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A Gas-Cooled Reactor Surface Power System

Description: A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.
Date: November 9, 1998
Creator: Harms, G.A.; Lenard, R.X.; Lipinski, R.J. & Wright, S.A.
Partner: UNT Libraries Government Documents Department

The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors: A preliminary assessment of experiments HRB-17, HFR-B1, HFR-K6 and KORA

Description: The effect of water vapor on the release of fission gas from the fuel elements of high temperature, gas-cooled reactors has been measured in different laboratories under both irradiation and post irradiation conditions. The data from experiments HRB-17, HFR-B1, HFR-K6, and in the KORA facility are compared to assess their consistency and complimentarily. The experiments are consistent under comparable experimental conditions and reveal two general mechanisms involving exposed fuel kernels embedded in carbonaceous materials. One is manifest as a strong dependence of fission gas release on the partial pressure of water vapor below 1 kPa and the other, as a weak dependence above 1 kPa.
Date: September 1, 1995
Creator: Myers, B.F.
Partner: UNT Libraries Government Documents Department

Licensed operating reactors: Status summary report data as of December 31, 1991. Volume 16

Description: The Nuclear Regulatory Commission`s annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar year 1991) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided.
Date: March 1, 1992
Partner: UNT Libraries Government Documents Department

On the benefits of an integrated nuclear complex for Nevada

Description: An integrated nuclear complex is proposed for location at the Nevada Test Site. In addition to solving the nuclear waste disposal problem, this complex would tremendously enhance the southern Nevada economy, and it would provide low cost electricity to each resident and business in the affected counties. Nuclear industry and the national economy would benefit because the complex would demonstrate the new generation of safer nuclear power plants and revitalize the industry. Many spin-offs of the complex would be possible, including research into nuclear fusion and a world class medical facility for southern Nevada. For such a complex to become a reality, the cycle of distrust between the federal government and the State of Nevada must be broken. The paper concludes with a discussion of implementation through a public process led by state officials and culminating in a voter referendum.
Date: January 1, 1994
Creator: Blink, J.A. & Halsey, W.G.
Partner: UNT Libraries Government Documents Department

Formulation of a possible advanced reactor legislative strategy and proposal

Description: A number of initiatives have been taken to date regarding the formulation of legislation to support in various ways the DOE advanced nuclear reactor program. Among the more prominent of these are bills that have been introduced by Sen. Johnston (D-La) and Rep. Udall (D-Az) as well as a draft bill put together by the nuclear industry and that could be introduced by Rep. Stallings (D-Id). These legislative initiatives are presented in this paper.
Date: December 31, 1994
Partner: UNT Libraries Government Documents Department

Validation of NESTLE against static reactor benchmark problems

Description: The NESTLE advanced modal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE`s geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs) and CANDU heavy- water reactors (HWRs).
Date: February 1, 1996
Creator: Mosteller, R.D.
Partner: UNT Libraries Government Documents Department

Proceedings of the Third International Workshop on the implementation of ALARA at nuclear power plants

Description: This report contains the papers presented and the discussions that took place at the Third International Workshop on ALARA Implementation at Nuclear Power Plants, held in Hauppauge, Long Island, New York from May 8--11, 1994. The purpose of the workshop was to bring together scientists, engineers, health physicists, regulators, managers and other persons who are involved with occupational dose control and ALARA issues. The countries represented were: Canada, Finland, France, Germany, Japan, Korea, Mexico, the Netherlands, Spain, Sweden, the United Kingdom and the United States. The workshop was organized into twelve sessions and three panel discussions. Individual papers have been cataloged separately.
Date: March 1995
Creator: Khan, T. A. & Roecklein, A. K.
Partner: UNT Libraries Government Documents Department

New concept of small power reactor without on-site refueling for non-proliferation

Description: Energy demand in developing countries is increasing to support growing populations and economies. This demand is expected to continue growing at a rapid pace well into the next century. Because current power sources, including fossil, renewable, and nuclear, cannot meet energy demands, many developing countries are interested in building a new generation of small reactor systems to help meet their needs. The U.S. recognizes the need for energy in the developing countries. In its 1998 Comprehensive Energy Strategy, the Department of Energy calls for research into low-cost, proliferation- resistant, nuclear reactor technologies to ensure that this demand can be met in a manner consistent with U.S. non-proliferation goals and policies. This research has two primary thrusts: first, the development of a small proliferation-resistant nuclear system (i.e., a technology focus); second, the continuation of open communication with the international community through early engagement and cooperation on small reactor development. A system that meets developing country requirements must: (1) achieve reliably safe operation with a minimum of maintenance and supporting infrastructure; (2) offer economic competitiveness with alternative energy sources available to the candidate sites; and (3) demonstrate significant improvements in proliferation resistance relative to existing reactor systems. These challenges are the most significant driving forces behind the LLNL proposed program for development of a new, small nuclear reactor system. This report describes a technical approach for developing small nuclear power systems for use in developing countries. The approach being proposed will establish a preliminary set of requirements that, if met, will cause new innovative approaches to system design to be used. The proposed approach will borrow from experience gained over the past forty years with four types of nuclear reactor technologies (LWR, LMR, HTGR, and MSR) to develop four or more pre-conceptual designs. The pre-conceptual designs will be used to confirm the ...
Date: July 13, 1998
Creator: Brown, N.W., LLNL
Partner: UNT Libraries Government Documents Department

