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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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Supporting Analysis for Thermal Suitability of Fuel Elements for SM-1A Core I Loading

Description: A recommended SM-1A Core I loading chart was derived from available, metallurgically acceptable elements at the SM-1A and SM-1 sites. The derivation was based on local thermal and hydraulic considerations of minimum elementto- element coolant channel clearances. These clearances were determined from field inspection measurements of outer fuel plate spacing, as modified by analytical calculations of plate ripple growth during exposure to reactor operating thermal stresses. (auth)
Date: January 10, 1962
Creator: Brondel, J. O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Analysis of LOFT pressurizer spray and surge nozzles to include a 450/sup 0/F step transient

Description: This report presents the analysis of the LOFT pressurizer spray and surge nozzles to include a 450/sup 0/F step thermal transient. Previous analysis performed under subcontract by Basic Technology Incorporated was utilized where applicable. The SAASIII finite element computer program was used to determine stress distributions in the nozzles due to the step transient. Computer results were then incorporated in the necessary additional calculations to ascertain that stress limitations were not ex… more
Date: January 18, 1978
Creator: Nitzel, M.E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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PNL technical review of pressurized thermal-shock issues. [PWR]

Description: Pacific Northwest Laboratory (PNL) was asked to develop and recommend a regulatory position that the Nuclear Regulatory Commission (NRC) should adopt regarding the ability of reactor pressure vessels to withstand the effects of pressurized thermal shock (PTS). Licensees of eight pressurized water reactors provided NRC with estimates of remaining effective full power years before corrective actions would be required to prevent an unsafe operating condition. PNL reviewed these responses and the r… more
Date: July 1, 1982
Creator: Pedersen, L. T.; Apley, W. J.; Bian, S. H.; Defferding, L. J.; Morgenstern, M. H.; Pelto, P. J. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Detailed Stress Analysis of SM-1 Steam Generator Tube Sheet

Description: The detailed stress analysis of the SM-1 steam generator tube sheet showed it to be safe from strain cycling damage. However, the pressure stresses were greater than the yield strength during the hydrostatic test. The differential between pressure stresses and yield strength indicates that some initial deformation may have taken place in the tube sheet. (auth)
Date: July 11, 1962
Creator: Busuttil, J. J. & Chittum, R. A.
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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Thermal-Stress and Strain-Fatigue Analyses of the MSRE Fuel and Coolant Pump Tanks

Description: Thermal-stress and strain-fatigue analyses of the MSRE fuel and coolant pump tanks were completed for determining the quantity of coollng air required to obtain the maximum life of the pump tanks and to determine the acceptability of the pump tanks for the intended service of 100 heating cycles from room temperature to 1200 deg F and 500 reactor power-change cycles from zero to 10 Mw. A cooling-air flow rate of 200 cfm for the fuel pump tank was found to be an optimum value that provided an amp… more
Date: October 1, 1962
Creator: Gabbard, C.G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Thermal strains in titanium aluminide and nickel aluminide composites

Description: Neutron diffraction was used to measure residual thermal strains developed during postfabrication cooling in titanium aluminide and nickel aluminide intermetallic matrix composites. Silicon carbide /Ti 14Al-21Nb, tungsten and sapphire/NiAl, and sapphire and SiC-coated sapphire/NiAl{sub 25}Fe{sub 10} composites were investigated. The thermal expansion coefficient of the matrix is usually greater than that of the fibers. As such, during cooldown, compressive residual strains are generated in the … more
Date: January 1, 1992
Creator: Saigal, A. (Tufts Univ., Medford, MA (United States). Dept. of Mechanical Engineering) & Kupperman, D.S. (Argonne National Lab., IL (United States))
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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DEVELOPMENT OF RING-JOINT FLANGES FOR USE IN THE HRE-2

Description: Ring-joint flanges were studied in thermal-cycle tests as part of the development work associated with Homogeneous Reactor Experiment No. 2 (HRE-2). The purpose of this study was to provide criteria for design, installation, and operation of joints that would remain leaktight under reactor operating temperatures and pressures. Joints ranging from 1/2 in., l500 lb to 4 in., 2500 lb and with various initial bolt loadings were cycled between room temperature and 636 deg F. It was demonstrated that… more
Date: December 21, 1961
Creator: Robinson, J. N.; Lundin, M. I. & Spiewak, I.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Seismic velocities and attenuation in an underground granitic waste repository subjected to heating

Description: The behavior of a granitic rock mass subjected to thermal load has been studied by an acoustic cross-hole technique between four boreholes, over a period of some two years. Velocities between boreholes were obtained from the times-of-flight of pulses of acoustic waves between transducers clamped to the borehole wall. The attenuation was obtained by a spectral ratios technique. When the heater was turned on, the velocities increased rapidly to an asymptotic value. When the heater was turned off,… more
Date: March 1, 1984
Creator: Paulsson, B.N.P. & King, M.S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Test PCM-5 rod bowing and bow direction reversal. [PWR]

