Reactor thermal-hydraulic analysis

LANL, Los Alamos National Laboratory, 1986. TRAC-PFl/MODl An Advanced Best-Estimate Computer Program for Pressurized Water Reactor Thermal-Hydraulic Analysis. NUREG/ CR-3858, LA-10157-MS. 

Yadigaroglu, G., and M. Andreani, 1989, Two Fluid Modeling of Thermal-Hydraulic Phenomena for Best Estimate LWR Safety Analysis, Proc. 4th Int. Topical Meeting on Nuclear Reactors Thermal-Hydraulics, Karlsruhe, U. Mueller, K. Rehnee, and K. Rust, Eds., Rep. NURETH-4, pp. 980-996. (3)... 

Anglart H, Andersson S, Podowski MZ, Kurul N (1993) An Analysis of Multidimensional Void Distribution in Two-Phase Flows. Proc 6th Int Topical Meeting on Nuclear Reactor Thermal Hydraulics NURETH-6, Grenoble, October 5-8, 1 139-153... 

The reactor coolant pumps are sized to deliver flow that equals or exceeds the design flow rate utilized in the thermal hydraulic analysis of the Reactor Coolant System. Analysis of steady-state and anticipated transients is performed assuming the minimum design flow rate. Tests are performed to evaluate reactor coolant pump performance during the post-core load hot functional testing to verify adequate flow. 

W. J. METE VIA, A. HUSAIN, and T. R. HENCEy, Relap-4 Thermal/Hydraulic Analysis of the Maine Yankee Spent Fuel Pit, Conf. Reactor Operating Experience, Trans. Am.Nucl. Soc., 21, SuppL 1, 18 (1975). 

The thermal-hydraulic analysis is performed using special codes that receive the fission power spatial deposition from the reactor analysis pin power deconstruction and perform a heat transfer and fluid flow calculation to calculate the expected temperature profile throughout the core. Because neutron cross sections depend on temperature, this calculation must be iteratively linked into the previous six steps. In many cases, this linkage results in an expansion of the parametric expansions of the neutron cross sections (described as part of Step 6) to be a function of temperature and total assembly fission power. 

The overall objectives of this chapter are to (1) provide a background on heat transfer in reactor systems (2) describe methods of analysis employed in the reactor thermal-hydraulics and safety with basic analysis processes and tools and (3) provide analysis examples, sources of information, and computer codes used for detailed reactor thermal-hydraulics and safety analysis. 

Though the fission in the reactor core is analyzed by the nuclear consideration, the heat generated by the reactor through fission and its use in the generation of power for a given reactor core is largely limited by thermal processes and material properties rather than by nuclear considerations. The safety analysis of the reactor during normal and abnormal operational conditions involves detailed thermal-hydraulics analysis. 

RETRAN (REactor TRansient ANalysis) is a best-estimate transient thermal-hydraulic analysis computer program (sponsored by EPRI) designed to provide analysis capabilities for BWR and PWR transients, small-break LOCAs, balance-of-plant modeling, and anticipated transients without scram (ATWS). 

Reactor physics and thermal hydraulics analysis methods that have been developed for earlier HTGR designs will need to be updated and qualified for apphcation to the GT-MHR... 

Reactor physics and thermal hydraulics analysis methods and codes... 

COBRAIIIc is a widely used code for the thermal-hydraulic analysis of nuclear reactor cores. Its approach is based on subchannel analyses where conservation equations are axially solved and coupled through mixing coefficients (Todreas Kazimi 2001). 

Ambrosini, W., 2008. Lesson learned from the adoption of numerical techniques in the analysis of nuclear reactor thermal-hydraulic phenomena. Progress in Nuclear Energy 50, 866-876. 

Richards, D.J., Hanna, B.N., Hobson, N., Ardron, K.H., 1985. ATHENA, a two-fluid code for CANDO LOCA analysis. In Third International Topical Meeting on Reactor Thermal Hydraulics, Newport, Rhode Island, USA, October 15—18, 1985, 7E-1 thru 7E-14 (CATHENA Formerly Named ATHENA). 

Zhou, W., Liu, R., Revankar, S.T., 2015. Fabrication methods and thermal hydraulics analysis of enhanced thermal conductivity U02-Be0 fuel in light water reactors. Journal of Annals of Nuclear Energy 81, 240—248. 

