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Modularization and nuclear power. Report by the Technology Transfer Modularization Task Team

Description: This report describes the results of the work performed by the Technology Transfer Task Team on Modularization. This work was performed as part of the Technology Transfer work being performed under Department of Energy Contract 54-7WM-335406, between December, 1984 and February, 1985. The purpose of this task team effort was to briefly survey the current use of modularization in the nuclear and non-nuclear industries and to assess and evaluate the techniques available for potential application to nuclear power. A key conclusion of the evaluation was that there was a need for a study to establish guidelines for the future development of Light Water Reactor, High Temperature Gas Reactor and Liquid Metal Reactor plants. The guidelines should identify how modularization can improve construction, maintenance, life extension and decommissioning.
Date: June 1, 1985
Partner: UNT Libraries Government Documents Department

Advances in HTGR fuel performance models

Description: Fuel performance models based on empirical evidence are used to predict particle failure and fission product release in the design of high-temperature gas-cooled reactors (HTGRs). Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10/sup -4/ fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW(t) HTGR.
Date: February 1, 1985
Creator: Stansfield, O.M.; Goodin, D.T.; Hanson, D.L. & Turner, R.F.
Partner: UNT Libraries Government Documents Department

Gaseous and metallic fission product release characteristics of a modular pebble bed HTGR during loss of core cooling accidents

Description: A quantitative safety criteria for the high-temperature gas-cooled reactor (HTGR) is to limit the radiological consequences for a wide spectrum of accidents to a level not requiring public sheltering. This leads to reliance on passive safety characteristics for improbable loss of core cooling accidents. Models have been developed to predict the transport of metallic and gaseous fission products (FPs) through the multilayered fuel particle coatings and the graphite matrix of the core under accident conditions. Using these models, FP transport and releases were calculated for a loss of core convective cooling accident in a 250-MW(t) 3.8-W/cc pebble bed HTGR. Fission-product transport through the particle kernel and coatings, the graphite pebbles/reflectors, the reactor vessel, and the confinement were assessed. The results of this study show that the most effective barrier to fission products is the coated fuel particle. The reactor vessel and the confinement provide additional attenuation for the small amount released from the core. The small release to the environment occurs over a period of days and is so low that the safety criterion of 5 rem thyroid dose (to avoid offsite sheltering) is satisfied with a margin of more than an order of magnitude. 6 figs.
Date: May 1, 1985
Creator: Inamati, S.B.; Richards, M.B. & Hoot, C.G.
Partner: UNT Libraries Government Documents Department

Analysis and evaluation of recent operational experience from the Fort St. Vrain HTGR

Description: The Fort St. Vrain operating experience to be discussed here includes notable safety-related events which have occurred since late 1981 when ORNL was first contracted to provide technical assistance to AEOD. Earlier Fort St. Vrain operating experience through the time of successful full-power testing in November 1981 has been summarized by the licensee and the reactor vendor, GA Technologies, Inc. (GA), in papers presented at several different forums during 1982. In addition, extensive and very useful detailed evaluations of preoperational and startup testing and of the rise-to-power operating experience through completion of the first refueling outage in August 1979 have been compiled into a series of reports under the sponsorship of the Electric Power Research Institute (EPRI). Finally, the US Department of Energy's Fort St. Vrain Improvement Plan provides a summary of the major operational limits which have affected the plant since start-up. The events discussed here are categorized based on the major systems affected, namely, (1) primary system and reactor vessel, (2) electrical systems, and (3) the reactor building. In all cases to be discussed, the lessons to be learned are vigilance and prevention. These lessons translate into the need for the recognition and control of unexpected situations and of their potential for branching effects. At Fort St. Vrain, these lessons are found in the effects of moisture ingress, in the challenges experienced to the supply of essential electrical power, and in controlling the environment of the reactor building. 13 refs.
Date: May 1, 1985
Creator: Moses, D.L. & Lanning, W.D.
Partner: UNT Libraries Government Documents Department

Development of high-strength concrete mix designs in support of the prestressed concrete reactor vessel design for a HTGR steam cycle/cogeneration plant

