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Analysis of metal fuel transient overpower experiments with the SAS4A accident analysis code

Description: The results of the SAS4A analysis of the M7 TREAT Metal fuel experiment are presented. New models incorporated in the metal fuel version of SAS4A are described. The computational results are compared with the experimental observations and this comparison is used in the interpretation of physical phenomena. This analysis was performed using the integrated metal fuel SAS4A version and covers a wide range of events, providing an increased degree of confidence in the SAS4A metal fuel accident analy… more
Date: January 1, 1990
Creator: Tentner, A. M.; Kalimullah & Miles, K. J.
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Partner: UNT Libraries Government Documents Department
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Analysis of reactivity-insertion accidents in the TREAT Upgrade reactor

Description: The expansion of the experimental capabilities of the TREAT Upgrade (TU) reactor also tends to increase the potential risks associated with off-normal reactivity insertion incidents compared to the TREAT reactor. To provide adequate prtection for the public and the facility, while meeting experimenter's requirements, a specialized Reactor Trip System (RTS) with energy-dependent scram trips on reactor power and period has been developed. With this protection strategy, the consequences of reactiv… more
Date: January 1, 1983
Creator: Rudolph, R.R. & Bhattacharyya, S.K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Analysis of the TREAT loss-of-flow tests L6 and L7 using SAS3D

Description: The TREAT loss-of-flow tests L6 and L7 have been analyzed using the SAS3D accident analysis code. The impetus for the analysis was the need for experimentally supported fuel motion modeling in whole core accident studies performed in support of licensing of the Clinch River Breeder Reactor Project. The input prescription chosen for the SAS3D/SLUMPY fuel motion model gave reasonable agreement with the test results. Tests L6 and L7, each conducted with a cluster of three fuel pins, were planned t… more
Date: January 1, 1985
Creator: Morris, E. E.; Simms, R. & Gruber, E. E.
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Partner: UNT Libraries Government Documents Department
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Analytical and experimental investigation of the dispersion process during rapid transients for the aluminum-based nuclear fuel plates

Description: A thermally induced fuel-plate dispersion model was developed to analyze for dispersive potential and determine the onset of fuel plate dispersion for aluminum-based research and test reactor fuels. The effect of rapid energy deposition in a fuel plate was simulated. Several data types for aluminum-based fuels tested in the Nuclear Safety Research Reactor (NSRR) facility in Japan and in the Transient Reactor Test (TREAT) facility in Idaho, US, were reviewed. Analyses of experiments show that th… more
Date: June 1, 1995
Creator: Georgevich, V.; Taleyarkhan, R. P.; Kim, S. H.; Fuketa, T.; Soyama, K. & Ishijima, K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Application of the INSTANT-HPS PN Transport Code to the C5G7 Benchmark Problem

Description: INSTANT is the INL's next generation neutron transport solver to support high-fidelity multi-physics reactor simulation INSTANT is in continuous development to extend its capability Code is designed to take full advantage of middle to large cluster (10-1000 processors) Code is designed to focus on method adaptation while also mesh adaptation will be possible. It utilizes the most modern computing techniques to generate a neutronics tool of full-core transport calculations for reactor analysis a… more
Date: June 1, 2011
Creator: Wang, Y.; Zhang, H.; Szilard, R. H. & Martineau, R. C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Axial location of cladding failure during a slow transient overpower TREAT test. [LMFBR]

Description: The axial location of cladding failure following a transient overpower accident is of importance in fast reactor safety studies in that it is a determining factor in the relocation of fuel, and therefore in the possibility of inherent neutronic shutdown of the reactor. In-pile experimental data on the axial location of cladding failure of fuel in bundles of pins is sparse since, in general, the experimental fuel pin bundles are largely destroyed during the in-pile test. The post-test examinatio… more
Date: January 1, 1983
Creator: Page, R. J.; Murphy, W. F. & Holland, J. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Behavior of metallic fuel in treat transient overpower tests

