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Diffusion of gases in solids: rare gas diffusion in solids; tritium diffusion in fission and fusion reactor metals. Final report

Description: Major results of tritium and rare gas diffusion research conducted under the contract are summarized. The materials studied were austenitic stainless steels, Zircaloy, and niobium. In all three of the metal systems investigated, tritium release rates were found to be inhibited by surface oxide films. The effective diffusion coefficients that control tritium release from surface films on Zircaloy and niobium were determined to be eight to ten orders of magnitude lower than the bulk diffusion coe… more
Date: September 1, 1976
Creator: Abraham, P. M.; Chandra, D.; Mintz, J. M.; Elleman, T. S. & Verghese, K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Oxidation of 316 stainless steel and other alloys in prototypic GCFR environments

Description: The oxidation behavior of type 316 stainless steel and candidate advanced alloys for the gas-cooled fast reactor (GCFR) is being investigated at General Atomic Company. The test program consists of oxidation tests in prototypic GCFR environments. Two tests have been completed to date and a third test is under way. The first test was performed in an environment containing a hydrogen/water ratio of 10. The oxidation behavior of all the alloys was good to excellent in this environment. Preferentia… more
Date: May 1, 1980
Creator: Acharya, R.T.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Gas-cooled fast breeder reactor steady-state irradiation testing program

Description: The requirements for the gas-cooled fast breeder reactor irradiation program are specified, and an irradiation program plan which satisfies these requirements is presented. The irradiation program plan consists of three parts and includes a schedule and a preliminary cost estimate: (1) a steady-state irradiation program, (2) irradiations in support of the design basis transient test program, and (3) irradiations in support of the GRIST-2 safety test program. Data from the liquid metal fast bree… more
Date: August 1, 1980
Creator: Acharya, R.T.; Campana, R.J. & Langer, S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Survey of the Hazards Involved in Processing Liquid Metal Bonded Fuels

Description: A survey of the character and magnitude of hazards involved in processing liquid metal bonded fuels was made and the scope of a preliminary experimental program outlined. Processing of SRE and CPPD fuels by mechanical decladding followed by controlled reaction of the collected methods. Simdlarly, shearing of PRDC fuel and controlled exposure of the Na in the severed portions to water appears more desfrable than chethical dissolution of the metallic cladding. (auth)
Date: August 14, 1961
Creator: Adams, J. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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COMPARATIVE COST STUDY OF PROCESSING STAINLESS STEEL-JACKETED UO$sub 2$ FUEL: MECHANICAL SHEAR-LEACH VS SULFEX-CORE DISSOLUTION

Description: The economics of mechanical shear-leach and Sulfex decladding-core dissolution head end treatments for processing typical tubular bundles of stainless steel-jacketed UO/sub 2/ nuclear fuels were compared. A 2.66 metric ton U/day head end portion of a plant was designed for each process and capital and operating costs were developed. For plants of this size and larger, mechanical shear-leach processing has the advantage of ~20% lower capital cost and 50% lower operating cost. Processing costs of… more
Date: April 23, 1962
Creator: Adams, J. B.; Benis, A. M. & Watson, C. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Fuel Elements for the Argonne Advanced Research Reactor

Description: The core design and the fuel element concept for the high-flux Argonne Advanced Research Reactor are presented. The core is cooled and moderated by light water and utilizes beryllium as a reflector. The fuel element assembly is rhomboidal in cross section and consists of 27 plates fastened together at their edges by dovetailed locking keys, and at each end by end fittings. Each fuel plate is 40 mils thick and contains a uniform dispersion of highly enriched UO/ sub 2/ particles, up to a maximum… more
Date: January 1, 1962
Creator: Adolph, N.R.; Silberstein, M.S. & Weinstein, A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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EFFECT ON POWER COST OF SUBSTITUTING ALUMINUM FOR ZIRCONIUM FUEL CLADDING IN A BOILING WATER REACTOR

Description: The substitution of Al-clad for Zr-clad fuel necessitates reducing the operating temperature. The amount of saving that must be made in fabrication and chemical reprocessing to overcome the cost of operating at reduced thermal efficiency is estimated for a boiling water reactor. (auth)
Date: July 24, 1958
Creator: Albrecht, W.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Westinghouse Hanford Company recommended strategy for K Basin sludge disposition

Description: The objective of this document is to present the recommended strategy for removal of sludges from the K Basins. This document ties sludge removal activities to the plan for the K Basin spent nuclear fuel (SNF) described in WHC-EP-0830, Hanford Spent Nuclear Fuel Project Recommended Path Forward and is consistent with follow-on direction provided in February 1995. Solutions and processes for resolving sludge removal technical and management issues to meet accelerated K Basin deactivation objecti… more
Date: May 1, 1995
Creator: Alderman, C.J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Aluminum-Oxide Temperatures on the Mark VB, VE, VR, 15, and Mark 25 Assemblies

