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Accident analysis of supercritical water reactors during startup
Abstract
To analyze the typical characteristics of accidents occurring during the startup of a supercritical water reactor (SCWR), a comprehensive SCWR system model was established with the SCTRAN analysis code, based on models of the high-performance light-water reactor (HPLWR) steam cycle, SCWR circulation startup loop, passive safety systems, and CSR1000 core parameters. According to the sequence of the loss of coolant flow accident (LOFA), without a safety system and trigger signal of the safety system under the rated condition, a new trigger signal during the startup was designed and then used to assess “LOFA,” “uncontrolled CR withdrawal accident,” and “loss of coolant accident (LOCA)” during the startup process. The results show that the new trigger signal can ensure the effective and timely response of the safety system, as well as reactor safety during the startup process. Moreover, the maximum cladding surface temperature (MCST) of the reactor peaks (850 °C, well below the safety criterion of 1260 °C) at the end of the fourth stage of the startup process in the case of LOCA.
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In this paper the authors present the review of numerous Super-Critical Water Reactors (SCWRs) designs currently under development all over the world. The current Generation III/III + nuclear reactors operate with working fluid under subcritical conditions (below Tcr = 374 °C, pcr = 22.115 MPa). The efficiency, limited by the ratio of source and sink temperatures, is low due to operating below the critical temperature. The SCWRs are able to rise efficiency limit while operating at the supercritical conditions. The amount of energy carried by working fluid is higher leading to potential efficiency improvement of nearly 30% above current nuclear stations. Therefore, SCWRs are promising advanced nuclear systems and one of the most interesting Generation IV nuclear reactor technology, as a direct upgrade from Gen III light water reactors (LWRs). The SCWRs are a natural evolution of the LWRs just like it happened in conventional thermal power plants years ago.
However, the evolution must be done ensuring safe operation during normal and accidental conditions. The more compact size makes it possible to be used for marine applications such as propulsion system or floating nuclear power plant. Therefore, review of safety analysis has been prepared based on different reactors designs and approaches found in the literature. It was found that the conceptual designs are suitable for deployment and satisfy the design-basis safety criteria. The Japanese investigation is the most advanced one, especially the safety analysis is nearly done. Many possible evaluations of transient and accident condition had been conducted. The other teams are usually limited to ensure safety and stability during normal operating conditions.
A small scale test fuel assembly, having the characteristic design features of an SCWR, is planned to be installed inside a pressure tube to be operated at a pressure of 25 MPa and at coolant temperatures up to around 400°C inside the core of a research reactor in the Czech Republic. The pressure tube shall replace one of the ordinary fuel assemblies which are cooled with water at ambient pressure. The coolant inside the pressure tube is connected with a closed coolant loop removing the fissile power and the residual heat under normal and accidental conditions. Being a small scale nuclear facility operated with supercritical water, the safety strategy of the fuel qualification test is following a defense in depth strategy similar to the strategy used to design nuclear power plants. As the first barrier against release of radioactive material, the integrity of the fuel claddings shall not be endangered, and the safety systems needed to protect the first barrier during design basis accidents have been analyzed with system codes. The primary system with its pressure tube containing the test section is the second barrier against release of radioactive material. A rupture or even explosion of the primary system inside the reactor could eventually damage the reactor core or the control rods and must be avoided. The paper summarizes the planned safety analysis and reports about the status which has been reached up to now.
The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.
The leak before break (LBB) concept plays an important role in reactor safety. A code (LEABLE) for leak before break (LBB) leakage estimation was developed in this paper. Complicated crack morphologies and different conditions were taken into account, covering subcritical and supercritical fluids, single phase and multi-phases, choked and unchoked flow. Precise and reliable criteria were proposed for proper model selection. A modified two-phase critical flow model was presented and a single-phase critical flow model for superheat gas was established. Besides, modifications on flashing inception, flow resistance and particulate plugging were conducted in the LEABLE code. Moreover, the code was validated and verified by experimental data for both artificial and natural cracks. Reasonable agreements between experiment values and code predictions are shown for all the conditions. And in the present study the LEABLE shows a higher precision than existing models and commercial softwares (PICEP and SQUIRT). The parameter sensitivity study of LBB leakage was performed as well. The mass flux increases with crack opening crack opening displacement (COD), crack length and the entrance resistance coefficient, and decreases with global and local roughness, number and angle of turns. However, the entrance resistance coefficient and local roughness have limited effects on LBB leakage.