Circulating water subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

Description: The Circulating Water System is a subsystem within the Heat Rejection Group (HRG). The Circulating Water System consists of two independent loops to remove waste heat from the turbine building closed cooling water system and from the condensers associated with each turbine generator set. In normal plant operation circulating water is pumped from the cooling tower basin through the condensers and heat exchangers and back to the cooling tower where the waste heat is released to the atmosphere via mechanical draft cooling towers. The system consists of two flow paths with two half-size, vertical pumps associated with each path.
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

USA/FRG umbrella agreement for cooperation in GCR [Gas Cooled Reactor] development: Fuel, fission products and graphite subprogram. Part 1, Management meeting report: Part 2, Revised subprogram plan, Revision 10

Description: This Subprogram Plan describes cooperative work in the areas of HTR fuel and graphite development and fission product studies that is being carried out under US/FRG/Swiss Implementing Agreement for cooperation in Gas Cooled Reactor development. Only bilateral US/FRG cooperation is included, since it is the only active work in this subprogram area at this time. The cooperation has been in progress since February 1977. A number of Project Work Statements have been developed in each of the major areas of the subprogram, and work on many of them is in progress. The following specific areas are included in the scope of this plan: fuel development; graphite development; fission product release; and fission product behavior outside the fuel elements.
Date: May 1, 1986
Partner: UNT Libraries Government Documents Department

High-Temperature Gas-Cooled Reactor Technology Development Program: Annual progress report for period ending December 31, 1987

Description: The High-Temperature Gas-Cooled Reactor (HTGR) Program being carried out under the US Department of Energy (DOE) continues to emphasize the development of modular high-temperature gas-cooled reactors (MHTGRs) possessing a high degree of inherent safety. The emphasis at this time is to develop the preliminary design of the reference MHTGR and to develop the associated technology base and licensing infrastructure in support of future reactor deployment. A longer-term objective is to realize the full high-temperature potential of HTGRs in gas turbine and high-temperature, process-heat applications. This document summarizes the activities of the HTGR Technology Development Program for the period ending December 31, 1987.
Date: March 1, 1989
Creator: Jones, J.E.,Jr.; Kasten, P.R.; Rittenhouse, P.L. & Sanders, J.P.
Partner: UNT Libraries Government Documents Department

Experimental test plan: USDOE/JAERI collaborative program for the coated particle fuel performance test

Description: This document describes the coated-particle fuel performance test agreed to under Annex 2 of the arrangement between the US Department of Energy and the Japan Atomic Energy Research Institute on cooperation in research and development regarding high-temperature gas-cooled reactors (HTGRs). The test will evaluate the behavior of reference fuel compacts containing coated-particle fuels fabricated according to the specifications for the US Modular HTGR and the Japanese High-Temperature Engineering Test Reactor (HTTR) concepts. Two experimental capsules, HRB-21 and HRB-22, are being tested. Capsule HRB-21 contains only US reference fuel, and HRB-22 contains only JAERI reference fuel. Both capsules will be irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Capsule HRB-21 will be operated at a mean volumetric fuel temperature of 975{degrees}C and will achieve a peak fissile burnup of 26% fissions per initial metal atom (FIMA) and a fast fluence of {le}4.5 {times} 10{sup 25} neutrons/m{sup 2}. Capsule HRB-22 will be operated at a mean centerline fuel temperature of 1250 to 1300{degrees}C and will achieve a peak fissile burnup of 5.5% FIMA and a fast fluence of 1.7 {times} 10{sup 25} neutrons/m{sup 2}. Performance of the fuels during irradiation will be closely monitored using on-line fission gas surveillance. Following irradiation, both capsules will undergo detailed examinations and core heatup simulation testing. Results from in-reactor monitoring and postirradiation testing will be analyzed to comparatively assess US and Japanese coated-particle fuel performance. 3 refs., 9 figs., 10 tabs.
Date: December 1989
Creator: Kania, M. J. & Fukuda, K.
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

Makeup water treatment and auxiliary boiler building structural design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

Description: The Makeup Water Treatment and Auxiliary Boiler Building (MWABB) is a grade-founded, single-story, steel-framed structure with insulated sheet metal exterior walls and roof decking. It houses the electrically-heated auxiliary boiler and related equipment, and the Raw Water Treatment System. The Makeup Water Treatment and Auxiliary Boiler building is located adjacent to the Maintenance Building in the Energy Conversion Area of the plant.
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

Liquid radioactive waste subsystem design description

Description: The Liquid Radioactive Waste Subsystem provides a reliable system to safely control liquid waste radiation and to collect, process, and dispose of all radioactive liquid waste without impairing plant operation. Liquid waste is stored in radwaste receiver tanks and is processed through demineralizers and temporarily stored in test tanks prior to sampling and discharge. Radwastes unsuitable for discharge are transferred to the Solid Radwaste System.
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

