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Fission product retention in TRISCO coated UO sub 2 particle fuels subjected to HTR simulated core heating tests

Description: Results of the examination and analysis of 25,730 individual microspheres from spherical fuel elements HFR-K3/1 and HFR-K3/3 are reported. The parent spheres were irradiated in excess of end-of-life exposure and subsequently subjected to simulated core heating tests in a special high-temperature furnace at Forschungszentrum, Juelich, GmbH (KFA). Following the heating tests, the spheres were electrolytically deconsolidated to obtain unbonded fuel particles for Irradiated Microsphere Gamma Analyzer (IMGA) analysis. For sphere HFR-K3/1, which was heated for 500 h at 1600{degree}C, only four particles were identified as having released fission products. The remaining particles from the sphere showed no statistical evidence of fission product release. Scanning Electron Microscopy (SEM) examination showed that three of the defect particles had large sections of the TRISO coating missing, while the fourth appeared normal. For sphere HFR-K3/3, which was heated for 100 h at 1800{degree}C, the IMGA data revealed that fission product release (cesium) from individual particles was significant and that there was large particle-to-particle variation in retention capabilities. Individual particle release (cesium) averaged ten times the KFA-measured integral spherical fuel element release value. In addition, the bimodal distribution of the individual particle data indicated that two distinct modes of failure at fuel temperatures of 1800{degree}C and above may exist. 6 refs., 6 figs., 4 tabs.
Date: November 1, 1990
Creator: Baldwin, C.A. & Kania, M.J.
Partner: UNT Libraries Government Documents Department

Diffusion modeling of fission product release during depressurized core conduction cooldown conditions

Description: A simple model for diffusion through the silicon carbide layer of TRISO particles is applied to the data for accident condition testing of fuel spheres for the High-Temperature Reactor program of the Federal Republic of Germany (FRG). Categorization of sphere release of {sup 137}Cs based on fast neutron fluence permits predictions of release with an accuracy comparable to that of the US/FRG accident condition fuel performance model. Calculations are also performed for {sup 85}Kr, {sup 90}Sr, and {sup 110m}Ag. Diffusion of cesium through SiC suggests that models of fuel failure should consider fuel performance during repeated accident condition thermal cycling. Microstructural considerations in models in fission product release are discussed. The neutron-induced segregation of silicon within the SiC structure is postulated as a mechanism for enhanced fission product release during accident conditions. An oxygen-enhanced SiC decomposition mechanism is also discussed. 12 refs., 11 figs., 2 tabs.
Date: January 1, 1990
Creator: Martin, R.C.
Partner: UNT Libraries Government Documents Department

IMGA [Irradiated Microsphere Gamma Analyzer] examination of the Set No. 4 fuel under project work statement FD-20

Description: Results of an examination of over 10,800 unbonded fuel particles from three irradiated spherical fuel elements by the Irradiated Microsphere Gamma Analyzer system are reported. The investigation was initiated to assess fission product behavior in LEU UO{sub 2} TRISO-coated fuel particles at elevated temperatures. Of the three spheres considered, one was reserved as a control and the other two were subjected to simulated accident-condition temperatures of 1600{degree}C and 1800{degree}C, respectively. For the control sphere and the sphere tested at 1600{degree}C, no statistical evidence of fission product release (cesium) from individual particles was observed. At fuel temperatures of 1800{degree}C, however, fission product release (cesium) from individual particles was significant and there was large particles-to-particle variation. At 1800{degree}C, individual particle release (cesium) was on average ten times the Kernforschungsanlage-measured integral spherical fuel element release value. Particle release data from the sphere tested at 1800{degree}C indicate that there may be two distinct modes of failure at fuel temperatures of 1800{degree}C and above. 5 refs., 9 figs., 9 tabs.
Date: March 1, 1990
Creator: Baldwin, C.A. & Kania, M.J.
Partner: UNT Libraries Government Documents Department

Capsule HRB-21 postirradiation examination plan

Description: Irradiation capsule HRB-21 is a test capsule designed to provide Modular High-Temperature Gas-Cooled Reactor (MHTGR) coated particle fuel performance data under test reactor conditions representative of normal MHTGR operation. The irradiated fuel will also be used for postirradiation heating in a controlled atmosphere allowing acquisition of fission product release data at sustained high temperatures. The in-reactor performance data, the postirradiation examination data, and the postirradiation heating data will be used for the validation of fuel performance models under normal and off-normal operating conditions. The accelerated irradiation is to take place in the High Flux Isotope Reactor (HFIR) at ORNL. This report identifies the procedures to be followed in carrying out the postirradiation disassembly and examination of HRB-21. Included is a description of the capsule, a detailed sequence of steps for disassembly of the capsule, a description of the postirradiation examination techniques to be employed, and specifications for the storage of capsule components and the reporting of results. 9 refs., 6 figs., 2 tabs.
Date: March 1, 1990
Creator: Packan, N.H.; Kania, M.J. & Shrader, L.G.
Partner: UNT Libraries Government Documents Department

Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

Description: Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.
Date: May 1, 1990
Partner: UNT Libraries Government Documents Department

Development of a detailed core flow analysis code for prismatic fuel reactors

Description: The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs.
Date: January 1, 1990
Creator: Bennett, R.G.
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

