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Evaluation of EPRI nuclear power division research topics supportive of HTGR technology

Description: For HTGR commercialization studies, an LWR/HTGR Technology Transfer program was devised. Candidate programs were identified out of a total of 208 EPRI NPD (Nuclear Power Division) projects. Of these, 26 project areas presented the highest probability for technology transfer. (DLC)
Date: October 6, 1978
Partner: UNT Libraries Government Documents Department

Gas-cooled reactors

Description: Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.
Date: January 1, 1976
Creator: Schulten, R. & Trauger, D. B.
Partner: UNT Libraries Government Documents Department

Medium-size high-temperature gas-cooled reactor

Description: This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760/sup 0/C (1400/sup 0/F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics (a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant) and engineered safety features (core auxiliary cooling, relief valve, and steam generator dump systems).
Date: August 1, 1980
Creator: Peinado, C.O. & Koutz, S.L.
Partner: UNT Libraries Government Documents Department

Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, January 1, 1979-March 31, 1979

Description: This report presents the results of work performed from January 1, 1979 through March 31, 1979 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. Work covered in this report includes the activities associated with the creep-rupture testing of the test materials for the purpose of verifying the stresses selected for the screening creep test program, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment.
Date: July 19, 1979
Partner: UNT Libraries Government Documents Department

Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

Description: The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas.
Date: September 1, 1978
Partner: UNT Libraries Government Documents Department

New small HTGR power plant concept with inherently safe features - an engineering and economic challenge

Description: Studies are in a very early design stage to establish a modular concept High-Temperature Gas-Cooled Reactor (HTGR) plant of about 100-MW(e) size to meet the special needs of small energy users in the industrialized and developing nations. The basic approach is to design a small system in which, even under the extreme conditions of loss of reactor pressure and loss of forced core cooling, the temperature would remain low enough so that the fuel would retain essentially all the fission products and the owner's investment would not be jeopardized. To realize economic goals, the designer faces the challenge of providing a standardized nuclear heat source, relying on a high percentage of factory fabrication to reduce site construction time, and keeping the system simple. While the proposed nuclear plant concept embodies new features, there is a large technology base to draw upon for the design of a small HTGR.
Date: January 1, 1983
Creator: McDonald, C.F. & Sonn, D.L.
Partner: UNT Libraries Government Documents Department

Near-isotropic petroleum-coke based graphites for high temperature gas-cooled reactor core components

Description: The standard covers procurement requirements for extruded graphite logs, 15 in. (381 mm) or greater in diameter, manufactured with near-isotropic petroleum cokes and coal-tar pitch binders which are candidates or reference materials for replaceable fuel and reflector blocks for High-Temperature Gas-Cooled Reactors (HTGRs). The requirements are designed to produce the degree of lot-to-lot reproducibility which is required to ensure consistent and predictable properties and irradiation performance for specific graphite grades and to ensure traceability of the graphite logs to production processes and raw materials that affect performance. The standard is intended for use in the procurement of developmental and commercial grades of nuclear graphite which are to be evaluated on Department of Energy (DOE) funded programs for use as core components in HTGRs.
Date: October 1, 1977
Partner: UNT Libraries Government Documents Department

1170 MW/sub t/ HTGR steamer cogeneration plant: design and cost study

Description: A conceptual design and cost study is presented for intermediate size high temperature gas-cooled reactor (HTGR) for industrial energy applications performed by United Engineers and Constructors Inc., (UE and C) and The General Atomic Company (GAC). The study is part of a program at ORNL and has the objective to provide support in the evaluation of the technical and economic feasibility of a single unit 1170 MW/sub t/ HTGR steam cycle cogeneration plant (referred to as the Steamer plant) for the production of industrial process energy. Inherent in the achievement of this objective, it was essential to perform a number of basic tasks such as the development of plant concept, capital cost estimate, project schedule and annual operation and maintenance (O and M) cost.
Date: August 1, 1980
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

Transient diffusion through a spherical shell into its finite spherical core. [HTGR]

Description: The problem of the diffusion of gas through a spherical shell into its spherical core is solved by using the Laplace transform technique. The transformed solution is obtained and then inverted to the corresponding solution in real space. This real space solution is slowly converging for small times. Therefore, an alternate expression is obtained that is appropriate and more rapidly convergent for small times. As an example, the first solution is used to calculate the theoretical gas content of the core of High-Temperature Gas-Cooled Reactor (HTGR) biso-coated fuel particles from fuel-particle batch OR2261/HT after 850/sup 0/C anneals in helium.
Date: August 1, 1979
Creator: Reeves, M. & Tolliver, J.S.
Partner: UNT Libraries Government Documents Department

