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Proposal for 2020 program

Description: This document is a proposal to do an analysis of the use of light water reactors (LWRs) in Weapons Material Production. The objective of this study is to examine the major issues associated with using a LWR to produce weapons materials. The central focus of the study will be on the design and safety issues relating to tritium production in LWRs using uranium fuel with both low and high enrichment. This study will identify the problems and propose solutions. This study will analyze the technology of tritium production in an LWR. The first step will identify and quantify the major problems and the worst credible accident condition. The next step will consist of identifying and evaluating engineering changes to the LWR design.
Date: March 26, 1982
Creator: Dingee, D. A.
Partner: UNT Libraries Government Documents Department

A STUDY OF THE PRIMARY SHIELD FOR THE PRDC REACTOR

Description: Temperature distributions, irradiation effects, stacking arrangements, voidage, and economics for the borated-graphite shield of the PRDC reactor were investigated. Of the shield systems considered, four are reported here. System 1 contalns 30 in. of 1% borated graphite, with either ordinary graphite or a cement as a filler for the remaindcr of the volume. The maximum temperature at the flex plates in this system was calculated to be 5OO deg F. Systems 2 and 3 consist of 2 in. of 5% borated graphite near the core vessel and 1/2 in. of Boral at the primary-shield tank. A filler material of carbon blocks is used in System 2 and graphite in System 3. The calculated maximum temperatures were 700 deg F and 35O deg F, respectively. System 4 consists of a laminated structure of Boral and graphite near the primary-shield tank and carbon-block filler. It was calculated to have a maximum temperature of 600 deg F at the flex plates. The maximum temperature at the flex plates recommended by APDA is 500 deg F. Energy storage and radiation damage were found to be within permissible limits in all four systems. However, these conclusions are based on experimental data from the Hanford reactor in which the neutron-energy spectrum differs considerably from the PRDC spectrum. A porosity of less than 740 cu ft is required in order that a sodium leak from the core vessel does not expose the core. The voidages in any of the systems mentioned above is about 400 cu ft excluding absorption effects. these are believed to be small. The systems coataining Boral were found to be less expensive than the ones using only borated graphite. Over-all material costs range between 0,000 for Boral systems and 0,000 for borated- graphite systems. (auth)
Date: April 15, 1957
Creator: Epstein, H.M.; Dingee, D.A. & Chastain, J.W.
Partner: UNT Libraries Government Documents Department

HAZARDS SUMMARY REPORT FOR THE BATTELLE PLASTIC REACTOR FACILITY

Description: Experiments using a plastic-moderated assembly as a radiation source are described, and the hazards attendant to these experiments are evaluated. The critical assembly, designated Battelle's Plastic Reactor Facility, is constructed in the form of a cube. A portion of the plastic and fuel will be removed from the center of this reactor to provide a radiation environment for exposing fission-chamber devices. This central void region will approximate a flux trap in which a bigh neutron-to-gamma ratio is expected. The fuel-element assemblies are composed of strips of aluminum and Teflon-coated uranium sandwiched with plastic and encased in aluminum boxes. One portion of the core is on a movable table, while the other part is on a fixed table. The core is assembled remotely by driving the movable table against the fixed table. Primary control and safety of the assembly are achieved by insenting or withdrawing regulating and safety elements and by increasing or decreasing the distance between the two core halves. For safety, the maximum normal rate of reactivity addition has been limited to 0.04% DELTA k/k per sec for control-element withdrawal and table closure. The system is interlocked so that only one control element can be withdrawn at a time and so that the movable table and elements cannot be moved simultaneously. In addition, criticality cannot be reached by only moving the tables together. No experiments with the Plastic Reactor Facility are planned since it will serve only as a source of radiation operating at a constant power level. Consequently, the hazards of a nuclear excursion are minimized. However, the safety aspects of the operations and possible power excursions have been analyzed. The analysis indicates that these operations present no significant hazard to the public persons or operating staff. (auth)
Date: January 20, 1961
Creator: Dingee, D.A. & Chastain, J.W. Jr.
Partner: UNT Libraries Government Documents Department

