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Errors associated with standard nodal diffusion methods as applied to mixed oxide fuel problems

Description: The evaluation of the disposition of plutonium using light water reactors is receiving increased attention. However, mixed-oxide (MOX) fuel assemblies possess much higher absorption and fission cross- sections when compared to standard UO2 assemblies. Those properties yield very high thermal flux gradients at the interfaces between MOX and UO2 assemblies. It has already been reported that standard flux reconstruction methods (that recover the homogeneous intranodal flux shape using the converged nodal solution) yield large errors in the presence of MOX assemblies. In an accompanying paper, we compare diffusion and simplified PN calculations of a mixed-oxide benchmark problem to a reference transport calculation. In this paper, we examine the errors associated with standard nodal diffusion methods when applied to the same benchmark problem. Our results show that a large portion of the error is associated with the quadratic leakage approximation (QLA) that is commonly used in the standard nodal codes.
Date: July 24, 1998
Creator: Brantley, P. S., LLNL
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

Analysis of Hydrogen Depletion Using a Scaled Passive Autocatalytic Recombiner

Description: Hydrogen depletion tests of a scaled passive autocatalytic recombine (pAR) were performed in the Surtsey test vessel at Sandia National Laboratories (SNL). The experiments were used to determine the hydrogen depletion rate of a PAR in the presence of steam and also to evaluate the effect of scale (number of cartridges) on the PAR performance at both low and high hydrogen concentrations.
Date: October 28, 1998
Creator: Blanchat, T.K. & Malliakos, A.
Partner: UNT Libraries Government Documents Department

Uncertainties in the analysis of plutonium fueled light water moderated assemblies

Description: A theoretical analysis of UO/sub 2/-- PuO/sub 2/ fueled, light-water- moderated lattice experiments has been performed to aid in establishing technical bases and design criteria for the utilization of plutonium bearing fuel in thermal power reactors. Results for UO/sub 2/ and Al-- Pu lattices are included in order to understand the effects due to uranium and plutonium separately. The problems involved in calculating high leakage critical experiments are discussed. Estimates of the effects of various approximations inherent in the theoretical methods and/or analysis procedures are included along with the consequence on the results of the correlation. Uncertainties in the measurements and the neutron crosssection data are related to uncertainties in the calculated values K/sub eff/ .The results of other studies which bear on evaluating the calculational methods are summarized. Areas which should be investigated in future analyses are also identified. (111 references) (auth)
Date: May 1, 1973
Creator: Liikala, R.C.; Uotinen, V.O. & Jenquin, U.P.
Partner: UNT Libraries Government Documents Department

{sup 16}O neutron cross section evaluation

Description: This work has resulted from a need to compute more accurately the neutron scattering cross sections and angular distributions for {sup 16}O. Several oxygen evaluations have been performed in the past with R-Matrix theory, including ENDF/B-V and ENDF/B-VI. ENDF/B-VI is an improvement over ENDF/B-V, but still underpredicts in general the forward scattering of neutrons below 2.5 MeV. R-Matrix theory is used in describing cross sections at and near the resonance energies; but may not always be adequate in describing cross sections between resonances, especially when they are widely spaced. The optical (potential well) model of the nucleus is very good in representing cross sections that vary smoothly with energy, but not at describing all of the detailed resonance cross sections. A combination of the potential well model and R-Matrix theory was used for this work to represent cross sections with isolated resonances with large spacings between them. The total neutron cross section of oxygen-16 below 3.0 MeV has widely separated resonances and a dip in the cross section at 2.35 MeV. In the vicinity of resonances, where cross sections vary rapidly with energy, R-Matrix theory has been successful in fitting experimental data. In the region between resonances, an analytical procedure with physical basis is needed that agrees with data over a wide range of energies bracketing regions where experimental measurements are lacking.
Date: June 1, 1998
Creator: Caro, E.
Partner: UNT Libraries Government Documents Department

Nuclear energy and materials in the 21st century

Description: The Global Nuclear Vision Project at the Los Alamos National Laboratory is examining a range of long-term nuclear energy futures as well as exploring and assessing optimal nuclear fuel-cycle and material strategies. An established global energy, economics, environmental (E{sup 3}) model has been adopted and modified with a simplified, but comprehensive and multi-regional, nuclear energy module. Consistent nuclear energy scenarios are constructed, where future demands for nuclear power are projected in price competition with other energy sources under a wide range of long-term ({approx}2100) demographic, economic, policy, and technological drivers. A spectrum of futures is examined at two levels in a hierarchy of scenario attributes in which drivers are either external or internal to nuclear energy. The result reported examine departures from a basis scenario and are presented in the following order of increasing specificity: (a) definition and parametric variations of the basis scenario; (b) comparison of the basis scenario with other recent studies; (c) parametric studies that vary upper-level hierarchical scenario attributes (external drivers); and (d) variations of the lower-level scenario attributes (internal drivers). Impacts of a range of nuclear fuel-cycle scenarios are reflected back to the higher-level scenario attributes that characterize particular nuclear energy scenarios. Special attention is given to the role of nuclear materials inventories (in magnitude, location, and form) and their contribution to the long-term sustainability of nuclear energy, the future competitiveness of both conventional and advanced nuclear reactors, and proliferation risk.
Date: May 1, 1997
Creator: Krakowski, R.A.; Davidson, J.W. & Bathke, C.G.
Partner: UNT Libraries Government Documents Department

