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Gas cooled fast reactor

Description: Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.
Date: June 1, 1972
Partner: UNT Libraries Government Documents Department

Central worth and spectral measurements in the GCFR. Phase I assembly

Description: Central fission and capture rates, the central neutron spectrum and the reactivity worths of small samples were measured at the core center of the GCFR Phase I Assembly, the initial benchmark GCFR mockup assembly. Results of these measurements and comparisons with calculations are reported. (auth)
Date: November 1, 1975
Creator: Morman, J.A.; Bhattacharyya, S.K.; Smith, D.M.; McKnight, R.D.; Yule, T.J. & Bohn, E.M.
Partner: UNT Libraries Government Documents Department

Gas cooled fast reactor benchmark critical assembly

Description: The GCFR Phase I assembly is the initial Gas-Cooled Fast Reactor Benchmark assembly on the ZPR-9 reactor at Argonne National Laboratory. It represents the first full scale mockup of a GCFR ever assembled. It is a clean, simple geometry benchmark reference for the 300 MW(e) GCFR Demonstration Plant designed by General Atomic Company. A description and the evaluated specifications (zero-excess reactivity critical mass and dimensions) of the benchmark assembly are presented. (auth)
Date: November 1, 1975
Creator: Bhattacharyya, S.K.; McKnight, R.D.; Robinson, W.R.; Bohn, E.M.; Rusch, G.K.; Martens, F.H. et al.
Partner: UNT Libraries Government Documents Department

GCFR shielding design and supporting experimental programs

Description: The shielding for the conceptual design of the gas-cooled fast breeder reactor (GCFR) is described, and the component exposure design criteria which determine the shield design are presented. The experimental programs for validating the GCFR shielding design methods and data (which have been in existence since 1976) are also discussed.
Date: May 1, 1980
Creator: Perkins, R.G.; Hamilton, C.J. & Bartine, D.
Partner: UNT Libraries Government Documents Department

Integration of safety considerations into the design of GCFR safety systems

Description: Under DOE sponsorship, the GCFR Program is preparing a program to integrate reliability into the engineering and design of safety related systems, subsystems and components. This program is considered consistent with the NRC licensing position for CRBR which established a formal reliability objective for the prevention of core damage. The objective of the program is to provide assurance that reliability goals established for systems and subsystems are met consistent with the overall plant goals. Special consideration is given to components for which only a generic data base exists. Based on evaluations of past reliability test programs, it is concluded that full scale reliability test programs are not cost effective but that extended DV and S testing may be warranted in special circumstances. The major elements of the program, their relationship and benefits to the design of safety systems are discussed.
Date: October 1, 1979
Creator: Torri, A.; Kelley, A.P.; Emon, D.E. & Goetzmann, C.
Partner: UNT Libraries Government Documents Department

Refueling system for the gas-cooled fast breeder reactor

Description: Criteria specifically related to the handling of Gas-Cooled Fast Breeder Reactor (GCFR) fuel are briefly reviewed, and the most significant requirements with which the refueling system must comply are discussed. Each component of the refueling system is identified, and a functional description of the fuel handling machine is presented. An illustrated operating sequence describing the various functions involved in a typical refueling cycle is presented. The design status of components and subsystems selected for conceptual development is reviewed, and anticipated refueling time frames are given.
Date: May 1, 1980
Creator: Hawke, B.C.
Partner: UNT Libraries Government Documents Department

Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

Description: The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.
Date: September 1, 1978
Creator: None
Partner: UNT Libraries Government Documents Department

GCFR radial blanket and shield experiment: objectives, preanalysis, and specifications

Description: An integral experiment has been designed for the verification of radiation transport methods and nuclear data used for the design of the radial shield for the proposed 300 MW(e) gas-cooled fast breeder reactor (GCFR). The scope of the experiment was chosen to include a thorium oxide radial blanket mockup as well as several shield configurations in order to reduce the uncertainties in the calculated source terms for the radial shield, and to reduce the uncertainties in the calculated radiation damage to the prestressed concrete reactor vessel (PCRV). Additionally, the measurements are intended to bound the uncertainties in calculated gamma-ray heating rates within the blanket and shield. Although designed specifically for the GCFR, the experiment will provide generic data regarding deep penetration in ThO/sub 2/ and common shield materials, which should also benefit LMFBR designers.
Date: September 1, 1979
Creator: Ingersoll, D.T.; Bartine, D.E. & Muckenthaler, F.J.
Partner: UNT Libraries Government Documents Department

