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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

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

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

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

Experimental basis for parameters contributing to energy dissipation in piping systems

Description: The paper reviews several pipe testing programs to suggest the phenomena causing energy dissipation in piping systems. Such phenomena include material damping, plasticity, collision in gaps and between pipes, water dynamics, insulation straining, coupling slippage, restraints (snubbers, struts, etc.), and pipe/structure interaction. These observations are supported by a large experimental data base. Data are available from in-situ and laboratory tests (pipe diameters up to about 20 inches, response levels from milli-g's to responses causing yielding, and from excitation wave forms including sinusoid, snapback, random, and seismic). A variety of pipe configurations have been tested, including simple, bare, straight sections and complex lines with bends, snubbers, struts, and insulation. Tests have been performed with and without water and at zero to operating pressure. Both light water reactor and LMFBR piping have been tested.
Date: January 1, 1985
Creator: Ibanez, P. & Ware, A.G.
Partner: UNT Libraries Government Documents Department

Ex-vessel core catcher design requirements and preliminary concepts evaluation. [LMFBR]

Description: As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant.
Date: June 14, 1974
Creator: Friedland, A.J. & Tilbrook, R.W.
Partner: UNT Libraries Government Documents Department

COVERX service module of the FORSS system. [LMFBR]

Description: The COVERX Service Module includes seven execution paths to aid in understanding and using multigroup cross-section covariance matrices contained in the standard interface file COVERX. The execution paths provide the following operations on COVERX file(s): list the contents of a COVERX file; allow adding new multigroup cross-section covariance matrices to an existing COVERX file; allow deletion of multigroup covariance matrices from an existing COVERX file; merge two COVERX files and creates a new file; change the mode of a file from unformatted to formatted and conversely; allow modification of the records contained in a COVERX file; and selectively edits or copies a file.
Date: April 1, 1980
Creator: Drischler, J.D.
Partner: UNT Libraries Government Documents Department

Core design and optimization of high performance low sodium void 1000 MWe heterogeneous oxide LMFBR cores

Description: Radially heterogeneous core configurations are effective means to reduce sodium void reactivity. In general, radially heterogeneous cores can be designed as tightly or loosely coupled cores with center core or center blanket arrangements. Core height, number of core regions and number of fuel pins per assembly are additional variables in an optimization of basic heterogeneous core configurations. An extensive study was carried out to optimize the core configurations for 1000 MWe LMFBRs. All cores were subject to a common set of nuclear, mechanical, and thermal-hydraulic design assumptions. They were restrained by an upper sodium void reactivity limit of $2.50 and a doubling time of approximately 15 to 18 years. The screening and optimization procedures employed lead to two core layouts which were both tightly coupled. A complete nuclear analysis of these two cores (derived from a loosely coupled configuration/derived from a tightly coupled configuration) determined the fissile inventories (4268.4/4213.4 kg at BOEC), burnups (83.90/100.7 MWd/t peak), reactivity swings (0.49/1.8% ..delta..k total), power and flux distributions for different control insertion patterns, the breeding performance (15.7/15.3 yrs CSDT), the safety parameters, such as sodium void reactivity ($2.38/$2.23 at EOEC), isothermal Doppler coefficients for both sodium-in (45.6/46.1 T dk/dT x 10/sup -4/ core at EOEC) and sodium-out conditions (28.6/28.2 T dk/dT x 10/sup -4/ core at EOEC), and the transient behavior which shows very little space-dependence during a 60 cent reactivity step insertion.
Date: January 1, 1979
Creator: Barthold, W.P.; Orechwa, Y.; Su, S.F.; Beitel, J.C.; Turski, R.; Lam, P.S.K. et al.
Partner: UNT Libraries Government Documents Department

Clinch River Breeder Reactor Plant Project: construction schedule

Description: The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule.
Date: January 1, 1982
Creator: Purcell, W.J.; Martin, E.M. & Shivley, J.M.
Partner: UNT Libraries Government Documents Department

Concept and preliminary design of double tubesheet connector region for design and steady-state conditions. [LMFBR]

Description: In this analysis the structural integrity of the double tubesheet connector for the LMFBR demonstration plant steam generator is investigated for design and steady state conditions. Although the evaporator and superheater are to be interchangeable, only the steam outlet end of the evaporator is investigated. This was selected because the mean temperature differences between the tubesheets, tubes and connector are largest here at steady state, thus yielding the highest thermal loads. Combined with the thermal loads are superheater pressures in order to also use the highest mechanical loads. Although this is a conservative approach for the evaporator, the superheater must be analyzed separately at a later date in order to assure interchangeability. Evaluation of the results is based upon ASME Code Case 1331-8.
Date: June 1, 1974
Creator: Rinne, W.A.
Partner: UNT Libraries Government Documents Department

Analysis of SBTF quadelliptical furnaces. [LMFBR]

