Thermal hydraulics

The OASIS code has been used for the thermal hydraulics analyses. A large number of events were analysed. The objective of these studies was to identify the worst-case scenario. Therefore, rupture position, rupture time, and break size spectra must be analysed. Currently, the detailed thermal hydraulics analysis for the primary coolant pipe break accident as a function of the break position is being performed. This study is based on conservative assumptions, as required for Category 4 DBE analyses. The main results of this analysis are the core inlet and outlet temperatures, the break flow rate, the IHX inlet and outlet temperatures, as well as the and clad temperature distribution. The results of the calculations show that the hotspot temperatures for the cladding and for the coolant are below the design safety limits (DSL). These temperatures are lower than the sodium boiling temperature. The regulator accepted the methods applied, as well as the results of the analysis. 

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N. E. Todreas and M. S. Ka2imi, Nuclear Systems I Thermal Hydraulic Fundamentals, 1989, II Elements of Thermal Hydraulic Design, 1990, Hemisphere Publishing Corp., New York. 

R. T. Lahey, Jr. and E. J. Moody, The Thermal-Hydraulics of a Boiling Water Nuclear Reactor, 2nd ed., American Nuclear Society, La Grange Park, lU., 1993. 

The combinations of failures and non-failed conditions define the state of the pJani at the right branches. The damage associated with these plant damage states are calculated using thermal-hydraulic analyses to determine temperature profiles that are related to critical chemical reactions, explosions and high pressure. These end-states serve as initiators fot breaking confinement that leads to release in the plant and aquatic and atmospheric release outside ol the plant,... 

Given the damage states, the analysis flows much as shown in Figure 6.3-1, depending on the process. For a nuclear power plant, thermal-hydraulic analyses determine the spatial temperature of the damaged core, and consequently the ability of the core to retain radioactive materials. Analysis of the physical processes reveals the amounts of hazardous materials that may be released. 

Section 8.1 provided a description of a core melt. This section backs up to describe thermal-hydraulic calculations of the phenomena before, during, and after the accident, and other calculations to estimate the radioactive release from containment. In this accident physics cannot be analyzed separately from in-plant transport. 

Of these phenomena, the first three in particular, involve thermal hydraulics beginning with the pre-accident conditions. Items 4 through 7 address the meltdown of the core and its influence on (1) hydrogen production, which affects containment loads, (2) fuel temperatures, which affect in-vessel fission product releases, (3) thermal-... 

Thermal-Hydraulic Analysis must be extended to at least 24 hours into the accident. This study used calculations for feed-and-bleed operation with a charging pump, and with gravity teed from the refueling water storage tank (RWST),... 

Gentry, G. G., R. K. Young, and W. M. Small, RODbaffle Heat Exchanger Thermal-Hydraulic Predictive Methods for Bare and Low-Finned Tubes, National Heat Transfer Gonference, Niagara Falls, NY, Aug. 5-8, (1984). 

Calame JP, Myers RE, Binari SC, Wood FN, Garven M (2007) Experimental investigation of micro-channel coolers for the high heat flux thermal management of GaN-on-SiC semiconductor devices. Int J Heat Mass Transfer 50 4767-4779 Celata GP, Cumo M, Zummo G (2004) Thermal-hydraulic characteristics of single- phase flow in capillary pipes. Exp Thermal Fluid Sci 28 87-95 Celata GP (2004). Heat transfer and fluid flow in micro-channels. Begell House, N.Y. 

Celata GP, Gumo M, Zummo G (2004) Thermal-hydraulic characteristics of single-phase flow in capillary pipes. Exp Thermal Fluid Sci 28 87-95... 

Kakac, S., Bergles, A. E. and Mayinger, F. (eds) Heat Exchangers, thermal-hydraulic fundamentals and design (Hemisphere, 1981). 

This section describes some of the boiling phenomena that occur in water reactors with respect to safety analyses that require thermal hydraulic considerations. 

System thermal-hydraulic phenomena associated with instrument tube breaks at the reactor vessel and seal table are comparable. 

