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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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Low strain diameter expansion of internally pressurized Zircaloy-4 tubing at high temperatures

Description: Tests of closed-end, internally pressurized, Zircaloy-4 tubing specimens were utilized to develop low strain creep characteristics as a function of time at temperatures in the range of 1475/sup 0/F to 2000/sup 0/F (802/sup 0/C to 1093/sup 0/C) and hoop stresses in the range of 250 to 2500 psi for use in loss-of-coolant accident (LOCA) analyses. The strain rate above the start of the alpha to beta phase transformation region, approximately 1490/sup 0/F (810/sup 0/C), was found to be sensitive to… more
Date: March 1978
Creator: White, L.S. & Busby, C.C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Susceptibility of unirradiated recrystallized Zircaloy-4 tubing to stress corrosion cracking

Description: Stress corrosion cracking (SCC) in unirradiated recrystallized Zircaloy-4 internally pressurized tubing specimens in atmospheres containing iodine vapor, cesium, or combinations of iodine and cesium is evaluated experimentally in terms of the effects of internal surface flaw morphology, iodine and cesium concentrations, tubing hydrogen content, test temperature, and test atmosphere water vapor content on the time to failure. The iodine vapor SCC data are analyzed in the framework of a fracture … more
Date: December 1977
Creator: Polan, N. W. & Tucker, R. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Nuclear materials research progress reports for 1979

Description: Research is presented concerning iodide stress corrosion cracking of zircaloy, self-diffusion of oxygen in hypostoichiometric urania, surface chemistry of epitaxial silicon deposition by thermal cracking of silane, kinetics of laser pulse vaporization of UO/sub 2/, gas laser model for laser induced evaporation, solubility of hydrogen in uranium dioxide, thermal gradient migration of metallic inclusions in UO/sub 2/, molecular beam studies of atomic hydrogen reduction of oxides, and thermal grad… more
Date: December 1, 1979
Creator: Olander, D.R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

Description: This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls … more
Date: November 1, 1977
Creator: Kullen, B J; Levitz, N M & Steindler, M J
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Compact nuclear power systems based on particle bed reactors

Description: Compact, low cost nuclear power systems with an extremely low radioactive inventory are described. These systems use the Particle Bed Reactor (PBR), in which HTGR particle fuel is contained in packed beds that are changed daily. The small diameter particle fuel (500 ..mu..m) is directly cooled utilizing the large heat transfer area available (7.8 m/sup 2//liter), thus allowing high bed power densities (MW/liter).
Date: January 1, 1986
Creator: Horn, F. L.; Powell, J. R.; Steinberg, M. & Takahashi, H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Load relaxation studies of zircaloy-4

Description: The results of the present work have shown (1) the load relaxation data of Zircaloy-4 in the temperature range of 200/sup 0/C to 385/sup 0/C can be represented by the phenomenological model based on Hart's plastic equation of state; (2) the room temperature load relaxation data reflect the effects of deformation twinning; (3) at higher temperatures the load relaxation data suggest the contribution of grain boundary sliding; and (4) the effects of strain aging can be identified based on the load… more
Date: January 1, 1976
Creator: Huang, F. H.; Sabol, G. P.; McDonald, S. G. & Li, C. Y.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Accurate determination of the parameters of the 292. 4-eV resonance of /sup 91/Zr and the 301. 1-eV resonance of /sup 96/Zr

Description: High-resolution transmission measurements of zirconium metal samples have been carried out at ORELA using the 80-m flight path and an improved /sup 6/Li-glass scintillation neutron detector. Four different thicknesses of the pure zirconium metal and one sample of zircaloy were used in four separate experiments. The transmission data for these samples were measured at room temperature, and one sample was cooled to liquid nitrogen temperature. These data were analyzed using the multilevel R-matri… more
Date: January 1, 1985
Creator: Salah, M.M.; Harvey, J.A.; Hill, N.W.; Hussein, A.Z. & Perey, F.G.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Cladding creepdown under compression. [BWR; PWR]

