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Severe accident core heatup transients in modular high temperature gas-cooled reactors without operating Reactor Cavity Cooling Systems

Description: The ultimate decay heat removal system for the current Modular High Temperature Gas-Cooled reactors is a completely passive natural convection air cooling loop. This paper considers an extremely remote accident scenario, where even this passive system fails, and heat rejection is only via a layer of thermal insulation to the reactor silo structure and the surrounding soil. The results show that even in this case the peak fuel temperatures remain well within safe limits. However, vessel and conc… more
Date: January 1, 1988
Creator: Kroeger, P. G.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Characterization of solids in the Three Mile Island Unit 2 Reactor defueling water: Addendum

Description: Shortly after ORNL/TM-10362 was issued, it was discovered that a series of 31 figures had been inadvertently omitted. These figures, which consist of scanning electron microscope (SEM) photographs and energy-dispersive X-ray fluorescence scans, provide significant information about the results obtained in the tests performed with water sample W3. This Addendum includes these figures. Details of and comments on the SEM photographs may be found in ORNL/TM-10362.
Date: March 1, 1988
Creator: Campbell, D.O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Nuclear plant-aging research on reactor protection systems

Description: This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show th… more
Date: January 1, 1988
Creator: Meyer, L.C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Coarse-node computations with an adaptive node structure

Description: The analysis with COMMIX of liquid metal reactor (LMR) intermediate heat exchanger (IHX) transients that are characterized by low flows, and especially imbalanced low flows, shows that if a coarse-node structure is used the predicted temperatures are significantly different than those given by a fine-node structure. If a fine-node structure is used, for problems that involve a large part of the plant, the computation time becomes excessive. This paper presents an improved version of an adaptive… more
Date: January 1, 1988
Creator: Tzanos, C.P.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Transient thermal analysis of three fast-charging latent heat storage configurations for a space-based power system

Description: A space-based thermal storage application must accept large quantities of heat in a short period of time at an elevated temperature. A model of a lithium hydride phase change energy storage system was used to estimate reasonable physical dimensions for this application which included the use of a liquid metal heat transfer fluid. A finite difference computer code was developed and used to evaluate three methods of enhancing heat transfer in the PCM energy storage system. None of these three met… more
Date: January 1, 1988
Creator: Stovall, T. K. & Arimilli, R. V.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Criticality experiments with mixed oxide fuel pin arrays in plutonium-uranium nitrate solution

Description: A series of critical experiments was completed with mixed plutonium-uranium solutions having a Pu/(Pu + U) ratio of approximately 0.22 in a boiler tube-type lattice assembly. These experiments were conducted as part of the Criticality Data Development Program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. A complete description of the experiments and data are included in this report. The experiments were perf… more
Date: August 1, 1988
Creator: Lloyd, R.C. (Pacific Northwest Lab., Richland, WA (United States)) & Smolen, G.R. (Oak Ridge National Lab., TN (United States))
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Passive safety and the advanced liquid metal reactors

Description: Advanced Liquid Metal Reactors being developed today in the USA are designed to make maximum use of passive safety features. Much of the LMR safety work at Argonne National Laboratory is concerned with demonstrating, both theoretically and experimentally, the effectiveness of the passive safety features. The characteristics that contribute to passive safety are discussed, with particular emphasis on decay heat removal systems, together with examples of Argonne's theoretical and experimental pro… more
Date: January 1, 1988
Creator: Hill, D.J.; Pedersen, D.R. & Marchaterre, J.F.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Benefits of vertical and horizontal seismic isolation for LMR (liquid metal reactor) nuclear reactor units

Description: Seismic isolation has been shown to be able to reduce transmitted seismic force and lower response accelerations of a structure. When applied to nuclear reactors, it will minimize seismic influence on the reactor design and provide a design which is less site dependent. In liquid metal reactors where components are virtually at atmospheric pressure but under severe thermal conditions, thin-walled structures are generally used for primary systems. Thin-walled structures, however, have little inh… more
Date: January 1, 1988
Creator: Wu, Ting-shu; Chang, Y. W. & Seidensticker, R. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Integral fast reactor safety features

