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Fuel pin failure in the PFR/TREAT experiments

Description: The PFR/TREAT safety testing programme involves the transient testing of fresh and pre-irradiated UK and US fuel pins. This paper summarizes the experimental and calculational results obtained to date on fuel pin failure during transient overpower (resulting from an accidental addition of resolivity) and transient undercooling followed by overpower (arising from an accidental stoppage of the primary sodium circulating pumps) accidents. Companion papers at this conference address: (I) the progress and future plans of the programme, and (II) post-failure material movements.
Date: January 1, 1986
Creator: Herbert, R.; Hunter, C.W.; Kramer, J.M.; Wood, M.H. & Wright, A.E.
Partner: UNT Libraries Government Documents Department

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 resultant fuel dispersals is described. Simple models are presented which describe observed cladding failures and pre-failure axial expansions yet are general enough to apply to all metal fuel types.
Date: January 1, 1988
Creator: Bauer, T.H.; Wright, A.E.; Robinson, W.R. & Klickman, A.E.
Partner: UNT Libraries Government Documents Department

Safety characteristics of the integral fast reactor concept

Description: The Integral Fast Reactor (IFR) concept is an innovative approach to liquid metal reactor design which is being studied by Argonne National Laboratory. Two of the key features of the IFR design are a metal fuel core design, based on the fuel technology developed at EBR-II, and an integral fuel cycle with a colocated fuel cycle facility based on the compact and simplified process steps made possible by the use of metal fuel. The paper presents the safety characteristics of the IFR concept which derive from the use of metal fuel. Liquid metal reactors, because of the low pressure coolant operating far below its boiling point, the natural circulation capability, and high system heat capacities, possess a high degree of inherent safety. The use of metallic fuel allows the reactor designer to further enhance the system capability for passive accommodation of postulated accidents.
Date: January 1, 1985
Creator: Marchaterre, J.F.; Cahalan, J.E.; Sevy, R.H. & Wright, A.E.
Partner: UNT Libraries Government Documents Department

Recent metal fuel safety tests in TREAT

Description: In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described.
Date: January 1, 1986
Creator: Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R. & Palm, R.G.
Partner: UNT Libraries Government Documents Department

First TREAT (Transient Reactor Test Facility) transient overpower tests on U-Pu-Zr fuel: M5 and M6

Description: Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) tests of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the Integral Fast Reactor concept. Similar tests on U-Fs fueled EBR-II driver pins were previously performed and reported (1,2). Results from these earlier tests indicated a margin to failure of about 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuel types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety related fuel performance characteristics of U-Fs and U-Pu-Zr. This report described conditions, results, and conclusions of testing of these fuel types.
Date: January 1, 1987
Creator: Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S. & Klickman, A.E.
Partner: UNT Libraries Government Documents Department

In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

Description: Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio.
Date: January 1, 1983
Creator: Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R. & Woods, W.J.
Partner: UNT Libraries Government Documents Department

PFR/TREAT Tests L04 and L06: irradiated versus fresh LMFBR fuel under TUCOP accident conditions

Description: Test L06, closely following L04 in the PFR/TREAT series, was a multi-pin simulation of a LMFBR transient under cooling/overpower (TUCOP) accident using full-length prototypic UK fast reactor fuel. In L04 the test fuel had been pre-irradiated to effect some fuel restructuring and fission gas retention. By contrast, in L06 the test fuel ws fresh. The pre-failure test fuel power and coolant flow histories in L04 were duplicated as closely as possible in L06 to make the L04/L06 pair a direct comparison of the performance of fresh versus irradiated fuel under TUCOP conditions. Identical 7-pin gridded test bundles in identically outfitted MK-3 integral flowing sodium loops also contributed to test environments in L04 and L06 that were as close as possible.
Date: January 1, 1984
Creator: Bauer, T.H.; Tylka, J.P.; Fink, C.L.; Stanford, G.S.; Wright, A.E. & Herbert, R.
Partner: UNT Libraries Government Documents Department

Comparison of L04, L05, and L07: three irradiated 7-pin bundle TUCOP tests. [LMFBR]

Description: Four transient-undercooling-driven overpower (TUCOP) tests on seven-pin bundles have been performed in the PFR/TREAT program. All were on full-length, bottom-plenum UK-design fuel. Three of them (tests L04, L05, and L07) tested sibling fuel elements having had the same preirradiation in PFR; one (L06) tested fresh fuel. The three tests on preirradiated fuel were designed to determine the differences in the motions of reactor-core materials that would result from the variation in power-to-flow mismatch conditions across the core of a commercial-size reactor during a hypothetical TUCOP accident. By initiating the overpower bursts at different fuel-coolant thermal-hydraulic states, the three tests yielded distinct differences in fuel and coolant response, providing a wide range of behavior useful in verifying accident models and codes.
Date: January 1, 1984
Creator: Wright, A.E.; Robinson, W.R.; Bauer, T.H.; Klickman, A.E.; Woods, W.J.; Cooper, A.A. et al.
Partner: UNT Libraries Government Documents Department

