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Physics of gas cooled reactors

Description: From meeting on new developments in reactor physics and shieiding calculations; Lake Kiamesha, New York, USA (12 Sep The temperature coefficient of the HTOR is composed of a strong negative Doppler coefficient and a positive moderator coefficient, the net effect being about -10 x 10/sup -5// deg C at operating temperature. the positive moderator coefficient is a result of the absence of significant density coefficients and the nuclear characteristics of U- 233. In addition, selected iission products, principally Xe-135, contribute positive components to the temperature coefficient. Current estimates of the value of the temperature coefficient in the HTGR are presented along with recent experimental data pertinent to the subject. (11 references) (auth)
Date: July 15, 1972
Creator: Dahlberg, R.C.
Partner: UNT Libraries Government Documents Department

HTGR Mechanistic Source Terms White Paper

Description: The primary purposes of this white paper are: (1) to describe the proposed approach for developing event specific mechanistic source terms for HTGR design and licensing, (2) to describe the technology development programs required to validate the design methods used to predict these mechanistic source terms and (3) to obtain agreement from the NRC that, subject to appropriate validation through the technology development program, the approach for developing event specific mechanistic source terms is acceptable
Date: July 1, 2010
Creator: Moe, Wayne
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

Description: The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.
Date: July 1, 2010
Creator: Wright, J. K. & Wright, R. N.
Partner: UNT Libraries Government Documents Department

Next Generation Nuclear Plant Resilient Control System Functional Analysis

Description: Control Systems and their associated instrumentation must meet reliability, availability, maintainability, and resiliency criteria in order for high temperature gas-cooled reactors (HTGRs) to be economically competitive. Research, perhaps requiring several years, may be needed to develop control systems to support plant availability and resiliency. This report functionally analyzes the gaps between traditional and resilient control systems as applicable to HTGRs, which includes the Next Generation Nuclear Plant; defines resilient controls; assesses the current state of both traditional and resilient control systems; and documents the functional gaps existing between these two controls approaches as applicable to HTGRs. This report supports the development of an overall strategy for applying resilient controls to HTGRs by showing that control systems with adequate levels of resilience perform at higher levels, respond more quickly to disturbances, increase operational efficiency, and increase public protection.
Date: July 1, 2010
Creator: Stevens, Lynne M.
Partner: UNT Libraries Government Documents Department

Energy Economic Data Base (EEDB) Program. Phase III. Final report and third update

Description: The objective of the USDOE EEDB Program is to provide periodic updates of technical and cost (capital, fuel and operating and maintenance) information of significance to the US Department of Energy. This information is intended to be used by USDOE in evauating and monitoring US civilian nuclear power programs, and to provide them with a consistent means of evaluating the nuclear option and proposed alternatives. The data tables, which make up the bulk of the report, are updated to January 1, 1980. The data in these tables and in the backup data file supercede the information presented in the Second Update (1979). Where required, new descriptive information is added in the text to supplement the data tables.
Date: July 1, 1981
Partner: UNT Libraries Government Documents Department

HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1983

Description: This document reports the technical accomplishments of the HTGR Fuel Technology Program at GA Technologies Inc. during the first half of FY 83. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, core component verification, and core technology transfer tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR.
Date: July 1, 1983
Partner: UNT Libraries Government Documents Department

Gas-cooled reactor programs: High-Temperature Gas-Cooled Reactor Base-Technology Program. Annual progress report for period ending December 31, 1979

Description: Progress in HTGR studies is reported in the following areas: HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; and evaluation of the pebble-bed HTR.
Date: July 1, 1980
Partner: UNT Libraries Government Documents Department

Analysis of reactor strategies to meet world nuclear energy demands

Description: A number of reactor deployment strategies for long-term nuclear system development are analyzed from a global perspective in terms of resource utilization and economic benefits. Two time frames are chosen: 1975 - 2025 and 1975 - 2050. Uranium demand for various strategies is compared with uranium supply assuming different production capabilities and resource base. The analysis shows that a given reactor deployment strategy could strongly influence the extent of uranium exploration and production. Power systems cost comparisons are made to identify clearly competitive or non-competitive reactors. The sensitivity of power cost to different uranium price projections and nuclear demands is also examined. The results indicate that breeders are necessary to support a long-term nuclear power system. Advanced converter-breeder symbiotic systems, particularly those operating on the Th/U-233 cycle, have clear advantages in terms of resources and economics.
Date: July 1, 1979
Creator: Ligon, D.M. & Brogli, R.H.
Partner: UNT Libraries Government Documents Department

Influences of the couple-stresses on the pure-bending of a circular cylinder. [HTGR]