Description: Test PCM-5 was the first bundle test in the PCM Test Series being conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc. as part of the Nuclear Regulatory Commission's Fuel Behavior Program. The experiment was performed in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory. The bundle consisted of nine previously unirradiated PWR-type fuel rods, arranged in a 3 x 3 array within a square cross section flow shroud, with rod-to-rod spacing typical … more
Date: January 1, 1980
Creator: Kerwin, D. K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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MHD air heater development technology. Progress report, November 26, 1979-March 31, 1980

Description: Work on the development of the directly-fired high temperature air heater (HTAH) for MHD power plants is reported. Progress is reported on three tasks: (1) materials selection, evaluation, and development, (2) operability, performance, and materials testing, and (3) full-scale design concepts. Under Task 1, efforts were carried out in several areas. Work on the computer data base for material properties was begun. Data were compiled for several HTAH materials. Materials selections for Valve Tes… more
Date: May 1, 1980
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Thermal-stress analysis of a Fort St. Vrain core-support block under accident conditions

Description: A thermoelastic stress analysis of a graphite core support block in the Fort St. Vrain High Temperature Gas Cooled Reactor is described. The support block is subjected to thermal stresses caused by a loss of forced circulation accident of the reactor system. Two- and three-dimensional finite element models of the core support block are analyzed using the ADINAT and ADINA codes, and results are given that verify the integrity of this structural component under the given accident condition.
Date: January 1, 1982
Creator: Carruthers, L. M.; Butler, T. A. & Anderson, C. A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

FUEL ELEMENT DEVELOPMENT PROGRAM FOR THE PEBBLE BED REACTOR. Final Report

Description: >The basic fuel element consisted of a uniform dispersion of fuel in a 1 1/2 inch diameter graphite sphere. Ceramic coatings for the retention of fission products were studied. It was found-that molecularly deposited'' ceramics such as alumina, siliconized silicon carbide, and pyrolytic carbon were excellent barriers to fission product leakage. The most advantageous location for ceramic coatings was found to be on the individual fuel particles, where the coating was subject to smaller forces an… more
Date: April 30, 1961
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Postirradiation cladding strength under biaxial loading with an increasing temperature ramp. [LMFBR]

Description: The flow behavior of unirradiated 20% cold worked AISI 316 tubing during constant pressure, increasing temperature tests was modeled with a constitutive relation approach; strain below approximately 0.2% came predominantly from an anelastic portion of the model while higher strains were predominantly plastic. The flow of cladding sections from irradiated fuel pins was largely restricted to the strain region attributed to anelastic deformation due to reduced ductility compared to unirradiated tu… more
Date: April 1, 1980
Creator: Duncan, D. R. & Hunter, C. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Thermal responses of tokamak reactor first walls during cyclic plasma burns

Description: The CINDA-3G computer code has been adapted to analyze the thermal responses and operating limitations of two fusion reactor first-wall concepts under normal cyclic operation. A component of an LMFBR computer has been modified and adapted to analyze the ablative behavior of first-walls after a plasma disruption. The first-wall design concepts considered are a forced-circulation water-cooled stainless steel panel with and without a monolithic graphite liner. The thermal gradients in the metal wa… more
Date: January 1, 1977
Creator: Smith, D. L. & Charak, I.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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COOLING OF THE HFIR BERYLLIUM REFLECTOR FOLLOWING A REACTOR SCRAM OR AN ELECTRICAL POWER OUTAGE

Description: Thermal stresses in the HFIR beryllium reflector were computed for the unlikely case where the reactor is scrammed with a simultaneous loss of coolant flow and for the case following an electrical power outage where the reactor power level and the coolant flow rate are reduced simultaneously. For the case where the reactor is scrammed with a sudden loss of the coolant flow, the resulting maximum tensile thermal stress following the scram is 22,500 psi. In case of an electrical power outage, the… more
Date: December 12, 1961
Creator: McLain, H. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Thermal Stress Testing of Type 1 Fuel Plates

Description: Thermal stress tests on Type 1, SM-1A Core II fuel ele-ment sections were performed to study plate distortion and determine its dependency on temperature distribution, temperature differential, initial flatness, and ripple length. Test results will be correlated with the analytical model and used to predict ripple growth in other plate-type fuel elements. The tests showed that ripple growth is dependent on initial flatness of the plate and that the characteristic shape of ripples is maintained … more
Date: June 27, 1962
Creator: Gebhardt, F. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Time-dependent behavior of concrete

Description: This paper is a condensed version of the material presented at the International Workshop on Finite Element Analysis of Reinforced Concrete, Session 4 -- Time Dependent Behavior, held at Columbia University, New York on June 3--6, 1991. Dr. P.A. Pfeiffer presented recent developments in time-dependent behavior of concrete and Professor T. Tanabe presented a review of research in Japan on time-dependent behavior of concrete. The paper discusses the recent research of time-dependent behavior of c… more
Date: January 1, 1992
Creator: Pfeiffer, P.A. (Argonne National Lab., IL (United States)) & Tanabe, Tada-aki (Nagoya Univ. (Japan). Dept. of Civil Engineering)
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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