US/FRG joint report on the pebble bed high temperature reactor resource conservation potential and associated fuel cycle costs

Description: Independent analyses at ORNL and KFA have led to the general conclusion that the flexibility in design and operation of a high-temperature gas-cooled pebble-bed reactor (PBR) can result in favorable ore utilization and fuel costs in comparison with other reactor types, in particular, with light-water reactors (LWRs). Fuel reprocessign and recycle show considerable promise for reducing ore consumption, and even the PBR throwaway cycle is competitive with fuel recycle in an LWR. The best performance results from the use of highly enriched fuel. Proliferation-resistant measures can be taken using medium-enriched fuel at a modest ore penalty, while use of low-enriched fuel would incur further ore penalty. Breeding is possible but net generation of fuel at a significant rate would be expensive, becoming more feasible as ore costs increase substantially. The /sup 233/U inventory for a breeder could be produced by prebreeders using /sup 235/U fuel.
Date: November 1, 1979
Creator: Teuchert, E.; Ruetten, H.J.; Worley, B.A. & Vondy, D.R.
Partner: UNT Libraries Government Documents Department

Pebble Bed Reactor: core physics and fuel cycle analysis

Description: The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.
Date: October 1, 1979
Creator: Vondy, D.R. & Worley, B.A.
Partner: UNT Libraries Government Documents Department

Performance of HTGR fuel in HFIR capsule HT-33

Description: Irradiation capsule HT-33 was a cooperative effort between General Atomic Company (GA) and Oak Ridge National Laboratory (ORNL). In this capsule ThO/sub 2/ particles (fabricated by GA), low-enriched uranium particles, inert carbon particles, and various fuel rod matrices were tested under accelerated irradiation in the High-Flux Isotope Reactor. Visual examination showed good irradiation behavior for fuel rods with slug-injected matrices (using a pitch binder) and warm-molded matrices (using a thermosetting resin binder). Rod debonding improved somewhat with fuel rods that used GLCC H-451 ground graphite shim particles rather than Speer fluid coke shim particles. Measurements of permeability (by inert gas intrusion) of the pyrocarbon on the inert particles showed that the disorder created by the neutron flux did not increase the inert gas permeability. Metallographic examination of Triso-coated particles irradiated both with and without an outer pyrocarbon coating revealed that the outer coating is necessary to suppress SiC degradation at temperatures above approximately 1375/sup 0/C. The fission product behavior (determined by the electron microprobe) was similar in both low-enriched and high-enriched uranium particles made from weak-acid resins. Furthermore, fission product palladium caused severe SiC corrosion at time-averaged temperatures above 1400/sup 0/C.
Date: June 1, 1979
Creator: Tiegs, T.N. & Robbins, J.M.
Partner: UNT Libraries Government Documents Department

PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

Description: This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.
Date: September 1, 1981
Creator: Vondy, D.R.
Partner: UNT Libraries Government Documents Department

Structural model testing for prestressed concrete pressure vessels: a study of grouted vs nongrouted posttensioned prestressing tendon systems. [HTGR]

Description: Nongrouted tendons are predominantly used in this country as the prestressing system for prestressed concrete pressure vessels (PCPVs) because they are more easily surveyed to detect reductions in prestressing level and distress such as results from corrosion. Grouted tendon systems, however, offer advantages which may make them cost-effective for PCPV applications. Literature was reviewed to (1) provide insight on the behavior of grouted tendon system, (2) establish performance histories for structures utilizing grouted tendons, (3) examine corrosion protection procedures for prestressing tendons, (4) identify arguments for and against using grouted tendons, and (5) aid in the development of the experimental investigation. The experimental investigation was divided into four phases: (1) grouted-nongrouted tendon behavior, (2) evaluation of selected new material systems, (3) bench-scale corrosion studies, and (4) preliminary evaluation of acoustic emission techniques for monitoring grouted tendons in PCPVs. The groutability of large tendon systems was also investigated.
Date: April 1, 1979
Creator: Naus, D.J.
Partner: UNT Libraries Government Documents Department

