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Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

Description: The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.
Date: October 1, 1978
Creator: Philbin, J.S.; Nelson, W.E. & Rosenstroch, B.
Partner: UNT Libraries Government Documents Department

Projections of ENDF/B version V performance for fast and thermal reactors using sensitivity coefficients

Description: Proposed reductions to /sup 235/U(anti ..nu..) and /sup 235/U(n,f) in the fast energy range have significant impact for uranium fueled fast critical assemblies. The long-standing LMFBR /sup 28/c//sup 49/f calculated overprediction is not resolved by proposed Version 5 cross section modifications for /sup 238/U(n,..gamma..) and /sup 239/Pu(n,f). The upward evaluation for the /sup 239/Pu(n,f)//sup 235/U(n,f) ratio improves criticality predictions for Pu fueled fast assemblies. For thermal reactors, changes to the /sup 238/U resonance parameters significantly reduce the long-standing /sup 28/rho discrepancy. Reduced resonance capture in the 1 eV /sup 240/Pu resonsnce has significant implications for LWR fuel cycle studies.
Date: January 1, 1978
Creator: Weisbin, C.R.; Marable, J.H.; Hardy, J. Jr. & McKnight, R.D.
Partner: UNT Libraries Government Documents Department

Physics and feasibility study of the Fast-Mixed Spectrum Reactor concept

Description: Reactor physics and fuel cycle studies, coordinated with heat transfer and material science and structural analysis work has indicated the feasibility potential of the coupled Fast-Mixed Spectrum Reactor (FMSR) concept. This concept employs what are considered reasonable extrapolations of present fast breeder reactor technology to achieve a once-through-and-store reactor fuel cycle. Since the fuel cycle for this reactor is intended to use only natural or depleted uranium for its equilibrium feed, the resultant reactor would have excellent anti-proliferation characteristics. It would also extend utilization of natural uranium resources by a factor of about 15 relative to LWR reactors when on its equilibrium fuel cycle; startup requirements would of course reduce this factor.
Date: January 1, 1979
Creator: Fischer, G.J.; Kouts, H.J.C.; Cerbone, R.J.; Shenoy, S.; Durston, C.; Ludewig, H. et al.
Partner: UNT Libraries Government Documents Department

Relative consistency of ENDF/B-IV and -V with fast-reactor benchmarks

Description: The consistency of eleven selected fast-reactor and two neutron-field integral experiments with the ENDF/B-IV and -V libraries was examined by considering contributions to chi-square. Integral measurements least consistent with each given library were identified one at a time, and the particular cross sections which were significantly adjusted at each step were also identified. The results of this analysis demonstrate that, with respect to twelve out of the thirteen integral measurements considered, ENDF/B-V is a marked improvement over Version IV in the sense that (a) the integral data are significantly more consistent with Version V than with Version IV, (b) fewer cross sections need to be adjusted to achieve agreement between all the experimental data and the corresponding values calculated by Version V, and (c) the necessary adjustments are smaller.
Date: January 1, 1979
Creator: Yeivin, Y.; Wagschal, J.J.; Marable, J.H. & Weisbin, C.R.
Partner: UNT Libraries Government Documents Department

Fast mixed spectrum reactor concept

Description: The Fast Mixed Spectrum Reactor is a highly promising concept for a fast reactor with improved features of proliferation resistance, and excellent utilization of uranium resources. In technology, it can be considered to be a branch of fast breeder development, though its operation and implications are different from those of FBR'S in important respects. Successful development programs are required in several areas to bring FMSR to reality, but the payoff from a successful program can be high.
Date: April 1, 1979
Creator: Kouts, H.J.C.; Fischer, G.J. & Cerbone, R.J.
Partner: UNT Libraries Government Documents Department

Experiences with fast breeder reactor education in laboratory and short course settings

Description: The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented.
Date: January 17, 1983
Creator: Waltar, A.E.
Partner: UNT Libraries Government Documents Department

Physics constraints on the design of fast reactor safety test facilities

Description: This paper discusses the physics foundations common to all fast reactor safety test facilities and the constraints which they impose on the design. While detailed design discussions are confined to the experience with six ANL designs, available data from other designs are used to confirm the validity of the considerations and to broaden the scope of the discussion. This helps to view the various designs as a unified effort, to define their potential capabilities, and to assess how they could best complement each other.
Date: January 1, 1976
Creator: Travelli, A.; Meneghetti, D.; Matos, J.; Snelgrove, J.; Shaftman, D. H.; Tzanos, C. et al.
Partner: UNT Libraries Government Documents Department

Integral data for fast reactors

Description: Requirements at Argonne National Laboratory to establish the best estimates and uncertainties for LMR design parameters have lead to an extensive evaluation of the available critical experiment database. Emphasis has been put upon selection of a wide range of cores, including both benchmark, assemblies covering a range of spectra and compositions and power reactor mock-up assemblies with diverse measured parameters. The integral measurements have been revised, where necessary, using the most recent reference data and a covariance matrix constructed. A sensitivity database has been calculated, embracing all parameters, which enables quantification of the relevance of the integral data to parameters calculated with ENDF/B-V.2 cross sections.
Date: January 1, 1988
Creator: Collins, P. J.; Poenitz, W. P. & McFarlane, H. F.
Partner: UNT Libraries Government Documents Department