Overview of the GCFR core engineering development program

Description: The gas-cooled fast breeder reactor (GCFR) core development plan has been described as an incremental extension of the LMFBR program. The basis for this contention is reviewed along with some additional gas-cooled reactor precedence. Ideally, in the development of a reactor concept, it would be desirable to demonstrate the perforance of a full-sized fuel assembly under the anticipated irradiation and coolant conditions. Practically, however, this is seldom possible, particularly in fast reactors where fluence-to-burnup ratios increase with reactor power. Because of this limitation, the GCFR development program is based on qualifying the assemblies for the low distortions that occur during the swelling incubation period (approx. 50 MWd/kg). Extended burnup will be demonstrated in the operation of the GCFR demonstration plant core. The overall core development program and the related assumptions are discussed and the design goals stated. A multistage development approach is described and illustratd with an abbreviated program network. Major areas of the development program are briefly described.
Date: July 1, 1979
Creator: Snyder, H.J. Jr.
Partner: UNT Libraries Government Documents Department

Analysis of gas-cooled fast reactor shield designs

Description: In its shielding program for the Gas-Cooled Fast Reactor (GCFR) as conceived by General Atomic, Oak Ridge National Laboratory has developed an advanced shielding analysis system that incorporates the latest analysis techniques for converging to a shield design compatible with other design parameters such as cooling and structural requirements or material compatibility. Basically the system consists in applying the various techniques in a logical sequence to a given design, thereby generating a large body of data to serve as an information base for subsequent redesign. As an illustration, this system is applied to successive typical models for the GCFR, resulting in a reduction in the thickness of the radial shield and redesign of the lower shield region. In principle, the design-analysis-redesign iterations would continue until they converge upon an acceptable configuration.
Date: January 1, 1978
Creator: Bartine, D.E. & Williams, L.R.
Partner: UNT Libraries Government Documents Department

Evaluation of molten fuel containment concepts for gas-cooled fast breeder reactors

Description: Four in-vessel molten fuel containment concepts for the GCFR were compared, namely, (1) a ceramic crucible, (2) a borax bath, (3) a heavy metal bath, and (4) a steel bath. The ceramic crucible is the simplest but depends on substantial upward heat removal. The borax bath and the heavy metal bath concepts offer better performance but would require design changes and an increased experimental effort. The steel bath concept is a good compromise and has potential for further improvement by combining it with the essential features of other concepts, i.e., the crucible or the heavy metal bath. It is concluded that several concepts could potentially exploit the normally provided cooled liner barrier in the PCRV cavity for post-accident fuel containment.
Date: October 1, 1979
Creator: Kang, C.S. & Torri, A.
Partner: UNT Libraries Government Documents Department

Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

Description: A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.
Date: October 1, 1979
Partner: UNT Libraries Government Documents Department

Breakdown voltage at the electric terminals of GCFR-core flow test loop fuel rod simulators in helium and air

Description: Tests were performed to determine the ac and dc breakdown voltage at the terminal ends of a fuel rod simulator (FRS) in helium and air atmospheres. The tests were performed at low pressures (1 to 2 atm) and at temperatures from 20 to 350/sup 0/C (68 to 660/sup 0/F). The area of concern was the 0.64-mm (0.025-in.) gap between the coaxial conductor of the FRS and the sheaths of the four internal thermocouples as they exit the FRS. The tests were prformed to ensure a sufficient safety margin during Core Flow Test Loop (CFTL) operations that require potentials up to 350 V ac at the FRS terminals. The primary conclusion from the test results is that the CFTL cannot be operated safely if the terminal ends of the FRSs are surrounded by a helium atmosphere but can be operated safely in air.
Date: December 1, 1979
Creator: Huntley, W.R. & Conley, T.B.
Partner: UNT Libraries Government Documents Department

Assessment of a core meltdown in the gas-cooled fast breeder reactor with an upflow core

Description: This paper discusses the chronological sequence of events and supporting analysis of a postulated total loss of all coolant circulation in the gas-cooled fast breeder reactor (GCFR) with an upflow core. Redundant and diverse cooling systems are provided for decay heat removal, including pressurized natural circulation in the core auxiliary cooling system, which reduce the probability of this postulated event below the range of plant design bases. Nevertheless, this postulated accident has been considered so that the potential for consequence mitigation and containment margin could be investigated. Two distinct phases of the sequence are discussed: (1) the core response to a total loss of forced and natural coolant circulation and (2) the capability of the prestressed concrete reactor vessel (PCRV) to retain molten fuel debris. Specific design features of the GCFR which prevent recriticality and fuel vaporization due to fuel slumping are under investigation. Analytical work has been initiated to determine the potential for consequence mitigation in the PCRV and the containment. Several concepts for postaccident fuel containment have been identified and appear technically feasible.
Date: July 1, 1979
Creator: Torri, A.; Frank, M.V. & Kang, C.
Partner: UNT Libraries Government Documents Department