Class 1E dc power subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

Description: The Class 1E DC Power System of the Electrical Group provides reliable and regulated 125 V dc electric power to the plant safety-related dc loads connected to the Four redundant and independent 125 V dc Class 1E buses to ensure plant safe shutdown or mitigate the effects of a design basis event. These four dc buses comprise the plant four Class 1E dc control and instrument channels (A, B, C and D).
Date: June 1, 1986
Partner: UNT Libraries Government Documents Department

MHTGR-Nuclear Island Engineering: Final summary report for the period November 30, 1987 through December 1, 1988

Description: This report summarizes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) - Nuclear Island Engineering (NIE) design and development work performed by General Atomics (GA) for the period November 30, 1987 through December 1, 1988, under the Department of Energy (DOE) Contract AC03-88SF17367. The scope of the report includes work performed by Bechtel National Inc. (BNI), Combustion Engineering Inc. (C-E), and James Howden Company, as major subcontractors to GA.
Date: December 1, 1988
Partner: UNT Libraries Government Documents Department

Pressure relief subsystem design description

Description: The primary function of the Pressure Relief Subsystem, a subsystem of the Vessel System, is to provide overpressure protection to the Vessel System. When the overpressure setpoint is reached, pressure is reduced by permitting the flow of primary coolant out of the Vessel System. This subsystem also provides the flow path by which purified helium is returned to the vessel system, either as circulating purge/flow from the Helium Purification Subsystem or make-up helium from the Helium Storage and Transfer Subsystem.
Date: July 1, 1986
Partner: UNT Libraries Government Documents Department

Fission product plateout/liftoff/washoff test plan. Revision 1

Description: A test program is planned in the COMEDIE loop of the Commissariat a l`Energy Atomique (CEA), Grenoble, France, to generate integral test data for the validation of computer codes used to predict fission product transport and core corrosion in the Modular High Temperature Gas-Cooled Reactor (MHTGR). The inpile testing will be performed by the CEA under contract from the US Department of Energy (DOE); the contract will be administered by Oak Ridge National Laboratory (ORNL). The primary purpose of this test plan is to provide an overview of the proposed program in terms of the overall scope and schedule. 8 refs, 3 figs.
Date: May 1, 1988
Creator: Acharya, R. & Hanson, D.
Partner: UNT Libraries Government Documents Department

Comparison of US/FRG accident condition fuel failure and release models

Description: Although there are many differences between the High-Temperature Gas Cooled Reactor (HTGR) concepts being developed in the US and the High Temperature Reactor (HTR) concepts in the Federal Republic of Germany (FRG), the coated fuel particles are very similar. Significant benefits are achievable through cooperative research and exchange of information and data on the fuel performance and radionuclide retention in the coated fuel particles. This draft report describes cooperative work on HTGR safety research as agreed to in the "USA/FRG Umbrella Agreement for Cooperation in GCR Development: Safety Research Subprogram Plan," specifically, this work was conducted under Project Work Statement (PWS) S-6 titled "Fission Product Retention in Fuel," 9 refs., 12 figs., 4 tabs.
Date: May 29, 1989
Creator: Bolin, J. & Dunn, T.
Partner: UNT Libraries Government Documents Department

An evaluation of the suitability of laser-induced fluorescence for measurements of fission-product iodine sorptivity in the MHTGR [modular high-temperature gas-cooled reactor]

Description: Experiments and calculations indicate that laser-induced fluorescence (LIF) lacks the sensitivity needed for sorptivity measurements of I{sub 2} or other molecular species at partial pressures below 10{sup {minus}11} atm. Although the technique may have sufficient sensitivity for measurements of atomic species, the species of interest are, in all likelihood, not atomic. Methods of measurement which would allow the determination of species are proposed. 9 refs., 6 figs.
Date: July 1989
Creator: Sherrow, S. A.
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

Predicted discharge plutonium isotopics for LEU [low-enriched uranium] test pebble irradiated in the AVR [Arbeitsgemeinschaft Versuchsreaktor]

Description: A Subprogram Plan related to the Arbeitsgemeinschaft Versuchsreaktor (AVR) Test Program is in place and describes cooperative work being carried out under the United States/Federal Republic of Germany (US/FRG) Implementing Agreement for Cooperation in Gas-Cooled Reactor Development. The AVR information to be provided as described in the plan will provide a basis for examining the accuracy of computational methods used for performance and safety analysis. The purpose of the cooperation is to obtain experimental information from the AVR relevant to the performance and safety of modular gas-cooled reactors, and to compare measured results with predictions of analytical tools. This report provides a progress report on the prediction of plutonium buildup in LEU fuel in a high-enriched uranium (HEU) core and also describes the method for calculating the U-238 resonance integral (cross section). 4 refs., 5 figs., 11 tabs.
Date: June 1, 1988
Creator: Lane, R.K. & Lefler, W.L.
Partner: UNT Libraries Government Documents Department