Coolant density feedback

The setpoint of the main steam temperature increases stepwise from 500 to 504°C The results are shown in Figs. 4.27 and 4.28. The feedwater flow rate is decreased so as to increase the main steam temperature. Although the power decreases by the coolant density feedback, it is only about 2%, and then, the power returns to the... 

Only the partial loss of reactor coolant flow is accompanied by a decrease in the main coolant flow rate before initiating the ADS. The ADS is actuated at 5 s by the ATWS signal which is reactor coolant pump trip and reactor power ATWS permissive for 5 s . The calculation results are shown in Fig. 6.48. A decrease in the flow rate leads to an increase in the coolant temperature due to the power and flow mismatch. The cladding temperature increases due to the coolant heat-up and a decrease in the heat transfer coefficient. The net reactivity and the reactor power decrease due to coolant density feedback. The increase in the cladding temperature is about 120°C, which is the highest value of all the ATWS events with the alternative action. 

As noted above, uncertainties in the nuclear data, the calculation model, etc., are taken into account when establishing an error band of 20% of the Doppler coefficient. The range of —1.17 to 2.40 pcm/K is assumed for the sensitivity analysis. Even if the Doppler coefficient is 1.17 pcm/K, the maximum fuel enthalpy of the reactivity insertion events without the alternative action is 148 cal/g, which is higher than that of the reference case by only 1% because the contribution of coolant density feedback becomes higher. The peak power of the flow increasing events without the alternative action is 154% of the rated power with a Doppler coefficient of — 1.17 pcm/K. It is higher than the reference case but other parameters, i.e., peak temperature and peak pressure, are almost the same. 

Since the reactor power temporarily decreases due to the coolant density feedback in this event the change in the feedwater flow rate is slowed down by the second term of (7.29), so that the main steam temperature reaches the new setpoint without overshoot with Control system (B). On the other hand, the change in the feedwater flow rate is accelerated by the second term of (7.30), so that the settling time is shorter with Craitrol system (C) than that with the reference control system. 

Depending on the core conditions (power level), the reactivity feedback coefficient due to the coolant density may either be negative or slightly positive. Ingress of gas into the core is likely during injection by the hydroaccumulators. This also adds to the coolant density variations, which are bound to affect the reactivity. 

Reactivity control mechanism - Shutdown rod for reactor start-up and shutdown. - During operation, reactor power autonomously load follows by means of inherent physical processes without the need for any motion of control rods or any operator actions. - System temperatures change corresponding to reactivity feedbacks from fuel Doppler, fuel and cladding axial expansion, core radial expansion, and coolant density effects. - Control rods for possible line reactivity compensation during cycle. - Control rods also provide for diverse and independent shutdown. ... 

System temperatures change corresponding to reactivity feedbacks from fuel Doppler, fuel and cladding axial expansion, core radial expansion, and coolant density effects. 

The performance of the SPINNOR reactors under ULOF is shown in Fig. XXVI-3 to XXVI-6. Figure XXVI-3 indicates that following the loss of pumping power in the primary system, the flow rates in primary system and at the primary side of the steam generator (SG) decrease and progress toward the level of natural circulation. It causes the increase of temperatures as shown in Fig. 4, and results in the negative feedbacks including, in the order of importance, coolant density decrease, fuel axial expansion, Doppler effect, and core radial expansion, as shown in Fig. XXVI-5. These feedbacks assure the decrease of reactor power to match the new coolant flow rate, as shown in Fig. XXVI-6. 

The elimination or at least minimization of the positive coolant density component of the temperature reactivity effect is a favourable factor in limitating the consequences of ULOF and UTOP. The negative reactivity feedback caused by a thermal expansion of the control rod drive lines also plays an important role. Analysis of a ULOF accident shows that sodium boiling and fuel melt are excluded because the core outlet temperature does not exceed 800 C. Nevertheless a refractory sodium-cooled tray beneath the core is provided to prevent release of corium beyond the reactor vessel boundary and formation of a critical mass. Preliminary safety analysis for the BN-1600M reactor plant shows that ... 

A positive reactivity of 0.1 is inserted stepwise as a reactivity perturbation. The feedwater flow rate and the turbine control valve opening are kept constant. The results are shown in Figs. 4.9 and 4.10. The power quickly increases to 111% of the initial value. It is consistent with the analytical solution of prompt jump. Then, the power decreases due to reactivity feedbacks from Doppler and coolant density. The main steam temperature changes by following the power. The main steam pressure and the core pressure increase due to increases in the temperature and hence the volume flow rate of the main steam. The fuel channel inlet flow rate changes with the core pressure due to the relation between the feedwater flow rate and the core pressure shown in Fig. 4.4. The plant almost reaches a new steady state in 40 s. 

Since the coolant temperature and hence the density ratio decrease, the density feedback effect is reduced and the stability is improved (neutronic feedback). 