Characterization of SiC coatings on HTGR fuel particles: preliminary report

Description: Fuel particles for the HTGR contain a layer of pyrolytic silicon carbide to act as a pressure vessel and fission product barrier. The SiC is deposited by the thermal decomposition of methyltrichlorosilane (CH/sub 3/SiC/sub 3/ or MTS) in an excess of hydrogen. Coatings deposited at temperatures from 1500 to 1700/sup 0/C and coating rates of 0.4 to 1.2 ..mu..m/min have been studied by optical microscopy, x-ray diffraction, electron microscopy, and density measurements. Transmission electron microscopy (TEM) has shown the microstructural features to be extremely complex and much finer than the ability of optical microscopy to resolve. X-ray diffraction has detected traces of ..cap alpha..-SiC in some coatings, and those conditions of deposition temperature and coating rate that give rise to this phase have been determined.
Date: August 1, 1979
Creator: Lauf, R.J.; Braski, D.N. & Tennery, V.J.
Partner: UNT Libraries Government Documents Department

Interface-currents integral transport model for treating doubly-heterogeneous, multisystem geometries

Description: An analytical model for calculating neutron spectra in the doubly-heterogeneous fuel-moderator geometries of the pebble bed reactor concept is presented. The model is capable of simultaneously treating more than one type of fuel grain in the fuel matrix and more than one type of pebble in the reactor core. The model was developed to assess the need for treating various levels of material heterogeneity in processing neutron multigroup cross sections in the resolved resonance energy range. The slowing-down calculation is performed over a pointwise energy mesh tailored to the cross section structure for the nuclides present in the problem. Isotropic, elastic scattering theory is applied in an explicit calculation of down-scattered sources due to neutron interaction with all materials in all zones. At each energy point, neutron transport between zones is calculated with the interface-currents integral transport technique. Here, this technique is extended to include the simultaneous treatment of coupled, one-dimensional, multiregion systems. The coupling between the two levels of heterogeneity (grain systems and pebble systems) is accomplished by a sequence of source normalization and cross section averaging treatments. The equations applied in the slowing-down and spatial transport models are presented. Results from the analyses of single pebble and double pebble systems indicate the importance of resonance shielding as a function of fuel kernel diameter, fuel loading in each pebble, and the presence of more than one type of pebble in the system.
Date: January 1, 1979
Creator: Westfall, R.M. & Bjerke, M.A.
Partner: UNT Libraries Government Documents Department

Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

Description: This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation.
Date: November 1, 1979
Partner: UNT Libraries Government Documents Department

Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

Description: This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned.
Date: September 1, 1978
Partner: UNT Libraries Government Documents Department

Nonlinear response to the multiple sine wave excitation of a softening--hardening system. [HTGR]

Description: In studying the earthquake response of the HTGR core, it was observed that the system can display softening--hardening characteristics. This is of great consequence in evaluating the structural safety aspects of the core. In order to obtain a better understanding of the governing parameters, an investigation was undertaken with a single-degree-of-freedom system having a softening--hardening spring characteristic and excited by multiple sine waves. A parametric study varying the input amplitudes and the spring characteristic was performed. Transients were introduced into the system, and the jump phenomena between the lower softening characteristics to the higher hardening curve was studied.
Date: January 1, 1979
Creator: Koplik, B.; Subudhi, M. & Curreri, J.
Partner: UNT Libraries Government Documents Department

Gas-cooled reactor power systems for space

Description: Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system.
Date: January 1, 1987
Creator: Walter, C.E.
Partner: UNT Libraries Government Documents Department

Numerical determination of inert gas permeability parameters of High-Temperature Gas-Cooled Reactor (HTGR) fuel-particles

Description: The high temperature diffusion of inert gases through an outer layer of dense carbon into the ThO/sub 2/ core of biso-coated High-Temperature Gas-Cooled Reactor (HTGR) fuel-particles is studied numerically. A mathematical model of diffusion through a dense spherical shell into a spherical core is used to numerically calculate the theoretical gas content of the core. This theoretical calculation, in tandem with an optimizing computer code and experimental data, is used to determine the diffusion coefficient of the shell and the porosities of the shell and inner core. The activation energy is also determined for use in an Arrhenius relationship between the diffusion coefficients and absolute temperature.
Date: November 1, 1979
Creator: Tolliver, J.S.
Partner: UNT Libraries Government Documents Department