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Analysis of the DNB Ratio and the Loss-Of- Flow Accident (LOFA) of the 3 MW TRIGA MARK 11 Research Reactor

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Analysis of the DNB Ratio and the Loss-Of- Flow Accident (LOFA) of the 3 MW TRIGA MARK 11 Research Reactor BDO410004 INTERN,.,IL REPORT INST-90IRPED-22, MA Y 2003 Analysis of the DNB ratio and the Loss-of- Flow Accident (LOFA) of the 3 MW TRIGA MARK 11 Research Reactor M. Q. Huda, M. S. Mahmood, T. K. Chakrobortty, M. Rahman and M. M. Sarker REACTOR PHYSICS AND ENGINEERING DIVISION (RPED) INSTITUTE OF NUCLEAR SCIENCE TECHNOLOGY ATOMIC ENERGY RESEARCH ESTABLISHMENT GANAKBARI, SAVAR, GPO BOX 3787, DHAKA-1000 BANGLADESH ,q! 101- 9mWfw BANGLADESH ATOMIC ENERGY COMMISSION INST-90/RPED-22 CONTENTS Page 1. Introduction 1 2. Analysis of DNB Ratio 1 2.1. Effect of Operating Power 3 2.2. Effect of the Flow Rate 3 2.3. Effect of the Hot-Rod Factor 3 2.4. Effect of the Inlet Temperature 4 3. Loss-of-Flow Accident (LOFA) Analysis 4 3.1. Relative Power and Flow during LOFA 4 3.2. Fuel Temperature Distribution during LOFA 4 3.3. Occurrence of Flow Reversal 5 4. Conclusions 5 5. Acknowledgments 6 6. References 6 INST-90/RPED-22 ABSTRACT The PARET code was used to analyze two most important thermal hydraulic design parameters of the 3 MW TRIGA MARK research reactor. The first design parameter is the DNB (departure from nucleate boiling) ratio, which is defined as the ratio of the critical heat flux to the beat flux achieved in the core and was computed by means of a suitable correlation as defined in PARET code. The reactor core should be designed so as to prevent the DNBR from dropping below a chosen value under a high heat flux transient condition for the most adverse set of mechanical and coolant conditions. Over the length 0381 in of the hottest channel the DNB ratio vaiies, starting from 38951 to 5403 1, with a minimum of .785 1. The peak heat flux occurs at the axial center of the fue e therefore the DNB ratio is M1111mum at this location. The second design parameter is the loss-of-flow accident scenario of te TRIGA reactor. The Bergles-Rohsenow correlation was selected for detecting onset of nucleate boiling, te transition model with the McAdams correlation was included for fully developed two-phase flow, and the Seider-Tate correlation was used for the single-phase forced convection regime. The lss-of-flow transient after a trip time of 408 sec at 85% of loss of normal flow for the TRIGA core shows a peak temperature of 709.220C in the fuel centerline and 131.94'C in the clad and 46.63'C in the coolant exit of the hottest channel. The transient was terminated at 15% of nominal flow after about 48.0 sec. The time at which the reversal of coolant flow starts is about 67.0 sec. INST-90/RPED-22 1. Introduction The 3MW TRIGA MARK II research reactor was commissioned in late 1986 at Savar near Dhaka, Bangladesh. The principal objective of this study is to provide thorough analyses of two most important thermal hydraulic design parameters of the reactor, namely the departure from nucleate boiling ratio (DN-BR) and the loss-of-flow accident (LOFA). These analyses are an extension of some other studies reported recently 131. The core neutronic parameters neede& f6i the analyses were obtained from the work of Bhuiymi et al 4]. This investigation will provide a basis for updating the Final Safety Analysis Report (FSAR) of the TRIGA reactor and also for upgrading the current core configuration through bericbmarking the available TRIGA experimental and operational data both for steady-state and transient mode of reactor operations. The PARET [5] computer code that has already been used successfully elsewhere in the world for te analysis of TRIGA research reactor 6] was applied for this analysis. The important behaviors that have been studied relating DNB analysis are: (i) effect of operating power, (ii) effect of the flow rate, (iii) effect of the hot-rod factor, and (iv) the effect of the inlet temperature. For LOFA analysis the parameters those have been studied area (i) relative power and flow for the exponential 16ss-of-flow transients, (ii) the resulting fuel centerfine, clad surface and coolant outlet temperatures after the loss-of-flow transients, and (iii) the occurring of coolant flow reversal. 