Description: A computer model was developed which predicts the axial temperature profile and heat flux at the outer surface of the test section of the Sodium Boiling Test Facility constructed by the Engineering Technology Division at ORNL. The model was in agreement with observed temperature profiles at furnace power levels representative of single phase, dual phase, and dry-out operations. A parametric study demonstrated the effect of sodium flow rate and surface emissivities on the predicted temperature profile. It was concluded that axial conduction in the Hastelloy tube and sodium must be incorporated into the model to improve accuracy.
Date: February 23, 1979
Creator: Anderson, F.E. & Schulz, R.B.
Partner: UNT Libraries Government Documents Department

Analysis of fission-product effects in a Fast Mixed-Spectrum Reactor concept

Description: The Fast Mixed-Spectrum Reactor (FMSR) concept has been proposed by BNL as a means of alleviating certain nonproliferation concerns relating to civilian nuclear power. This breeder reactor concept has been tailored to operate on natural uranium feed (after initial startup), thus eliminating the need for fuel reprocessing. The fissile material required for criticality is produced, in situ, from the fertile feed material. This process requires that large burnup and fluence levels be achievable, which, in turn, necessarily implies that large fission-product inventories will exist in the reactor. It was the purpose of this study to investigate the effects of large fission-product inventories and to analyze the effect of burnup on fission-product nuclide distributions and effective cross sections. In addition, BNL requested that a representative 50-group fission-product library be generated for use in FMSR design calculations.
Date: February 1, 1980
Creator: White, J.R. & Burns, T.J.
Partner: UNT Libraries Government Documents Department

Analytical approach for confirming the achievement of LMFBR reliability goals

Description: The approach, recommended by GE-ARSD, for confirming the achievement of LMFBR reliability goals relies upon a comprehensive understanding of the physical and operational characteristics of the system and the environments to which the system will be subjected during its operational life. This kind of understanding is required for an approach based on system hardware testing or analyses, as recommended in this report. However, for a system as complex and expensive as the LMFBR, an approach which relies primarily on system hardware testing would be prohibitive both in cost and time to obtain the required system reliability test information. By using an analytical approach, results of tests (reliability and functional) at a low level within the specific system of interest, as well as results from other similar systems can be used to form the data base for confirming the achievement of the system reliability goals. This data, along with information relating to the design characteristics and operating environments of the specific system, will be used in the assessment of the system's reliability.
Date: September 30, 1981
Creator: Ingram, G.E.; Elerath, J.G. & Wood, A.P.
Partner: UNT Libraries Government Documents Department

Document control and information retrieval system for the Fast Flux Test Facility (FFTF)

Description: A description is given of the FFTF Document Control and Information Retrieval System. The system utilizes a mini-computer along with various microfilm equipment and is designed to accommodate an anticipated 50 million pages of text and 750,000 drawings. The system is simple, uncluttered, eliminates duplication, and provides quick retrievability of documents for all technical and administrative personnel.
Date: March 1, 1976
Creator: Theo, M.G.
Partner: UNT Libraries Government Documents Department

26 - LMFBR flexible pipe joint development

Description: Objective is the qualification of a PLBR-size primary loop flexible piping joint to the ASME Band PVC rules. Progress and activities are reported for: Class 1 flexible joint code approval support, engineering and design, material development, component testing, and manufacturing development. (DLC)
Date: May 1, 1978
Creator: Anderson, R.V.
Partner: UNT Libraries Government Documents Department

Producing a unified progress report with inputs from several contractors

Description: The project management organization in which the author works produces an annual technical progress report for the Clinch River Breeder Reactor Plant Project. The report has to be integrated and edited from the inputs of six major Project participants scattered from coast to coast. The integrated report manuscript then has to be submitted for two formal reviews, and the report must be published in a readable and attractive form. Accomplishing those steps in a reasonable length of time, with a high degree of accuracy, and at minimum expense requires careful planning and close supervision. Planning includes scheduling in such a way as to perform operations in parallel, where possible, instead of in series. Exploiting the capabilities of word processing saves much keyboarding and proofreading time. Art from previous reports is reused when possible. Many of these methods can be applied to other reports that require integration and editing of material from several sources.
Date: January 1, 1981
Creator: Nelson, O.A.
Partner: UNT Libraries Government Documents Department

Evaluation of the magnitude and effects of bundle duct interaction in fuel assemblies at developmental plant conditions

Description: Purpose of this evaluation is to estimate the magnitude and effects of irradiation and creep induced fuel bundle deformations in the developmental plant. This report focuses on the trends of the results and the ability of present models to evaluate the assembly temperatures in the presence of bundle deformation. Although this analysis focuses on the developmental plant, the conclusions are applicable to LMFBR fuel assemblies in general if they have wire spacers.
Date: September 1, 1980
Creator: Serell, D.C. & Kaplan, S.
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