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Some Scoping Experiments for a Space Reactor

Description: Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failur… more
Date: July 7, 1983
Creator: Alexander, C. A. & Ogden, J. S.
Partner: UNT Libraries Government Documents Department
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PWR Blowdown Heat Transfer Separate-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 103

Description: Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 103, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system.
Date: March 7, 1978
Creator: Clemons, V. D.; White, M. D.; Moore, P. A. & Hedrick, R. A.
Partner: UNT Libraries Government Documents Department
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LOFT suppression tank spray system piping: heat exchanger BS-H-31 piping modifications

Description: A stress analysis of the piping modification, resulting from relocation of heat exchanger BS-H-31 of the LOFT Blowdown Suppressing Tank Spray System, was performed. The piping, fittings, and supports were found to comply with the criteria of Section III of the ASME Boiler and Pressure Vessel Code, 1974.
Date: November 7, 1977
Creator: Blandford, R.K.
Partner: UNT Libraries Government Documents Department
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Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

Description: The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures,… more
Date: March 7, 1980
Partner: UNT Libraries Government Documents Department
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On the Use of Rossi Alpha Critical Assembly Measurements for Validating and Constraining Nuclear Data

Description: Critical assemblies are exquisitely sensitive to details of the microscopic nuclear reactions that govern neutron multiplication. For this reason experimental studies of critical assemblies represent a cornerstone in the process of validating nuclear data. Several different characteristics of a critical system can be measured. The most commonly considered is the so-called effective k eigenvalue, k{sub eff}. Another well-measured property of these systems is {alpha}{sub 0}, the inverse e-folding… more
Date: March 7, 2007
Creator: Pruet, J & Sleaford, B
Partner: UNT Libraries Government Documents Department
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Praise enhancements to include general strain hardening exponents and mid-life residual stress and water chemistry changes

Description: The purpose of this document is to describe some recent changes made to the PRAISE Code to provide some additional capabilities. The major changes are associated with the new capability to analyze cases where there is a mid-life change in residual stresses and/or water chemistry. Such changes have been proposed as a means of improving the reliability of BWR piping by reducing the oxygen content of the coolant (or other favorable chemistry changes) or altering the residual stresses near welds to… more
Date: June 7, 1990
Partner: UNT Libraries Government Documents Department
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SPR-9 concept definition study: FY 1986 summary

Description: This report summarizes the results of work undertaken to advance design of the SPR-9 MMW power reactor. Gas-cooled thermal reactors coupled with Brayton cycle power conversion appear feasible and have several attractive features for multi-megawatt space power systems. The high temperature capability of the reactor removes it as the limiting component on system performance. The low fuel inventories required make systems based on thermal reactors a lesser safeguard risk. However, an advanced radi… more
Date: October 7, 1986
Creator: Walter, C.E.
Partner: UNT Libraries Government Documents Department
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(Centralized reliability data organization for liquid metal reactor components)

Description: The Centralized Reliability Data Organization is a data bank and data analysis center, focusing on reliability, availability, and maintainability (RAM) data for components (e.g., valves, pumps, etc.) operating in liquid metal reactors and test facilities. PNC staff are using the CREDO data base as a resource for their development of a probabilistic safety assessment (PSA) for the MONJU fast breeder reactor now under construction. The meeting was held to discuss: (a) progress in FY 1989, (b) res… more
Date: February 7, 1991
Creator: Smith, M.S. & Manneschmidt, J.F.
Partner: UNT Libraries Government Documents Department
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Thermo-physics technical note No. 37: SNAP- 10A, Stainless Steel-316 vessel wall ablation. Final report

Description: The altitudes and times of ablation have been determined for the SNAP-10A, SS-316 vessel wall reentering under various conditions. The results are confined to one typical location on the reactor and to one typical reentry trajectory. The location is the side wall of the vessel and the trajectory is the one used in NAA-SR-8303.
Date: December 7, 1964
Creator: Montgomery, L.D.
Partner: UNT Libraries Government Documents Department
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Thermo-physics technical note No. 56: a parameter study of properties and mechanisms affecting the reentry ablation of SNAP fuel particles

Description: The results are presented of a parameter survey of the properties and mechanisms affecting the reentry ablation of SNAP fuel particles. The following properties and mechanisms were studied: aerodynamic heating, oxidation, oxide vaporization, metal vaporization, particle emissivity, specific heat, density, velocity, and initial size. These studies were made for particle release altitudes ranging from 170,000 to 228,000 feet.
Date: December 7, 1965
Creator: Sayles, C.W.
Partner: UNT Libraries Government Documents Department
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Standpipe-Bubbler pump level control study: 1E-3862, controlled clearance pump/system model analysis (Task A) W. P. 7373. Engineering memorandum 0190. [LMFBR]

Description: Computer simulation results of the Standpipe-Bubbler pump level control system/controlled clearance pump configuration 1E-3862 are presented. Fluid level control system behavior is presented in graphical form for normal plant loading, unloading, and trip transients, under varying conditions of cover gas mass flow rate, tank impedances, temperature, and flow rates through discharge seal and bearing leakage paths. Satisfactory fluid level control can be attained for this configuration with a Stan… more
Date: June 7, 1976
Creator: Salant, R.F. & Cook, M.E.
Partner: UNT Libraries Government Documents Department
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Interim report No.1, PT-105-403-P, installation and initial operation