2. Analysis of DNB Ratio 'flee power output of the reactor is limited by three dependent thermal and hydrodynamic variables: the DNBR, the maximum fuel temperature (Tma,), and the core pressure drop (A ). The variations of these variables are discussed in the following sections. The first design limitation is related to the DNB flux, which can be computed by means of a suitable correlation. When the heat flux become sufficiently large, the small bubble forined in nucleate boiling coalesce into a vapor film that covers the surface, since then the heat4ransfer efficiency drops dramatically, and the clad surface temperature will rise by several hundred degrees. This situation is known as the DNBR. The DNBR is defined as the ratio of the ctical heat flux to the heat flux achieved the core. The reactor core should be designed so as to prevent the DNBR from dropping below a chosen value under a high heat flux transient condition for the most adverse set of mechanical and coolant conditions. The PARET code was used to analyze the thermal and hydraulic design for the 3 W TRIGA MARK 11 research reactor. Two-channel model of PARET code was used where one channel represents the hottest rod and associated flow channel, and other average channel to represent the remaining 99 fuel rods a volume weighted sense. The axial source distribution was represented by 15 axial regions. The correlations from Dittus-Boelter 71 for single-phase, McAdams [8] for two-phase and Bernath 9] for ctical heat flux analysis were used in PARET calculation and the mode of transition model was selected. The Nusselt number used in the revised single-p ase heat transfer coefficient subroutine was 140 for laminar flow and the enhance effects of laminar flow were also included. Detailed input preparations were reported by Huda et al [ 1 3]. in forced convection, the heat flux and the wall and bulk temperatures are related by q j -_h (T, - T) INST-90/RPED-22 where, q = forced convection heat flux h forced convection heat transfer coefficient T, = wall temperature Tb coolant bulk temperature The forced-convection heat transfer coefficient h was determined with the Dittus-Boelter correlation as recom ended by Tong and Weisman 7], namely, Nu = .023 Re 0.1 r0.4 where, Nu, Re, and Pr are the Nusselt, Reynolds, ad Prandtl numbers, respectively, based on the bulk water properties, At higher ,eat fluxes, the surface is uniformly covered by a dynamic bubble layer, and the heat is removed by fully developed, subcooled ucleate boiling. The correlation used for this mode of heat transfer is from McAdams eal. [8], namely, q, 0074 A Ts'-" where, qfd is the heat flux for fully developed subcooled nucleate boiling, and AT, is the difference between te surface and saturation temperature. In the fully developed nucleate boiling regime it is possible to increase the heat flux further without an appreciable change the surface temperature, unti te bubble motion o the Surface becomes so violent that a hydrodynamic crisis occurs with the formation of a continuous vapor Film in the surface. This is termed DN13, and the heat flux is the critical heat fILIX (CHF). The ratio of the CHF to the actual heat flux is the DNBR. In subcooled boiling the CHF is a function of the coolant velocity, the degree of subcooling, and the pressure. The correlation used to predict the CHF is from Bernath 9] and is given by q = k (Tw, TO where, h, = 10,890 [D,/(D, + Di)] + sV s = 48/DO-' for D < 003048 n = 90 I De) for D > 003048 n Tw, I S 57 In P 54[P/(P + 15) - V/41 32, (critical wall temperature) D, hydraulic diameter of coolant channel Di diameter of heated surface V coolant velocity P pressure TB bulk coolant temperature The design flow Tate has a lower limit determined by the value of the CHF at that flow rate- the larger the flow rate the larger will be the CHF and hence the safety margin. he flow rate also has an upper limit, which is determined by the maximum allowable pressure drop through the core to avoid cavitation in the flow system. The design conditions used for the PARET calculation are tabulated in Table . Since the enthalpy of te coolant increases as i flows through the core, the computed DNB flux decreases correspondingly, as depicted in Fig. 1, The heat flux along the hot channel and the DNBR are also sown in the figure qualitatively. Over the length 03 8 1m: of the hottest channel the DNBR varies, starting from 38951 to 5403 1, with a minimum of 2785 1. The peak heat flux occurs at the axial center of the fuel elements-, therefore the DNB ratio is minimum at this location. 2 INST-90/RPED-22 Table 1. The design conditions used for the PARET calculation for DNB analysis Operating Power 3MW Flow rate 3208.9 kg/mz-s 3500 gpm) Inlet temperature 40.5556 C (105 OF) Pressure L60654XI0`kPa Hot Rod Factor (BFR) 1.854 (C4 element) Inlet pressure loss coefficient 1.81 Outlet pressure loss coefficient 2.12 2-1. Effect of Operating Power The mm'imum DNBR and the corresponding maximum fuel temperature as a function of operating power are shown in Fig. 2 The correlations [5] used for calculating the critical heat flux/DNBR in the PARET code were Tong, Bemath and Mirshak among which Bemath gives the minimum value, and according to the suggestions made by the General Atomics (GA) the Bemath correlation was selected [5].
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