This is a typical flow increasing transient. The demand of the main coolant flow rate is assumed to rise stepwise up to 138% of the rated flow as is assumed in the feedwater control system failure of Japanese ABWRs. Since increase in the core coolant flow rate is mild in ABWRs due to the large recirculation flow, the feed-water flow rate is assumed to increase stepwise. This assumption is too conservative for the Super LWR. The main coolant flow rate is gradually increased by the control system in the safety analysis. The calculation results are shown in Fig. 6.31. The reactor power increases with the flow rate due to water density feedback. A scram signal is released when the reactor power reaches 120% of the rated power. The maximum power is 124% while the criterion is 182%. The increase in the pressure is small. The sensitivity analysis is summarized in Table 6.15. 

The Super LWR has self-controllability of the reactor power against loss of flow and reactivity insertion, like LWRs, due to coolant density and Doppler feedbacks although reverse-flow in the downward-flow water rods slightly complicates the behavior of density feedback. The wide-range sensitivity analyses show that variation of the feedback coefficients does not significantly influence the self-controllability or the safety margin. 

A positive reactivity ( 0.1) is inserted stepwise by withdrawing the CRs. The feedwater pump speed and the turbine control valve stroke are kept constant. The results are shown in Fig. 7.68 [31]. The reactor power increases about 10% almost stepwise due to the prompt jump and then gradually decreases due to the reactivity feedbacks from the fuel temperature and coolant density. This behavior implies that the Super FR also has inherent self controllability of the reactor power despite the much smaller density reactivity coefficient compared to that of the Super LWR. The main steam temperature increases, which leads to an increase in the main steam and core pressures because the specific volume of the main steam increases. The increase in the core pressure leads to a decrease in the feedwater and core flow rates, which increases the main steam temperature further. As a result, the maximum increase in the main steam temperature is nearly 40°C while that in the Super LWR is only 9°C (see Fig. 4.10). 

With Control system (B), the plant response is almost the same as the reference case. This is because the reactor power does not change significantly due to the small reactivity feedback from the coolant density and hence the second term of (7.29) has almost no effect compared to the first term. 

The reactor power is not sensitive to the flow rate because the Super FR is a fast reactor with small reactivity feedback from coolant density. The reactor power is mainly regulated by the CRs. Therefore, the responses of the reactor power do not significantly differ with the four control systems including the reference one. Since the responses of the core and main steam pressures are very fast and determined by only the turbine control valves, they are almost the same with the four control systems. The changes in the main steam temperature obtained by the plant stability analyses are summarized in Table 7.37 [31]. The advantages and the issues of each control system are discussed below. 

The feedwater controller of Control system (C) also makes the flow rate follow the reactor power. When the change in the power is caused by CRs, this control system works better than the reference control system as evinced in Fig. 7.76 [31]. When it is caused by the reactivity feedback from the coolant density, the change in the main steam temperature is larger than the reference cases as in Figs. 7.79 [31] and 7.80 [31]. 

The analysis results of the partial loss of reactor coolant flow (transient) are shown in Fig. 7.98. The inlet flow rates of the three hot channels similarly decrease with the main coolant flow rate. The highest cladding temperature appears in the upward flow seed channel due to its higher value initially. The AMCST in the upward flow seed channel is about 80°C, which is higher than that in the Super LWR by 20°C. This is because the Super FR has a smaller heat capacity and also because the smaller reactivity feedback from the coolant density makes the decrease in the power slower before initiating the reactor scram. However, there is still a margin of over 60°C to the criterion. 

The analysis results of the reactor coolant flow control system failure (transient) are shown in Fig. 7.100. Since the reactivity feedback from the coolant density is smaller than that in the Super LWR, the increase in the power is smaller and hence the reactor is not tripped. 

Besides the high power density, the small reactivity feedback from the coolant density makes the ATWS behavior of the Super FR severer than that of the Super LWR. However, the ATWS events of the Super FR can be mitigated by effective alternative action(s) as those of LWRs are. 

Autonomous reactor operation requires intrinsic self-regulation of reactor power. This behaviour can be obtained by negative reactivity feedback such as that provided promptly by Doppler broadening of neutron absorption resonances in resulting from increased fuel temperature, and quickly by efrects associated with coolant temperature increases and density decreases (i.e., negative void coefficient), and fuel thermal expansion. However, adequate safety consideration would also have to be given to possible reactivity insertion transients initiated by overcooling events. 

This scenario is more severe than ULOF described in (A) and, therefore, it is categorized as an AWS. Primary flow is suddenly lost because of the dielectric breakdown in the EM pumps. Even though there are two EM pump units arranged in series, both of them are assumed to suddenly fail in this scenario. A flow rate of approximately 20% of the nominal is assured by natural circulation, and a temperature rise in the fuel, coolant and steel produces negative feedbacks then, the power decreases. The rate of temperature rise is lower than in typical fast breeder reactors (FBRs) because the power density is lower in the 4S. The inherently decreased power and the convection flow rate are balanced into a steady state. As a consequence, neither coolant boiling nor fuel melting occurs. 

The AHTR has the potential to provide a highly robust safety case because of various inherent and passive safety characteristics. Inherent safety characteristics include a low-core-power density, high-heat-capacity core, and high-temperature-margin fuel. Other inherent safety characteristics of the AHTR include atmospheric pressure operation and efficient liquid-coolant heat transfer. Reactor physics for the AHTR are similar to other graphite based, coated-particle fuel systems (GT-MHR) where negative feedback comes from the high-temperature Doppler effect within the fuel. 

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