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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

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

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

Response Matrix Method Development Program at Savannah River Laboratory

Description: The Response Matrix Method Development Program at Savannah River Laboratory (SRL) has concentrated on the development of an effective system of computer codes for the analysis of Savannah River Plant (SRP) reactors. The most significant contribution of this program to date has been the verification of the accuracy of diffusion theory codes as used for routine analysis of SRP reactor operation. This paper documents the two steps carried out in achieving this verification: confirmation of the accuracy of the response matrix technique through comparison with experiment and Monte Carlo calculations; and establishment of agreement between diffusion theory and response matrix codes in situations which realistically approximate actual operating conditions.
Date: January 1, 1976
Creator: Sicilian, J. M.
Partner: UNT Libraries Government Documents Department

Preliminary design concept of a subcritical reactor using available resources

Description: During the Fall 1993 semester, a project was initiated within the Nuclear Engineering Department of the University of Tennessee with the objective of developing a design for a subcritical reactor with maximized multiplication factor using materials currently available. Such a device, if constructed, would serve as a teaching tool for the Department of Nuclear Engineering. Design work was conducted as a large number of computer calculations, with trial pile configurations based on fundamental nuclear engineering principles, in an effort to maximize multiplication factor through fuel element geometry, moderator type, fissile/moderator ratio, and reflector character. The principal objective of the design group for the early phase of this project was to present several possible ``baseline`` reactor designs and identify directions for improvements. For the sake of calculational ease, the cores analyzes to date have been of nearly cubic shape. The SCALE CSAS25 software which runs KENO.Va, a Monte Carlo code, was used for all neutronics calculations. The baseline reactors resulting from work to date are cuboidal in shape and graphite reflected. Two types of fuel element geometries are proposed, a typical triangular pitch rod lattice and an arrangement of discrete fuel slugs placed in a lattice corresponding to body centered cubic packing. The latter arrangement provides slightly higher multiplication factors than the former. Calculations were performed for both graphite and heavy water moderation with heavy water moderation producing considerably higher multiplication factors, as expected. In general, the maximum k{sub eff} for the reactors are in the range of 0.5 to 0.9, well subcritical, except in the cases of the extreme possible values of fuel assay where critical configurations are possible. In these cases, designs with reduced fuel loading are recommended to assure a subcritical multiplication factor.
Date: December 31, 1993
Creator: Churnetski, E. L.; Hoyny, V.; Chaudhuri, B. R.; Taprantzis, A. & Yavas, A.
Partner: UNT Libraries Government Documents Department

Advanced neutron source reactor thermal-hydraulic test loop facility description

Description: The Thermal-Hydraulic Test Loop (THTL) is a facility for experiments constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory. The ANSR is both cooled and moderated by heavy water and uses uranium silicide fuel. The core is composed of two coaxial fuel-element annuli, each of different diameter. There are 684 parallel aluminum-clad fuel plates (252 in the inner-lower core and 432 in the outer-upper core) arranged in an involute geometry that effectively creates an array of thin rectangular flow channels. Both the fuel plates and the coolant channels are 1.27 mm thick, with a span of 87 mm (lower core), 70 mm (upper core), and 507-mm heated length. The coolant flows vertically upwards at a mass flux of 27 Mg/m{sup 2}s (inlet velocity of 25 m/s) with an inlet temperature of 45{degrees}C and inlet pressure of 3.2 MPa. The average and peak heat fluxes are approximately 6 and 12 MW/m{sup 2}, respectively. The availability of experimental data for both flow excursion (FE) and true critical heat flux (CHF) at the conditions applicable to the ANSR is very limited. The THTL was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of thermal limits under the expected ANSR thermal-hydraulic conditions. For these experimental studies, the involute-shaped fuel plates of the ANSR core with the narrow 1.27-mm flow gap are represented by a narrow rectangular channel. Tests in the THTL will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat-transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat-transfer coefficients, (4) two-phase pressure-drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) CHF limits.
Date: February 1, 1994
Creator: Felde, D. K.; Farquharson, G.; Hardy, J. H.; King, J. F.; McFee, M. T.; Montgomery, B. H. et al.
Partner: UNT Libraries Government Documents Department

Probabilistic risk assessment support of emergency preparedness at the Savannah River Site

Description: Integration of the Probabilistic Risk Assessment (PRA) for K Reactor operation into related technical areas at the Savannah River Site (SRS) includes coordination with several onsite organizations responsible for maintaining and upgrading emergency preparedness capabilities. Major functional categories of the PRA application are scenario development and source term algorithm enhancement. Insights and technologies from the SRS PRA have facilitated development of: (1) credible timelines for scenarios; (2) algorithms tied to plant instrumentation to provide best-estimate source terms for dose projection; and (3) expert-system logic models to implement informed counter-measures to assure onsite and offsite safety following accidental releases. The latter methodology, in particular, is readily transferable to other reactor and non-reactor facilities at SRS and represents a distinct advance relative to emergency preparedness capabilities elsewhere in the DOE complex.
Date: December 31, 1992
Creator: O`Kula, K. R.; Baker, W. H.; Simpkins, A. A.; Taylor, R. P.; Wagner, K. C. & Amos, C. N.
Partner: UNT Libraries Government Documents Department

