Thermal hydraulics design

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. 

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. 

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)... 

Baffle pitch, or distance between baffles, normally is 0.2-1.0 times the inside diameter of the shell. Both the heat transfer coefficient and the pressure drop depend on the baffle pitch, so that its selection is part of the optimization of the heat exchanger. The window of segmental baffles commonly is about 25%, but it also is a parameter in the thermal-hydraulic design of the equipment. 

M. Richards, A. Shenoy, Y. Kiso, N. Tsuji, N. Kodochigov, and S. Shepelev Thermal Hydraulic Design of a Modular Helium Reactor Core Operating at 1 000°C Coolant Outlet Temperature, Proceedings of the 6 International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6), October 4-8, 2004, Nara, Japan, Atomic Energy Society of Japan, Tokyo, Japan (2004). 

Within this Report it is impossible to discuss all details and strategies of LMFR thermal hydraulic design. Therefore, only the most important aspects will be treated in the following subsections. 

Thermal hydraulic design. The preliminary core thermal hydraulic design reported here needs to be continued using multidimensional methods to establish the heat removal capability and coolant pump sizing for the target power level and for assessing the reactor s performance in anticipated transients. 

The thermal-hydraulic design of the rack provides sufficient natural convection cooling flow to remove 19,929 W/bundle (68,000 Btu/hrAjundle) of decay heat. 

Rail transportability imposes a size limitation upon the reactor vessel and guard vessel of 6.1 m in diameter and 18.9 m in height [XXIII-25]. The fission gas plenum height is based upon an assumed conservative gas release from nitride fuel of 2.5% per atom % of bum-up. The fuel volume fraction was held fixed in the thermal hydraulic design analyses at the value of 0.215 determined by the core design. The fuel rod outer diameter and pitch-to-diameter ratio were varied to determine an optimum combination. Figure XXIII-5 shows the relationship... 

Coupling of a LFR to the S-CO2 Brayton cycle is dependent upon the thermal hydraulic design of in-reactor lead-to-C02 heat exchangers (HXs). The Pb-to-C02 HXs must meet several requirements and constraints ... 

The thermal-hydraulic design bases form the acceptance criteria for the thermal-hydraulic... 

The method used to imdertake the thermal hydraulic design of the reactor and the references to the design information is provided in Section 4.4.2 of Reference 6.1. The analysis of the thermal hydraulic design shows that there is a sufficient flow rate and flow distribution through the reactor to adequately remove the reactor core heat. The analysis covers normal operation and Design Basis transients, and it has considered ... 

The thermal hydraulic design of CAREM reactor core was carried out using an improved version of the 3D two fluid model realized in the THERMIT code. In order to take into account strong coupling of the thermal-hydraulic and neutronic characteristics of the core, THERMIT was linked with the neutronic code CITVAP. This coupled model makes it possible to produce a 3D map of power and thermal-hydraulic parameters at any moment of the bum-up cycle. 

Hundsbedt, A., Experiments and Analyses in Support of the US ALMR Thermal-Hydraulic Design, Proc. of the IAEA Specialists Meeting on Evaluation of Decay Heat Removal by Natural Convection, Oarai, Japan, 1993. 

Thermal-hydraulics design of the clad failure detection system. This system is used to detect possible clad failures in the fuel subassemblies while the plant is operating. Clad failure causes a release of fission products emitting delayed neutrons, that are transported to the hot plenum and to the detectors. In SPXl the detector itself is placed outside the hot plenum with a continuous poped sampling system for analysis of the primary sodium was set up other systems in which neutron detectors are placed near the intermediate heat exchanger inlets and enabled activity to be measured directly, have been studied. In both cases thermal-hydraulics studies were necessary to measure the hydraulic transfer functions between the various core... 

Design approaches and calculation technologies for thermal-hydraulic and safety design Thermal-hydraulic design of the core PEACER-300 and 550 core analyses completed. 

Todreas, N.E. M S. Kazimi (2001). Nuclear Systems II—Elements of Thermal Hydraulic Design. US Taylor and Francis. 

All the features listed above had to be represented in the nuclear design analysis. It was clear at the outset that particular attention would have to be devoted to the problem of core representation If the overall objectives of predicting reactivity levels and void coefficients accurately, as already achieved in simple lattices, were still to be attained in the actual reactor core. Moreover, there was the problem of predicting the power distribution to an accuracy of about 5 of the peak value In order to satisfy the requirements of thermal hydraulic design. 

The development of methods of calculation for thermal-hydraulic design has proceeded in close association with the work on nuclear design methods. The complete system of digital codes now available for steady state performance analysis is known as PATRIARCH (see Pig. 1). Descriptions of the functions of the reactor physics codes appear in a companion paper. The approaches used in the thermal-hydraulic codes will be reviewed briefly below. In the space available, it is only possible to outline the basis of the more important codes. 

Fig. 2.1. Integration of nuclear design, thermal-hydraulic design, and materials design. Source Adapted from Nuclear Power Plant Design Analysis, Alexander Sesonshe, TID-26241, 1973.

From Table 2, mass flow distributions to the three core Mets exhibit (see Figure 2) a nearly inverted profile to those specified for the in-core thermal/hydraulic design. That is, more flow is going to the innermost fuel region of the core than was assumed by the in-core analysis), leaving the outer region with less flow. 

The neutronic and thermal-hydraulic design criteria (limits for normal operations) of the Super LWR core are described next. 

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