This paper summarizes and illustrates the overall design issues of the CSR1000, including overall technical requirements, main technical selections, main technical parameters and important technical issues discussion and etc.
CATHARE is the French best estimate code used for safety analysis of Light Water Reactors. Two-phase flow is modelled with a two-fluid six-equation model in 0-D, 1-D and 3-D modules. A few months ago, the range of applications was limited to the sub-critical two-phase region, but thanks to the extension of the water and steam properties up to 26 MPa and recent modifications of the standard code version, it is now possible to simulate transients where both supercritical and sub-critical regimes are encountered. This paper describes these recent developments in the modelling of supercritical water flows using the two-fluid model implemented into the CATHARE code and shows the results of the preliminary assessment of this model successfully achieved by running several calculations for 1-D applications: - Steady state in supercritical conditions followed by blow-down - Stationary isenthalpic expansion - Stationary heat-up of supercritical fluid Then, using the modified CATHARE version, first steady-state and transient calculations of a Supercritical Light Water Reactor operating above the critical point were carried out, among them, Loss of Coolant Accidents (LOCAs) considering simple sequences of events. The predicted transient behaviours seem to be correct from supercritical to sub-critical regimes, which confirms the CATHARE capabilities.
The reactor system response of the mixed neutron energy spectrum of supercritical water cooled reactor (SCWR-M) under two types of reactivity initiated accidents (inadvertent control rod withdrawal and rod ejection) was studied. Firstly, applying the modified system code RELAP5 which is available for supercritical conditions, a model of SCWR-M system was established. Then, in the process of the accident under operation conditions, some important parameters, such as power, flow rate and cladding temperature were calculated and analyzed. Finally, analysis of parameters, such as worth and withdrawal speed of the control rod and negative feedback coefficient of the core was carried out. The results show that under the design condition, SCWR-M can effectively remove the decay heat from the core to ensure the integrity of the fuel rods. In addition, reactivity parameters have little influence on the safety under the control rod withdrawal accident. But they have strong effect on the peak cladding temperature under the control rod ejection accident. The safety margin will be greatly reduced when too conservative parameters are taken into the safety analysis.
Super Light Water Reactors and Super Fast Reactors provides an overview of the design and analysis of nuclear power reactors. Readers will gain an understanding of the conceptual design elements and specific analysis methods for supercritical-pressure light water cooled reactors. Nuclear fuel, reactor core, plant control, plant stand-up and stability are among the topics discussed, in addition to safety system and safety analysis parameters. Providing the fundamentals of reactor design criteria and analysis, this volume is a useful reference for engineers, industry professionals, and graduate students involved with nuclear engineering and energy technology. © Springer Science+Business Media, LLC 2010. All rights reserved.
The SCWR fuel qualification test is an in-pile test of a 4 rod fuel assembly at supercritical pressure inside a research reactor, which is operated at atmospheric pressure. The risk of radioactive release from this new test facility should not exceed the accepted risk of the existing research reactor. A large number of safety analyses have been performed to assess this risk, which are summarized in this paper. Among them are studies of design basis accidents, assuming different failure modes of the high-pressure system, as well as an assessment of consequences of postulated accidents beyond the design basis. Results show that the safety objectives can be met.
The safety design and the analyses of pressure- and flow-induced accidents and transients of the direct-cycle supercritical-water-cooled fast reactor (SCFR) are discussed. The coolant system of the SCFR is the once-through type like a fossil-fired power plant. Maintaining the core flow is adopted as a fundamental safety requirement. The coolant flow rate is measured for the safety signal instead of the water level of a boiling water reactor. To guarantee the core flow, the plant is equipped with four high-pressure auxiliary feedwater systems, four low-pressure coolant injection systems, turbine bypass valves, and an automatic depressurization system.
Behaviors at pressure- and flow-induced events are analyzed by a computer code for assessing the safety of the reactor. Total loss of flow and pump seizure are considered as accidents, and the maximum cladding temperature criterion for stainless steel is satisfied. Flywheels are needed for the main feedwater pumps for prolonging the coastdown time more than 10 s. Six events are considered as transients. All results satisfy the minimum deterioration heat flux criterion. The loss of feedwater heating is not a severe transient, although there is no recirculation coolant. The loss of turbine load is not severe because the coolant flow is stagnated by closing turbine control valves, and the core power is reduced because of the decrease of the coolant density. The SCFR tolerates the pressure- and flow-induced events.