Description: Results and analyses are reported for TREAT in-pile transient overpower tests of margin to cladding failure and pre-failure axial expansion of metallic fuel. In all cases the power rise was exponential on an 8 s period until either incipient or actual cladding failure was achieved. Test fuel included EBR-II driver fuel and ternary alloy, the reference fuel of the Intergral Fast Reactor concept. Test pin burnup spanned the widest range available. The nature of the observed cladding failure and r… more
Date: January 1, 1988
Creator: Bauer, T. H.; Wright, A. E.; Robinson, W. R. & Klickman, A. E.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Chemical characteristics of material released during Source Term Experiments Project (STEP) in-pile tests: Part 1

Description: A series of four experiments aimed at characterizing the radiological source term associated with postulated severe light water reactor (LWR) accidents has been conducted at Argonne's Transient Reactor Test Facility (TREAT). The STEP tests drove fuel elements to the point of severe cladding disruption in steam environments by fission heating and oxidation of the Zircaloy cladding. The released fission products and volatile cladding constituents entrained in the steam/hydrogen flows were capture… more
Date: September 1, 1986
Creator: Schlenger, B. J.; Dunn, P. F.; Herceg, J. E.; Simms, R.; Horton, E. L.; Baker, L., Jr. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Common cause analysis of the TREAT upgrade reactor protection system

Description: A triply redundant reactor scram system (RSS) has been designed for the upgraded TREAT facility. The independent failures reliability goal for the RSS is <10/sup -9/ failures per demand. An independent failures analysis indicated that this goal would be met. In addition, however, recognizing that in heavily redundant systems common-cause failures dominate, a common cause analysis of the TREAT upgrade RSS was done. The objective was to identify those common-cause initiators which could affect th… more
Date: September 1, 1984
Creator: Page, R.J.; Kamis, G.J.; Marbach, R.A. & Mueller, C.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Comparative analysis of internal fuel motion in annular fuel designs

Description: In this paper, the whole-core reactivity consequences of internal fuel motion in three annular fuel designs during a hypothetical 3 dollars/s transient overpower (TOP) accident are compared to determine the effect of geometric design variations. The PINEX-2 and PINEX-3 experiments were performed in the TREAT reactor using annular fuel pins irradiated in GETR. This paper investigates three combinations of solid and annular axial blankets and fission gas plena: top annular blanket and plenum, bot… more
Date: September 16, 1983
Creator: Smith, D. E.
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Partner: UNT Libraries Government Documents Department
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Core design of the upgraded TREAT reactor

Description: The upgrading of the TREAT reactor involves the replacement of the central 11 x 11 subzone of the 19 x 19 fuel assembly array by new, Inconel-clad, high-temperature fuel assemblies, and the additions of a new reactor control system, a safety-grade plant protection system, and an enhanced reactor filtration/coolant system. The final design of these modifications will be completed in early 1983. The TREAT facility is scheduled to be shut down for modification in mid-1984, and should resume the sa… more
Date: January 1, 1982
Creator: Wade, D.C.; Bhattacharyya, S.K.; Lipinski, W.C. & Stone, C.C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Coupled hydrodynamic-structural analysis of an integral flowing sodium test loop in the TREAT reactor

Description: A hydrodynamic-structural response analysis of the Mark-IICB loop was performed for the TREAT (Transient Reactor Test Facility) test AX-1. Test AX-1 is intended to provide information concerning the potential for a vapor explosion in an advanced-fueled LMFBR. The test will be conducted in TREAT with unirradiated uranium-carbide fuel pins in the Mark-IICB integral flowing sodium loop. Our analysis addressed the ability of the experimental hardware to maintain its containment integrity during the… more
Date: January 1, 1979
Creator: Zeuch, W.R. & A-Moneim, M.T.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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DENBIG and DARN. Codes for the reduction of digital pinex data