CASY: a dynamic simulation of the gas cooled fast breeder reactor core auxiliary cooling system. Volume 1. Technical discussion

Description: This report documents the development of the digital computer simulation code CASY. CASY is designed to evaluate alternative systems to perform emergency core cooling. It is modular in form such that modifications can be made easily. CASY was independently validated by comparing similar analysis results with existing core and core auxiliary cooling system transient codes. CASY is written in the preprocessor language available at General Atomic entitled SYSL.
Date: September 1, 1979
Partner: UNT Libraries Government Documents Department

CASY: a dynamic simulation of the gas-cooled fast breeder reactor core auxiliary cooling system. Volume II. Example computer run

Description: A listing of a CASY computer run is presented. It was initiated from a demand terminal and, therefore, contains the identification ST0952. This run also contains an INDEX listing of the subroutine UPDATE. The run includes a simulated scram transient at 30 seconds.
Date: September 1, 1979
Partner: UNT Libraries Government Documents Department

Life-limiting aspects of GCFR core assembly designs

Description: A review is made of the various potential life-limiting effects of current consideration in the GCFR core assemblies. The most life-limiting aspects are not related to individual fuel rod behavior characteristics, but rather to distortions within the rod bundles. Fuel rod internal fission gas is not limiting, as it is in most fast reactors, because of the GCFR pressure-equalization system. The life-limiting bundle distortions are the result of neutron-flux-enhanced creep and fast-fluence-induced metal swelling. These effects are somewhat unique to fast reactors and are material dependent. The current evaluations apply to the use of 20% cold-worked type 316 stainless steel as the core structural material. Alternate structural materials having lower neutron-fluence-induced metal swelling would reduce the distortions, in which case, the most life-limiting effect would likely be one of the more traditional effects such as fuel-clad mechanical interaction (FCMI).
Date: July 1, 1979
Creator: LaBar, M.P.
Partner: UNT Libraries Government Documents Department

Hydraulic design and performance of a GCFR core assembly orifice

Description: The design and performance of a core assembly orifice for gas-cooled fast-breeder reactors (GCFRs) are studied in this report. Successful reactor operation relies on adequate cooling, among other things, and orificing is important to cooling. A simple, yet effective, graphical design method for estimating the loss coefficient of an orifice and its associated opening area is presented. A numerical example is also provided for demonstration of the method. The effect of the orifice configuration on orifice hydraulic performance is discussed. The design method stated above provides a first estimate toward an orifice design. Hydraulic experiments are required for verification of the design adequacy.
Date: February 1, 1980
Creator: Tang, I.M.
Partner: UNT Libraries Government Documents Department

GCFR plant control system

Description: A plant control system is being designed for a gas-cooled fast breeder reactor (GCFR) demonstration plant. Control analysis is being performed as an integral part of the plant design process to ensure that control requirements are satisfied as the plant design evolves. The load control portion of the plant control system provides stable automatic (closed-loop) control of the plant over the 25% to 100% load range. Simulation results are presented to demonstrate load control system performance. The results show that the plant is controllable at full load with the control system structure selected, but gain scheduling is required to achieve desired performance over the load range.
Date: May 1, 1980
Creator: Estrine, E.A. & Greiner, H.G.
Partner: UNT Libraries Government Documents Department

GCFR residual heat removal capability

Description: The residual heat removal (RHR) capability for providing lines of protection (LOPS) 1 and 2 of the gas-cooled fast breeder reactor (GCFR) demonstration plant is described. Included are design criteria and system descriptions for the RHR cooling systems and the portion of the plant protection system that is related to initiation of the RHR system operation. The design features of these systems provide inherently redundant and diverse means of core cooling for the GCFR. The hierarchy in the selection of the RHR systems and the application of the systems to key transient events are discussed. Methods of RHR system operation, dynamic responses of the GCFR plant, and margins of safety in RHR operations are also presented.
Date: May 1, 1980
Creator: Chi, H.W.; Chung, H.S. & Shenoy, A.
Partner: UNT Libraries Government Documents Department