The supercritical water-cooled reactor (SCWR) is one of six reactor concepts under development in the Generation-IV International Forum (GIF). As a member of GIF, Canada is developing a pressure-tube type SCWR, which has the potential to fulfill all major GIF goals on enhanced safety, sustainability, economics, and proliferation resistance. The system thermalhydraulics code CATHENA will be used in the safety analyses for the Canadian SCWR. Based on the current conceptual design of the Canadian SCWR, a CATHENA idealization has been developed. This model includes all 336 fuel channels with a detailed model of heat transfer in the reactor core. Also modeled are the main pumps, inlet plenum, outlet plenum, turbines, and heavy water moderator. In this paper, the CATHENA idealization of the Canadian SCWR conceptual design is described. Simulation results for steady-state normal operations are also presented for the current Canadian SCWR conceptual design.
Safety analysis is one of the chief difficulties during the research and design of SCWRs. Currently, the development of the SCWR safety analysis code is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs, has been developed based on RETRAN-02, the best estimate code used for safety analysis of Light Water Reactors. The ability of SCTRAN to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with the APROS and RELAP5-3D codes, respectively. The results show that the SCTRAN code developed in this study is capable of performing safety analysis for SCWRs, and the results are reliable. SCTRAN developed in this study is of great value and engineering significance, and could be utilized in the Chinese SCWR research and development.
This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities.
The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST.
A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R&D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.
Anticipated-transient-without-scram (ATWS) of the supercritical-pressure light water cooled thermal reactor with downward-flow water rods (Super LWR) is analyzed to clarify its safety characteristics. At loss-of-flow, heat-up of the fuel cladding is mitigated by the water rods removing heat from the fuel channels by heat conduction and supplying their coolant inventory to the fuel channels by volume expansion. The average coolant density is not sensitive to the pressure due to the small density difference between “steam” and “water” at supercritical-pressure. Closure of the coolant outlet of the once-through system causes flow stagnation that suppresses an increase in the coolant density due to an increase in the temperature. Therefore, the increase in power is small for pressurization events. The coolant density and Doppler feedbacks provide good self-controllability of the power against loss-of-flow and reactivity insertion. An alternative action is not needed either to satisfy the safety criteria or to achieve a high-temperature stable condition for all ATWS events. Initiating the automatic depressurization system is a good alternative action that induces a strong core coolant flow and inserts a negative reactivity. It provides an additional safety margin for the ATWS events. Even the high core power rating of the Super LWR has excellent ATWS characteristics, providing a key reactor design advantage.
LOCA analysis of the supercritical-pressure light water cooled thermal reactor with downward-flow water rods (called Super LWR) are carried out to clarify its characteristics. 1–100% hot/cold leg breaks are analyzed. At the cold-leg large break, excessive core heat-up is mitigated by the automatic depressurization system (ADS) during blowdown because reactor depressurization induces core coolant flow. The coolant inventory in the top dome and the water rods is effectively used for core cooling. After blowdown, the core is slowly re-flooded by the low-pressure ECCS like PWR. The highest cladding temperature of the large LOCA is lower than the criterion (1,260°C) by about 430°C which appears during the reflooding phase. Small break of the cold-leg gives the higher cladding temperatures than that of the large break because the ADS are not actuated in the analysis. The highest cladding temperature is lower than the criterion by about 260°C. If the ADS actuation is assumed by the “drywell pressure high” signal, the cladding temperature is lower. The hot-leg break is less severe than the cold-leg break because it increases the core coolant flow rate and forced flooding is expected after blowdown.
Based on a revised version of RELAP5, which can be used for super-critical pressure calculation, a model of mixed spectrum SCWR (SCWR-M) system is established. To analyze the transient behavior of SCWR-M and develop mitigation measures during loss of flow accident (LOFA), some important parameters, e.g. reactor coolant pump (RCP) coast-down time, Reactor Pressure Vessel (RPV) upper water volume and safety injection flow, etc., are chosen for the parametric analysis. The results achieved so far indicate that the SCWR-M system design is feasible and promising. Three important mitigation measures for LOFA of SCWR-M are derived from the results: RCP coast-down time of more than 15s, RPV upper water volume of more than 27m3, and safety injection of more than 5% of the system design flow.