Description: The task was to compute the maximum aluminum-oxide and oxide-coolant temperatures of assemblies cladded in 99 plus percent aluminum. The assemblies considered were the Mark VB, VE, V5, 15 and 25. These assemblies consist of nested slug columns with individual uranium slugs cladded in aluminum cans. The CREDIT code was modified to calculate the oxide film thickness and the aluminum-oxide temperature at each axial increment. The information in this report will be used to evaluate the potential fo… more
Date: July 17, 2001
Creator: Aleman, S. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Aluminum-Oxide Temperatures on the Mark VB, VE, VR, 15, and Mark 25 Assemblies

Description: The task was to compute the maximum aluminum-oxide and oxide-coolant temperatures of assemblies cladded in 99+ percent aluminum. The assemblies considered were the Mark VB, VE, V5, 15 and 25. These assemblies consist of nested slug columns with individual uranium slugs cladded in aluminum cans. The CREDIT code was modified to calculate the oxide film thickness and the aluminum-oxide temperature at each axial increment. This information in this report will be used to evaluate the potential for c… more
Date: July 17, 2001
Creator: Aleman, S. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Effect of S8DR chromium diffusion treatment on the creep rupture properties of processed Hastelloy-N tubing

Description: Process development personnel have been seeking means of improving the (after thermal exposure) adhesion of the S8DR ceramic coating to chromized Hastelloy N tubing. This effort has led to the incorporation of an elevated temperature chromium diffusion treatment prior to the application of the ceramic coating. This report describes the results of tests performed to ascertain the influence of the additional thermal treatment on the 1400/sup 0/F creep-rupture properties of S8DR tubing.
Date: June 3, 1967
Creator: Allaria, G.G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Irradiation Effects Test Series: Test IE-2. Test results report. [PWR]

Description: The report describes the results of a test using four 0.97-m long PWR-type fuel rods with differences in diametral gap and cladding irradiation. The objective of this test was to provide information about the effects of these differences on fuel rod behavior during quasi-equilibrium and film boiling operation. The fuel rods were subjected to a series of preconditioning power cycles of less than 30 kW/m. Rod powers were then increased to 68 kW/m at a coolant mass flux of 4900 kg/s-m/sup 2/. Afte… more
Date: August 1977
Creator: Allison, C. M.; Croucher, D. W.; Ploger, S. A. & Mehner, A. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

SCDAP/RELAP5/MOD2 Code Manual

Description: The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A… more
Date: September 1, 1989
Creator: Allison, C. M.; Johnson, E. C.; Berna, G. A.; Cheng, T. C.; Hagrman, D. L.; Johnsen, G. W. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

PWR representative behavior during a LOCA

Description: To date, there has been substantial analytical and experimental effort to define the margins between design basis loss-of-coolant accident (LOCA) behavior and regulatory limits on maximum fuel rod cladding temperature and deformation. As a result, there is extensive documentation on the modeling of fuel rod behavior in test reactors and design basis LOCA's. However, modeling of that behavior using representative, non-conservative, operating histories is not nearly as well documented in the publ… more
Date: January 1, 1981
Creator: Allison, C.M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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NON-PRODUCTION FUELS REPROCESSING, CENTRIFUGATION STUDIES ON VARIOUS DISSOLVER EFFLUENT SOLUTIONS

Description: >The proposed flowsheets for reprocessing of nonproduction fuels include centrifugal separation of particulate matter from various dissolver effluent solutions. The settling characteristics of process solids were determined in water and in cold process solutions. Uranium dioxide particles will be recovered from Zirflex and Sulfex cladding waste solutions, and core-dissolver solutions will be centrifuged for removal of ZrO/sub 2/, metallic slimes, siliceous matter, and uranium-bearing solids.… more
Date: December 1, 1959
Creator: Amos, L.C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Surface and microstructural characterization of commercial breeder reactor candidate alloys exposed to 700/sup 0/C sodium

Description: Sodium compatibility screening tests were performed on several commercial austenitic alloys at 700/sup 0/C for 2000 hours for applications as breeder reactor fuel cladding. The sodium-exposed surfaces were characterized by Optical Metallography, Scanning Electron Microscopy (SEM) and Electron Probe Micro Analysis (EPMA). Sodium exposure generally resulted in the depletion of Ni, Cr, Ti, Si, Mn and Nb, and enrichment of Fe and Mo at the surface. The average thickness of the depleted zone was 5 .… more
Date: March 1, 1979
Creator: Anantatmula, R. P. & Brehm, W. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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