Review of recent ANL safety experiments in SLSF and TREAT. [LMFBR]

Description: Among the recent significant in-pile experiments conducted by ANL are Sodium Loop Safety Facility (SLSF) experiment P4 in the Engineering Test Facility (ETR) and TREAT experiments F3, F4, and J1. The P4 experiment, which had three heat-generating flow blockages each installed in six coolant channels in a 37-pin bundle of FTR (Fast Test Reactor)-type fuel elements, investigated the bounding consequences of severe local faults. The principal objectives were to eject molten fuel into the bundle geometry and, during subsequent extended operation, to characterize the behavior of (and response of instrumentation to) any subsequent blockage growth; secondary objectives included characterizing the severity of any molten-fuel/coolant interaction and the response of the coolant. The F3 and F4 experiments in TREAT were phenomenological tests to study the fuel-column disruption mode in loss-of-flow accidents. The J1 experiment was the first slow period (approx. 10 s) transient overpower experiment done in TREAT. Results of these experiments will be presented.
Date: January 1, 1982
Creator: Klickman, A.E.; Thompson, D.H.; Ragland, W.A.; Wright, A.E.; Palm, R.G. & Page, R.J.
Partner: UNT Libraries Government Documents Department

PFR/TREAT CO1 and LO1 experiments

Description: The CO1 and LO1 TREAT experiments were the first two in a series of international fast-reator safety experiments to be conducted as part of the PFR/TREAT program being performed by the US and the UK. Conditions for these two experiments simulated a large fast-reactor hypothetical 5$/second (150 ms period) transient overpower (TOP) accident. The CO1 configuration was a single fuel pin in a NaK filled capsule while LO1 used a seven-pin bundle in a flowing-sodium loop. The fuel pins in both experiments had no prior irradiation and represented the PFR driver fuel. The CO1 and LO1 experiments were performed on November 5 and November 24, 1980, respectively. This paper describes the objectives, test articles, experimental approach, and the major observations. Post-transient examination results and fuel-failure modeling studies are also presented.
Date: January 1, 1982
Creator: Tylka, J.P.; Wright, A.E.; Pember, L.A.; Culley, G.C.; Davies, A.L.; Herbert, R. et al.
Partner: UNT Libraries Government Documents Department

Instrument response during overpower transients at TREAT

Description: A program to empirically analyze data residuals or noise to determine instrument response that occurs during in-pile transient tests is out-lined. As an example, thermocouple response in the Mark III loop during a severe overpower transient in TREAT is studied both in frequency space and in real-time. Time intervals studied included both constant power and burst portions of the power transient. Thermocouple time constants were computed. Benefits and limitations of the method are discussed.
Date: January 1, 1982
Creator: Meek, C.C.; Bauer, T.H.; Hill, D.J.; Froehle, P.H.; Klickman, A.E.; Tylka, J.P. et al.
Partner: UNT Libraries Government Documents Department

Summary and Evaluation of Fuel Dynamics Transient-Overpower Experiments : Status 1974

Description: The report summarizes and evaluates experiments conducted in the Transient Reactor Test Facility (TREAT) using the Mark-II loop facility. The tests discussed are of the E and H series. Detailed descriptions of test conditions and test results as of February 1974 are presented. Since all data have not been acquired on all experiments, this report must be considered interim in nature. Particular emphasis is placed on data relevant to Fast Test Reactor (FTR) safety-analysis efforts.
Date: June 1977
Creator: Deitrich, L. W.; Doerner, R. C.; Hughes, T. H. & Wright, A. E.
Partner: UNT Libraries Government Documents Department

A risk characterization of safety research areas for Integral Fast Reactor program planning

Description: This paper characterizes the areas of Integral Fast Reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure of critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR Safety and related Base Technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorites.
Date: January 1, 1988
Creator: Mueller, C.J.; Cahalan, J.E.; Hill, D.J.; Kramer, J.M.; Marchaterre, J.F.; Pedersen, D.R. et al.
Partner: UNT Libraries Government Documents Department

Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

Description: Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs.
Date: January 1, 1990
Creator: Wright, A.E. (Argonne National Lab., IL (USA)); Dutt, D.S. (Westinghouse Hanford Co., Richland, WA (USA)) & Harrison, L.J. (Argonne National Lab., Idaho Falls, ID (USA))
Partner: UNT Libraries Government Documents Department

Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

Description: The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs.
Date: May 1, 1990
Creator: Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W. et al.
Partner: UNT Libraries Government Documents Department