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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

HELIOS: applications at the Los Alamos National Laboratory

Description: The Los Alamos National Laboratory (LANL) is involved in the analysis of many different types of nuclear systems. The nuclear systems that we have analyzed have included subcritical accelerator driven systems for the transmutation of waste, fusion systems, critical experiment systems, and space propulsion and power systems. We have also analyzed special purpose reactors such as the LANL Omega West reactor, production reactors, and conventional commercial light- and heavy-water reactors. Thus the systems that we analyze and the type of results desired, often vary considerably from those of a power company normally analyzing their PWR or BWR for fissile fuel burnup and production. The reactor geometries that we model are often quite complicated such as those of an RBMK or Savannah River Production Reactor. Rather than fissile fuel production and burnup, the goal of a calculation could be the production rate of some obscure isotope which has medical applications.
Date: October 1, 1997
Creator: Perry, R.T.; Mosteller, R.D.; Chodak, Paul III; Charlton, W. & Adams, B.T.
Partner: UNT Libraries Government Documents Department

Observations of the boiling process from a downward-facing torispherical surface: Confirmatory testing of the heavy water new production reactor flooded cavity design

Description: Reactor-scale ex-vessel boiling experiments were performed in the CYBL facility at Sandia National Laboratories. The boiling flow pattern outside the RPV bottom head shows a center pulsating region and an outer steady two-phase boundary layer region. The local heat transfer data can be correlated in terms of a modified Rohsenow correlation.
Date: June 1, 1995
Creator: Chu, T.Y.; Bentz, J.H. & Simpson, R.B.
Partner: UNT Libraries Government Documents Department

Studies of alternative nuclear technologies

Description: This report is a summary of tasks performed for the U.S. Arms Control and Disarmament Agency under Contract AC7NC114. The work is directly related to the Agency effort to examine potential alternative fuel cycles that might enhance uranium resource utilization, minimize plutonium production, and reduce the weapons proliferation risk from spent fuel reprocessing or early introduction of fast breeder reactors. Reported herein are summaries of various inter-related task assignments, including: fuel utilization in current light water reactors operating with the uranium fuel cycle; alternate fuel cycles, including the use of denatured fuel in LWRs and of the spectral shift concept for reactivity control; fuel utilization in high temperature graphite moderated reactors using the denatured fuel cycle; fuel utilization in heavy water reactors (CANDU type), including the use of enriched fuel, denatured fuel, and recycle of plutonium and U-233; the tandem fuel cycle (recovery of spent fuel and further irradiation in a CANDU type reactor); issues in the utilization of denatured fuel in LWRs; preliminary conceptual evaluation of a heavy water moderated reactor suitable for use in the United States.
Date: April 1, 1978
Creator: Turner, S.E.; Gurley, M.K.; Kirby, K.D.; Mitchell, W. III & Roach, K.E.
Partner: UNT Libraries Government Documents Department

450/sup 0/F step transient thermal analysis of the LOFT pressurizer surge and spray line piping

Description: The LOFT pressurizer spray and surge line piping was analyzed for a 450/sup 0/F step change in fluid temperature. This transient was chosen to conservatively represent several pressurizer operating transients that had not previously been analyzed. These include temperature transients resulting from a 300/sup 0/F ..delta..T between pressurizer temperature and cold leg temperature, injection of a cooled (70/sup 0/F) slug of stagnant fluid into the hot (540/sup 0/F) spray line piping, and inflow of 100/sup 0/F primary coolant system water into the hot (480/sup 0/F) surge line piping.
Date: July 7, 1977
Creator: Tolan, B.J.
Partner: UNT Libraries Government Documents Department

Critical heat flux experiments in a circular tube with heavy water and light water. (AWBA Development Program)

Description: Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on the average, 8 percent below the light water CHF values.
Date: May 1, 1980
Creator: Williams, C.L. & Beus, S.G.
Partner: UNT Libraries Government Documents Department

