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ADVANCED ENERGY SYSTEM WITH NUCLEAR REACTORS

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ADVANCED ENERGY SYSTEM WITH NUCLEAR REACTORS ( advanced-energy-system-with-nuclear-reactors )

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withdrawal operations with burnup time and lower reactivity insertion caused by rod withdrawal. In the S-CO2-cooled core, void reactivity is about 0.72 ∆k/kk’ at BOC and 0.38 ∆k/kk’ at EOC, which is about 3/4 of that in the Na-cooled Monju core (0.9% ∆k/kk’), although 237Np is added by content of 6.5 weight percent into the core fuel and the active core length (1.2 m) is longer by 0.27 m than that of Monju (0.93 m). In the Na-cooled core, the void reactivity is 1.83% ∆k/kk’ at BOC and 1.66% ∆k/kk’ at EOC, which is about 5 $ at BOC and meets with the void reactivity limit of 5 $. The Doppler coefficients are -3.31 × 10-3 at BOC and -0.34 × 10-3 at EOC in the S-CO2-cooled core. These values are -1.64 × 10-3 and -1.23 × 10-3, respectively, at BOC and EOC in the Na-cooled core with 10 weight percent content of 237Np. Hot-spot temperatures of fuel cladding were calculated, taking into account the uncertainties in evaluation by hot-spot factors into nominal temperature obtained from subchannel analyses. The hot-spot temperatures for the S-CO2 and Na cooled cores are less than the maximum permissible temperature (700°C) of the core structural material (SUS316FR) developed for Monju. III.D. Reactor System Figure 8a) shows the direct cycle reactor system.9 The reactor pressure vessel (RPV) is connected with the power conversion unit (PCU) using a horizontal double- walled cross pipe. Hot CO2 heated to an average temperature of 527°C in the core flows into the gas turbine in PCU and the cooled CO2 returns from the gas turbine system to the core through the annulus compartment of the double-walled pipe. The double- walled pipe is enclosed in a guard pipe to prevent CO2 leakage from the outer pipe into the containment vessel during an accident of the outer pipe failure. Core decay heat is removed through natural convection at a loss-of-forced-cooling accident using an auxiliary core cooling system (ACCS). This ACCS is located in the upper part of the RPV. All control rods are actuated from the bottom of the RPV by a control rod drive mechanism. Refueling is carried out using a fuel- handling machine (FHM) of a manipulator arm type installed at the top of the RPV. Figure 8b) shows the reactor system of the indirect cycle.10 The reactor core and the intermediate heat exchanger (IHX) are installed respectively into the reactor vessel (RV) and the two IHX vessels. The RV and the IHX vessels are connected with the two hot leg and the two cold leg pipes. This three-vessel concept minimizes the use of materials. Moreover, this concept simplifies isolation of the S-CO2 gas turbine system from the reactor system and minimizes intrusion of Na-CO2 reaction products and CO2 into the core at the IHX failure accident. The RV size (diameter and height) in the direct system is almost the same as that in the indirect cycle system, in spite of the 1.4-m-larger core diameter in the direct cycle than that in the indirect cycle. III.E. Safety Plant system feasibility can be judged briefly from analyses of hypothetical representative accidents: a depressurization accident in the direct cycle and a CO2 leak to Na accident in the indirect cycle.9–10 For the temperature transient analysis in the direct cycle system during the depressurization accident, the following assumptions were made: the coolant free-flow leak area opening from the guard pipes enclosing the cold leg pipes to CV is within 50 cm2; the CO2 inventory is 860 m3 and half time of the flow coastdown is about 16 min; and the forced circulation flow rate is 2.0 m3/s. The air-flow dampers of air-blast type heat exchangers in ACCS are opened within a delay time of 60 s. Figure 9 shows that the maximum pressure and temperature in CV are respectively about 1.0 MPa and 170°C during the depressurization accident. Figure 10 illustrates the hot spot cladding temperature transient during the depressurization accident. The hot spot cladding temperature is lower than the cladding faulted limit of 900°C. The forced circulation flow rate of ACCS is designed to be 5.8 m3/s by any two loops out of the four. This result shows that the plant system allows an adequate margin to integrity of the fuel cladding in depressurization accident conditions. In the direct cycle, if CO2 in the gas turbine system leaks from IHX to Na of the primary cooling system in the IHX vessel and CO2 reacts with Na, then CO gas and Na2CO3 are generated as reaction products and the IHX cover gas pressure rises. If that high pressure is not relieved and the reaction products are not discharged adequately, the primary cooling system, including the IHX vessel, might be damaged by the high-pressure CO2: intrusion of CO2 and CO into the core might cause Na voiding reactivity, and accumulation of solid Na2CO3 on the cladding surface of fuel pins might deteriorate core cooling. In conventional Na-core-cooled steam cycle FRs, a hypothetical failure of steam generator tubes is assumed for designing a relief system in which a guillotine failure of a tube occurs, with subsequent secondary failures of six circumjacent tubes.12 The major safety concern of the indirect cycle system is whether or not S-CO2 leaked to the primary Na in the IHX vessel might cause wastage of surrounding tubes and ultimately engender damage to the IHX. Although the Na-CO2 reaction was experimentally observed to be much milder than the Na-water one, it has a threshold temperature of around 600°C for initiation of the continuous reaction.13 Simulation analyses were carried out to confirm that the relief system for the Na-CO2 reaction can prevent the excessive pressure rise in the IHX cover gas and the excessive amount of CO2 and reaction product transport to the core. The SPROUT code was used for analyses modifying it for the indirect S-CO2 cycle; the code was 5

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