Pressure Startup Research Articles

Study on thermal hydraulic characteristics under startup of SCWR

To investigate the startup characteristics of the supercritical pressure water-cooled reactor (SCWR) system, a complete startup system model of the SCWR was established with the analysis code SCTRAN, based on the high-performance light water reactor (HPLWR) steam cycle and SCWR circulation startup loop. The correctness of the model was verified in comparison with the steady-state parameters of the steam cycle of the HPLWR. A sliding pressure startup procedure with a circulation loop that employs a control system was analyzed, and the transient performances of the core, steam drum, turbine, reheaters, steam extraction and heaters at each stage were obtained. The calculation results showed that the startup sequence and startup thermal parameters could fully meet the expectations: the system starts up stably and the core remains in the single phase; the inlet steam of the turbine stays superheated; the core inlet temperature can reach 280 °C after the adoption of the high-pressure and low-pressure heaters; and the inlet pressure of the high-pressure turbine can be kept constant. During the startup process, the maximum cladding surface temperature (MCST) remains below the limit of 650 °C, suggesting that the entire startup procedure is safe and reliable.

Control and thermal analysis for SCWR startup

Startup system, the design of startup sequences analysis is an important part of SCWR design. A thermal hydraulic system analysis code for supercritical water reactor named SCTRAN is used to model the entire startup system based on the circulation loop for startup and once-through direct cycle. The problem of the heat transfer coefficient (HTC) does not accurately capture deterioration phenomenon, the HTC is calculated as a discontinuity in the mode transfer region, its low prediction accuracy above the quasi-critical region have been solved by the new wall heat transfer model. Especially, the look-up table would not be used to obtain the HTC and achieves high prediction accuracy across the critical region, unless the pressure is higher than 19 MPa. After that, to get a smooth recirculation variable pressure startup process, the system model integrates the control system which can controls the temperature, the steam drum water level, the thermal power, and the coolant flow rate. Based on the CSR1000 core and entire once-through direct cycle and circulation loop for startup, four stages under control systems, from low pressure to full power condition in recirculation startup process, were analyzed with code SCTRAN and wall heat transfer model was modified. The calculation results show that the recirculation system can startup from subcritical state to full power state without issue of CHF. The control system can control the parameters quite well and maximum cladding temperature (MCST) can be limited under 650 °C in the startup process. The modified SCTRAN code in this paper can further expand the computational range and computational accuracy. The full-scale control system can meet the needs of parameters expected response.

Parametric study of CSR1000 thermal hydraulic stability

A parametric study is performed on thermal hydraulic stability of CSR1000 (a Chinese supercritical water-cooled reactor) using SCTHSAC (Single Channel Thermal Hydraulic Stability Analysis Code). This code uses RMM (system response matrix method) for analyzing the system stability in the frequency domain, a useful method to reveal detailed information about the operational thermodynamic parameters before and after onset of instability in the nuclear reactor. System response matrices are written for each node of the single coolant channel (by linearizing and perturbing conservation differential equations), which are solved for decay ratios. The maximum decay ratio represents stability of the whole system. The decay ratios are calculated for different values of reactor power and inlet orifice diameters. A parametric study of pressure, inlet temperature and moderator flow fraction is presented, which gives vital information about the thermal hydraulic stability bounds of CSR1000 in a sliding pressure startup scheme. The results show a similar trend with the stability charts of a typical SCWR, proving the novelty of RMM. Using this method, a stability surface is generated against power and flow fraction, giving a quick reference for designing a detailed startup scheme of CSR1000 in the future.

