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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

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

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

Safety Evaluation Report Restart of K-Reactor Savannah River Site

Description: In April 1991, the Department of Energy (DOE) issued DOE/DP-0084T, Safety Evaluation Report Restart of K-Reactor Savannah River Site.'' The Safety Evaluation Report (SER) documents the results of DOE reviews and evaluations of the programmatic aspects of a large number of issues necessary to be satisfactorily addressed before restart. The issues were evaluated for compliance with the restart criteria included in the SER. The results of those evaluations determined that the restart criteria had been satisfied for some of the issues. However, for most of the issues at least part of the applicable restart criteria had not been found to be satisfied at the time the evaluations were prepared. For those issues, open or confirmatory items were identified that required resolution. In August 1991, DOE issued DOE/DP-0090T, Safety Evaluation Report Restart of K-Reactor Savannah River Site Supplement 1.'' That document was the first Supplement to the April 1991 SER, and documented the resolution of 62 of the open items identified in the SER. This document is the second Supplement to the April 1991 SER. This second SER Supplement documents the resolution of additional open times identified in the SER, and includes a complete list of all remaining SER open items. The resolution of those remaining open items will be documented in future SER Supplements. Resolution of all open items for an issue indicates that its associated restart criteria have been satisfied, and that DOE concludes that the programmatic aspects of the issue have been satisfactorily addressed.
Date: October 1, 1991
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

Heat exchanger restart evaluation

Description: On December 24, 1991, the K-Reactor was in the shutdown mode with full AC process water flow and full cooling water flow. Safety rod testing was being performed as part of the power ascension testing program. The results of cooling water samples indicated tritium concentrations higher than allowable. Further sampling and testing confirmed a Process Water System to Cooling Water System leak in heat exchanger 4A (HX 4A). The heat exchanger was isolated and the plant shutdown. Heat exchanger 4kA was removed from the plant and moved to C-Area prior to performing examinations and diagnostic testing. This included locating and identifying the leaking tube or tubes, eddy current examination of the leaking tube and a number of adjacent tubes, visually inspecting the leaking tube from both the inside as well as the area surrounding the failure mechanism. In addition ten other tubes that either exhibited eddy current indications or would represent a baseline condition were removed from heat exchanger 4A for metallurgical examination. Additional analysis and review of heat exchanger leakage history was performed to determine if there are any patterns which can be used for predictive purposes. Compensatory actions have been taken to improve the sensitivity and response time to any future events of this type. The results of these actions are summarized herein.
Date: February 28, 1992
Creator: Morrison, J.M.; Hirst, C.W. & Lentz, T.F.
Partner: UNT Libraries Government Documents Department

Analysis of coolability of the control rods of a Savannah River Site production reactor with loss of normal forced convection cooling

Description: An analytical study of the coolability of the control rods in the Savannah River Site (SRS) K-Production Reactor under conditions of loss of normal forced convection cooling has been performed. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven flow over the control rods that occurs when forced cooling is lost, and the limit of critical heat flux that sets the acceptance criteria for the study. The objective of the study is to demonstrate that the control rods will remain cooled at powers representative of those anticipated for restart of the reactor. The study accomplishes this objective with a very tractable simplified analysis for the modest restart power. In addition, a best-estimate calculation is performed, and the results are compared to results from sub-scale scoping experiments. 5 refs.
Date: January 1, 1992
Creator: Easterling, T.C.; Hightower, N.T. (Westinghouse Savannah River Co., Aiken, SC (United States)); Smith, D.C. & Amos, C.N. (Science Applications International Corp., Albuquerque, NM (United States))
Partner: UNT Libraries Government Documents Department

Automatic diagnosis of alarms: a system to improve operator emergency response

Description: A system is being developed at the Savannah River Plant to help reactor operators respond to multiple alarms in a developing incident situation. The need for such systems has becme evident in recent years, particularly after the Three Mile Island incident.
Date: January 1, 1980
Creator: Olson, H P; Gimmy, K L; Nomm, E & Finley, R H
Partner: UNT Libraries Government Documents Department

Action Plan for updated Chapter 15 Accident Analysis in the SRS Production Reactor SAR

Description: This report describes the Action Plan for the upgrade of the Chapter 15 Accident Analysis in the SRS Production Reactor SAR required for K-Restart. This Action Plan will be updated periodically to reflect task accomplishments and issue resolutions.
Date: November 15, 1989
Creator: Hightower, N.T. III & Burnett, T.W.
Partner: UNT Libraries Government Documents Department

Safety rod latch inspection

Description: During an attempt to raise control rods from the 100 K reactor in December, one rod could not be withdrawn. Subsequent investigation revealed that a small button'' in the latch mechanism had broken off of the lock plunger'' and was wedged in a position that prevented rod withdrawal. Concern that this failure may have resulted from corrosion or some other metallurgical problem resulted in a request that SRL examine six typical latch mechanisms from the 100 L reactor by use of radiography and metallography. During the examination of the L-Area latches, a failed latch mechanism from the 100 K reactor was added to the investigation. Fourteen latches that had a history of problems were removed from K-Area and sent to SRL for inclusion in this study the week after the original seven assemblies were examined, bringing the total of latch assemblies discussed in this report to twenty one. Results of the examination of the K-Area latch that initiated this study is not included in this report.
Date: February 1, 1992
Creator: Leader, D.R.
Partner: UNT Libraries Government Documents Department

Calculation of Savannah River K Reactor Mark-22 assembly LOCA/ECS power limits

Description: This paper summarizes the results of TRAC-PF1/MOD3 calculations of Mark-22 fuel assembly of loss-of-coolant accident/emergency cooling system (LOCA/ECS) power limits for the Savannah River Site (SRS) K Reactor. This effort was part of a larger effort undertaken by the Los Alamos National Laboratory for the US Department of Energy to perform confirmatory power limits calculations for the SRS K Reactor. A method using a detailed three-dimensional (3D) TRAC model of the Mark-22 fuel assembly was developed to compute LOCA/ECS power limits. Assembly power was limited to ensure that no point on the fuel assembly walls would exceed the local saturation temperature. The detailed TRAC model for the Mark-22 assembly consisted of three concentric 3D vessel components which simulated the two targets, two fuel tubes, and three main flow channels of the fuel assembly. The model included 100% eccentricity between the assembly annuli and a 20% power tilt. Eccentricity in the radial alignment of the assembly annuli arises because axial spacer ribs that run the length of the fuel and targets are used. Wall-shear, interfacial-shear, and wall heat-transfer correlations were developed and implemented in TRAC-PF1/MOD3 specifically for modeling flow and heat transfer in the narrow ribbed annuli encountered in the Mark-22 fuel assembly design. We established the validity of these new constitutive models using separate-effects benchmarks. TRAC system calculations of K Reactor indicated that the limiting ECS-phase accident is a double-ended guillonite break in a process water line at the pump discharge (i.e., a PDLOCA). The fuel assembly with the minimum cooling potential is identified from this system calculation. Detailed assembly calculations then were performed using appropriate boundary conditions obtained from this limiting system LOCA. Coolant flow rates and pressure boundary conditions were obtained from this system calculation and applied to the detailed assembly model.
Date: January 1, 1992
Creator: Fischer, S.R.; Farman, R.F. & Birdsell, S.A.
Partner: UNT Libraries Government Documents Department