ML-1-1A CORE STUDIES WITH THE GCRE CRITICAL ASSEMBLY

Description: Critical assembly studies were conducted tc provide physics and engineering data to aid in developing the Mobile Low-Power Reactor (ML-1). The ML-1-lA core was critical with 59 elements containing 17,906.71 g of U/sup 235/ and had an excess reactivity of 0.381 x 10/sup -2/ DELTA k/k at a moderator temperature of 24.91 deg C. The ratio of maximum element power to core-averaged power was approximately 1.09. The ratio of maximum to core-averaged thermal flux was approximately 1.10. At an 18-deg separation, the shutdown worth of the cadmium-covered control-blade mock-up was 1.14 x 10/sup -2/ DELTA k/k for a 69 element core. Radial and upper axial reflector-moderator void coefficients were - 0.59 plus or minus 0.07 and -0.36S plus or minus 0.015 x 10/sup -2/ DELTA k/ k.per in., respectively. Two lA production fuel elements were evaluated in the critical assembly core. The results predict that the production elements tested contained roughly the same fuel as the critical assembly element and an additional 772 g of stainless steel equivalent on the average. Radial power and neutron flux distributions were measured in a 19-pin lB fuel element. Fairly uniform distributions were observed. Data to evaluate the thermal utilization of this element were obtained. (auth)
Date: November 27, 1959
Creator: Egen, R.A.; Hogan, W.S.; Dingee, D.A. & Chastain, J.W.
Partner: UNT Libraries Government Documents Department

FURTHER STUDIES WITH THE GCRE CRITICAL ASSEMBLY

Description: Further engineering and physics data to aid in constructing GCRE-1 were obtained in critical-assembly studies. Four major experiments were performed to investigate the effect on reactivity caused by changes in axial reflector materials, the effect on reactivity and the power perturbation caused by fast safety control-blade guides, the effect of changes in fuel-element material composition, and the effect of changes in fuel-element spacing designed to produce uniform radial power-generation rates. All studies were performed with a 4-in.-thick lead reflector at the core perimeter. Axial-reflector-material studies employed combirations of aluminum and steel reflectors. The reactivity worth of a 2 3/4-in.-thick steel reflector was +0.414% DELTA k/k compared with 0.175% DELTA k/k for a similar aluminum reflector. The perturbation in the flux distribution caused by the safety-blade guides was localized, and affected only the regions immediately adjacent to the guides. The combined reactivity worth of two guides was -0.281% DELTA k/k. Fuel-element material compositions were changed by separate additions of fuel and stainless steel. An increase in uranium loading from an average value of 303 to 404 g per element would provide, based on extrapolations from experimental data, a reactivity of about 4.5% DELTA k/k. An increase in steel from 1708 to 2093 g per element decreased the core reactivity by abeut 1.1% DELTA k/k. A change in fuelelement spacing reduced the ratio of maximum to average power generation from 1.46 to 1.24. (auth)
Date: December 29, 1958
Creator: Dingee, D.A.; Ballowe, W.C.; Egen, R.A.; Jankowski, F.J. & Chastain, J.W. Jr.
Partner: UNT Libraries Government Documents Department

HAZARDS SUMMARY REPORT FOR THE VMR CRITICAL-ASSEMBLY EXPERIMENTS

Description: Moderator Reactor (VMR), a reactor concept under investigation by American-Standard for the AEC. The VMR is light-water moderated and cooled and is fueled with slightly enriched uranium dioxide pellets loaded into aluminum tubes. The core consists of 37 hexagonal fuel cans each loaded with 61 fuel pins. The cooling water, which flows upward around the pins inside the fuel can, boils in passing through the core. Reactor control in the prototype will be achieved by varying the moderator height. The site, laboratory, and the critical assembly, including control and safety mechanisms, are described in detuil. Special characteristics of the assembly pentinent to safety were calculated. The nuclesr energy released and the average and maximum fuel temperatures resulting from step reactivity increases up to 2% DELTA k/k are presented graphically for two cases. In the first case, fuel-temperature effects are considered to be the oniy shutdown mechanism; in the second radiolytic gas is considered to contribute to shutdown, in addition to fuel-temperature effects. The accident considered to be the maximum credible accident causes a step addition in reactivity of 1.5% DELTA k/k. The nuclear-energy release is between 160 and 310 megawatt-sec depending on the assumed shutdown mechanisms. This accident does not cause any fuel to be vapcrized (and probably none to be melted) and, hence, there does not appear to be a hazard from fission-product activity. It appears that this criticalassembly program can be conducted with reasonable assurance of safe operation and that no public persons will be jeopardized by its operation. (auth)
Date: June 10, 1960
Creator: Egen, R.A.; Hogan, W.S.; Dingee, D.A. & Chastain, J.W.
Partner: UNT Libraries Government Documents Department

HAZARDS SUMMARY REPORT FOR THE GCRE CRITICAL-ASSEMBLY EXPERIMENTS

Description: Critical experiments are described, and the hazards attendant to these experiments are evaluated for a gascooled water-moderated reactor design. Shutdown control of the critical assembly is achieved by cadmium control blades and by dumping the water into a storage tank. The maximum normal rats of reactivity addition is limited to an estimated 0.04 per cent per sec for all remotely controlled operations. The system is interlocked to insure that safe operational procedures are followed, Hazard calculations are made to determine the dosage from direct radiation, fall-out, and inhalation from a radioactive cloud resulting from an accident. The exclusion area is shown to be adequate for even the maximum hypothetical accident. (auth)
Date: July 22, 1958
Creator: Hogan, W.S.; Dingee, D.A.; Jankowski, F.J. & Chastain, J.W.
Partner: UNT Libraries Government Documents Department