The supercritical-pressure water-cooled fast reactor (SWFR) is a fast spectrum supercritical water-cooled reactor (SCWR) studied by the University of Tokyo. The SWFR is designed as a two-pass core with an outlet temperature 500°C. The SWFR has fuel channels cooled by downward flow, higher power density, and smaller coolant density reactivity feedback compared with Super LWR. This paper describes the safety analyses of abnormal events for the SWFR. SPRAT-F code is used for the safety analysis at supercritical pressure considering the downward flow cooled seed fuel channel. This code is based on a 1-D node junction model with point kinetics and decay heat calculations. Flow redistribution among parallel paths is calculated by pressure-loss balance and momentum conservation. The initiating events are selected from those of LWRs. For the safety analysis, nine abnormal transients and four accidents are selected with considering types of abnormality. By the numerical analyses, it was found that the loss of flow events can be mitigated by the “water source” effect of the downward flow blanket channels in the abnormal transients and accidents. All the abnormal events satisfy the criteria with margin.
The dynamic core behavior of the HPLWR (High Performance Light Water Reactor) concept (Schulenberg, 2006), which is the European version of the GEN4 Supercritical Water Cooled Reactors (SCWR) was investigated in case of RIA (Reactivity Initiated Accidents) events. Before the analyses, the heat transfer correlations and property table of the supercritical water, built into the ATHLET code, were verified by using the Kirillov single channel thermal hydraulic measurements. According to the comparison of the calculations and the measurements, the Watt's correlation proved to be an appropriate one for average HPLWR conditions. The presented results of the performed analyses are reporting on significant perturbations of the power distribution. Furthermore, in comparison to PWRs, a more effective moderator feedback was obtained. The acceptance criteria are fulfilled, however, in spite of the relatively strong feedback, in some cases the hot channel temperatures are not far from the limits, which points out the importance of the RIA analyses. The observations can be explained by the sensitivity of the neutronic characteristics on the moderator density, changing in a wide range and, as a consequence, the significant perturbation of power peaking factors during the transients resulting in some cases larger local power increase than that originating from the average reactor power rise.
This paper describes safety analysis of the high-temperature supercritical water-cooled thermal reactor with down-ward-flow water rods (called Super LWR) at supercritical pressure. Eleven transients and four accidents are chosen for the safety analysis considering types of abnormalities. The cladding temperature is taken as the important transient criterion instead of the heat flux ratio. The once-through cooling system and the downward-flow water rod system characterize safety of the Super LWR. "Loss of feedwater" is important because it is the same as "loss of reactor coolant flow" unlike BWR and PWR. However, the downward-flow water rods mitigate core heat-up before startup of the auxiliary feedwater system because they remove heat from the fuel channels by heat conduction and supply their water inventory to the fuel channels by volume expansion. During pressurization transients, the reactor power does not increase significantly unlike BWR due to no void collapse in single-phase flow and decrease in coolant density by flow stagnation in the once-through cooling system. All the transients and the accidents satisfy the criteria. Increases in the hottest cladding temperatures are about 50°C at transients and 250°C at accidents at maximum. The period of the high cladding temperature is very short at transients.
TACOS(Transient Analysis Code Of SCWR) code has been developed to analyze the safety characteristics of SCWR-M for present work. In the code TACOS, the single channel model, point kinetic model, heat transfer model and control system model were used. Four different heat transfer correlations were applied to the steady state calculation of SCWR with mixed spectrum core (SCWR-M). After the comparison of four correlations, Bishop Correlation was selected for transient analysis.Using TACOS code, preliminary analysis of loss of flow accident (LOFA) for SCWR-M has been performed. Partial loss of flow and total loss of flow accident were discussed respectively. For the total loss of flow, the results indicate that the maximum cladding surface temperature (MCST) in the thermal zone was much higher than that in the fast zone. A “special flow mode” similar to natural circulation between coolant channels and moderator channels in the thermal zone has been found. The co-current flow mode design in the thermal zone would cause disadvantage to safety performance because it has a negative effect on the flow rate of coolant channels after the startup of Auxiliary Feedwater System (AFS). The sensitivity analyses of pump coastdown time and AFS flow rate for total loss of flow accident were also carried out.
Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once- through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.
This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis.
This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical , an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.
The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density.
