Supercritical-pressure Light Water Reactor Research Articles

Thermo hydraulic analysis of narrow channel effect in supercritical-pressure light water reactor

The size of the gap between fuel rods has important effects on flow and heat transfer in a supercritical-pressure light water reactor. Based on thermal analysis at different coolant flow rates, the reasonable value range of gap size between fuel rods is obtained, for which the maximum cladding temperature safety limits and installation technology are comprehensively considered. Firstly, for a given design flow rate of coolant, thermal hydraulic analysis of supercritical pressure light water reactor with different gap sizes is provided by changing the fuel rod pitch only. The results show that, by means of reducing the gap size between fuel rods, the heat transfer coefficients between coolant and fuel rod, as well as the heat transfer coefficient between coolant and water rod, would both increase noticeably. Furthermore, the maximum cladding temperature will significantly decrease when the moderator temperature is decreased but coolant temperature remains essentially constant. Meanwhile, the reduction in the maximum cladding temperature in the inner assemblies is much larger than that in the outer assemblies. In addition, the maximum cladding temperature could be further reduced by means of increasing coolant flow rate for each gap size. Finally, the characteristics of narrow channels effect are proposed, and the maximum allowable gap between fuel rods is obtained by making full use of the enhancing narrow channels effect on heat transfer, and concurrently considering installation. This could provide a theoretical reference for supercritical-pressure light water reactor design optimization, in which the effects of gap size and flow rate on heat transfer are both considered.

Principle of rationalizing the criteria for abnormal transients of the Super LWR with fuel rod analyses

A detailed fuel rod design is carried out for the first time in the development of Supercritical-pressure Light Water Reactor (Super LWR). The fuel rod design is similar to that of LWR, consisting of UO 2 pellets, a gas plenum and a Stainless Steel Cladding. The principle of rationalizing the criteria for abnormal transients of the Super LWR is developed. The fuel rod integrities can be assured by preventing plastic strains on the cladding, preventing the cladding buckling collapse, and keeping the pellet centerline temperature below its melting point. The FEMAXI-6 fuel analysis code is used to evaluate the fuel rod integrities in abnormal transient conditions. Detailed analyses have shown that allowable limits to the maximum fuel rod power and maximum cladding temperature can be determined to assure the fuel integrities. These limits may be useful in the plant safety analyses to confirm the fuel integrities during abnormal transients.

Fuel and Core Design of Super Light Water Reactor with Low Leakage Fuel Loading Pattern

An equilibrium core for the High Temperature Supercritical-pressure Light Water Reactor, now called the Super Light Water Reactor (Super LWR), has been designed. The fuel assemblies loaded in the peripheral region of the core are cooled with descending flow to achieve a high average coolant core outlet temperature. This flow scheme is compatible with a low leakage fuel loading pattern (LLLP) in which 3rd cycle fuel assemblies are loaded in the core peripheral region. Stainless steel is used for fuel rod claddings and for structural materials. Watts correlations are used for predicting heat transfer in the core. They take into account the improved heat transfer for downward flow. It is found that the water rods with their downward flow need to be thermally insulated with thin ZrO2 layer to keep the moderator temperature below the pseudo critical temperature and to reduce the thermal stress in water rod walls. An average coolant core outlet temperature of 500°C is achieved. The effects of various heat transfer correlations on the cladding surface temperature are evaluated.

Development of Statistical Thermal Design Procedure to Evaluate Engineering Uncertainty of Super LWR

The thermal performance of a nuclear reactor core contains various engineering uncertainties. These uncertainties often arise from calculation, measurement, instrumentation, manufacture, fabrication and data processing. Statistical techniques are useful to evaluate and combine these uncertainties in the thermal design of nuclear reactors. In this paper, a statistical method is developed and employed in the thermal design of the supercritical pressure light water reactor (Super LWR) to evaluate the statistical engineering uncertainties. This method is referred as the Monte Carlo Statistical Thermal Design Procedure for Super LWR (MCSTDP). This method uses the maximum cladding surface temperature (MCST) as a crucial criterion and a sub-channel code is utilized to perform the core thermal hydraulic analysis. The engineering uncertainties are considered with strict respect to the 95/95 limit of Super LWR. The main purpose of this paper is to establish the statistical evaluation methodology. The engineering uncertain for the thermal design of Super LWR is evaluated by using this method to get an approximate quantification. The results are compared with those of the Revised Thermal Design Procedure (RTDP) of Super LWR.

