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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 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

Progress and future plans in the PFR/TREAT safety testing programs

Description: This paper briefly describes the progress to date on the joint UKAEA/USDOE program of fast reactor fuel safety testing and the definition of the future tests. The program involves transient tests in the TREAT reactor on fresh and irradiated mixed fuel pins. The tests simulate transient overpower (TOP) accidents, which result from an unintentional addition of reactivity and transient undercooling followed by overpower (TUCOP) accidents, which arise from an unintentional stoppage of the primary sodium circulating pumps, both with failure to scram. Thirteen tests have been performed to date, all on UK pins. Future plans include five tests, all on US pins which have been irradiated in FFTF. Much has been learned about the behavior of fuel driven to conditions well beyond those existing during normal reactor operation.
Date: January 1, 1986
Creator: Alter, H.; Cowking, C.B.; Klickman, A.E. & Wood, M.H.
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

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

Summary of Treat Experiments on Oxide Core-Disruptive Accidents

Description: A program of transient in-reactor experiments is being conducted by Argonne National Laboratory in the Transient Reactor Test (TREAT) facility to guide and support analyses of hypothetical core-disruptive accidents (HCDA) in liquid-metal fast breeder reactors (LMFBR). Test results provide data needed to establish the response of LMFBR cores to hypothetical accidents producing fuel failure, coolant boiling, and the movement of coolant, molten fuel, and molten cladding. These data include margins to fuel failure, the modes of failure and movements, and evidence for identification of the mechanisms which determine the failure and movements. A key element in the program is the fast-neutron hodoscope, which detects fuel movement as a function of time during experiments.
Date: February 1979
Creator: Dickerman, Charles Edward; Rothman, Alan B.; Klickman, A. E.; Spencer, B. W. & DeVolpi, Alexander
Partner: UNT Libraries Government Documents Department

Posttest examination results of recent treat tests on metal fuel

Description: A series of in-reactor transient tests is underway to study the characteristics of metal-alloy fuel during transient-overpower-without-scam conditions. The initial tests focused on determining the margin to cladding breach and the axial fuel motions that would mitigate the power excursion. The tests were conducted in flowing-sodium loops with uranium - 5% fissium EBR-II Mark-II driver fuel elements in the TREAT facility. Posttest examination of the tests evaluated fuel elongation in intact pins and postfailure fuel motion. Microscopic examination of the intact pins studied the nature and extent of fuel/cladding interaction, fuel melt fraction and mass distribution, and distribution of porosity. Eutectic penetration and failure of the cladding were also examined in the failed pins.
Date: January 1, 1986
Creator: Holland, J. W.; Wright, A. E.; Bauer, T. H.; Goldman, A. J.; Klickman, A. E. & Sevy, R. H.
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

Summary of treat experiments on oxide core-disruptive accidents. [LMFBR]

Description: A program of transient in-reactor experiments is being conducted by Argonne National Laboratory in the Transient Reactor Test (TREAT) facility to guide and support analyses of hypothetical core-disruptive accidents (HCDA) in liquid-metal fast breeder reactors (LMFBR). Test results provide data needed to establish the response of LMFBR cores to hypothetical accidents producing fuel failure, coolant boiling, and the movement of coolant, molten fuel, and molten cladding. These data include margins to fuel failure, the modes of failure and movements, and evidence for identification of the mechanisms which determine the failure and movements. A key element in the program is the fast-neutron hodoscope, which detects fuel movement as a function of time during experiments.
Date: February 1, 1979
Creator: Dickerman, C E; Rothman, A B; Klickman, A E; Spencer, B W & DeVolpi, A
Partner: UNT Libraries Government Documents Department

Nuclear safety research collaborations between the U.S. and Russian Federation International Nuclear Safety Centers

Description: The Russian Federation Ministry for Atomic Energy (MINATOM) and the US Department of Energy (USDOE) have formed International Nuclear Safety Centers to collaborate on nuclear safety research. USDOE established the US Center (ISINSC) at Argonne National Laboratory (ANL) in October 1995. MINATOM established the Russian Center (RINSC) at the Research and Development Institute of Power Engineering (RDIPE) in Moscow in July 1996. In April 1998 the Russian center became a semi-independent, autonomous organization under MINATOM. The goals of the center are to: Cooperate in the development of technologies associated with nuclear safety in nuclear power engineering; Be international centers for the collection of information important for safety and technical improvements in nuclear power engineering; and Maintain a base for fundamental knowledge needed to design nuclear reactors. The strategic approach is being used to accomplish these goals is for the two centers to work together to use the resources and the talents of the scientists associated with the US Center and the Russian Center to do collaborative research to improve the safety of Russian-designed nuclear reactors. The two centers started conducting joint research and development projects in January 1997. Since that time the following ten joint projects have been initiated: INSC databases--web server and computing center; Coupled codes--Neutronic and thermal-hydraulic; Severe accident management for Soviet-designed reactors; Transient management and advanced control; Survey of relevant nuclear safety research facilities in the Russian Federation; Computer code validation for transient analysis of VVER and RBMK reactors; Advanced structural analysis; Development of a nuclear safety research and development plan for MINATOM; Properties and applications of heavy liquid metal coolants; and Material properties measurement and assessment. Currently, there is activity in eight of these projects. Details on each of these joint projects are given.
Date: May 5, 2000
Creator: Hill, D. J.; Braun, J. C.; Klickman, A. E.; Bougaenko, S. E.; Kabonov, L. P. & Kraev, A. G.
Partner: UNT Libraries Government Documents Department

Joint nuclear safety research projects between the US and Russian Federation International Nuclear Safety Centers

