ending November 30, 1977

Description: The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel, the design and development of an integrated pilot line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials, and design work in support of Hot Engineering Tests (HET). Work is also described on trade-off studies concerning the required design of facilities and equipment for the large-scale recycle of HTGR fuels in order to guide the development activities for HTGR fuel recycle.
Date: December 1, 1977
Partner: UNT Libraries Government Documents Department

TRAFIC, a computer program for calculating the release of metallic fission products from an HTGR core

Description: A special purpose computer program, TRAFIC, is presented for calculating the release of metallic fission products from an HTGR core. The program is based upon Fick's law of diffusion for radioactive species. One-dimensional transient diffusion calculations are performed for the coated fuel particles and for the structural graphite web. A quasi steady-state calculation is performed for the fuel rod matrix material. The model accounts for nonlinear adsorption behavior in the fuel rod gap and on the coolant hole boundary. The TRAFIC program is designed to operate in a core survey mode; that is, it performs many repetitive calculations for a large number of spatial locations in the core. This is necessary in order to obtain an accurate volume integrated release. For this reason the program has been designed with calculational efficiency as one of its main objectives. A highly efficient numerical method is used in the solution. The method makes use of the Duhamel superposition principle to eliminate interior spatial solutions from consideration. Linear response functions relating the concentrations and mass fluxes on the boundaries of a homogeneous region are derived. Multiple regions are numerically coupled through interface conditions. Algebraic elimination is used to reduce the equations as far as possible. The problem reduces to two nonlinear equations in two unknowns, which are solved using a Newton Raphson technique.
Date: February 1, 1978
Creator: Smith, P.D.
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

Interim design report: fuel particle crushing. [Double-roll crusher]

Description: The double-roll fuel particle crusher was developed to fracture the silicon carbide coatings of Fort St. Vrain (FSV) fertile and fissile and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. The report details the design task for the fuel particle crusher, including historical test information on double-roll crushers for carbide-coated fuels and the design approach selected for the cold pilot plant crusher, and shows how the design addresses the equipment goals and operational objectives. Design calculations and considerations are included to support the selection of crusher drive and gearing, the materials chosen for crushing rolls and housing, and the bearing selection. The results of the initial testing for compliance with design objectives and operational capabilities are also presented. 8 figures, 4 tables.
Date: November 1, 1977
Creator: Baer, J.W.; Strand, J.B.; Cook, E.J. & Miller, C.M.
Partner: UNT Libraries Government Documents Department

Review of fission product plateout investigations at General Atomic. [HTGR]

Description: The status of fission product plateout studies at General Atomic is reviewed and suggestions are offered for future work. The deposition, or plateout, of condensible radionuclides in the primary circuits of gas-cooled reactors affects shielding requirements, maintenance procedures, and plant availability as well as representing a significant radiological source and/or sink for certain hypothetical accidents. Physical models and computer codes used to describe these plateout phenomena for reactor analysis are presented along with their limitations and possible refinements. The review includes portions of the recent AIPA study which sought to quantify the effects of uncertainties in input parameters on plateout code predictions. Major emphasis is placed upon the design methods verification program to assess the validity of plateout predictions by comparison of calculated behavior with experimental transport data.
Date: December 1, 1977
Creator: Hanson, D.L.
Partner: UNT Libraries Government Documents Department

MHTGR (Modular High-Temperature Gas-Cooled Reactor) technology development plan

Description: This paper presents the approach used to define the technology program needed to support design and licensing of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The MHTGR design depends heavily on data and information developed during the past 25 years to support large HTGR (LHTGR) designs. The technology program focuses on MHTGR-specific operating and accident conditions, and on validation of models and assumptions developed using LHTGR data. The technology program is briefly outlined, and a schedule is presented for completion of technology work which is consistent with completion of a Final Safety Summary Analysis Report (FSSAR) by 1992.
Date: January 1, 1988
Creator: Homan, F.J. & Neylan, A.J.
Partner: UNT Libraries Government Documents Department

The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

Description: High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, licensing, and risk analyses. During off normal events, where temperatures may reach up to 1600/degree/C, the data show that no significant radionuclide releases from the fuel will occur.
Date: January 1, 1988
Creator: Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W. & Verfondern, K.
Partner: UNT Libraries Government Documents Department

High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research. Quarterly progress report, April 1-June 30, 1980

Description: Development of the new steam turbine plant model ORTURB was completed. Further development, implementation, and verification work was done on the BLAST steam generator code.
Date: December 1, 1980
Creator: Ball, S J; Cleveland, J C; Harrington, R M & Conklin, J C
Partner: UNT Libraries Government Documents Department

Head-end reprocessing studies with irradiated high temperature gas-cooled reactor (HTGR) fuels

Description: Fifty (U-2.75 Th)C/sub 2/ and ThC/sub 2/ coated-particle fuel rods irradated in Peach Bottom were crushed and burned. The fertile and fissile fractions were separated using Thorex reagent and chemical analyses conducted for carbon, heavy metals, and fission products. Results were generally consistent with predictions, indicating that the reprocessing of TRISO-BISO fuel can be accomplished by the proposed flowsheet steps of crushing, fluidized-bed burning, coated particle separation and crushing, secondary burning, dissolution, clarification, and solvent extraction. (DLC)
Date: January 1, 1980
Creator: Fitzgerald, C.L. & Vaughen, V.C.A.
Partner: UNT Libraries Government Documents Department

Metals and Ceramics Division. Annual progress report, ending June 30, 1980

Description: Research is reported concerning: (1) engineering materials, including materials compatibility, mechanical properties, nondestructive testing, pressure vessel technology, and welding and brazing; (2) fuels and processes consisting of ceramic technology, fuel cycle technology, fuels evaluation, fuel fabrication and metals processing; and (3) materials science which includes, ceramic studies, physical metallurgy properties, radiation effects and microstructural analysis, metastable and superconducting materials, structure and properties of surfaces, theroretical research and x-ray research and applications. Highlights of the work of the metallographic group and the current state of the High-Temperature Materials Laboratory (HTML) and the Materials and Structures Technology Management Center (MSTMC) are presented. (FS)
Date: September 1, 1980
Partner: UNT Libraries Government Documents Department

HTGR technology development: status and direction

Description: During the last two years there has been an extensive and comprehensive effort expended primarily by General Atomic (GA) in generating a revised technology development plan. Oak Ridge National Laboratory (ORNL) has assisted in this effort, primarily through its interactions over the past years in working together with GA in technology development, but also through detailed review of the initial versions of the technology development plan as prepared by GA. The plan covers Fuel Technology, Materials Technology (including metals, graphite, and ceramics), Plant Technology (including methods, safety, structures, systems, heat exchangers, control and electrical, and mechanical), and Component Design Verification and Support areas (including the PCRV, control, fuel handling, service equipment, reactor core and internals, cooling and service systems).
Date: January 1, 1982
Creator: Kasten, P.R.
Partner: UNT Libraries Government Documents Department