The stability-analysis code FIAT development for density wave oscillations and its application to PV/PT SCWR

SuperCritical Water Reactor owns a high operation pressure and a wide coolant temperature range. The thermo-physical properties and transport properties of the coolant in the core would change greatly both in the operation condition and startup process, which may causes density wave oscillations. In this paper, a stability analysis code named FIAT is developed based on frequency domain method to analyze the density wave oscillations (DWO) for SCWR. Frequency domain method and assumptions of constant pressure drop between core inlet and outlet are applied in the code. The neutron/thermal-hydraulic coupling characteristic equation is derived through using linearization and laplace transformation for the mass, momentum and energy conservation equations. With analyzing the characteristic equation, the onset of DWO instabilities can be confirmed with non-dimensional parameters plane. The prediction accuracy of the FIAT code is verified through experiment data and code-to-code comparison under subcritical and supercritical pressure. The FIAT code is applied to evaluate the effects of basic core model on the stability for pressure vessel type SCWR and optimize the startup procedure for pressure tube type SCWR. The results show that for pressure vessel type SCWR, fuel rods heat transfer model and neutron kinetics model help stabilize the system while water rods model produces an opposite effect. For pressure tube type SCWR, two pressurization phases are required during sliding pressure start-up process to meet both the thermal–hydraulic and stability requirements.

Various startup system designs of HPLWR and their thermal analysis

This paper summarizes the results of various startup system designs and their thermal analysis of the high performance light water reactor (HPLWR) which is the European version of the various supercritical water cooled reactor proposals. In order to study the thermal-hydraulic characteristics of the HPLWR core, a simplified axial one-dimensional (1D) single channel model is developed, which consists of fuel, cladding, coolant and moderator. The model is verified by the related results of Seppälä (2008). Both constant pressure startup systems and sliding pressure startup systems of HPLWR are presented. In constant pressure startup system, the reactor starts at supercritical pressure. It appears that compared with other SCWR designs, the weight of the component required for constant pressure startup of HPLWR is medium and reasonable. Constant pressure startup systems are found feasible from thermal analysis. And for sliding pressure startup, the reactor starts at subcritical pressure. The adequate core power of 25% with 28% flow rate and a feedwater temperature of 280°C are determined during pressurization phase. The thermal analysis results show that the sliding pressure startup systems for HPLWR are also feasible. Considering the same flow rate as the supercritical-pressure light water-cooled fast reactor (SCFR), the component weight required is reduced in HPLWR.

Thermal–hydraulic performance analysis of IPWR during full pressure start-up mode

This study analyzes the structural characteristics and the dynamic performance of an integrated pressurized water reactor (IPWR). The primary loop of the reactor is relatively simplified. Thereafter a description of the system’s mathematical and physical models are established. The models of the primary loop including the core model, the simplified pump model, the once-through steam generator (OTSG), the flow friction and heat transfer models are built and then coded with Fortran 90. Meanwhile, all possible flow and heat transfer conditions are considered and the corresponding optional models are also presented. The GEAR method is adopted for the numerical solution of transient equations. The full pressure start-up mode is used to control the IPWR based on these models. The results show that the calculated parameters are in good agreement with the design values. Core inlet and outlet temperatures, OTSG secondary side temperature and pressurizer pressure are within the safety limit values, which shows that the mode of IPWR is reasonable. The proposed mode is suitable for not only the ship nuclear power systems but also the small or medium size nuclear power plants in both the economic and technological aspects. The present work provides an important theoretical basis and reference for the control strategy of an IPWR. Additionally, because of the adoption of modular programming techniques, both the steady- and transient-state analysis codes can be easily applied to other thermal–hydraulic analyses of special reactors and special cases by modifying and updating the corresponding function modules.

Investigation of Gearbox Vibroactivity with the Use of Vibration and Acoustic Pressure Start-Up Characteristics

Investigation of Gearbox Vibroactivity with the Use of Vibration and Acoustic Pressure Start-Up CharacteristicsThis article presents the results of investigating the influence of tooth contact ratio in helical cylindrical gears on vibroactivity of the gearbox. Based on the measurements carried out on a laboratory test stand, time-domain and frequency-domain start-up characteristics of vibrations and acoustic pressure were determined, and vibroactivity was assessed for a gearbox in which 4 pairs of gears were successively mounted with different face contact ratios equal to, respectively, εβ=1,001; 1,318; 1,574; 2,636.