Aging of nuclear station diesel generators: Evaluation of operating and expert experience: Phase 1, Study

Description: Pacific Northwest Laboratory evaluated operational and expert experience pertaining to the aging degradation of diesel generators in nuclear service. The research, sponsored by the US Nuclear Regulatory Commission (NRC), identified and characterized the contribution of aging to emergency diesel generator failures. This report, Volume I, reviews diesel-generator experience to identify the systems and components most subject to aging degradation and isolates the major causes of failure that may affect future operational readiness. Evaluations show that as plants age, the percent of aging-related failures increases and failure modes change. A compilation is presented of recommended corrective actions for the failures identified. This study also includes a review of current, relevant industry programs, research, and standards. Volume II reports the results of an industry-wide workshop held on May 28 and 29, 1986 to discuss the technical issues associated with aging of nuclear service emergency diesel generators.
Date: August 1, 1987
Creator: Hoopingarner, K.R.; Vause, J.W.; Dingee, D.A. & Nesbitt, J.F.
Partner: UNT Libraries Government Documents Department

CRITICAL-ASSEMBLY EXPERIMENTS ON A REFLECTOR CONTROL SYSTEM FOR A BOILING REACTOR

Description: Research directed toward an evaluation of a reflector control scheme for a small boiling heterogeneous reactor is reported. The control method uses the ambient reactor steam pressure to adjust the height of a radial water reflector. Critical-assembly experiments were conducted to measure the reflector worth and the effect of nonhomogeneous voids. The critical-assembly core approximated a right circular cylinder 24 in. high and 21 in. in diameter. The core was made up of polyethylene strips to simulate water channels, aluminum strips, and uranium foil. The reflector consisted of polyethylene blocks and two different densities were used to simulate two different water temperatures. The coldwater, hot non- power-producing, and boiling powerproducing reactor conditions were mocked-up. These conditions were obtained by voiding the moderator plastic strips. The results indicate that sufficient reactivity is available in the reflector to compensate for large power variations and equilibrium poisons. However, the reactivity worth is not great enough in the system considered to include startup from the cold condition or long-time fuel burnup. These experiments also show that in calculatioiis of critical mass and reflector worth the assumption of uniform void distributions seems valid for corcs of the size studied. (auth)
Date: December 20, 1957
Creator: Dingee, D.A.; Hogan, W.S.; Wilson, R.G.; Ballowe, W.C.; Jankowski, F.J. & Chastain, J.W.
Partner: UNT Libraries Government Documents Department

GCRE CRITICAL-ASSEMBLY STUDIES

Description: Critical-assembly studies were made to provide engineering and physics data to aid in developing the Gas Cooled Reactor Experiment-1 (GCRE-l). Measurements of critical mass, flux and power distributions, and shutdown worth of the GCRE-1 mock-up safety and control blades were obtained. The critical assembly consists of aluminum tubes containing four concentric stainless steel cylinders wrapped wtth highly enriched uranium foil. These tubes are supported at each end by grid plates aiid arranged to approximate a right circular cylinder. The entire core structure is supported within a tank which can be filled remotely with moderator water. A void beneath the core structure and the air above it represent the gas plenums of the GCRE-1 reactor. Critical mass and flux power distributions were determined for several cases of four basic cores. The thermal utilization, measured in two core configurations, was 0.780 for the reactor wtthout burnable poison or a lead reflector. The temperature coefficient of reactivity, measured in two cores, was positive at room temperature but negative in the proposed operatingtemperature range. The total reactivity effect in going from 20 to 80 deg C was a positive 0.22% DELTA k/k. The worth of control and safety-shutdown blades of several compositions and sizes in various locations was determined for a number of core configurations; blades having a sandwich construction of cadmium, tungsten, and indium had the greatest shutdown worth. To date these studies have resulted in a 61-element core (critical with 54 elements) with a 4-in. lead reflector which has 2.5% DELTA k/k excess reactivity for startup and for override of poison and burnup. A mock-up blade configuration has been determined which will control the excess reactivity introduced by flooding the elements with mineral oil (to simulate water flooding when fuel elements are changed in GCRE-1). (auth)
Date: September 10, 1958
Creator: Dingee, D. A.; Ballowe, W. C.; Klingensmith, R. W.; Egen, R. A.; Jankowski, F. J. & Chastain, J. W.
Partner: UNT Libraries Government Documents Department