Podoxvski, M.Z., Lahey, R.T., Jr and Burger, J.M., "The Analysis of Core Meltdown Progression in BWRs", Proceedings of the Fourth International Topical Meeting on Nuclear Reactor Thermal- Hydraulics, Karlsruhe, FRG, 1989. 

J. Gou, Y. Ishiwatari, Y. Oka, M. Yamakawa and S. Ikejiri, Research and Development of Super Fast Reactor (5) Thermal Hydraulic Analysis of Tight-laiice Subchannels, Proc. 16th PBNC, Aomori, Japan, October 13-18, 2008, P16P1294 (2008)... 

Preliminary reactor core concepts have been proposed at various institutions. Among them, the concept with mixed neutron spectrum proposed by SJTU [95,96] has achieved the special attention of Chinese researchers. The mixed spectrum SCWR core combines the merits of both thermal and fast spectra as far as possible. The basic idea is to divide the reactor core into two zones with different neutron spectra. In the outer zone, the neutron energy spectrum is similar to that of PWRs. To assess the performance of the reactor core, a coupled neutron-physics and thermal-hydraulics analysis was conducted [96]. 

Y. S. Tang. Ph.D has more than 35 years of experience in the field of thermal and fluid flow. His research interests have covered aspects of thermal hydraulics that are related to conventional and nonconventional power generation systems, with an emphasis on nuclear reactor design and analysis that focuses on liquld-meta -cooled reactors. Dr. Tang is co-author of Radioactive Waste Management published by Taylor 8 Francis, and Thermal Analysis of Liquid Metal Fast Breeder Reactors, He received a B5. from National Central University In China and an MS. in mechanical engineering from the University of Wisconsin. He earned his Ph.D. 

JAEA conducted an improvement of the RELAP5 MOD3 code (US NRC, 1995), the system analysis code originally developed for LWR systems, to extend its applicability to VHTR systems (Takamatsu, 2004). Also, a chemistry model for the IS process was incorporated into the code to evaluate the dynamic characteristics of process heat exchangers in the IS process (Sato, 2007). The code covers reactor power behaviour, thermal-hydraulics of helium gases, thermal-hydraulics of the two-phase steam-water mixture, chemical reactions in the process heat exchangers and control system characteristics. Field equations consist of mass continuity, momentum conservation and energy conservation with a two-fluid model and reactor power is calculated by point reactor kinetics equations. The code was validated by the experimental data obtained by the HTTR operations and mock-up test facility (Takamatsu, 2004 Ohashi, 2006). 

Kelly JM, Stewart CW, Cuta JM (1992) VIPRE-02 - A two-fluid thermal-hydraulics code for reactor core and vessel analysis Mathematical modeling and solution methods. Nucear Technology 100 246-259... 

Experiments were carried out on power to measure the fuel SA outlet temperature with CCPM stuck at 80 mm position and a temperature attenuation of 7% (average) was found in Mark I SA. However, this attenuation is large for Mark II SA where sodium flow is less. Studies were conducted to find out the probability of plugging during reactor operation and found to be acceptable. 3D analysis of outlet plenum thermal hydraulic was carried out to establish the level of plugging that can be detected viz., allowable plugging for fuel clad integrity. 

Determination of the loads through analysis of the reactor operating parameters combined with thermal hydraulics calculations for the specific thermal stratification and fluctuation areas. For this purpose, the recent R D progress made in the EFR (European Fast Reactor) project was used and transposed to a real installation. 

Computer code CALPER — a thermal hydraulic subchannel analysis code for the assessment of coolant local conditions in the fuel assemblies and in the core of PWRAVWER-type nuclear reactors... 

The database information consists of the thermal-hydraulic/radiological variables of the core, reactor coolant system, containment system, and safety systems which are generated from integrated severe accident analysis code, such as MAAP or MELCOR program. 

It was revealed that most of the events due to a work planning problem where a work procedure is provided occurred during low-power states or startup operations. The reason for this can be inferred as the variable characteristics of plant configuration and dynamics of the low-power states of NPP, which may cause the identification of human error potentials and prediction of physical transition to be difficult. Therefore, the identification of human error possibilities or potentials during low-power states seems not to be an easy task to be accomplished by a list of simple checklist items, but belongs to a hard task that requires a careful investigation on the potential of human error by a concerted effort between experts in plant systems and hiunan errors, and, as necessary, may requires the use of thermal-hydraulic and reactor analysis computer codes. 

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