Design Parameters for a Natural Uranium UO{sub 3} or U{sub 3}O{sub 8} Fueled Nuclear Reactor

Description: A recent Oak Ridge National Laboratory report provided preliminary analyses to propose alternative design parameters for a nuclear reactor that could be fueled with natural UO{sub 3} or U{sub 3}O{sub 8} and moderated with either heavy water or reactor-grade graphite. This report provides more specific reactor design and operating parameters for a heavy water-moderated reactor only. The basic assumptions and analytical approach are discussed together with the results of the analysis.
Date: November 15, 2002
Creator: Hopper, C.M.
Partner: UNT Libraries Government Documents Department

Shielding methods for evaluating the versatility of proposed shipping casks

Description: After a shipping cask has been designed for a certain type of spent fuel, the number of assemblies it can carry is fixed, as are the thicknesses of the various steel shells, the neutron shield, and the gamma shield. The question then becomes ''What other types of spent fuel may be shipped in the cask.'' Using the same neutron and gamma source terms, miscellaneous correlations, and one-group cross sections found in the CAPSIZE program, a fast new interactive shielding program called KWIKDOSE has been written for the IBM-PC which computes and displays a 2-D table showing the total dose rate ten feet from the centerline of a cask, as a function of the spent fuel's burnup and cooling time.
Date: January 1, 1987
Creator: Bucholz, J.A.
Partner: UNT Libraries Government Documents Department

ASME N510 test results for Savannah River Site AACS filter compartments

Description: The K-Reactor at the Savannah River Site recently implemented design improvements for the Airborne Activity Confinement System (AACS) by procuring, installing, and testing new Air Cleaning Units, or filter compartments, to ASME AG-1, N509, and N510 requirements. Specifically, these new units provide documentable seismic resistance to a Design, Basis Accident earthquake, provide 2 in. adsorber beds with 0.25 second residence time, and meet all AG-1, N509, and N510 requirements for testability and maintainability. This paper presents the results of the Site acceptance testing and discusses an issue associated with sample manifold qualification testing.
Date: July 1, 1994
Creator: Paul, J. D. & Punch, T. M.
Partner: UNT Libraries Government Documents Department

Startup data report for NRC/PNL Halden Assembly IFA-513

Description: This report presents data from the first month of operation of IFA-513, which is a heavily instrumented 6-rod test assembly in the Halden Reactor in Norway. The assembly is jointly sponsored by the Halden Project and the Nuclear Regulatory Commission (NRC), and is part of a series of irradiation tests sponsored by the NRC to verify its single-rod fuel modeling computer programs. All the rods in the series are of the basic BWR-6 design with variations in gap size, fuel type, fill gas composition, and fill gas pressure. The first two tests in the series were IFA-431 and IFA-432. These were identical 6-rod assemblies, each containing the same variations of gap size and fuel pellet types, but operating at different power levels and burnups. The present assembly, IFA-513, is the third in the series; its 6 rods are all identical, except for variations in fill gas composition and pressure. The fourth and last assembly, designated IFA-527, is yet to be built, and will study the effects of fuel pellet cracking and relocation. The measurements made in IFA-513 and the earlier tests include: (1) fuel temperature and power (both steady-state and transient), (2) total cladding elongation, and (3) fill gas pressure. The measurements were made on a continuous basis, providing a record of their variation with both power and burnup. Along with the data, this report includes some analysis to put the IFA-513 startup data in perspective to similar data from IFA-431 and IFA-432.
Date: July 1, 1979
Creator: Lanning, D.D. & Cunningham, M.E.
Partner: UNT Libraries Government Documents Department

Monte Carlo analyses of simple U233 O/sub 2/-ThO/sub 2/ and U235 O/sub 2/-ThO/sub 2/ lattices with ENDF/B-IV data (AWBA development program)

Description: A number of water-moderated Th-U235 and Th-U233 lattice integral experiments were analyzed in a consistent manner, with ENDF/B-IV data and detailed Monte Carlo methods. These experiments provide a consistent test of the nuclear data. The ENDF/B-IV data are found to perform reasonably well. Adequate agreement is found with integral measurements of thorium capture. Calculated K/sub eff/ values show a generally coherent pattern which is consistent with K/sub eff/ results obtained for homogeneous aqueous critical assemblies. Harder prompt fission spectra for U233 and U235 can correct the principal discrepancy observed with ENDF/B-IV, a bias trend in K/sub eff/ attributed to an underprediction of leakage.
Date: September 1, 1980
Creator: Hardy, J. Jr. & Ullo, J.J.
Partner: UNT Libraries Government Documents Department

Plant systems/components modularization study. Final report. [PWR]