Description: Code DENBIG was written to reduce the digital pinex data received from the Nevada Test Site (NTS) and Transient Reactor Test Facility (TREAT) experiments. Output modes are density plots, contour plots, splatter plots, equivalent radius calculations, line cuts, and reduced data in matrix form. Provisions are included to incorporate subtraction of dry-run data, transfer function corrections, and pinhole absorption corrections. Code DARN processes pinex data in which electromagnetic pulse shifts h… more
Date: October 1, 1977
Creator: Beyer, L.M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Design analysis of the upgraded TREAT reactor

Description: The TREAT reactor, fueled by a dilute dispersion of fully enriched UO/sub 2/ in graphite, has been a premier transient testing facility since 1959. A major Upgrade of the reactor is in progress to enhance its transient testing capability in support of the LMFBR safety program. The TREAT Upgrade (TU) reactor features a modified central zone of the core with higher fissile loadings of the same fuel, clad in Inconel to allow operation at higher temperatures. The demanding functional requirements o… more
Date: January 1, 1982
Creator: Bhattacharyya, S.K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Determination of optimal core fissile loadings in the TREAT Upgrade reactor

Description: The TREAT Upgrade (TU) reactor design is presently nearing completion. The reactor will be used to test LMFBR fuel under simulated accident conditions. The physics of the TU core is complicated by a number of factors related to the planned application of the facility. In this paper the design approach used to produce the core fissile loading spatial distribution needed to satisfy the requirement to test a number of different test clusters in various test loops in a transient operating mode is d… more
Date: January 1, 1983
Creator: Bhattacharyya, S. K.; Lell, R. M.; Ulrich, A. J. & Yang, S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Determination of the design excess reactivity for the TREAT Upgrade reactor

Description: The excess reactivity designed to be built into a reactor core is a primary determinant of the fissile loadings of the fuel rods in the core. For the TREAT Upgrade (TU) reactor the considerations that enter into the determination of the excess reactivity are different from those of conventional power reactors. The reactor is designed to operate in an adiabatic transient mode for reactor safety in-pile test programs. The primary constituent of the excess reactivity is the calculated reactivity r… more
Date: January 1, 1983
Creator: Bhattacharyya, S.K. & Hanan, N.A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Developments in the LASL Fuel Pin Imaging System: PINEX-3A

Description: The LASL Fuel Pin Imaging System was evaluated using a series of 10 TREAT transients, each of approx. 240-MW peak power. HEDL provided the fuel-ejection type capsule with annular fuel pellets. The pin visibility threshold was determined to be approx. 20-MW of TREAT power (approx. 130 W/g), almost an order of magnitude improvement over our PINEX-2 threshold. The impact of changes in instrumentation, imaging apertures, and fluors that produced the improved sensitivity are reported. Results of a t… more
Date: January 1, 1979
Creator: Lumpkin, A. H.; Berzins, G. J.; Cosimi, R. A.; O'Hare, T. E. & Davidson, J. R.
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Partner: UNT Libraries Government Documents Department
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Direct hodoscope monitoring of flux and power-coupling at TREAT

Description: The most important parameter in calibrating reactor-safety experiments at TREAT is test-fuel energy deposition during irradiation transients. The coupling of the reactor power to the test fuel depends on conditions that vary from experiment to experiment and vary during the transient. In order to more accurately determine the power coupling and better understand its time and space dependencies, data has been obtained by the fast-neutron hodoscope and compared to experiment calibrations, reactor… more
Date: January 1, 1983
Creator: DeVolpi, A.; Fink, C.L.; Rhodes, E.A. & Stanford, G.S.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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The effect of carbon crystal structure on treat reactor physics calculations

Description: The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory-West (ANL-W) is fueled with urania in a graphite and carbon mixture. This fuel was fabricated from a mixture of graphite flour, thermax (a thermatomic carbon produced by ''cracking'' natural gas), coal-tar resin and U/sub 3/O/sub 8/. During the fabrication process, the fuel was baked to dissociate the resin, but the high temperature necessary to graphitize the carbon in the thermax and in the resin was avoided. Therefore… more
Date: January 1, 1988
Creator: Swanson, R. W. & Harrison, L. J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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