Description: Light-water power reactors use Zircaloy tubing as cladding to contain the UO/sub 2/ fuel pellets. In-service operating conditions impose an external hydrostatic force on the cladding, causing it to creep down into eventual contact with the fuel. Knowledge of the rate of such creepdown is of great importance to modelers of fuel element performance. An experimental system was devised for studying creepdown that meets several severe requirements by providing (1) correct stress state, (2) multiple … more
Date: November 9, 1977
Creator: Hobson, D.O.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Boundary effects on Zircaloy-4 cladding deformation in LOCA simulation tests. [PWR; BWR]

Description: Deformation behavior of Zircaloy-4 cladding under simulated loss-of-coolant accident (LOCA) conditions is being investigated in the Multirod Burst Test (MRBT) program in single rod and multirod tests. In these tests, internally-pressurized unirradiated Zircaloy-4 tubes containing internal electrical heaters are heated to failure in a low-pressure, superheated-steam environment (200 < Re < 800). The results provide a data base for evaluating deformation and blockage models employed with design-b… more
Date: January 1, 1982
Creator: Longest, A.W.; Chapman, R.H. & Crowley, J.L.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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ORNL irradiation creep facility

Description: A machine was developed at ORNL to measure the rates of elongation observed under irradiation in stressed materials. The source of radiation is a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). This choice allows experiments to be performed which simulate the effects of fast neutrons. A brief review of irradiation creep and experimental constraints associated with each measurement technique is given. Factors are presented which lead to the experimental choices ma… more
Date: September 1, 1980
Creator: Reiley, T. C.; Auble, R. L.; Beckers, R. M.; Bloom, E. E.; Duncan, M. G.; Saltmarsh, M. J. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Analysis of the ballooning deformation of an internally pressurized thin-wall tube during fast thermal transients

Description: A large-strain time-dependent thermoplastic analysis has been developed for the ballooning deformation of a thin-wall tube subjected to internal pressure, axial loading, and fast thermal transients. This deformation initiates with the onset of plastic instability in the material, the onset being determined by a plastic-instability criterion for strain-rate sensitive materials. The interaction among the local ballooning geometry, the state of stress, and the plastic flow process was considered, … more
Date: January 1, 1977
Creator: Lin, E.I.H.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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SSMS near surface analysis of B in irradiated Zircaloy-2: ion implantation standards as a calibration technique

Description: Purpose of this study was to determine the amount of /sup 10/B contamination on the surface of Zircaloy-2 clad irradiated fuel elements that had been stored in an aqueous solution containing 5000 wt. ppM enriched B. SMSS indicated that the contamination was less than 0.06 ..mu..g/cm/sup 2/. (DLC)
Date: January 1, 1980
Creator: Christie, W. H.; Carter, J. A.; Eby, R. E.; Landau, L. & Musick, W. R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Assessment of molten debris freezing in a severe RIA in-pile test. [PWR; BWR]

Description: An understanding of the freezing of molten debris on cold core structures following a hypothetical core meltdown accident in a light water reactor (LWR) is of importance to reactor safety analysis. The purpose of the present investigation was to analyze the transient freezing of the molten debris produced in a severe reactivity initiated accident (RIA) scoping test, designated RIA-ST-4, which was performed in the Power Burst Facility and simulated a BWR control rod drop accident. In the RIA-ST-… more
Date: January 1, 1980
Creator: El-Genk, Mohamed S. & Moore, Richard L.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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LWR Fuel Rod Behavior Observed During Postulated Accident Conditions: A Comparison of FRAP-T Calculated and PBF Experimental Results

Description: Light water reactor (LWR) fuel rod behavior during transient experiments conducted in the Power Burst Facility is reviewed. The experiments examined simulated hypothetical reactivity initiated accidents (RIA) and power-cooling-mismatch (PCM) events. Fuel rod behavior calculated by the Fuel Rod Analysis Program-Transient (FRAP-T) is compared with the test data. Important physical phenomena observed during the tests and not presently incorporated into the FRAP-T code are: (a) fuel swelling in the… more
Date: January 1, 1980
Creator: Charyulu, M. K.; Hobbins, R. R. & MacDonald, P. E.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Post-irradiation recovery of growth in Zircaloy. [Fast neutron irradiation]