Advances in fast reactor sensitivity and uncertainty analysis

Description: A review of present methods and existing computer codes indicates an enormous capability not only to calculate sensitivity coefficients but also to apply them to a variety of purposes. However, there are still many limitations to our present capabilities. One of these limitations has been our inability to include in a complete and systematic way the effect of methods uncertainties on the determination of adjusted data, which depends, in general, not only on experimental measurements, but also on estimates of covariances associated with the measurements and the methods. Also, the uncertainty in adjusted data contains contributions from uncertainties in covariance estimates which contributions we have heretofore neglected. A new and comprehensive approach to include effects of methods uncertainties is presented here, and all sources which contribute to the uncertainty of the adjusted data are considered. This new approach is demonstrated using rough estimates for the methods uncertainties as applied to a simplified representation of the ZPR-6/7 fast benchmark. The results indicate that it may be essential to include methods uncertainties if integral experiments are to be used for the creation of adjusted nuclear data libraries. A careful evaluation of methods bias and uncertainties must still be performed.
Date: January 1, 1978
Creator: Marable, J.H. & Weisbin, C.R.
Partner: UNT Libraries Government Documents Department

Resonance region neutronics of unit cells in fast and thermal reactors

Description: A method has been developed for generating resonance-self-shielded cross sections based upon an improved equivalence theorem, which appears to allow extension of the self-shielding-factor (Bondarenko f-factor) method, now mainly applied to fast reactors, to thermal reactors as well. The method is based on the use of simple prescriptions for the ratio of coolant-to-fuel region-averaged fluxes, in the equations defining cell averaged cross sections. Linearization of the dependence of these functions on absorber optical thickness is found to be a necessary and sufficient condition for the existence of an equivalence theorem. Results are given for cylindrical, spherical and slab geometries. The functional form of the flux ratio relations is developed from theoretical considerations, but some of the parameters are adjusted to force-fit numerical results. Good agreement over the entire range of fuel and coolant optical thicknesses is demonstrated with numerical results calculated using the ANISN program in the S/sub 8/P/sub 1/ option.
Date: May 1, 1977
Creator: Salehi, A.A.; Driscoll, M.J. & Deutsch, O.L.
Partner: UNT Libraries Government Documents Department

State-of-the-art surveys on sodium-water reaction products cleanup methods and equipment. [LMFBR]

Description: This report describes the basic cleaning methods and equipment which have been used to clean and requalify specimens and components that have been exposed to sodium. Data obtained from laboratory cleaning of test specimens which were earlier exposed to sodium, have been included for various sodium removal methods. Development programs on cleanup methods for removing sodium-water reaction products along with sodium from surfaces and the purification of Intermediate Heat Transport System (IHTS) sodium after emergency dump have been identified.
Date: January 1, 1975
Creator: Anand, R.K.
Partner: UNT Libraries Government Documents Department

Cost-competitive, inherently safe LMFBR pool plant

Description: The Cost-Competitive, Inherently Safe LMFBR Pool Plant design was prepared in GFY 1983 as a joint effort by Rockwell International and the Argonne National Laboratory with major contributions from the Bechtel Group, Inc.; Combustion Engineering, Inc.; the Chicago Bridge and Iron Company; and the General Electric Company. Using current LMFBR technology, many innovative features were developed and incorporated into the design to meet the ultimate objectives of the Breeder Program, i.e., energy costs competitive with LWRs and inherent safety features to maintain the plant in a safe condition following assumed accidents without requiring operator action. This paper provides a description of the principal features that were incorporated into the design to achieve low cost and inherent safety.
Date: January 1, 1984
Creator: McDonald, J.S.; Brunings, J.E.; Chang, Y.I.; Seidensticker, R.W. & Hren, R.R.
Partner: UNT Libraries Government Documents Department

Seismic criteria studies and analyses. Quarterly progress report No. 3. [LMFBR]

Description: Information is presented concerning the extent to which vibratory motions at the subsurface foundation level might differ from motions at the ground surface and the effects of the various subsurface materials on the overall Clinch River Breeder Reactor site response; seismic analyses of LMFBR type reactors to establish analytical procedures for predicting structure stresses and deformations; and aspects of the current technology regarding the representation of energy losses in nuclear power plants as equivalent viscous damping.
Date: January 3, 1975
Partner: UNT Libraries Government Documents Department

U. S. fast reactor materials and structures program

Description: The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program.
Date: January 1, 1984
Creator: Harms, W.O. & Purdy, C.M.
Partner: UNT Libraries Government Documents Department

Influence of core-assembly refueling requirements on LMFBR core-system design

Description: Liquid metal fast breeder reactor (LMFBR) core assemblies are exposed to an operational environment which induces permanent distortions in their main structural members. These distortions have a substantial impact on core assembly refueling since the distortions are large compared to the available spaces. Core assembly refueling requirements demand that refueling be accomplished without damage to the adjacent core components or to the refueling equipment. This paper describes the core assembly refueling requirements and the design procedures used to demonstrate compliance with the requirements. This paper also provides an assessment relative to the influence of these requirements on LMFBR core system design.
Date: March 1, 1979
Creator: Fox, J.N.
Partner: UNT Libraries Government Documents Department

