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TECHNIQUES FOR MONITORING PLUTONIUM IN THE ENVIRONMENT

Description: Plutonium is one of the principal materials of both commercial and military nuclear power. It is produced primarily in fission reactors that contain uranium fuel, and its importance arises from the fact that a large portion of the plutonium produced is fissile: like uranium 235, the mass 239 and 241 isotopes of plutonium can be caused to fission by neutrons, including those with low energy. Because such fission events also release neutrons, substantial amounts of energy can be extracted from plutonium in a controlled or an explosive nuclear chain reaction. Now that commercial nuclear reactors provide a noticeable fraction of United States (and world) electrical energy, these reactors account for most plutonium production. For the most part, this material now remains in the irradiated fuel after removal from reactors, but should this fuel be reprocessed, the plutonium could be recycled to provide part and even most of the fissile content of fresh fuel. For the current generation of water-cooled reactors, the amount of plutonium to be recycled is substantial. In fast breeder reactors, designed to produce more fissile material than they destroy, considerably larger quantities of plutonium would be recycled. In other types of advanced reactors, particularly those which depend heavily on thorium as the material from which fissile material (primarily uranium 233) is produced, the amount of plutonium to be handled would be considerably reduced. Because plutonium is a highly toxic substance, great care is taken to contain it at the sites and facilities where it is stored or handled. In addition, it is necessary that devices be available to monitor any releases from these facilities into environmental media and to measure concentrations of plutonium in these media. The radiation protection standards are so strict for plutonium that only small releases and low concentrations can be tolerated. Such considerations, ...
Date: July 1, 1978
Creator: Nero, A. V., Jr.
Partner: UNT Libraries Government Documents Department

Evaluation of materials for retention of sodium and core debris in reactor systems. Annual progress report, September 1977-December 1978. [LMFBR]

Description: This report considers some of the consequences of a hypothetical core disruptive accident in a nuclear reactor. The interactions expected between molten core debris, liquid sodium, and materials that might be employed in an ex-vessel sacrificial-bed or in the reactor building are discussed. Experimental work performed for NRC by Sandia Laboratories and Hanford Engineering Development Laboratory on the interactions between liquid sodium and basalt concrete is reviewed. Studies of molten steel interactions with concrete at Sandia Laboratories and molten UO/sub 2/ interactions with concrete at The Aerospace Corporation are also discussed. The potential of MgO for use in core containment is discussed and refractory materials other than MgO are reviewed. Finally, results from earlier experiments with molten core debris and various materials performed at The Aerospace Corporation are presented.
Date: December 1, 1978
Creator: Swanson, D.G.; Zehms, E.H.; McClelland, J.D.; Meyer, R.A. & van Paassen, H.L.L.
Partner: UNT Libraries Government Documents Department

Thermal insulation system design and fabrication specification (nuclear) for the Clinch River Breeder Reactor plant

Description: This specification defines the design, analysis, fabrication, testing, shipping, and quality requirements of the Insulation System for the Clinch River Breeder Reactor Plant (CRBRP), near Oak Ridge, Tennessee. The Insulation System includes all supports, convection barriers, jacketing, insulation, penetrations, fasteners, or other insulation support material or devices required to insulate the piping and equipment cryogenic and other special applications excluded. Site storage, handling and installation of the Insulation System are under the cognizance of the Purchaser.
Date: July 21, 1978
Partner: UNT Libraries Government Documents Department

Leakage and motion detection system for the flexible-joint assembly, large-scale LMFBR

Description: Flexible joint assemblies in the primary sodium piping to the large scale LMFBR pressure vessel are designed to accommodate thermal expansion of the piping system. To monitor the performance of the flexible joint assemblies during reactor operation, sodium leakage and flexible joint motion detection/sensors are specified. This report describes the mechanical/hydraulic portions only of the leak detection and motion monitoring system. A tentative choice of displacement transducer and its readout equipment is presented. However, the required EI and C portion of the monitoring system is not covered.
Date: August 1, 1978
Creator: Gaspar, N.L.
Partner: UNT Libraries Government Documents Department

26 - LMFBR flexible pipe joint development

Description: Objective is the qualification of a PLBR-size primary loop flexible piping joint to the ASME Band PVC rules. Progress and activities are reported for: Class 1 flexible joint code approval support, engineering and design, material development, component testing, and manufacturing development. (DLC)
Date: May 1, 1978
Creator: Anderson, R.V.
Partner: UNT Libraries Government Documents Department

Evaporative removal of sodium: interim progress report and preliminary facility specification

Description: A summary of the current Evaporative Removal of Sodium (ERNA) activities at the Energy Systems Group is presented. Also included is a review of earlier work on sodium evaporation. As a result of this work it was concluded that the ERNA process was extremely successful and worthy of future consideration as a recognized process for reactor components. Also included in the report is a Preliminary Outline Specification for a large facility to remove sodium from full size CRBR fuel rod assemblies.
Date: September 27, 1978
Creator: Welch, F.H.
Partner: UNT Libraries Government Documents Department