ORNL small-break LOCA tests Experimental investigation of heat transfer and reflood analysis was made under conditions similar to those expected in a small-break LOCA. These tests were performed in a large, high-pressure, electrically heated test loop of the ORNL Thermal Hydraulic Test Facility. The analysis utilized a heat transfer model that accounts for forced convection and thermal radiation to steam. The results consist of a high-pressure, high-temperature database of experimental heat transfer coefficients and local fluid conditions. 

In the microscopic analysis of CHF, researchers have applied classical analysis of the thermal hydraulic models to the CHF condition. These models are perceived on the basis of physical measurements and visual observations of simulated tests. The physical properties of coolant used in the analysis are also deduced from the operating parameters of the test. Thus the insight into CHF mechanisms revealed in microscopic analysis can be used later to explain the gross effects of the operating parameters on the CHF. 

Enhancement of CHF subcooled water flow boiling was sought to improve the thermal hydraulic design of thermonuclear fusion reactor components. Experimental study was carried out by Celata et al. (1994b), who used two SS-304 test sections of inside diameters 0.6 and 0.8 cm (0.24 and 0.31 in.). Compared with smooth channels, an increase of the CHF up to 50% was reported. Weisman et al. (1994) suggested a phenomenological model for CHF in tubes containing twisted tapes. 

March-Leuba (1990) presented radial nodalization effects on the stability calculations. March-Leuba and Blakeman (1991) reported on out-of-phase power instabilities in BWRs. BWR stability analyses were reported by Anegawa et al. (1990) and by Haga et al. (1990). The experience and safety significance of BWR core-thermal-hydraulic stability was presented by Pfefferlen et al. (1990). 

Agee, L. J., 1978, Power Series Solutions of the Thermal-Hydraulic Conservation Equations, in Transient Two-Phase Flew, Proc. 2nd Specialists Meeting, OECD Committee for the Safety of Nuclear Installations, Paris, Vol. 1, pp. 385-410. (3)... 

Akoski, J., R. D. Watson, P. L. Goranson, A. Hassanian, and J. Salmanson, 1991, Thermal Hydraulic Design Issues and Analysis for ITER Diverters, Fusion Technol. 79.-1729—1735. (4)... 

ANS/ASME/NRC, 1980, Proc. of Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Saratoga, FL, Hemisphere, New York. (3)... 

Celata, G. P, and A. Mariani, 1993, A Data Set of CHF in Water Subcooled Flow Boiling, Specialists Workshop on the Thermal-Hydraulics of Hight Heat Flux Components in Fusion Reactors, Coordinated by G. P. Celata/A. Mariani, ENEA-C.R.E. Casaccia, Italy. (5)... 

Chu, K. J., and A. E. Dukler, 1974, Statistical Characteristics of Thin Wavy Films, AIChE J. 20 695. (3) Chu, P. T., H. Chelemer, and L. E. Hochreiter, 1973, THINC IV An Improved Program for Thermal Hydraulic Analysis of Rod Bundle Cores, WCAP-7965, Westinghouse Electric Corporation, Pittsburgh, PA. (App.)... 

Nelson, R., and C. Unal, 1992, A Phenomenological Model of the Thermal Hydraulics of Convective Boiling during the Quenching of Hot Rod Bundles, Part I Thermal Hydraulic Model, Nuclear Eng. Design 756 277-298. (4)... 

Pfefferlen, H., R. Raush, and G. Watford, 1990, BWR Core Thermal-Hydraulic Stability Experience and Safety Significance, Proc. Int. Workshop on BWR Instability. Hottsville, NY, CSNI Rep. 178, OECD-NEA, Paris. (6)... 

Rowe, D. S, 1970, COBRA II Digital Computer Program for Thermal Hydraulic Subchannel Analysis of Rod Bundle Nuclear Fuel Elements, BNWL 1229, Battelle Northwest Laboratory, Richland,... 

Todreas, N. E., and M. S. Kazimi, 1990, Two-Phase Flow Dynamics, in Nuclear Systems I Thermal Hydraulic Fundamentals, pp. 476-479, Hemisphere, Washington, DC. (3)... 

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)... 

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