Description: This report presents the solution to the pure-bending of a circular cylinder with the couple-stress theory of linear elasticity. A linear bending moment-curvature relation is derived parallel to the classical beam theory. The section modulus (or the proportional coefficient) associated with the couple-stress theory is always greater than that predicted by the classical theory, and the ratio of the former to the latter increases as the radius of the beam decreases. These aspects clearly agree with the observed behavior of nuclear-grade graphite. Based on the solution, it is further estimated that the characteristic length l/sub 2/ of the couple-stress theory for H-451 graphite ranges from 0.62 to 1.54 mm. This range of l/sub 2/ concurs with the magnitude of the grain size (maximum is 1.57 mm for H-451 graphite) and agrees with an aspect of the couple-stress theory.
Date: July 1, 1979
Creator: Kao, B.G.; Tzung, F.K. & Ho, F.H.
Partner: UNT Libraries Government Documents Department

Investigation of stick-slip (chatter) phenomenon of HTGR thermal barrier attachment fixture sliding interfaces. Phase I test: Class A thermal barrier hardware and environment

Description: This test program was performed to investigate if significant chatter (stick-slip) would occur at the thermal barrier sliding surfaces. Given such conditions, cyclic loads could be induced in the thermal barrier attachment fixture and studs. A representative section of thermal barrier was tested with realistic HTGR temperature cycles in a high purity helium environment. No significant chatter was detected and there was no visible deterioration of the hardware after testing.
Date: July 1, 1979
Creator: Middleton, A.
Partner: UNT Libraries Government Documents Department

Fission-product SiC reaction in HTGR fuel

Description: The primary barrier to release of fission product from any of the fuel types into the primary circuit of the HTGR are the coatings on the fuel particles. Both pyrolytic carbon and silicon carbide coatings are very effective in retaining fission gases under normal operating conditions. One of the possible performance limitations which has been observed in irradiation tests of TRISO fuel is chemical interaction of the SiC layer with fission products. This reaction reduces the thickness of the SiC layer in TRISO particles and can lead to release of fission products from the particles if the SiC layer is completely penetrated. The experimental section of this report describes the results of work at General Atomic concerning the reaction of fission products with silicon carbide. The discussion section describes data obtained by various laboratories and includes (1) a description of the fission products which have been found to react with SiC; (2) a description of the kinetics of silicon carbide thinning caused by fission product reaction during out-of-pile thermal gradient heating and the application of these kinetics to in-pile irradiation; and (3) a comparison of silicon carbide thinning in LEU and HEU fuels.
Date: July 13, 1981
Creator: Montgomery, F.
Partner: UNT Libraries Government Documents Department

Design of an HTGR for high-temperature process heat applications

Description: The high-temperature gas-cooled reactor (HTGR) offers a unique heat source for process heat applications since its operating temperature is substantially higher than that of other types of nuclear reactors. This paper discusses a design study of an advanced 842-MW(t) very high temperature reactor (VHTR) coupled with a chemical process whose product is hydrogen (or a mixture of hydrogen and carbon monoxide) generated by steam reforming of a light hydrocarbon mixture. A key feature of the plant is the nuclear reactor core, which utilizes helium as its primary coolant, has ceramic-coated fuel particles containing uranium and thorium, and employs graphite as the moderator and structural material. As in other HTGR designs, the VHTR has its entire primary coolant system contained in a prestressed concrete reactor vessel (PCRV) which provides the necessary biological shielding and pressure containment. The high-temperature nuclear thermal energy is transported to the externally located process plant by a secondary helium transport loop. In addition to providing the thermal driving potential required for the reforming process, the nuclear heat is also used to generate high-temperature, high-pressure steam to satisfy both the process and electrical generation needs for the operation of the nuclear plant and reforming process plant.
Date: July 1, 1979
Creator: Vrable, D.L.; Quade, R.N. & Stanley, J.D.
Partner: UNT Libraries Government Documents Department

Beta and gamma dose calculations for PWR and BWR containments

Description: Analyses of gamma and beta dose in selected regions in PWR and BWR containment buildings have been performed for a range of fission product releases from selected severe accidents. The objective of this study was to determine the radiation dose that safety-related equipment could experience during the selected severe accident sequences. The resulting dose calculations demonstrate the extent to which design basis accident qualified equipment could also be qualified for the severe accident environments. Surry was chosen as the representative PWR plant while Peach Bottom was selected to represent BWRs. Battelle Columbus Laboratory performed the source term release analyses. The AB epsilon scenario (an intermediate to large LOCA with failure to recover onsite or offsite electrical power) was selected as the base case Surry accident, and the AE scenario (a large break LOCA with one initiating event and a combination of failures in two emergency cooling systems) was selected as the base case Peach Bottom accident. Radionuclide release was bounded for both scenarios by including spray operation and arrested sequences as variations of the base scenarios. Sandia National Laboratories used the source terms to calculate dose to selected containment regions. Scenarios with sprays operational resulted in a total dose comparable to that (2.20 /times/ 10/sup 8/ rads) used in current equipment qualification testing. The base case scenarios resulted in some calculated doses roughly an order of magnitude above the current 2.20 /times/ 10/sup 8/ rad equipment qualification test region. 8 refs., 23 figs., 12 tabs.
Date: July 1, 1989
Creator: King, D.B.
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