Description: The final results are summarized of a Plant Systems/Components Modularization Study based on Stone and Webster's Pressurized Water Reactor Reference Design. The program has been modified to include evaluation of the most promising areas for modular consideration based on the level of the Sundesert Project engineering design completion and the feasibility of their incorporation into the plant construction effort.
Date: July 1, 1977
Partner: UNT Libraries Government Documents Department

Critical experiments supporting close proximity water storage of power reactor fuel. Technical progress report, October 1-December 31, 1978

Description: Experimental measurements are being taken on critical configurations of clusters of fuel rods mocking up LWR-type fuel elements in close proximity water storage. The results will serve to benchmark the computer codes used in designing nuclear power reactor fuel storage racks.
Date: March 1, 1979
Creator: Baldwin, M.N.; Hoovler, G.S.; Eng, R.L. & Welfare, F.G.
Partner: UNT Libraries Government Documents Department

Army Gas-Cooled Reactor Systems program: alternator final design report

Description: The development and testing of a demonstration brushless alternator for the ML-1 mobile nuclear power plant is described. The brushless concept was selected after it became apparent that a conventional power generator could not satisfy the ML-1 weight and size requirements. The demonstration alternator fabricated and tested under this program did not meet all performance specifications; the efficiency was low and the unit could not be operated for significant periods of time without overheating. However, a large body of useful data was accumulated during the extensive development program. Of special interest are data on the rotor and stator design, the cooling requirements and on the distribution of eddy current losses. Analysis of the data indicates that a brushless alternator, only slightly larger and heavier than was specified for the ML-1, could be developed with a modest additional effort.
Date: June 1, 1964
Partner: UNT Libraries Government Documents Department

Limit cycles and bifurcations in nuclear systems

Description: This work provides a basis for scoping calculations to determine the dynamic behavior - both linear and nonlinear - of BWRs. Additional work is now underway to establish the feasibility of routine operation of nuclear systems in the nonlinear (limit-cycle) regime.
Date: January 1, 1986
Creator: Cacuci, D.G.; March-Leuba, J. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

Determination of statistically based design limits associated with engineering models. (LWBR Development Program)

Description: This report provides a usable reference of methods and procedures for the construction of both one-sided and two-sided ..gamma../P statistical tolerance limits for design application to both linear and nonlinear models in any number of variables.
Date: February 1, 1980
Creator: Ginsburg, H.
Partner: UNT Libraries Government Documents Department

Fuel performance improvement program: description and characterization of HBWR Series H-2, H-3, and H-4 test rods

Description: The fabrication process and as-built characteristics of the HBWR Series H-2 and H-3 test rods, as well as the three packed-particle (sphere-pac) rods in HBWR Series H-4 are described. The HBWR Series H-2, H-3, and H-4 tests are part of the irradiation test program of the Fuel Performance Improvement Program. Fifteen rods were fabricated for the three test series. Rod designs include: (1) a reference dished pellet design incorporating chamfered edges, (2) a chamfered, annular pellet design combined with graphite-coated cladding, and (3) a sphere-pac design. Both the annular-coated and sphere-pac designs include internal pressurization using helium.
Date: March 1, 1980
Creator: Guenther, R.J.; Barner, J.O. & Welty, R.K.
Partner: UNT Libraries Government Documents Department

Fuel Performance Improvement Program. Quarterly progress report, April-June 1979

Description: Two series of test rods are under irradiation in the Halden Boiling Water Reactor (HBWR Series H-1 and Series H-4). Fuel rods for Series H-2 and H-3 have been fabricated and delivered to Halden. Plans for the first series of demonstration fuel assemblies for irradiation in the Big Rock Point Reactor have been modified to substitute two BRPR Series S-2 assemblies containing segmented rods, some with sphere-pac fuel, for Series S-1 assemblies that will be irradiated later. These assemblies are in the current BRPR reload and reactor startup is scheduled for September-October. An effective method for applying graphite coating to the inside surface of cladding tubes has been demonstrated. Rod loading procedures for sphere-pac fuel that assure a uniform axial fuel density have been developed. Special hardware has been designed and tested that assures the spherical fuel can be retained in the fuel column during loading and handling operations. The measured centerline fuel temperature in the pressurized sphere-pac fuel rod (HBWR Series H-4) is lower (approx. 145/sup 0/C) than for the reference pellet fuel rod operating at comparable linear heat ratings.
Date: July 1, 1979
Creator: Crouthamel, C.E. (comp.)
Partner: UNT Libraries Government Documents Department

Evaluation of tight-pitch PWR cores

Description: The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - /sup 235/U/UO/sub 2/ : Pu/ThO/sub 2/ : /sup 233/U/ThO/sub 2/ - and the conventional recycle-mode uranium system - /sup 235/U/UO/sub 2/ : Pu/UO/sub 2/. The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) < F/M < 4.0 are limited by the scarcity of experiments with F/M > 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments.
Date: August 1, 1979
Creator: Correa, F.; Driscoll, M.J. & Lanning, D.D.
Partner: UNT Libraries Government Documents Department