Cover gas seals: 26-cover gas seal components

Description: Progress during the report period included CRBRP inflatable seal vendor qualification (Presray Inc.), static inflatable seal development, high-temperature seal degradation, and alternate seal concepts (supported and cantilevered rings). (DLC)
Date: January 1, 1978
Creator: Steele, O.P. III & Horton, P.H.
Partner: UNT Libraries Government Documents Department

Commercial LMFBR steam generator design comparison. Final report for period from 1 October 1977 through 30 September 1978

Description: This report presents results obtained from the commercial LMFBR Steam Generator Design Comparison Study from 1 October 1977 through 30 September 1978 relative to selecting the preferred steam generator design for a commercial-size plant using a Benson, Sulzer, or saturated steam cycle. The primary emphasis was placed on identifying potential problem areas in each design for each steam cycle. The study indicates the hockey stick design as the preferred concept for each steam cycle.
Date: September 30, 1978
Creator: Newburn, F.
Partner: UNT Libraries Government Documents Department

Fabrication details for wire wrapped fuel assembly components. [LMFBR]

Description: Extensive hydraulic testing of simulated LMFBR blanket and fuel assemblies is being carried out under this MIT program. The fabrication of these test assemblies has involved development of manufacturing procedures involving the wire wrapped pins and the flow housing. The procedures are described in detail in the report.
Date: September 1, 1978
Creator: Bosy, B.J.
Partner: UNT Libraries Government Documents Department

Interim report on cold trap alternatives, sodium technology

Description: This is an interim report on the modifications which are being made to an existing sodium loop so that a new method of removing hydrogen from sodium can be evaluated in a flowing sodium system. Some preliminary results on the performance of this type of cold trap alternative in a static sodium system are reported. Some tentative reactor design parameters for this type of getter device are presented based on the static test results. The life of such a unit is calculated to be substantially greater than that of a cold trap of equal volume.
Date: September 18, 1978
Creator: Hill, E.F.
Partner: UNT Libraries Government Documents Department

Simple representations of spectrum averaged cross sections in LMFBR blankets

Description: The setting up of correlations to represent spectrum averaged cross sections in the core-blanket interface and blanket region of LMFBRs is described. Analytical representations of concentrations as functions of position are used. Sample calculations using a CRBR core composition are presented.
Date: January 1, 1978
Creator: Badruzzaman, A.; Wiley, R. & Becker, M.
Partner: UNT Libraries Government Documents Department

Turbulent sweeping flow mixing model for wire wrapped LMFBR assemblies

Description: A physical model is proposed to derive the sweeping flow for LMFBR triangular array wire-wrapped assemblies under the turbulent flow condition. Two correlations are suggested for the sweeping flow through two different types of gaps between subchannels, the gap between the interior subchannels and the gap between the wall subchannels. These two sweeping flow correlations are evolved by calibrating the constants in the proposed model against the available experimental data. Agreement between the correlations and all the experimental data of +- 35% is obtained over the assembly design range of 1.315 > P/D > 1.067 and 52 > H/D > 4. Based on these correlations, flow sweeping input parameters for two popular computer codes, i.e., COBRA and SUPERENERGY, are recommended.
Date: October 1, 1978
Creator: Chiu, C.; Rohsenow, W.M. & Todreas, N.E.
Partner: UNT Libraries Government Documents Department

Large-systems improvement studies

Description: A set of five backup decay heat removal candidate schemes has been studied in sufficient depth to create conceptual arrangements, P and I diagrams, and equipment lists for each. A summary of the attributes of each system arrangement is developed. System costs were also developed but do not show significant differences.
Date: September 1, 1978
Creator: Riedel, R.H.
Partner: UNT Libraries Government Documents Department

Cell liner design for LMFBR plants

Description: Those areas or cells within LMFBR plants that contain radioactive sodium systems are provided with certain design features which eliminate or limit potential sodium/concrete reaction and thus protect the concrete structure in the event of an accidental sodium spill. The principal design feature within these cells that controls sodium spill effects is the cell liner system. The description, requirements and analysis of such a system design for the Clinch River Breeder Reactor Plant (CRBRP) is presented in this paper. The information included in this paper can be utilized directly or can formulate the basis for design of cell liners for commercial scale LMFBR's or future large scale liquid metal test facilities.
Date: September 1, 1978
Creator: Brolin, E.C. & Palm, R.E.
Partner: UNT Libraries Government Documents Department

Cell liner design for LMFBR plants

Description: Those areas or cells within LMFBR plants that contain radioactive sodium systems are provided with certain design features which eliminate or limit potential sodium/concrete reaction and thus protect the concrete structure in the event of an accidental sodium spill. The principal design feature within these cells that controls sodium spill effects is the cell liner system. The development of such a system design for the Clinch River Breeder Reactor Plant (CRBRP) is presented in this paper. The basis for cell liner designs is described, including general design criteria, materials and welding requirements, system design requirements, load categories and loading combinations and allowable stress and strain levels. Results of stress analysis and design details of the cell wall and floor systems are presented, including provisions for protection of the concrete structure against high temperature effects.
Date: January 1, 1978
Creator: Palm, R.E. & Burrow, R.C.
Partner: UNT Libraries Government Documents Department