スーパー軽水炉 (高温超臨界圧軽水冷却炉) (2) 熱中性子炉改良炉心設計

スーパー軽水炉 (高温超臨界圧軽水冷却炉) (2) 熱中性子炉改良炉心設計

Thermal and Stability Considerations of Super LWR during Sliding Pressure Startup

The feasibility of the sliding pressure startup of a high-temperature supercritical-pressure light water reactor (super LWR, SCLWR-H) is assessed from both thermal and stability considerations. In the sliding pressure startup, nuclear heating starts at subcritical pressure and the reactor is pressurized to supercritical pressure at a low power and high enough flow rate. The reactor power and flow rate are then raised gradually to the rated normal values at constant supercritical operating pressure. During startup, the maximum cladding surface temperature must not exceed 620°C. For two-phase flow at subcritical pressures, the homogeneous equilibrium model is used. The thermal-hydraulic and coupled neutronic thermal-hydraulic stabilities during pressurization and power-raising are investigated by a frequency-domain linear analysis for both supercritical-pressure and subcritical-pressure operating conditions. The same stability criteria as those of BWRs are used. From the analysis results, a sliding pressure startup procedure is proposed for super LWR. The thermal criteria are satisfied by keeping the core power between the maximum allowable limit and minimum limit required for turbine startup and operation. The thermal-hydraulic stability and coupled neutronic thermal-hydraulic stability can be maintained by applying an orifice pressure drop coefficient at the inlet of fuel assembly and by controlling the power and flow rate during startup.

Improved core design of the high temperature supercritical-pressure light water reactor

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C. In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained. In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.

Three-dimensional Core Design of High Temperature Supercritical-Pressure Light Water Reactor with Neutronic and Thermal-Hydraulic Coupling

The equilibrium core of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H) is designed by three-dimensional neutronic and thermal-hydraulic coupled core calculations. The average coolant core outlet temperature of 500°C is accurately evaluated for the first time in the development of the SCLWR-H. The average coolant core outlet temperature is one of the key parameters, which must be accurately determined in order to establish the concept of this unique reactor design. However, it can only be determined by three-dimensional core design method, taking into account the control rod patterns, fuel loading patterns, coupling of the neutronic and thermal-hydraulic calculations, and burnup distribution of each fuel assembly. The R—Z two-dimensional core design method used in previous studies could not model or evaluate such parameters with sufficient accuracy. In this study, a three-dimensional equilibrium core design method for the SCLWR-H is established. This method can accurately evaluate the average coolant core outlet temperature and has permitted a comprehensive equilibrium core to be developed, which satisfies all design criteria. The design criteria are maximum fuel rod cladding surface temperature of 650°C, maximum linear heat generation rate of 39 kW/m, and a positive water density reactivity coefficient.

Three-dimensional Core Design of High Temperature Supercritical-Pressure Light Water Reactor with Neutronic and Thermal-Hydraulic Coupling

The equilibrium core of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H) is designed by three-dimensional neutronic and thermal-hydraulic coupled core calculations. The average coolant core outlet temperature of 500°C is accurately evaluated for the first time in the development of the SCLWR-H. The average coolant core outlet temperature is one of the key parameters, which must be accurately determined in order to establish the concept of this unique reactor design. However, it can only be determined by three-dimensional core design method, taking into account the control rod patterns, fuel loading patterns, coupling of the neutronic and thermal-hydraulic calculations, and burnup distribution of each fuel assembly. The R—Z two-dimensional core design method used in previous studies could not model or evaluate such parameters with sufficient accuracy. In this study, a three-dimensional equilibrium core design method for the SCLWR-H is established. This method can accurately evaluate the average coolant core outlet temperature and has permitted a comprehensive equilibrium core to be developed, which satisfies all design criteria. The design criteria are maximum fuel rod cladding surface temperature of 650°C, maximum linear heat generation rate of 39 kW/m, and a positive water density reactivity coefficient.