Description: Design optimization studies indicate that a significant reduction in the size of the PCRV for a 2240 MW(t) HTGR plant can be effected through utilization of high-strength concrete in conjunction with large capacity prestressing systems. A three-phase test program to develop and evaluate high-strength concretes (>63.4 MPa) is described. Results obtained under Phase I of the investigation related to materials selection-evaluation and mix design development are presented. 3 refs., 4 figs.
Date: January 1, 1985
Creator: Naus, D.J. & Oland, C.B.
Partner: UNT Libraries Government Documents Department

Containment vs confinement trade study, small HTGR plant PCRV [prestressed concrete reactor vessel] concept

Description: This trade study has been conducted to evaluate the differences between four different HTGR nuclear power plants. All of the plants use a prestressed concrete reactor vessel (PCRV) to house the core and steam generation equipment. The reactor uses LEU U/Th fuel in prismatic carbon blocks. All plant concepts meet the utility/user requirements established for small HTGR plants. All plants will be evaluated with regard to their ability to produce safe, economical power to satisfy Goals 1, 2, and 3 of the HTGR program and by meeting the MUST criteria established in the concept evaluation plan. Capital costs for each plant were evaluated on a differential cost basis. These costs were developed according to the ``NUS`` code of accounts as defined in the Cost Estimating and Control Procedure, HP-20901. Accounts that were identical in scope for all four plants were not used for the comparison. Table 1-1 is a list of capital cost accounts that were evaluated for each plant.
Date: March 1, 1985
Partner: UNT Libraries Government Documents Department

Report on chemical impurities in lots of 2020 graphite

Description: Chemical and physical studies on agglomerated and dispersed impurities found in three lots of 2020 graphite received from the Stackpole Company are described. Each lot consisted of three billets, one shipment of off-the-shelf 2020 and two of ``nuclear quality`` graphite for which efforts had been taken to decrease impurity levels. An earlier report described the results of chemical analyses of the three lots of graphite and the gross distribution of the impurities in the billets. The present report describes the chemical and physical studies on the detailed distribution and chemical identity of the agglomerates that led to the identification of their source. Agglomerated impurities were found in all lots of 2020. The particles possessed an average size range of about 0.3 mm. In addition to these particulates, other inorganic ash forming components were found to be rather more uniformly distributed internally throughout the graphite. 4 refs., 4 figs., 3 tabs.
Date: November 1, 1985
Creator: Strehlow, R.A.
Partner: UNT Libraries Government Documents Department

Analysis of iodine and cesium chemical forms evolved from graphite surfaces at temperatures from 425 to 1400{sup 0}C

Description: Information has been obtained to aid in the identification of the chemical forms of fission product cesium and iodine which are evolved from graphite surfaces heated to temperatures up to 1400{sup 0}C. Iodine and cesium were initially added to the graphite as adsorbed CsI; subsequently, more cesium was added as Cs{sub 2}0 to allow variations of the initial cesium/iodine mol ratio from 1 to 10. The identifications were determined, in part, by inference from the locations of cesium and iodine deposits on a graphite thermal gradient tube, the measured mol ratios of the deposits, and the results of electron surface chemical analyses. Cesium iodide was the most abundant of the chemical forms found; however, significant quantities of cesium-rich oxygen-bearing deposits (probably cesium oxide) and of iodine-rich deposits (mostly molecular I{sub 2}) were also present. The iodine species, CsI and I{sub 2}, were found to move downstream with time and/or gas flow from the hotter to the colder regions of the system. 3 refs., 8 figs., 5 tabs.
Date: August 1, 1985
Creator: Tallent, O. K.; Wichner, R. P.; Towns, R. L. & Godsey, T. T.
Partner: UNT Libraries Government Documents Department

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

Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

Description: This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed.
Date: November 1, 1985
Creator: Harrington, R.M.; Ball, S.J. & Cleveland, J.C.
Partner: UNT Libraries Government Documents Department