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

Preconceptual design of the new production reactor circulator test facility

Description: This report presents the results of a study of a new circulator test facility for the New Production Reactor Modular High-Temperature Gas-Cooled Reactor. The report addresses the preconceptual design of a stand-alone test facility with all the required equipment to test the Main Circulator/shutoff valve and Shutdown Cooling Circulator/shutoff valve. Each type of circulator will be tested in its own full flow, full power helium test loop. Testing will cover the entire operating range of each unit. The loop will include a test vessel, in which the circulator/valve will be mounted, and external piping. The external flow piping will include a throttle valve, flowmeter, and heat exchanger. Subsystems will include helium handling, helium purification, and cooling water. A computer-based data acquisition and control system will be provided. The estimated costs for the design and construction of this facility are included. 2 refs., 15 figs.
Date: June 1, 1990
Creator: Thurston, G.
Partner: UNT Libraries Government Documents Department

Cost estimate guidelines for advanced nuclear power technologies

Description: To make comparative assessments of competing technologies, consistent ground rules must be applied when developing cost estimates. This document provides a uniform set of assumptions, ground rules, and requirements that can be used in developing cost estimates for advanced nuclear power technologies. 10 refs., 8 figs., 32 tabs.
Date: March 1, 1990
Creator: Delene, J.G. & Hudson, C.R. II.
Partner: UNT Libraries Government Documents Department

Modeling and performance of the MHTGR (Modular High-Temperature Gas-Cooled Reactor) reactor cavity cooling system

Description: The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.
Date: April 1, 1990
Creator: Conklin, J.C. (Oak Ridge National Lab., TN (USA))
Partner: UNT Libraries Government Documents Department

MHTGR product distinction

Description: Commercial power reactor operating experience and regulatory practice in the United States is dominated by Light Water Reactor (LWR) technology. This experience base and regulatory structure is a point of departure for the development of any future nuclear option, and must be understood and utilized as appropriate. This report discusses experience with current generation plants considered relevant to MHTGR development, as well as the fundamental differences between the MHTGR and LWR technology, and quantifies avoided business risks relative to current generation plants. It does not attempt to compare the MHTGR with other advanced nuclear options, LWR or other technologies, which are also under development and must relate to the historical framework in accordance with their own unique characteristics. 14 refs., 5 figs., 5 tabs.
Date: December 1, 1990
Partner: UNT Libraries Government Documents Department

New Production Reactors Program Plan

Description: Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs.
Date: December 1, 1990
Partner: UNT Libraries Government Documents Department

Office of Analysis and Evaluation of Operational Data 1989 annual report, Power reactors

Description: The annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1989. The report is published in two separate parts. This document, NUREG-1272, Vol. 4, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports, diagnostic evaluations, and reports to the NRC's Operations Center. This report also compiles the status of staff actions resulting from previous Incident Investigation Team (IIT) reports. 16 figs., 9 tabs.
Date: July 1, 1990
Partner: UNT Libraries Government Documents Department

The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

Description: A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.
Date: December 1, 1990
Creator: Hollaway, W. R.; Lidsky, L. M. & Miller, M. M.
Partner: UNT Libraries Government Documents Department

Bounding core temperature transients for severe and rapid water ingress scenarios in modular high temperature gas-cooled reactors

Description: A rapid water ingress transient, resulting from steam generator tube or tube-sheet failures, could lead to a reactivity insertion and core heatup in the Modular High Temperature Gas-Cooled Reactors. This paper considers the effect of hypothetical rapid and severe water ingress scenarios of extremely low probability, and assesses the effect of such transients on potentially excessive fuel temperatures and subsequent fuel failures. The results indicate that for the worst postulated scenarios the conservatively set limiting fuel temperature of 1600{degree}C is indeed exceeded, but only for a few seconds, and then only in a small fraction of the core. Therefore, it appears that even the most severe and rapid water ingress transients would not lead to significant fuel failures. Parametric variations of the key variables indicate that the reactivity worth of water and the fuel thermal properties must be established with high confidence as the design progresses. 7 refs., 11 figs., 3 tabs.
Date: January 1, 1990
Creator: Kroeger, P.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

(Fuel, fission product, and graphite technology)

Description: Travel to the Forschungszentrum (KFA) -- Juelich described in this report was for the purpose of participating in the annual meeting of subprogram managers for the US/DOE Umbrella Agreement for Fuel, Fission Product, and Graphite Technology. At this meeting the highlights of the cooperative exchange were reviewed for the time period June 1989 through June 1990. The program continues to contribute technology in an effective way for both countries. Revision 15 of the Subprogram Plan will be issued as a result of the meeting. There was interest expressed by KFA management in the level of support received from the NPR program and in potential participation in the COMEDIE loop experiment being conducted at the CEA.
Date: July 25, 1990
Creator: Stansfield, O.M.
Partner: UNT Libraries Government Documents Department

A vectorized heat transfer model for solid reactor cores

Description: The new generation of nuclear reactors includes designs that are significantly different from light water reactors. Among these new reactor designs is the Modular High-Temperature Gas-Cooled Reactor (MHTGR). In addition, nuclear thermal rockets share a number of similarities with terrestrial HTGRs and would be amenable to similar types of analyses. In these reactors, the heat transfer in the solid core mass is of primary interest in design and safety assessment. One significant safety feature of these reactors is the capability to withstand a loss of pressure and forced cooling in the primary system and still maintain peak fuel temperatures below the safe threshold for retaining the fission products. To accurately assess the performance of gas-cooled reactors during these types of transients, a Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions. Also, HERA has been used to analyze a depressurized loss of forced cooling transient in a HTGR with a very detailed three-dimensional input model. The results compare favorably with other means of analysis and provide further validation of the models and methods. 18 refs., 11 figs.
Date: January 1, 1990
Creator: Rider, W. J.; Cappiello, M. W. & Liles, D. R.
Partner: UNT Libraries Government Documents Department