Description: The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on… more
Date: January 1, 1988
Creator: Cahalan, J. E.; Kramer, J. M.; Marchaterre, J. F.; Mueller, C. J.; Pedersen, D. R.; Sevy, R. H. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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The passive safety characteristics of modular high temperature gas-cooled reactor fuel elements

Description: High-Temperature Gas-Cooled Reactors (HTGR) in both the US and West Germany use an all-ceramic, coated fuel particle to retain fission products. Data from irradiation, postirradiation examinations and postirradiation heating experiments are used to study the performance capabilities of the fuel particles. The experimental results from fission product release tests with HTGR fuel are discussed. These data are used for development of predictive fuel performance models for purposes of design, lice… more
Date: January 1, 1988
Creator: Goodin, D.T.; Kania, M.J.; Nabielek, H.; Schenk, W. & Verfondern, K.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Experimental and analytical studies of passive shutdown heat removal from advanced LMRs (liquid metal reactors)

Description: A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) has investigated the heat transfer performance of the GE/PRISM passive design. This initial series of experiments simulates the air-side geometry of the PRISM Radiant Reactor Vessel Auxiliary… more
Date: January 1, 1988
Creator: Pedersen, D.; Heineman, J.; Stewart, R.; Anderson, T.; Lottes, P. & Tessier, J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Pour stream breakup of molten IFR (Integral Fast Reactor) metal fuel in sodium

Description: Tests have been conducted in which the breakup behavior of kilogram quantities of molten uranium, uranium-zirconium alloy, and uranium-iron alloy pour streams in 600C sodium was studied. A sodium depth of less than 0.3 m was required for hydrodynamic breakup and freezing of 25-mm pour streams of uranium and uranium-zirconium alloy with up to 400C melt superheat. The breakup material was primarily in the form of filaments and sheets with a settled bed voidage on the order of 0.9. The uranium-iro… more
Date: January 1, 1988
Creator: Gabor, J. D.; Purviance, R. T.; Aeschlimann, R. W. & Spencer, B. W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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A scoping study of fission product transport from failed fuel during N Reactor postulated accidents

Description: This report presents a scoping study of cesium, iodine, and tellurium behavior during a cold leg manifold break in the N Reactor. More detail about fission product behavior than has previously been available is provided and key parameters that control this behavior are identified. The LACE LA1 test and evidence from the Power Burst Facility Severe Fuel Damage tests are used to test the key model applied to determine aerosol behavior. Recommendations for future analysis are also provided. The pr… more
Date: January 1, 1988
Creator: Hagrman, D.L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Sloshing of coolant in a seismically isolated reactor

Description: During a seismic event, the liquid coolant inside the reactor vessel will have sloshing motion which is a low-frequency phenomenon. In a reactor system incorporated with seismic isolation, the isolation frequency usually is also very low. There is concern on the potential amplification of sloshing motion of the liquid coolant. This study investigates the effects of seismic isolation on the sloshing of liquid coolant inside the reactor vessel of a liquid metal cooled reactor. Based on a syntheti… more
Date: January 1, 1988
Creator: Wu, Ting-shu; Gvildys, J. & Seidensticker, R.W.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

Description: The uncertainty in predicting the effectiveness of inherent shutdown in innovative liquid metal cooled reactors with metallic fuel results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require inherent shutdown; (2) the approximation of representing all such events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to th… more
Date: January 1, 1988
Creator: Mueller, C.J. & Wade, D.C.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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The US Liquid Metal Reactor Development Program