Helium-cooled high temperature reactors

Description: Experience with several helium cooled reactors has been favorable, and two commercial plants are now operating. Both of these units are of the High Temperature Graphite Gas Cooled concept, one in the United States and the other in the Federal Republic of Germany. The initial helium charge for a reactor of the 1000 MW(e) size is modest, approx.15,000 kg.
Date: January 1, 1985
Creator: Trauger, D.B.
Partner: UNT Libraries Government Documents Department

Assessment of modelling needs for safety analysis of current HTGR concepts

Description: In view of the recent shift in emphasis of the DOE/Industry HTGR development efforts to smaller modular designs it became necessary to review the modelling needs and the codes available to assess the safety performance of these new designs. This report provides a final assessment of the most urgent modelling needs, comparing these to the tools available, and outlining the most significant areas where further modelling is required. Plans to implement the required work are presented. 47 refs., 20 figs.
Date: December 1, 1985
Creator: Kroeger, P.G. & Van Tuyle, G.J.
Partner: UNT Libraries Government Documents Department

ORNL analyses of AVR performance and safety

Description: Because of the high interest in modular High Temperature Reactor performance and safety, a cooperative project has been established involving the Oak Ridge National Laboratory (ORNL), Arbeitsgemeinschaft Versuchs Reaktor GmbH (AVR), and Kernforschungsanlage Juelich GmbH (KFA) in reactor physics, performance and safety. This paper presents initial results of ORNL's examination of a hypothetical depressurized core heatup accident and consideration of how a depressurized core heatup test might be conducted by AVR staff. Also presented are initial analyses of a test involving a reduction in core flow and of a test involving reactivity insertion via control rod withdrawal.
Date: January 1, 1985
Creator: Cleveland, J.C.
Partner: UNT Libraries Government Documents Department

Very high flux research reactors based on particle fuels

Description: A new approach to high flux research reactors is described, the VHFR (Very High Flux Reactor). The VHFR fuel region(s) are packed beds of HTGR-type fuel particles through which coolant (e.g., D/sub 2/O) flows directly. The small particle diameter (typically on the order of 500 microns) results in very large surface areas for heat transfer (approx. 100 cm/sup 2//cm/sup 3/ of bed), high power densities (approx. 10 megawatts per liter), and minimal ..delta..T between fuel and coolant (approx. 10 K) VHFR designs are presented which achieve steady-state fluxes of approx. 2x10/sup 16/ n/cm/sup 2/sec. Deuterium/beryllium combinations give the highest flux levels. Critical mass is low, approx. 2 kg /sup 235/U for 20% enriched fuel. Refueling can be carried out continuously on-line, or in a batch process with a short daily shutdown. Fission product inventory is very low, approx. 100 to 300 grams, depending on design.
Date: January 1, 1985
Creator: Powell, J.R. & Takahashi, H.
Partner: UNT Libraries Government Documents Department

Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

Description: One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600/sup 0/C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth.
Date: January 1, 1985
Creator: Harrington, R.M. & Ball, S.J.
Partner: UNT Libraries Government Documents Department

ORNL's NRC-sponsored HTGR safety and licensing analysis activities for Fort St. Vrain and advanced reactors

Description: The ORNL safety analysis program for the HTGR was established in 1974 to provide technical assistance to the USNRC on licensing questions for both Fort St. Vrain and advanced plant concepts. The emphasis has been on development of major component and system dynamic simulation codes, and use of these codes to analyze specific licensing-related scenarios. The program has also emphasized code verification, using Fort St. Vrain data where applicable, and comparing results with industry-generated codes. By the use of model and parameter adjustment routines, safety-significant uncertainties have been identified. A major part of the analysis work has been done for the Fort St. Vrain HTGR, and has included analyses of FSAR accident scenario re-evaluations, the core block oscillation problem, core support thermal stress questions, technical specification upgrade review, and TMI action plan applicability studies. The large, 2240-MW(t) cogeneration lead plant design was analyzed in a multi-laboratory cooperative effort to estimate fission product source terms from postulated severe accidents.
Date: January 1, 1985
Creator: Ball, S.J.; Cleveland, J.C. & Harrington, R.M.
Partner: UNT Libraries Government Documents Department