Description: The Russian Federation Ministry for Atomic Energy (MINATOM) and the US Department of Energy (USDOE) formed international Nuclear Safety Centers in October 1995 and July 1996, respectively, to collaborate on nuclear safety research. Since January 1997, the two centers have initiated the following nine joint research projects: (1) INSC web servers and databases; (2) Material properties measurement and assessment; (3) Coupled codes: Neutronic, thermal-hydraulic, mechanical and other; (4) Severe accident management for Soviet-designed reactors; (5) Transient management and advanced control; (6) Survey of relevant nuclear safety research facilities in the Russian Federation; (8) Advanced structural analysis; and (9) Development of a nuclear safety research and development plan for MINATOM. The joint projects were selected on the basis of recommendations from two groups of experts convened by NEA and from evaluations of safety impact, cost, and deployment potential. The paper summarizes the projects, including the long-term goals, the implementing strategy and some recent accomplishments for each project.
Date: August 1, 1998
Creator: Bougaenko, S.E.; Kraev, A.E.; Hill, D.L.; Braun, J.C. & Klickman, A.E.
Partner: UNT Libraries Government Documents Department

Computer network that assists in the planning, execution and evaluation of in-reactor experiments

Description: For over 20 years complex, in-reactor experiments have been performed at Argonne National Laboratory (ANL) to investigate the performance of nuclear reactor fuel and to support the development of large computer codes that address questions of reactor safety in full-scale plants. Not only are computer codes an important end-product of the research, but computer analysis is also involved intimately at most stages of experiment planning, data reduction, and evaluation. For instance, many experiments are of sufficiently long duration or, if they are of brief duration, occur in such a purposeful sequence that need for speedy availability of on-line data is paramount. This is made possible most efficiently by computer assisted displays and evaluation. A purposeful linking of main-frame, mini, and micro computers has been effected over the past eight years which greatly enhances the speed with which experimental data are reduced to useful forms and applied to the relevant technological issues. This greater efficiency in data management led also to improvements in the planning and execution of subsequent experiments. Raw data from experiments performed at INEL is stored directly on disk and tape with the aid of minicomputers. Either during or shortly after an experiment, data may be transferred, via a direct link, to the Illinois offices of ANL where the data base is stored on a minicomputer system. This Idaho-to-Illinois link has both enhanced experiment performance and allowed rapid dissemination of results.
Date: January 1, 1985
Creator: Bauer, T.H.; Froehle, P.H.; August, C.; Baldwin, R.D.; Johanson, E.W.; Kraimer, M.R. et al.
Partner: UNT Libraries Government Documents Department

NUMERICAL SOLUTION OF FUEL-ELEMENT THERMAL-STRESS PROBLEMS

Description: In developing a method of numerical analysis for the solution of thermal- stress problems special emphasis was given to fuel elements with internal coolant channels. Numerical techniques ior reducing the partial differential equation system te a form suitable for numerical solution and a new iteratlve method of solving large systems of linear algebraic equations were employed. Computer codes were devised to obtain the numerical solution of the thermal-stress problems and were used to obtain numerical results for single-hole and seven-hole hexagonal elements and plate-type elements. Comparisons were made between analytical results and numerical results for the case of t:.ie simple annulus shape. (auth)
Date: February 26, 1960
Creator: Redmond, R.F.; Pollack, H.; Klickman, A.E.; Hogan, W.S.; Epstein, H.M. & Chastain, J.W.
Partner: UNT Libraries Government Documents Department

PFR/TREAT programme: objectives, progress and future work

Description: The PFR/TREAT collaborative program of fast-reactor fuel testing is described and the objectives are illustrated in terms of the parameters selected for the irradiation of US and UK full-length fuel pins in PFR, followed by safety testing in TREAT. The measurements being made before, during, and after testing are outlined and the equipment and facilities being used in the UK and USA are described. An outline is given of the progress made and results obtained since the beginning of the collaboration in November 1979, together with future schedules for irradiation and testing. More-detailed results from the first two tests are given in a companion paper.
Date: January 1, 1982
Creator: Cowking, C.B.; Alter, H.; Hewison, R.; Borys, S.S.; Wood, M.H.; Culley, G.E. 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

PFR/TREAT program: objectives, accomplishments, and plans. [LMFBR]

Description: The PFR-TREAT collaborative program of transient safety testing of fast reactor fuel was established in 1979 to provide mutual advantage to USDOE and the UKAEA through irradiation of US and UK full-length fuel pins in PFR, followed by safety testing in TREAT. The tests which were planned include Transient Over-Power (TOP) and Transient Under-Cooling with Over-Power (TUCOP) tests to fuel destruction and re-distribution; the results will provide significant new information on fuel and cladding behavior in hypothetical reactor faults. The information obtained in both US and UK fuel pins is to be interpreted by both partners and published jointly when mutually agreed. Thirteen tests, on fresh and irradiated fuel, in single-pin and 7-pin test sections, were completed by the end of 1983. The test matrix, which is currently being re-evaluated, calls for additional tests to be run under the present agreement. There has been an extensive program of post irradiation examination of sibling pins in both the UK and the US to characterize the test fuel prior to destructive irradiation, including testing of irradiated cladding to determine its failure characteristics.
Date: January 1, 1984
Creator: Cowking, C.B.; Alter, H.; Stillwell, J.; Wood, M.H.; Woods, W.J.; Culley, G.E. et al.
Partner: UNT Libraries Government Documents Department

Design and Economic Evaluation of Mobile Blankets for Fast Reactors

Description: Report evaluating the design characteristics and limitations of mobile blankets for breeder reactors. This also includes economic considerations for each tested blanket. Appendices begin on page 40.
Date: March 10, 1964
Creator: Klickman, A. E.; Ball, G. L.; Edwards, J. J.; Jens, W. H.; Segal, B. M.; Amorosi, A. et al.
Partner: UNT Libraries Government Documents Department