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

Vessel support subsystem design description. Revision 1

Description: The Vessel Support Subsystem is one of three subsystems comprising the Vessel System of the Modular High Temperature Gas-Cooled Reactor 4 x 350 MW(t) Plant. The design of this subsystem has been developed by means of the Integrated Approach. This document establishes the functions and system design requirements of the Vessel Support Subsystem from the Functional Analysis, and includes institutional requirements from the Overall Plant Design Specification and the Vessel System Design Description. A description of the subsystem design which satisfies these requirements is presented. Lower-tier requirements at the subsystem level are next defined for the component design. This document also includes information on aspects of subsystem construction, operation, maintenance, and decommissioning.
Date: July 1, 1987
Creator: Perry, R.A. & Mehta, D.D.
Partner: UNT Libraries Government Documents Department

Helium Storage and Transfer Subsystem design description. Revision

Description: The Helium Storage and Transfer Subsystem (HSTS) consists of two parts. The first consists of nine (9) high pressure storage tanks containing helium at 15.6 MPa (2250 psig). These tanks provide makeup and purge helium at a rate of 1216 kg per y (2680 lb/y) to the various helium users, including circulator bearing seals, analysis packages, and cooling system surge tanks. The second, larger part of the system, provides for the low pressure storage of 6078 kg (13,400 lb) of primary coolant helium in 180 storage tanks at 7.0 MPa (1000 psig). The system serves all four (4) reactor modules. The low pressure storage part of the system receives helium from the discharge of Helium Purification Subsystem (HPS) and is activated during depressurization and pumpup operations only. It is not required to operate continuously. Storage capacity is provided for primary helium coolant from two reactor modules. However, since depressurization and pumpup operations are performed for only one reactor module at a time, two 50% capacity low pressure transfer compressors are provided having a total transfer capacity of 340 am{sup 3}/h (200 acfm) which is sufficient to service one module. High pressure helium is supplied continuously to all the four reactor modules simultaneously from the high pressure storage tanks. These tanks are replaced periodically with fresh tanks.
Date: July 1, 1987
Partner: UNT Libraries Government Documents Department

Metals design handbook

Description: This report gives an approved set of material properties over a range of environmental conditions which are sufficient to design the metallic components in the reactor system and hot duct assembly. Table 1-1 list these metallic components together with the reference design material chosen for each component. Table 1-2 summarizes the structural criteria of each metallic component taken from the component specifications. In all cases, the criteria references the ASME B&PV Code. The ASME-Code includes the material properties of Coded material. The Code does not, however, include environmental effects (such as irradiation, corrosion, or thermal aging), and for some components the material maximum allowable temperature is below that of the design and/or postulated ``safety-related`` accident conditions. Table 1-3 gives the Code limits for the portions of the Code given in Table 1-2. This document includes the effects of the radiation environment, chemical impurity effects (in the primary coolant), and the effects of thermal aging and corrosion on the metallic properties. The design information introduced in this document includes that available from the ASME B&PV Code High-Temperature Code Cases plus material information from General Atomics (GA) and Oak Ridge National Laboratories (ORNL) that is published.
Date: July 1, 1988
Creator: Betts, W.S.
Partner: UNT Libraries Government Documents Department

Site fuel handling subsystem design description. Revision

Description: The Site Fuel Handling Subsystem (SFHS) consists of equipment and facilities located in the reactor Service Building which are used to handle hexagonal graphite fuel and reflector blocks. This equipment interfaces closely with the core refueling equipment. The SFHS uses some of the equipment in the Core Refueling System to transfer fuel elements between the spent fuel storage facility (part of Core Refueling Subsystem, HFD-43413) and the fuel sealing and inspection facility (FSIF).
Date: July 1, 1987
Partner: UNT Libraries Government Documents Department

Hot Service Facility subsystem design description. Revision

Description: The Hot Service Facility Subsystem, which is also referred to as the Reactor Equipment Service Facility (RESF), is located in an environmentally controlled shielded vault and provides inspection, maintenance, care, and repair of reactor service equipment and tools. The shielded vault is located in the Reactor Service Building.
Date: July 1, 1987
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

Core refueling subsystem design description. Revision 1

Description: The Core Refueling Subsystem of the Fuel Handling and Storage System provides the mechanisms and tools necessary for the removal and replacement of the hexagonal elements which comprise the reactor core. The Core Refueling Subsystem is not "safety-related." The Core Refueling Subsystem equipment is used to prepare the plant for element removal and replacement, install the machines which handle the elements, maintain control of air inleakage and radiation release, transport the elements between the core and storage, and control the automatic and manual operations of the machines. Much of the element handling is performed inside the vessel, and the entire exchange of elements between storage and core is performed with the elements in a helium atmosphere. The core refueling operations are conducted with the reactor module shutdown and the primary coolant pressure slightly subatmospheric. The subsystem is capable of accomplishing the refueling in a reliable manner commensurate with the plant availability requirements.
Date: July 1, 1987
Creator: Anderson, J.K. & Harvey, E.C.
Partner: UNT Libraries Government Documents Department