Thermal consideration of CANDU-SCWR sliding pressure startup through subchannel analysis

A startup system and startup procedures are designed based on the subchannel analysis for CANDU-SCWR sliding pressure startup. Lookup tables are selected to predict the CHF and PDO heat transfer due to their wide application range. Plant parameters are analyzed in detail. The results show that the maximum cladding surface temperature can be well restricted lower than criterion (850 °C), and the proposed startup procedure is feasible for CANDU-SCWR from the point of view of thermal-hydraulics.

Hot-Channel Stability of Supercritical Water-Cooled Reactors—I: Steady-State and Sliding Pressure Startup

The drastic change of fluid density in the reactor core of a supercritical water-cooled reactor (SCWR) gives rise to a concern about density-wave stability. Using a single-channel thermal-hydraulic model, stability boundary maps for the U.S. reference SCWR design have been constructed for both steady state and sliding pressure startup conditions.The supercritical water flow in the reactor core has been simulated using a three-region model: a heavy fluid with constant density, a mixture of heavy fluid and light fluid similar to a homogeneous-equilibrium two-phase mixture, and a light fluid, which behaves like an ideal gas or superheated steam. Two important nondimensional numbers, namely, a pseudosubcooling number Npsub and an expansion number Nexp, have been identified for the supercritical region. The stability map in the supercritical region is then plotted in the plane made of these two numbers. The U.S. reference SCWR design operates in a stable region with a large margin. Sensitivity studies produced results consistent with the findings of the earlier research done for the subcritical two-phase flow.During the sliding pressure startup of the SCWR, a two-phase steam-water mixture at subcritical pressure will appear in the reactor core. A nonhomogeneous (e.g., drift-flux) nonequilibrium two-phase flow model was applied. The characteristic equation was numerically integrated, and stability boundary maps were plotted on the traditional subcooling number versus phase change number (or Zuber number) plane. These maps have been used to develop a sliding pressure SCWR startup strategy avoiding thermal-hydraulic flow instabilities.

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.

Thermal and Stability Considerations of Super LWR during Sliding Pressure Startup

Thermal and Stability Considerations of Super LWR during Sliding Pressure Startup

A Linear Stability Analysis of Supercritical Water Reactors, (I)

This paper summarizes the analysis results of the thermal-hydraulic stability of a high-temperature reactor cooled and moderated by supercritical-pressure light water (SCLWR-H). A linear stability analysis code in the frequency domain was developed to study the thermal-hydraulic stability of SCLWR-H at constant supercritical pressure. The analysis method is based on linearization by perturbation of numerically-discretized one-dimensional single-channel single-phase conservation equations. The effect of water rods on stability is considered. The thermal-hydraulic stability of SCLWR-H for full-power and partial-power normal operations was investigated by frequency domain method. Our analysis reveals that though SCLWR-H has low coolant flow rate and large density change in the core, the thermal-hydraulic stability can be maintained both at normal operation and during power raising phase of constant pressure startup by applying an orifice pressure drop coefficient at the inlet of the fuel assemblies. A parametric study was also carried out to determine the parameters affecting the stability.

Startup Thermal Analysis of a High-Temperature Supercritical-Pressure Light Water Reactor

The startup systems of a high-temperature supercritical-pressure light-water-cooled thermal reactor (SCLWR-H), in which the core outlet temperature is 500°C and downward-flowing water rods are used as moderators, are studied by thermal-hydraulic analysis. The thermal analyses are carried out for various startup phases and detailed procedures for these phases are investigated. In constant pressure startup system, the reactor starts at supercritical pressure. A flash tank and pressure-reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at subcritical pressure. A steam-water separator and a drain tank are required. The separator is designed by referring to those of supercritical fossil-fired power plants (FPPs). The maximum cladding surface temperature is restricted not to exceed the rated value of 620°C. The minimum flow rate is 25% for constant pressure startup and 35% for sliding pressure startup. Both constant pressure and sliding pressure startup systems are found feasible from thermal analysis. Because of lower flow rate than SCFR, of which the core outlet temperature is about 430°C, the component weight required is reduced in SCLWR-H. The sliding pressure startup system should be used to reduce the component weight and to simplify the plant system.