高温超臨界圧軽水炉 (8) 改良炉心設計

高温超臨界圧軽水炉 (8) 改良炉心設計

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.

Core Design of a Direct-Cycle, Supercritical-Pressure, Light Water Reactor with Double Tube Water Rods

Core design of a direct-cycle, supercritical-pressure, light water reactor (SCLWR) is carried out. Double tube water rods are used for enhancing the moderation. Whole coolant flows upward in the inner tube and downward in the outer tube of the water rod, and then the fuel is cooled. The double tube water rods flatten the axial power distribution easier than the single tube water rods. Since the hot coolant from the fuel channel does not mix with the cold coolant from the water rod and keeps high temperature, the thermal efficiency is not deteriorated. Whole coolant flows through the fuel channel and the mass flow rate is high. The critical heat flux, hence, increases and the enthalpy difference between the inlet and the outlet becomes large. The core height decreases to 3.7m from the core using the single tube water rod of 5.7m. The axial power distribution was further flattened by changing the gadolinia concentration and the fuel enrichment. Clustered control rods like PWR are adopted as the primary reactivity control system. Borated water injection system is adopted as the backup system. Each system attains cold shutdown independently. The designed plant has thermal efficiency of 40.8%, gross electric power of 1,006 MW.

Core Design of a Direct-Cycle, Supercritical-Pressure, Light Water Reactor with Double Tube Water Rods.

Core design of a direct-cycle, supercritical-pressure, light water reactor (SCLWR) is carried out. Double tube water rods are used for enhancing the moderation. Whole coolant flows upward in the inner tube and downward in the outer tube of the water rod, and then the fuel is cooled. The double tube water rods flatten the axial power distribution easier than the single tube water rods. Since the hot coolant from the fuel channel does not mix with the cold coolant from the water rod and keeps high temperature, the thermal efficiency is not deteriorated. Whole coolant flows through the fuel channel and the mass flow rate is high. The critical heat flux, hence, increases and the enthalpy difference between the inlet and the outlet becomes large. The core height decreases to 3.7m from the core using the single tube water rod of 5.7m. The axial power distribution was further flattened by changing the gadolinia concentration and the fuel enrichment. Clustered control rods like PWR are adopted as the primary reactivity control system. Borated water injection system is adopted as the backup system. Each system attains cold shutdown independently. The designed plant has thermal efficiency of 40.8%, gross electric power of 1,006 MW.

Design of water rod cores of a direct cycle supercritical-pressure light water reactor

A conceptual design of a direct-cycle supercritical-pressure light water reactor with water rods is presented. Three types of water rods are analyzed: single, semi-double and full-double tubes. A water rod replaces seven fuel rods in a triangular lattice. The coolant density change in the water rods and the fuel channel is calculated using a code developed in the present study. The full double tube is most superior in terms of the distribution of the moderator. The number of fuel rods to water rods is 198:19, which makes optimum moderation. The average enrichment becomes 4.13%. The axial power flattening is finally achieved by partial length fuel rods and enrichment split of 0.25%. The discharge burnup is 45 GWd/t.

Large-break loss-of-collant accident analysis of a direct-cycle supercritical-pressure light water reactor

Large-break loss-of-coolant accident (LOCA) was analyzed in the course of the design study concerning direct-cycle supercritical-pressure light water reactor (SCLWR). The advantages of SCLWR are a higher thermal efficiency and simpler reactor system than the current light water reactors (LWRs). A computer code was prepared for the analysis of the blowdown phase from the supercritical pressure. The calculation was connected to the REFLA-TRAC code when the system pressure decreased to around atmospheric pressure. The analyzed accidents are 100, 75, 50 and 25% cold-leg and 100% hot-leg breaks. First, blowdown and heatup phases without an emergency core cooling system (ECCS) were evaluated. A low-pressure coolant injection system (LPCI) was designed to fill the core with water before the cladding (stainless-steel) temperature reached a limit of 1260°C. The LPCI consists of four units, each of which has the capacity 805 kg/s. An automatic depressurization system (ADS) was designed to release the steam generated in the core in the case of cold-leg breaks and to permit operation of LPCI in the case of LOCAs of less than 100% break. For all cases analyzed, the peak cladding temperatures were lower than the limit when the designed ECCS is implemented.