Pump, sodium, inducer, intermediate size (ISIP) (impeller/inducer/diffuser retrofit)

Description: This specification defines the requirements for the Intermediate-Size Inducer Pump (ISIP), which is to be made by replacing the impeller of the FFTF Prototype Pump with a new inducer, impeller, diffuser, seal, and necessary adapter hardware. Subsequent testing requirements of the complete pump assembly are included.
Date: April 21, 1978
Creator: Paradise, D.R.
Partner: UNT Libraries Government Documents Department

Attenuation of airborne debris from LMFBR accidents

Description: Experimental and theoretical studies have been performed to characterize the behavior of airborne particulates (aerosols) expected to be produced by hypothetical core disassembly accidents (HCDA's) in liquid metal fast breeder reactors (LMFBR's). These aerosol studies include work on aerosol transport in a 20-m high, 850-m/sup 3/ closed vessel at moderate concentrations; aerosol transport in a small vessel under conditions of high concentration (approx. 1000 g/m/sup 3/), high turbulence, and high temperature (approx. 2000/sup 0/C); and aerosol transport through various leak paths. These studies have shown that little, if any, airborne debris from LMFBR HCDA's would reach the atmosphere exterior to an intact reactor containment building.
Date: January 1, 1978
Creator: Morewitz, H. A.; Johnson, R. P.; Nelson, C. T.; Vaughan, E. U.; Guderjahn, C. A.; Hilliard, R. K. et al.
Partner: UNT Libraries Government Documents Department

Prediction of thermal consequences of plenum fission gas release in an LMFBR fuel assembly

Description: One of the types of hypothetical accidents important in the safety analysis of Liquid Metal Fast Breeder Reactors is that in which fission gas release is postulated to occur from the plena of fuel pins at such a rate that significant perturbation to coolant flow occurs. Past consideration of such events has employed highly conservative calculations in the interest of simplicity and expediency. This paper describes a method developed to predict more accurately the thermal consequences of such fission gas release events.
Date: January 1, 1978
Creator: Kadambi, N.P.
Partner: UNT Libraries Government Documents Department

Postirradiation Examinations of Fuel Pins from the GCFR F-1 Series of Mixed-Oxide Fuel Pins at 5. 5 at. % Burnup

Description: Postirradiation examinations were performed on five fuel pins from the Gas-Cooled Fast-Breeder Reactor F-1 experiment irradiated in EBR-II to a peak burnup of approximately 5.5 at. %. These encapsulated fuel pins were irradiated at peak-power linear ratings from approximately 13 to 15 kW/ft and peak cladding inside diameter temperatures from approximately 625 to 760°C. The maximum diametral change that occurred during irradiation was 0.2% .delta.D/D₀. The maximum fuel-cladding chemical interaction depth was 2.6 mils in fuel pin G-1 and 1 mil or less in the other three pins examined destructively. Significant migration of the volatile fission products occurred axially to the fuel-blanket interfaces. The postirradiation examination data indicate that fuel melted at the inner surface of the annular fuel pellets in the two highest power rating fuel pins, but little axial movement of fuel occurred.
Date: May 1978
Creator: Johnson, C. E. & Strain, R. V.
Partner: UNT Libraries Government Documents Department

Investigations of Materials Compatibility Relevant to the EBR-II system : FY 1977 and the Transition Period

Description: Report on yearly investigations of components of the EBR-II reactor plant and on out-of-reactor experiments in which materials or techniques are tested before they are used in the EBR-II reactor system.
Date: February 1978
Creator: Longua, K. J.; Ruther, W. E.; Shields, J. A. & Clark, A. F.
Partner: UNT Libraries Government Documents Department

Approximations of Gamma Cross Sections for Fast Nuclear Reactors

Description: The report shows a method to approximate a P₁ scattering solution for the flux in a fast reactor, using an isotropic, but not a diagonal-transport-approximation scattering matrix. Presented are flux errors relative to a P₁ solution for different levels of transport approximation in an EBR-II type of core surrounded by a stainless steel reflector. Problems associated with the use of the method are also presented.
Date: 1978?
Creator: Grimm, K. N. & Meneghetti, D.
Partner: UNT Libraries Government Documents Department

Tritium and Hydrogen Transport in LMFBR Systems: EBR-II, CRBR, and FFTF

Description: A tritium and hydrogen transport model has been employed to simulate concentration profiles, tritium losses to auxiliary containment systems, and cold trap burdens for EBR-II, CRBR, and FFTF. Experimental data from EBR-II were found to correlate well with calculated tritium and hydrogen profiles. A major change relative to previous transport models, namely, the inhibiting effect of oxide coatings on tritium permeation through reactor structural surfaces, has been incorporated into the current model. Tritium release rates to auxiliary systems where oxide barrier effects were included were predicted to be approximately two orders of magnitude lower than those for the reference case where structural surfaces were assumed to be totally oxide-free. Tritium releases during operation of large LMFBRs are expected to present essentially no hazard to the environment.
Date: September 1978
Creator: Renner, T. A. & McPheeters, C. C.
Partner: UNT Libraries Government Documents Department