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SASSYS analysis of EBR-II SHRT experiments

Description: SASSYS is a general purpose one-dimensional thermal-hydraulic LMR systems analysis code. The application of SASSYS to studies of protected and unprotected transient analysis of LMR designs, such as PRISM and SAFR, is commonplace. This usage requires that a strong validation base exist for SASSYS. Validation of a code such as SASSYS is made up of a variety of activities such as, application of separate component models to model problems where the solution is analytically known, cross comparison with other codes and, most importantly, comparison with experiment. The EBR-II reactor is being used to perform experiments that address a wide range of LMR safety issues. The most important of these so far have been the SHRT series of experiments, begun in 1984, which culminated in the unprotected loss-of-flow and loss-of-heat sink transients performed as a demonstration to an international audience of fast reactor experts. The SASSYS-1 systems thermal hydraulic code for LMR's has been applied to the analysis of selected EBR-II SHRT experiments. The results show that SASSYS can predict the transient thermal hydraulic and reactivity behavior of EBR-II under these conditions. This contributes to the validation of the modelling of EBR-II with SASSYS and of the code itself.
Date: January 1, 1987
Creator: Hill, D.J.
Partner: UNT Libraries Government Documents Department

DEFORM-4: fuel pin characterization and transient response in the SAS4A accident analysis code system

Description: The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, fission gas bubble induced swelling, fuel cracking and healing, fission gas release, cladding swelling, and the thermal-mechanical state of the fuel and cladding. In the transient state, the module continues the thermal-mechanical response calculation, including fuel melting and central cavity pressurization, until cladding failure is predicted and one of the failed fuel modules is initiated. Comparisons with experimental data have demonstrated the validity of the modeling approach.
Date: January 1, 1986
Creator: Miles, K.J. & Hill, D.J.
Partner: UNT Libraries Government Documents Department

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 programs in this area.
Date: January 1, 1988
Creator: Hill, D.J.; Pedersen, D.R. & Marchaterre, J.F.
Partner: UNT Libraries Government Documents Department

Fuel relocation modeling in the SAS4A accident analysis code system

Description: The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs.
Date: January 1, 1986
Creator: Tentner, A.M.; Miles, K.J.; Kalimullah & Hill, D.J.
Partner: UNT Libraries Government Documents Department

Status and future direction of the melt attack and coolability experiments (MACE) program at Argonne National Laboratory.

Description: The Melt Attack and Coolability Experiments (MACE) program has been underway at Argonne National Laboratory addressing the ability of water to quench and thermally stabilize a molten core concrete interaction (MCCI) when the interaction is flooded from above. In this program, which has been sponsored by the EPRI-headed Advanced Containment Experiments (ACE) international consortium, large scale reactor material integral effects experiments have been conducted, in parallel with related modeling efforts. Plans are currently being developed for continued utilization of the MACE facility under the sponsorship of the Nuclear Energy Agency (NEA) to achieve the following objectives: (i) resolution of the ex-vessel debris coolability issue through a redirected program which focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests; and (ii) address remaining uncertainties related to long-term two-dimensional MCCI under dry cavity conditions. In terms of the ex-vessel debris coolability issue, separate effects tests are planned to provide data on key melt coolability mechanisms identified in MACE integral effects tests. The results of these tests will provide both confirmatory evidence and test data to support development of validated models for extrapolation to plant conditions. In terms of dry cavity conditions, reactor material tests are planned to address remaining uncertainties related to long-term 2-D MCCI; in particular, lateral vs. axial power split. This paper describes the essential elements of the program to address these two remaining important LWR safety issues.
Date: February 2, 2001
Creator: Farmer, M.T.; Spencer, B.W.; Binder, J.L. & Hill, D.J.
Partner: UNT Libraries Government Documents Department

Post-failure material movement in the PFR/TREAT experiments

Description: In the PFR/TREAT experiments the fast neutron hodoscope has provided experimental data on post-failure material movement during Transient Overpower and Transient Under-Cooled Overpower events. Analyses of selected experiments have been done, using SIMMER-II and SAS4A, to validate these codes. Full results of these analyses are presented in the paper. The general conclusion that can be drawn from the analyses that have been made so far is that the hodoscope is providing data which has appropriate resolution in time and space and that the computer models which are being used by the USDOE and the UKAEA provide good representations of fuel redistribution.
Date: January 1, 1986
Creator: Bauer, T.H.; Morman, J.A.; Hill, D.J.; DeVolpi, A. & Goldman, A.J.
Partner: UNT Libraries Government Documents Department

Thermal-hydraulic development a small, simplified, proliferation-resistant reactor.

Description: This paper addresses thermal-hydraulics related criteria and preliminary concepts for a small (300 MWt), proliferation-resistant, liquid-metal-cooled reactor system. A main objective is to assess what extent of simplification is achievable in the concepts with the primary purpose of regaining economic competitiveness. The approach investigated features lead-bismuth eutectic (LBE) and a low power density core for ultra-long core lifetime (goal 15 years) with cartridge core replacement at end of life. This potentially introduces extensive simplifications resulting in capital cost and operating cost savings including: (1) compact, modular, pool-type configuration for factory fabrication, (2) 100+% natural circulation heat transport with the possibility of eliminating the main coolant pumps, (3) steam generator modules immersed directly in the primary coolant pool for elimination of the intermediate heat transport system, and (4) elimination of on-site fuel handling and storage provisions including rotating plug. Stage 1 natural circulation model and results are presented. Results suggest that 100+% natural circulation heat transport is readily achievable using LBE coolant and the long-life cartridge core approach; moreover, it is achievable in a compact pool configuration considerably smaller than PRISM A (for overland transportability) and with peak cladding temperature within the existing database range for ferritic steel with oxide layer surface passivation. Stage 2 analysis follows iteration with core designers. Other thermal hydraulic investigations are underway addressing passive, auxiliary heat removal by air cooling of the reactor vessel and the effects of steam generator tube rupture.
Date: July 2, 1999
Creator: Farmer, M. T.; Hill, D. J.; Sienicki, J. J.; Spencer, B. W. & Wade, D. C.
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

Code validation with EBR-II test data

Description: An extensive system of computer codes is used at Argonne National Laboratory to analyze whole-plant transient behavior of the Experimental Breeder Reactor 2. Three of these codes, NATDEMO/HOTCHAN, SASSYS, and DSNP have been validated with data from reactor transient tests. The validated codes are the foundation of safety analyses and pretest predictions for the continuing design improvements and experimental programs in EBR-II, and are also valuable tools for the analysis of innovative reactor designs.
Date: January 1, 1992
Creator: Herzog, J.P.; Chang, L.K.; Dean, E.M.; Feldman, E.E.; Hill, D.J.; Mohr, D. et al.
Partner: UNT Libraries Government Documents Department

Code validation with EBR-II test data

Description: An extensive system of computer codes is used at Argonne National Laboratory to analyze whole-plant transient behavior of the Experiment Breeder Reactor 2. Three of these codes, NATDEMO/HOTCHAN, SASSYS, and DSNP have been validated with data from reactor transient tests. The validated codes are the foundation of safety analyses and pretest predictions for the continuing design improvements and experimental programs in EBR-2, and are also valuable tools for the analysis of innovative reactor designs. 29 refs., 6 figs.
Date: January 1, 1991
Creator: Herzog, J.P.; Chang, L.K.; Dean, E.M.; Feldman, E.E.; Hill, D.J.; Mohr, D. et al.
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