Identification of neutron noise sources in a boiling water reactor

Description: Measurements were made at Units 2 and 3 of the TVA Browns Ferry nuclear power plant in order to characterize the neutron and process signal noise signatures, to determine the degree of correlation between selected pairs of signals, and to assess the usefulness of such signatures for monitoring and anomaly detection in BWR-4s. Measurements were made in a power plant during normal operation at full power to determine the usefulness of the neutron and process signals from sensors and instrumentation in the plant which have been contaminated by plant electrical noise interference. It is concluded that the signals derived from existing plant sensors and instrumentation could be used to diagnose anomalies. The neutron signals could be used to monitor the stability of the core and to diagnose anomalies involving the reactor pressure, core flow, and steam flow.
Date: January 1, 1977
Creator: Sides, W.H. Jr. & Mathis, M.V.
Partner: UNT Libraries Government Documents Department

Improved numerical techniques for processing Monte Carlo thermal scattering data

Description: As part of a Thermal Benchmark Validation Program sponsored by the Electric Power Research Institute (EPRI), the National Nuclear Data Center has been calculating thermal reactor lattices using the SAM-F Monte Carlo Computer Code. As part of this program a significant improvement has been made in the adequacy of the numerical procedures used to process the thermal differential scattering cross sections for hydrogen bound in H/sub 2/O.
Date: January 1, 1980
Creator: Schmidt, E & Rose, P
Partner: UNT Libraries Government Documents Department

LOFT DTT rake pin stress analysis

Description: A stress analysis of the 3/8-inch and 1/4-inch pins which hold the rake assembly to the flange was performed and shows stresses to be lower than the Class 1 allowables of the ASME Boiler and Pressure Vessel Code. The alternating pin stresses were found to be below the endurance limit and fatigue failure will not occur. The rake assembly was assumed to be loaded by steady drag and lift forces and alternating vortex shedding forces.
Date: January 23, 1979
Creator: Mosby, W.R.
Partner: UNT Libraries Government Documents Department

Fuel utilization improvements in a once-through PWR fuel cycle. Final report on Task 6

Description: In studying the position of the United States Department of Energy, Non-proliferation Alternative Systems Assessment Program, this report determines the uranium saving associated with various improvement concepts applicable to a once-through fuel cycle of a standard four-loop Westinghouse Pressurized Water Reactor. Increased discharged fuel burnup from 33,000 to 45,000 MWD/MTM could achieve a 12% U/sub 3/O/sub 8/ saving by 1990. Improved fuel management schemes combined with coastdown to 60% power, could result in U/sub 3/O/sub 8/ savings of 6%.
Date: June 1, 1979
Creator: Dabby, D.
Partner: UNT Libraries Government Documents Department

Finite deformation analysis of continuum structures with time dependent anisotropic elastic plastic material behavior (LWBR/AWBA Development Program)

Description: A finite element procedure is presented for finite deformation analysis of continuum structures with time-dependent anisotropic elastic-plastic material behavior. An updated Lagrangian formulation is used to describe the kinematics of deformation. Anisotropic constitutive relations are referred, at each material point, to a set of three mutually orthogonal axes which rotate as a unit with an angular velocity equal to the spin at the point. The time-history of the solution is generated by using a linear incremental procedure with residual force correction, along with an automatic time step control algorithm which chooses time step sizes to control the accuracy and numerical stability of the solution.
Date: March 1, 1980
Creator: Hutula, D.N.
Partner: UNT Libraries Government Documents Department

Piping-reliability analysis for pressurized-water-reactor feedwater lines

Description: This paper presents a piping reliability analysis for feedwater lines at five PWR plants; the analysis is based on probabilistic fracture mechanics. On the basis of observed pipe cracks in these feedwater lines, the crack is modeled with an initial semi-elliptical shape along the pipe inner circumference. Initial crack samples are generated using the Monte Carlo technique in conjunction with an importance sampling scheme. The fatigue model for crack growth employs a Paris-type growth-rate equation. Leak probabilities for feedwater lines of five PWR plants are estimated through the first ten years of the design life. Comparison of estimated leak probabilities and leak data in the five PWR plants led to the conclusion that the piping reliability model used in the present analysis can provide a reasonable estimate of the reliability of PWR feedwater lines.
Date: January 1, 1982
Creator: Woo, H.H. & Chou, C.K.
Partner: UNT Libraries Government Documents Department