Liquid Metal Fast Breeder Reactors: a bibliography

Description: This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.
Date: November 1, 1980
Creator: Raleigh, H.D. (ed.)
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
Partner: UNT Libraries Government Documents Department

Liquid Metal Fast Breeder Reactors: a bibliography

Description: This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.
Date: November 1, 1980
Creator: Raleigh, H.D. (ed.)
Partner: UNT Libraries Government Documents Department

Status of SACRD: a data base for fast reactor safety computer codes

Description: In 1975 work was initiated to provide a central computerized data collection of evaluated data for use in fast reactor safety computer codes. This data base is called SACRD and is intended to encompass handbook and other nonproblem-dependent data related to LMFBR's, especially at extreme conditions where little or no experimental data are available. Version 1 of the data base was released in the latter part of 1978 and remained the standard version until Version 81, which was released in October 1981.
Date: January 1, 1982
Creator: Greene, N.M.; Flanagan, G.F. & Alter, H.
Partner: UNT Libraries Government Documents Department

Method for reliability analysis of complex reactor systems. [LMFBR]

Description: A method and a computer code for efficient and accurate reliability analyses of complex reactor systems are described and illustrated through an example. The method permits realistic analyses through its ability to accurately model and evaluate instantaneous and average unavailabilities for large systems with dependencies. The component models can include continuously monitored, non-repairable, and periodically tested components which are subject to failures resulting from components which are subject to failures resulting from component demands, stand-by conditions, human errors associated with testing and repair, as well as failures during actual operation. The numerical process used is efficient and allows analysis of general system configurations with arbitrary scheduling of maintenance operations.
Date: January 1, 1982
Creator: Elerath, J.G.; Vaurio, J.K. & Wood, A.P.
Partner: UNT Libraries Government Documents Department

Benefits of vertical and horizontal seismic isolation for LMR (liquid metal reactor) nuclear reactor units

Description: Seismic isolation has been shown to be able to reduce transmitted seismic force and lower response accelerations of a structure. When applied to nuclear reactors, it will minimize seismic influence on the reactor design and provide a design which is less site dependent. In liquid metal reactors where components are virtually at atmospheric pressure but under severe thermal conditions, thin-walled structures are generally used for primary systems. Thin-walled structures, however, have little inherent seismic resistance. The concept of seismic isolation therefore offers a viable and effective approach that permits the reactor structures to better withstand thermal and seismic loadings simultaneously. The majority of published work on seismic isolation deals with use of horizontal isolation system only. In this investigation, however, local vertical isolation is also provided for the primary system. Such local vertical isolation is found to result in significant benefits for major massive components, such as the reactor cover, designed to withstand vertical motions and loadings. Preliminary estimations on commodity savings of the primary system show that, with additional local vertical isolation, the savings could be twice that estimated for horizontal isolation only. The degree of effectiveness of vertical isolation depends on the diameter of the reactor vessel. As the reactor vessel diameter increases, the vertical seismic effects become more pronounced and vertical isolation can make a significant contribution.
Date: January 1, 1988
Creator: Wu, Ting-shu; Chang, Y.W. & Seidensticker, R.W.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY. Design and development quality assurance requirements for the FFTF

Description: The document is presented to provide general management requirements for Pacific Northwest Laboratory (PNL) and contractor design and development quality assurance programs to assure the required quality level of the various items required for the FFTF. The document is applicable as imposed by the contract to FFTF contractors and subcontractors. The document is also applicable to PNL design and development activities related to the FFTF.
Date: October 23, 1968
Creator: Albert, W.G.
Partner: UNT Libraries Government Documents Department

Computational procedures for multidimensional core analysis

Description: Comparisons made by the Large Core Code Evaluation Working Group (LCCEWG) show that excellent agreement is obtained by different diffusion theory codes run on different computers at different installations in computing representative models of large fast reactors. It is noted that finite difference diffusion theory computer codes are still not fast enough to run the largest problems, but the coarse-mesh nodal expansion methods now seem capable not only of providing accurate global results in very short computing times, but also of recovering consistent local flux shapes. A new multigrid method due to Brandt gives promise of very effective acceleration of diffusion equation iterations and also makes possible adaptive mesh solution procedures. The latest version of the discrete ordinates code DOT-IV allows energy and spatial variation of the discrete ordinates quadrature order as well as a form of line deletion. The three-dimensional discrete ordinates code THREETRAN (hex,.z) has been successfully used to analyze problems representative of fast reactor systems during the approach-to-critical loading phase. New discrete ordinates difference schemes are described that include a more accurate, positive weighed diamond scheme, a positive geometric mean scheme, and the analog of the diamond scheme for triangular meshes.
Date: January 1, 1978
Creator: Lathrop, K.D.
Partner: UNT Libraries Government Documents Department