Description: The previously reported model for fast neutron induced growth in Zircaloy has been modified to predict post-irradiation recovery of the growth. The predicted results are compared to growth recovery data obtained by Adamson. Agreement is good if the binding energy between vacancies and depleted zones is reduced from the in-pile value of 1.1 eV to a post-irradiation value of 1.0 eV. Such a reduction is reasonable.
Date: December 1977
Creator: Dollins, C. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Cladding hull decontamination and densification process. Part 2. Densification by inductoslag melting

Description: The Inductoslag melting process was developed to densify Zircaloy-4 cladding hulls. It is a cold crucible process that uses induction heating, a segmented water-cooled copper crucible, and a calcium fluoride flux. Metal and flux are fed into the furnace through the crucible, located at the top of the furnace, and the finished ingot is withdrawn from the bottom of the furnace. Melting rates of 40 to 50 kg/h are achieved, using 100 to 110 kW at an average energy use of 2.5 kWh/kg. The quality of … more
Date: April 1, 1980
Creator: Nelson, R. G. & Montgomery, D. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Effect of bundle size on cladding deformation in LOCA simulation tests. [PWR; BWR]

Description: Two LOCA simulation tests were conducted to investigate the effects of temperature uniformity and radial restraint boundary conditions on Zircaloy cladding deformation. In one of the tests (B-5), boundary conditions typical of a large array were imposed on an inner 4 x 4 square array by two concentric rings of interacting guard fuel pin simulators. In the other test (B-3), the boundary conditions were imposed on a 4 x 4 square array by a non-interacting heated shroud. Test parameters conducive … more
Date: January 1, 1982
Creator: Chapman, R.H.; Crowley, J.L. & Longest, A.W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Creepdown of Zircaloy fuel cladding: initial tests. [BWR; PWR]

Description: The report describes the initial creepdown tests of Zircaloy fuel cladding in which the surface displacements of the specimens have been measured as a function of time. Such measurements were made with a high accuracy, high precision eddy-current device capable of operation at elevated temperatures and pressures. Tests were conducted at 371/sup 0/C (700/sup 0/F) and external pressures of 14.5 to 17.2 MPa (2100 to 2500 psig) on specimens of reactor grade Zircaloy-4 fuel cladding in the unirradia… more
Date: April 1, 1978
Creator: Hobson, D.O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Irradiation effects test series, test IE-5. Test results report. [PWR]

Description: Test IE-5, conducted in the Power Burst Facility at the Idaho National Engineering Laboratory, employed three 0.97-m long pressurized water reactor type fuel rods, fabricated from previously irradiated zircaloy-4 cladding and one similar rod fabricated from unirradiated cladding. The objectives of the test were to evaluate the influence of simulated fission products, cladding irradiation damage, and fuel rod internal pressure on pellet-cladding interaction during a power ramp and on fuel rod be… more
Date: January 1, 1978
Creator: Croucher, D. W.; Yackle, T. R.; Allison, C. M. & Ploger, S. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods

Description: Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxi… more
Date: August 1985
Creator: Clayton, J. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Zircaloy-oxidation and hydrogen-generation rates in degraded-core accident situations

Description: Oxidation of Zircaloy cladding is the primary source of hydrogen generated during a degraded-core accident. In this paper, reported Zircaloy oxidation rates, either measured at 1500 to 1850/sup 0/C or extrapolated from the low-temperature data obtained at <1500/sup 0/C, are critically reviewed with respect to their applicability to a degraded-core accident situation in which the high-temperature fuel cladding is likely to be exposed to and oxidized in mixtures of hydrogen and depleted steam, ra… more
Date: February 1, 1983
Creator: Chung, H. M. & Thomas, G. R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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The Influence of the In-Situ Clad Staining on the Corrosion of Zircaloy in PWR Water Environment

Description: Zircaloy cladding tubes strain in-situ during service life in the corrosive environment of a Pressurized Water Reactor for a variety of reasons. First, the tube undergoes stress free growth due to the preferential alignment of irradiation induced vacancy loops on basal planes. Positive strains develop in the textured tubes along prism orientations while negative strains develop along basal orientations (Reference (a)). Second, early in life, free standing tubes will often shrink by creep in the… more
Date: June 21, 2001
Creator: Kammenzind, B.F., Eklund, K.L. and Bajaj, R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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