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Assessment of light water reactor fuel damage during a reactivity initiated accident

Description: This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalp… more
Date: January 1, 1980
Creator: MacDonald, P. E.; Seiffert, S. L.; Martinson, Z. R.; McCardell, R. K.; Owen, D. E. & Fukuda, S. K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Severe accident core heatup transients in modular high temperature gas-cooled reactors without operating Reactor Cavity Cooling Systems

Description: The ultimate decay heat removal system for the current Modular High Temperature Gas-Cooled reactors is a completely passive natural convection air cooling loop. This paper considers an extremely remote accident scenario, where even this passive system fails, and heat rejection is only via a layer of thermal insulation to the reactor silo structure and the surrounding soil. The results show that even in this case the peak fuel temperatures remain well within safe limits. However, vessel and conc… more
Date: January 1, 1988
Creator: Kroeger, P. G.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Fuel performance annual report for 1989

Description: This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulatory Commission evaluations are included.
Date: June 1, 1992
Creator: Bailey, W.J.; Berting, F.M. (Pacific Northwest Lab., Richland, WA (United States)) & Wu, S. (Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology)
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

Description: Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of i… more
Date: January 1, 1982
Creator: Wilson, W. B.; LaBauve, R. J. & England, T. R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, December 1, 1979-February 29, 1980

Description: Information is presented concerning bundle geometry with wrapped and bare rods; LMFBR outlet plenum flow mixing; and theoretical determination of local temperature fields in LMFBR fuel rod bundles.
Date: January 1, 1980
Creator: Todreas, N.E.; Golay, M.W. & Wolf, L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Analysis of fission product behavior in the Saclay Spitfire Loop Test SSL-1. [HTGR]

Description: The behavior of the fission metal cesium and the fission gases krypton and xenon in the Saclay Spitfire Loop SSL-1 test has been compared to that predicted using General Atomic reference data and computer code models. This is the first in a series of analyses planned in order to provide quantitative validation of HTGR fission product design methods. In this analysis, the first attempt to rigorously verify fission product design methods, the FIPERQ code was used to model the diffusion of cesium … more
Date: February 1, 1978
Creator: Jensen, D. D.; Haire, M. J. & Ballagny, A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Design configuration of GCFR core assemblies

Description: The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the … more
Date: May 1, 1980
Creator: LaBar, M.P.; Lee, G.E. & Meyer, R.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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FLOWTRAN benchmarking with onset of flow instability data from 1988 Columbia University single-tube OFI experiment

Description: Benchmarking FLOWTRAN, Version 16.2, with an Onset of Significant Voiding (OSV) criterion against measured Onset of Flow Instability (OFI) data from the 1988--89 Columbia University downflow tests has shown that FLOWTRAN with OSV is a conservative OFI predictor. Calculated limiting flow rates based on the Savannah River Site (SRS) OSV criterion were always higher than the measured flow rates at OFI. This work supplements recent FLOWTRAN benchmarking against 1963 downflow tests at Columbia Unive… more
Date: June 1, 1990
Creator: Chen, K.; Paul, P. K. & Barbour, K. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR]

Description: The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminar… more
Date: October 1, 1979
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Criticality experiments with mixed oxide fuel pin arrays in plutonium-uranium nitrate solution

Description: A series of critical experiments was completed with mixed plutonium-uranium solutions having a Pu/(Pu + U) ratio of approximately 0.22 in a boiler tube-type lattice assembly. These experiments were conducted as part of the Criticality Data Development Program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. A complete description of the experiments and data are included in this report. The experiments were perf… more
Date: August 1, 1988
Creator: Lloyd, R.C. (Pacific Northwest Lab., Richland, WA (United States)) & Smolen, G.R. (Oak Ridge National Lab., TN (United States))
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Summary of HTGR (high-temperature gas-cooled reactor) benchmark data from the high temperature lattice test reactor

Description: The High Temperature Lattice Test Reactor (HTLTR) was a unique critical facility specifically built and operated to measure variations in neutronic characteristics of high temperature gas cooled reactor (HTGR) lattices at temperatures up to 1000{degree}C. The Los Alamos National Laboratory commissioned Pacific Northwest Laboratory (PNL) to prepare this summary reference report on the HTLTR benchmark data and its associated documentation. In the initial stages of the program, the principle of th… more
Date: October 1, 1989
Creator: Newman, D.F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Fission product release from high gap-inventory LWR fuel under LOCA conditions

Description: Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200/sup 0/C. The basis for the model was tes… more
Date: January 1, 1980
Creator: Lorenz, R. A.; Collins, J. L.; Osborne, M. F. & Malinauskas, A. P.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Prototype arc saw design and cutting trials

Description: A program was initiated to develop the arc saw as a tool capable of removing the end fittings from spent nuclear fuel bundles. A special arc saw for this purpose was designed, installed at the Pacific Northwest Laboratory and satisfactorily operated to remove end fittings from simulated, nonradioactive fuel bundles. The design of the arc saw included consideration of the cutting environment, power supply size, control equipment, and work piece size. Several simulated fuel bundles were cut to de… more
Date: September 1, 1980
Creator: Allison, G.S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Quarterly Progress Report on Fission Product Behavior in LWRs for the Period October-December 1977

Description: Analysis of release data obtained during High Burnup Fuel Test 10 (HBU-10) has been completed. In this test the fuel rod segment was ruptured by internal pressurization at 900/sup 0/C, at which time the temperature was rapidly increased to 1200/sup 0/C and maintained at this temperature for 10 min. Approximately 0.061% of the total cesium inventory in the rod segment was released; this was accompanied by the release of about 1.69% of the total /sup 85/Kr inventory. Moreover, about 0.022% of the… more
Date: February 1, 1978
Creator: Malinauskas, A. P.; Lorenz, R. A.; Collins, J. L.; Osborne, M. F.; Whatley, S. K. & Towns, R. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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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.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Analysis of the thorium axial blanket experiments in the PROTEUS reactor

Description: An extensive program of reactor physics experiments in GCFR fuel pin lattices has been completed recently at the PROTEUS critical facility located at EIR laboratory in Switzerland. The PROTEUS reactor consists of a central test zone surrounded by a uranium buffer and thermal driver region. The test lattices included a PuO/sub 2//UO/sub 2/ fuel region with internal and axial blankets of UO/sub 2/, ThO/sub 2/, and thorium metal. Detailed analysis of the thorium-bearing lattices has been performed… more
Date: January 1, 1980
Creator: White, J.R.; Ingersoll, D.T. & Schmocker, U.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Integral fast reactor safety features

Description: The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on… more
Date: January 1, 1988
Creator: Cahalan, J. E.; Kramer, J. M.; Marchaterre, J. F.; Mueller, C. J.; Pedersen, D. R.; Sevy, R. H. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Nondestructive evaluation of LWR spent fuel shipping casks

Description: An analysis of nondestructive testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined.
Date: February 1, 1978
Creator: Ballard, D. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor

Description: The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process bec… more
Date: January 1, 1980
Creator: Welty, R.K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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HTGR fuel recycle program. Quarterly progress report for the period ending November 30, 1977

Description: The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel, the design and development of an integrated pilot line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials, and design work in support of Hot Engineering Tests (HET). Work is also described on trade-off studies concerning the required design of facilities and equipment for the large-scale recycle of HTGR fuels in order to … more
Date: December 1, 1977
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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