The proposed concepts for the fourth generation of nuclear reactors include a reactor operating with water at thermodynamically supercritical state, the Supercritical Water Reactor (SCWR). For the design and safety demonstrations of such a reactor, the possibility to accurately simulate the thermal hydraulics of the supercritical coolant is an absolute prerequisite. For this purpose, the one-dimensional two-phase thermal hydraulics solution of APROS process simulation software was developed to function at the supercritical pressure region. Software modifications included the redefinition of some parameters that have physical significance only at the sub-critical pressures, improvement of the steam tables, and addition of heat transfer and friction correlations suitable for the supercritical pressure region.
The supercritical-pressure light water cooled reactors aim at cost reduction by system simplification and higher thermal efficiency, and have flexibility for the fuel cycle due to technical feasibility for various neutron spectrum reactors. Since loss-of-coolant accident (LOCA) behavior at supercritical pressure conditions cannot be analyzed with the existing codes for the current light water reactors, a LOCA analysis code for the supercritical-pressure light water cooled reactor is developed in this study. This code, which is named SCRELA, is composed of two parts: the blowdown and reflood analysis modules. The blowdown analysis module is designed based on the homogeneous equilibrium model. The reflood analysis module is modeled by the thermal equilibrium relative velocity model. SCRELA is validated by the REFLA-TRAC code, which is developed in the Japan Atomic Energy Research Institute based on TRAC-PF1. A large break LOCA of the thermal neutron spectrum reactor (SCLWR) is analyzed by SCRELA. The result shows that the peak clad temperature (PCT) is nearly 980°C about 60s after the break and the PCT position is quenched at 170s This means that PCT is sufficiently lower than the safety limit of 1260°C. In conclusion, the developed code shows the safety of SCLWR under the large break LOCA, and is expected to be applied to LOCA analysis of other types of the supercritical-pressure light water cooled reactors.
International Conference on Advances in Nuclear Power Plants, ICAPP 2008. Anaheim, CA, USA, 8 - 12 June 2008 Vol.1, 649 - 659 The High Performance Light Water Reactor (HPLWR) is a Light Water Reactor (LWR) operating at supercritical pressure (25MPa). It belongs to the six reactors currently being investigated under the framework of the Generation IV International Forum. In Europe, investigations on the HPLWR have been integrated into a joint research project, called High Performance Light Water Reactor Phase 2 (HPLWR Phase 2), which is co-funded by the European Commission. Within 42 months, ten partners from eight European countries working on critical scientific issues shall show the feasibility of the HPLWR concept. The final goal is to assess the future potential of this reactor in the electricity market. Within the first year of the project, mainly design tasks of the reactor core, the pressure vessel and its internals have been carried out and codes were adapted for supercritical fluid. Material oxidation experiments were performed and heat transfer was investigated numerically. (24 refs.)
Safety of Super LWR, (Ⅱ) safety analysis at supercritical pressure
- Ishiwatrai
Safety analysis of reactivity initiated accidents in SCWR-M (Chinese)
- L I Dong
- X U Zhi-Hong
- L Xiao-Jing
Dong, L.I., Zhi-Hong, X.U., Xiao-Jing, L., et al., 2013. Safety analysis of reactivity
initiated accidents in SCWR-M (Chinese). Atomic Energy Sci. Technol. 47 (11),
2033-2040.
Safety analysis of a supercritical-pressure watercooled fast reactor under supercritical pressure
- S Ikejiri
- Y Ishiwatrai
- Y Oka
Ikejiri, S., Ishiwatrai, Y., Oka, Y., 2010. Safety analysis of a supercritical-pressure watercooled fast reactor under supercritical pressure. Nucl. Eng. Des. 240 (5), 1218-1228.
High Performance Light Water Reactor Phase 2
- J Starflinger
- T Schulenberg
- P Marsault
Starflinger, J., Schulenberg, T., Marsault, P., et al., 2008. Progress within the European
project "High Performance Light Water Reactor Phase 2". In: Proceedings of the
ICAPP 2008 Conference, Anaheim, California, 8-12 June. Paper id. 8247.
Overview on overall design of CSR1000 (Chinese). Nucl. Power Eng
- L I Xiang
- L I Qing
- B Y Xia
Xiang, L.I., Qing, L.I., Xia, B.Y., et al., 2013. Overview on overall design of CSR1000
(Chinese). Nucl. Power Eng.