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Abstract
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.
... More detailed results can be found in recent publications [6][7][8][9][10]. ...
... The elevated cladding temperatures in the core (up to 620°C), the supercritical steam parameters (up to 25 MPa, more than 500°C) and a large variation of moderator density in the core (more than a factor of 7) exceeds the range of experience gathered with PWR or BWR design up to now. Therefore as mentioned in [8] and [9], some design tools have to be fitted to these conditions and technologies need to be developed before a detailed reactor design can be performed [10]. These are: ...
A European R&D Roadmap of the High Performance Light Water Reactor for the years 2000-2020 is presented. It comprises feasibility studies for a thermal-and fast-core design, the reactor pressure vessel internals, the containment and for the balance of plant. Additionally, the definition of an appropriate safety approach is analyzed. Basic technologies for Supercritical-Water-Cooled Reactor Systems like materials and code development, and the related heat-transfer investigations, critical flow experiments and material testing, are major R&D tasks. Finally, the objective of having a conceptual design of the HPLWR ready to be built by the year 2020 is related to a series of coordinated R&D tasks.
... More detailed results can be found in recent publications [6][7][8][9][10]. ...
... The elevated cladding temperatures in the core (up to 620°C), the supercritical steam parameters (up to 25 MPa, more than 500°C) and a large variation of moderator density in the core (more than a factor of 7) exceeds the range of experience gathered with PWR or BWR design up to now. Therefore as mentioned in [8] and [9], some design tools have to be fitted to these conditions and technologies need to be developed before a detailed reactor design can be performed [10]. These are: ...
On July 30, 2003, EURATOM signed the charter to join the Generation IV International Forum and thus to contribute to innovative reactor design and development. Among other concepts, supercritical water cooled reactor systems shall be foreseen as their contribution. In order to support this international forum, a dedicated budget for R and D of innovative concepts is planned for the 6. European Framework Programme of the European Commission. Currently, a detailed work plan for supercritical water cooled reactor systems is being worked out, in order to be presented and decided by the European Commission thereafter. It shall include: - Design studies of a thermal reactor core, its reactor pressure vessel internals and of the balance of plant. - Study of a fast reactor option for sustainable use of fuel and for Plutonium management. - Study of the corrosion behavior and other performances of candidate materials at supercritical pressures. - Detailed investigations of heat transfer and pressure drop at supercritical pressures and at part load operation conditions of the reactor. - Design code improvements and verifications. - Conceptual design and analyses of a suitable safety system. As a result, the program shall enable a thorough assessment of the supercritical water cooled reactor system with a view to determine its future potential. This summary report shall give an overview of the contributions, which are planned to be provided by the EURATOM partners in France, Finland, Germany, Hungary, Switzerland, The Netherlands and The Czech Republic. (authors)
... MacDonald et al. (2004) modeled and analyzed the passive safety systems of the American SCWR concept with RELAP5-3D. CATHARE2 is an advanced version, which was extended based on the ''Best-Estimate'' system code CATHARE originally applied for PWR simulation in Europe, and was verified to own the ability to perform safety analysis of HPLWR (Antoni and Dumaz, 2003;Andreani et al., 2011). Through improving steam tables at subcritical pressure and adding constitutive equations at supercritical pressure, Markku and Joona (2008) expanded the application range of the code APROS from subcritical pressure to supercritical pressure. ...
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.
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.
This chapter aims to clarify the supercritical fluids thermal hydraulics characteristics including heat transfer, pressure drops and flow instabilities for the purpose of accurate design and efficient safe operation of advanced energy systems. At first, the convection heat transfer models considering the effect of nonlinear properties and the effect of buoyancy and acceleration have been provided and discussed. Secondly, the hydraulic resistance models for supercritical fluids have been selected and suggested for different conditions. Thirdly, the published models for supercritical flow instabilities based on four different regional partitions are summarized and clarified. At last, two typical case studies have been provided to further intuitively elaborate the thermal hydraulics of supercritical fluids within the advanced energy systems.
The paper summarizes the status of safety system development for supercritical water cooled reactors and thermal-hydraulic codes required to analyze them. While active safety systems are well-understood today and expected to perform as required, the development of passive safety systems will still require further optimization. Depressurization transients have successfully been simulated with some codes by a pseudo-two-phase flow simulation of supercritical water. Open issues of thermal-hydraulic codes include modeling of deteriorated heat transfer in one-dimensional system codes and predictions of heat transfer during depressurization transients from supercritical to sub-critical conditions.
Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards–O’Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the safety system. All the results achieved indicate that the modified ATHLET code has good feasibility in simulation of supercritical water cooled systems.
The Supercritical Water-cooled Reactor (SCWR) is one of the six concepts of the Generation IV International Forum. In Europe, investigations have been integrated into a joint research project, called High Performance Light Water Reactor (HPLWR). Due to the higher heat up within the core and a higher outlet temperature, a significant increase in turbine power and thermal efficiency of the plant can be expected.Besides the higher pressure and higher steam temperature, the design concept of this type of reactor differs significantly from a conventional LWR by a different core concept. In order to achieve the high outlet temperature of over 500 °C, a core with a three-step heat up and intermediate mixing is proposed to keep local cladding temperatures within today's material limits. A design for the reactor pressure vessel (RPV) and the internals has been worked out to incorporate a core arrangement with three passes. All components have been dimensioned following the safety standards of the nuclear safety standards commission in Germany. Additionally, a fuel assembly cluster with head and foot piece has been developed to facilitate the complex flow path for the multi-pass concept. The design of the internals and of the RPV is verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Furthermore, the reactor design ensures that the total coolant flow path remains closed against leakage of colder moderator water even in case of large thermal expansions of the components. The design of the RPV and internals is now available for detailed analyses of the core and the 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.