The Advanced Neutron Source (ANS) project: A world-class research reactor facility

Description: This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5{times}10{sup 19}m{sup {minus}2}{center_dot}sec{sup {minus}1}. Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities.
Date: July 1, 1993
Creator: Thompson, P. B. & Meek, W. E.
Partner: UNT Libraries Government Documents Department

Structural integrity evaluation of high activity moderator system evaporator

Description: The High Activity Moderator (HAM) system is wanted in a batch mode in which the evaporator tank is filled with 70{degrees}F cold moderator (D{sub 2}O) every 4 hours. This operation induces thermal shock to the wall of the tank. Thermal and structural analyses are performed to evaluate the impact of this thermal shock on the 220{degrees}F hot evaporator tank walls. Conservative thermal models are analyzed. Case 1 analyzes a 4 in. wide strip of D{sub 2}O running down the tank walls during the filling process and Case 2 analyzes the tank being filled instantly with 70{degrees}F D{sub 2}O. It is found that Case 1 results in larger temperature gradients are then input into the structural model for calculating the thermal stresses. The structural analysis shows that the maximum stress intensity due to combined pressure and thermal loading is about 17240 psi which is well below the yield stress (21000 psi) of the evaporator tank wall material, stainless steel 304L. The fatigue life is evaluated in accordance with the criteria given in ASME Code, Section VIII. It is found that at the stress level of 17240 psi plus any residual stresses that might be present at the welded attachments to the tank wall, the fatigue life is about 4{times}10{sup 6} cycles. If the evaporator tank is filled every 4 hours, the tank fatigue life is well above the anticipated batch operation period of 2 years.
Date: May 1, 1994
Creator: Gupta, N. K.
Partner: UNT Libraries Government Documents Department

Reactivity and parameter measurements in a coaxial uranium fuel--D/sub 2/O moderated critical lattice

Description: Reactivity and reaction rate parameters were measured for a 7-inch triangular pitch lattice of coaxial uranium fuel in a critical, D/sub 2/O-moderated reactor. The results were compared with RAHABR computations using ENDF/B-IV cross sections and with an earlier subcritical exponential measurement of the same lattice. Measured and calculated reactivity are in good agreement, however, the calculated ratio of epicadmium /sup 238/U captures to subcadmium /sup 238/U captures (rho/sup 28//sub 28/) was 10% lower than measured. Indirect verification of the rho/sup 28/ measurement was obtained by measurement of a new conversion ratio parameter (C/sup +/) defined as the ratio of /sup 238/U captures to total fissions. Agreement between measured and calculated inner-to-outer fuel activation ratios suggests that the discrepancy is caused by underprediction of epicadmium captures in /sup 238/U or by erroneously high measured values of rho/sup 28/ and C/sup +/. However, any adjustment of /sup 238/U captures to match the measured rho/sup 28/ will cause the calculated reactivity to be underpredicted. An unresolved concern is that the experimental results do not satisfy an internal consistency criterion based on the two group neutron balance equation even though all known systematic errors have been accounted for and well known experimental techniques are used. Reaction rate parameters from the critical experiment agree with the exponential measurements indicating that the neutron spectra in the measurement regions were closely identical. The measured buckling in the critical facility is 0.52 m/sup -2/ lower than in the exponential. Other studies have shown systematic differences in reactivity between D/sub 2/O critical and exponential measurements.
Date: January 1, 1978
Creator: Pellarin, D.J.; Ahlfeld, C.E. & Baumann, N.P.
Partner: UNT Libraries Government Documents Department

Survey of considerations involved in introducing CANDU reactors into the United States

Description: The important issues that must be considered in a decision to utilize CANDU reactors in the U.S. are identified in this report. Economic considerations, including both power costs and fuel utilization, are discussed for the near and longer term. Safety and licensing considerations are reviewed for CANDU-PHW reactors in general. The important issues, now and in the future, associated with power generation costs are the capital costs of CANDUs and the factors that impact capital cost comparisons. Fuel utilization advantages for the CANDU depend upon assumptions regarding fuel recycle at present, but the primary issue in the longer term is the utilization of the thorium cycle in the CANDU. Certain safety features of the CANDU are identified as intrinsic to the concept and these features must be examined more fully regarding licensability in the U.S.
Date: January 1, 1977
Creator: Till, C E; Bohn, E M; Chang, Y I & van Erp, J B
Partner: UNT Libraries Government Documents Department

Fuel-performance-improvement program. Semiannual progress report, October 1980-March 1981. [Sphere-pac and annular-coated-pressurized]

Description: Progress on the Fuel Performance Improvement Program's fuel test and demonstration irradiations is reported for the period of October 1980-March 1981. The purpose of the program is to test and demonstrate improved light water reactor fuel concepts that are more resistant to failure from pellet-cladding interaction during power increases than standard pellet fuel. This would also offer extended burnup potential and, hence, improved uranium utilization.
Date: April 1, 1981
Creator: Crouthamel, C E & Freshley, M D
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

Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

Description: The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant.
Date: June 1, 1977
Partner: UNT Libraries Government Documents Department

Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA]

Description: ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.
Date: June 1, 1977
Partner: UNT Libraries Government Documents Department