Description: The US Liquid Metal Reactor Development Program has been restructured to take advantage of the opportunity today to carry out R and D on truly advanced reactor technology. The program gives particular emphasis to improvements to reactor safety. The new directions are based on the technology of the Integral Fast Reactor (IFR). Much of the basis for superior safety performance using IFR technology has been experimentally verified and aggressive programs continue in EBR-II and TREAT. Progress has … more
Date: January 1, 1988
Creator: Till, C.E.; Arnold, W.H. & Griffith, J.D.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Fuel performance annual report for 1986

Description: This annual report, the ninth in a series, provides a brief description of fuel performance during 1986 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related U.S. Nuclear Regulatory Commission evaluations are included. 550 refs., 12 figs., 31 tabs.
Date: March 1, 1988
Creator: Bailey, W.J. & Wu, S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Heat pump augmented radiator for low-temperature space applications

Description: Closed-cycle, space-based heat rejection systems depend solely on radiation to achieve their heat dissipation function. Since the payload heat rejection temperature is typically 50 K above that of the radiation sink in near earth orbit, the size and mass of these systems can be appreciable. Size (and potentially mass) reductions are achievable by increasing the rejection temperature via a heat pump. Two heat pump concept were examined to determine if radiator area reductions could be realized w… more
Date: January 1, 1988
Creator: Olszewski, M. & Rockenfeller, U.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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MHTGR: New production reactor summary of experience base

Description: Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs.,… more
Date: March 1, 1988
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Uncertainties in source term estimates for a station blackout accident in a BWR with Mark I containment

Description: In this paper, attention is limited to a single accident progression sequence, namly a station blackout accident in a BWR with a Mark I containment building. Identified as an important accident in the draft version of NUREG-1150 a station blackout involves loss of both off-site power and dc power resulting in failure of the diesels to start and in the unavailability of the high pressure injection and core isolation cooling systems. This paper illustrates the calculated uncertainties (Probabilit… more
Date: January 1, 1988
Creator: Lee, M.; Cazzoli, E.; Liu, Y.; Davis, R.; Nourbakhsh, H.; Schmidt, E. et al.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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A simple model for radial expansion reactivity in LMRs (liquid metal reactors)

Description: Presented in this report is a simple analytical model developed for evaluating the radial expansion reactivity in small modular liquid metal reactors (LMRs). The present model is based on a non-leakage representation of the effective neutron multiplication factor. The resultant analytical expression for the radial expansion reactivity is simple and can be used directly in a system code for safety analyses. Applications of the present model to PRISM and SAFR resulted in a good agreement with the… more
Date: January 1, 1988
Creator: Cheng, H.S. & Van Tuyle, G.J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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ENDF/B-V cell comparisons and MCNP-3A

Description: This study compares calculated results from the CELL-2 spectrum code to calculated results from the MCNP Monte Carlo code for typical pressurized water reactor (PWR) fuel lattices at elevated temperatures. The MCNP calculations represent the first analysis of commerical fuel lattices by continuous energy Monte Carlo at realistic temperaturs with ENDF/B-V data. Eigenvalue results demonstrate that the two codes agree to within the statistical uncertainty of MCNP. No systematic variations or biase… more
Date: January 1, 1988
Creator: Eich, W. J.; Eisenhart, L. D.; Little, R. C.; Mosteller, R. D. & Chao, J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

Description: This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, ac… more
Date: February 1, 1988
Creator: Weiss, A. J.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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Interfacing system LOCAs at PWRs (pressurized water reactors): A new assessment

Description: Interfacing system loss-of-coolant accidents (ISL) were first identified and analyzed in the Reactor Safety Study (RSS). ISLs were found to be significant contributors to the risk estimates for the presurized water reactor (PWR) analyzed in the RSS. Some evaluations of ISL sequences in subsequent studies have also found that their contribution to risk is also significant. ISLs are potentially important to risk because a direct path is opened from the damaged reactor core which bypasses the cont… more
Date: January 1, 1988
Creator: Bozoki, G.; Kohut, P. & Fitzpatrick, R.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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