Description: The effect of temperature on the properties of graphite is being determined by exposing samples to pile neutrons at controlled temperatures in the range 75 C to 225 C. The first series of samples were charged into the newly installed dry, water-cooled test channel at B pile on January 23, 1951. This report presents details of the charging operation and observations noted during the initial operation of the test.
Date: February 7, 1951
Creator: Haussler, W. M. & Purcell, R. H.
Partner: UNT Libraries Government Documents Department
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Production test IP-243-A-6-FP evaluation of X-8001 alloy aluminum components fabricated from cast blanks

Description: This test is designed to accomplish two primary objectives: (1) to attempt to verify ex-reactor corrosion data which indicated improved corrosion resistance of cast blank X-8001 alloy material compared with wrought blank, and (2) to attempt to verify the resistance to groove pitting type of corrosion attack previously observed on M-388 components. Ex-reactor tests of cast blank M-388 alloy in autoclaves using water as the corrosive media up to 360 C, and in flow loops up to 120 C have indicated… more
Date: April 7, 1959
Creator: Hall, R. E.
Partner: UNT Libraries Government Documents Department
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CPTF run 6 flux monitoring

Description: At the present stage of reactor coolant studies it is possible to say that radiation affects the deposition behavior of corrosion products in the reactor. A quantitative relation between flux and deposition rate or amount is not available, nor is there information as to exactly which component, or combination thereof, of the radiation is most important. To determine these effects and those of other process parameters, Battelle-Northwest (BNW) is conducting an experimental program for the Knolls… more
Date: May 7, 1971
Creator: Divine, J. R.
Partner: UNT Libraries Government Documents Department
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Assessment of very high temperature reactors in process applications

Description: In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of … more
Date: April 7, 1976
Creator: Jones, J. E. Jr.; Spiewak, I. & Gambill, W. R.
Partner: UNT Libraries Government Documents Department
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Structural integrity of materials in nuclear service: a bibliography

Description: This report contains 679 abstracts from the Nuclear Safety Information Center (NSIC) computer file dated 1973 through 1976 covering material properties with respect to structural integrity. All materials important to the nuclear industry (except concrete) are covered for mechanical properties, chemical properties, corrosion, fracture or failure, radiation damage, creep, cracking, and swelling. Keyword, author, and permuted-title indexes are included for the convenience of the user.
Date: June 7, 1977
Creator: Heddleson, F. A.
Partner: UNT Libraries Government Documents Department
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CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

Description: CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered… more
Date: January 7, 1977
Creator: Cleveland, J. C.
Partner: UNT Libraries Government Documents Department
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ERDA LWR plant technology program: role of government/industry in improving LWR performance

Description: Information is presented under the following chapter headings: executive summary; LWR plant outages; LWR plant construction delays and cancellations; programs addressing plant outages, construction delays, and cancellations; need for additional programs to remedy continuing problems; criteria for government role in LWR commercialization; and the proposed government program.
Date: October 7, 1975
Partner: UNT Libraries Government Documents Department
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Reference Design Description for a Geologic Repository

Description: One of the current major national environmental problems is the safe disposal of large quantities of spent nuclear fuel and high-level radioactive waste materials, which are rapidly accumulating throughout the country. These radioactive byproducts are generated as the result of national defense activities and from the generation of electricity by commercial nuclear power plants. At present, spent nuclear fuel is accumulating at over 70 power plant sites distributed throughout 33 states. The saf… more
Date: October 7, 2000
Partner: UNT Libraries Government Documents Department
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Critical heat flux experimentation in an annular test section. [PWR]

Description: Steady-state critical heat flux experiments have been performed in the Forced Convection Test Facility (FCTF), an annular test section containing a single electrically heated rod, for the purpose of testing the applicability of existing critical heat flux correlations. Good accuracy has been obtained using the MacBeth-Barnett critical heat flux correlation for annuli, corrected for the ''stepped cosine'' power profile of the heater. The equivalent diameter of the test section, based on the wett… more
Date: March 7, 1978
Creator: White, J. D. & Levin, A. E.
Partner: UNT Libraries Government Documents Department
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FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004)

Description: The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor powe… more
Date: July 7, 2011
Creator: Bailey, Erika
Partner: UNT Libraries Government Documents Department
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Laboratory Pressure Drop Data for Support Charges Using Pierced-Solid Dummy Slugs: All Reactors

Description: This report presents Hydraulics Laboratory data of the pressure drop characteristics for various combinations of ``pierced-solid`` dummy slugs in the support charges in the K, BDF, and C reactor process tubes. ``Pierced-solid`` slugs are regular solid aluminum dummy slugs which have been concentrically drilled so that mating against an I and E slug would not block the hole flow channel. No operating peculiarities were observed while using ``pierced-solids`` in the support charge.
Date: March 7, 1958
Creator: Waters, E. D.
Partner: UNT Libraries Government Documents Department
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