Startup Thermal Analysis of a High-Temperature Supercritical-Pressure Light Water Reactor

The startup systems of a high-temperature supercritical-pressure light-water-cooled thermal reactor (SCLWR-H), in which the core outlet temperature is 500°C and downward-flowing water rods are used as moderators, are studied by thermal-hydraulic analysis. The thermal analyses are carried out for various startup phases and detailed procedures for these phases are investigated. In constant pressure startup system, the reactor starts at supercritical pressure. A flash tank and pressure-reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at subcritical pressure. A steam-water separator and a drain tank are required. The separator is designed by referring to those of supercritical fossil-fired power plants (FPPs). The maximum cladding surface temperature is restricted not to exceed the rated value of 620°C. The minimum flow rate is 25% for constant pressure startup and 35% for sliding pressure startup. Both constant pressure and sliding pressure startup systems are found feasible from thermal analysis. Because of lower flow rate than SCFR, of which the core outlet temperature is about 430°C, the component weight required is reduced in SCLWR-H. The sliding pressure startup system should be used to reduce the component weight and to simplify the plant system.

A Linear Stability Analysis of Supercritical Water Reactors, (I) Thermal-Hydraulic Stability

This paper summarizes the analysis results of the thermal-hydraulic stability of a high-temperature reactor cooled and moderated by supercritical-pressure light water (SCLWR-H). A linear stability analysis code in the frequency domain was developed to study the thermal-hydraulic stability of SCLWR-H at constant supercritical pressure. The analysis method is based on linearization by perturbation of numerically-discretized one-dimensional single-channel single-phase conservation equations. The effect of water rods on stability is considered. The thermal-hydraulic stability of SCLWR-H for full-power and partial-power normal operations was investigated by frequency domain method. Our analysis reveals that though SCLWR-H has low coolant flow rate and large density change in the core, the thermal-hydraulic stability can be maintained both at normal operation and during power raising phase of constant pressure startup by applying an orifice pressure drop coefficient at the inlet of the fuel assemblies. A parametric study was also carried out to determine the parameters affecting the stability.

Startup Thermal Considerations for Supercritical-Pressure Light Water-Cooled Reactors

Supercritical-pressure light water-cooled reactors (SCRs) are innovative systems aimed at high efficiency and cost reduction. The once-through direct-cycle plant system is the leading system of fossil-fired power plants (FPPs). Estimates of the coolability and necessary sizes of the SCR startup systems, sequences, and required equipment for startup are investigated with reference to supercritical FPPs. There are two types of supercritical boilers. One is a constant pressure boiler, and the other is a variable pressure boiler.First, startup of the constant pressure boiler is examined. The reactor starts at a supercritical pressure. A startup bypass system consisting of a flash tank and pressure-reducing valves is required. Second, startup of the variable pressure boiler is investigated. The reactor starts at a subcritical pressure, and the pressure increases with the load. A steam-water separator and a drain tank are required for startup.The results of computer calculations show that with both constant pressure and variable pressure startup, the peak cladding temperature does not exceed the operating limit through startup, and both startup sequences are feasible. The sizes of the components required for the startup systems are assessed. To simplify the plant system and to reduce the component size, variable pressure startup with steam separators in the bypass line appears desirable.

Thermo-hydraulic instability of boiling natural circulation loop induced by flashing (analytical consideration)

An analytical study is presented on the thermo-hydraulic stability of a boiling natural circulation loop with a chimney at low pressure start-up. The effect of flashing induced by the pressure drop in the channel and the chimney due to gravity head on the instability is considered. A method to analyze linear stability is developed, in which a drift-flux model is used. The analytical result of a stability map agrees very well with the experimental one obtained in a previous report. Instability does not occur when the heater power is too low to generate voids in the chimney and only natural circulation of single phase can be induced. Instability tends to occur when boiling occurs only near the chimney exit due to flashing. This instability phenomenon has some similarities with density wave oscillation, such as the phase difference of temperature between the boiling region and non-boiling region, and the oscillation period which is near to the time required for fluid to pass through the chimney. However, there are also some differences from density wave oscillation, such as the boiling region is very short, and pressure fluctuation can affect void fraction fluctuation.