Supercritical-pressure Water-cooled Reactor 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.

Neutronic Analysis of Heavy Water Application in HPLWR Emergency Core Cooling System

Since 2002 a generation IV nuclear energy system development project has been initiated to enhance the future role of nuclear energy systems. Challenging technology goals for generation IV nuclear reactor systems were defined in 4 areas: sustainability, economics, safety and reliability and proliferation resistance. Light water reactors at supercritical pressure, being currently under design, are the new generation of nuclear reactors. The high-performance light water reactor (HPLWR) is the European version of the supercritical-pressure water cooled reactor (SCWR). The aim of this study is to reliability analysis of heavy water emergency core cooling system (ECCS) (instead of light water ECCS) in HPLWRs during LOCA (loss of coolant accident) through neutronics calculations and applied feedback.. Reliability analysis (neutronic calculations) are performed using WIMS and CITATION codes. Changes in neutronic parameters, such as neutron cross-sections, effective multiplication factor and power peaking factor during LOCA, are discussed by injecting different concentrations of the heavy water coolant by ECCS.

NEUTRONIC INVESTIGATION OF SEMI-HEAVY WATER APPLICATION IN HPLWR NEW FLOW PATTERN

The high-performance light-water reactor (HPLWR) is the European version of the supercritical-pressure water-cooled reactor. Light water reactors at supercritical pressure, being currently under design, are the new generation of nuclear reactors. The variable moderation high performance light-water reactor (VMHWR) is the new version of HPLWR conceptual design. In this design, the fluid passing through the moderator channels and the assembly gaps (known as a moderator) flows in channels detached from coolants. This paper presents a neutronic analysis of semi-heavy water in VMHWR. Neutronic calculations were made through WIMSD5 and CITATION codes. Neutronic parameters such as effective multiplication factor, power peaking factor, and the effect of semi-heavy water on these parameters were investigated.

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Thermal–hydraulic analysis of HPLWR fuel assembly cluster

The High-Performance Light Water Reactor (HPLWR) is the European version of the Supercritical-pressure Water Cooled Reactor (SCWR). Light water reactors at supercritical pressure, being currently under design, are the new generation of nuclear reactors. The present article is a thermal–hydraulic analysis of the HPLWR fuel assembly cluster. The existing thermal hydraulic codes are not capable of analyzing and designing the HPLWR fuel assembly. It is necessary to introduce a new thermal–hydraulic code that can incorporate the parameters and specifications of the HPLWR fuel assembly. The governing equations include energy, mass and momentum equations beside thermal–hydraulic equations are solved numerically using the finite volume (upwind) method. The recent thermal–hydraulic code will be coupled with netronic code to yield a complete analysis of the HPLWR fuel assembly cluster.

Heat transfer to a supercritical pressure fluid flowing in sub-bundle channels

Supercritical pressure water cooled reactor (SCWR) has been regarded as an innovative nuclear reactor. For the design and development of the SCWR, heat transfer performance under supercritical pressure is one of the most important indicators. In this paper, experimental data are presented on the heat transfer to a supercritical pressure fluid flowing vertically upward and downward in a small diameter heated tube and two sub-bundle channels with three heater rods and seven heater rods, using HCFC22 as the test fluid. Downstream of grid spacer for the sub-bundles, heat transfer enhancement was observed in the upward flow, but not in the downward flow. The enhancement was remarkable especially when the heat transfer deterioration occurs in the fully developed region unaffected by the spacer. The heat transfer correlation for the downstream region of the spacer was developed in the normal heat transfer of sub-bundles. In the fully developed region for the sub-bundle, occurrence of the heat transfer deterioration was suppressed or degree of the deterioration was moderated. At high mass velocity for downward flow in the seven rod sub-bundle, oscillation of heat transfer was observed in the region of the enthalpy over the pseudocritical point.

D124 臨界圧近傍における限界熱流束の特性(OS-15: 沸騰・凝縮伝熱および混相流の最近の進展(2))

Supercritical pressure water cooled reactor (SCWR), which is operated at supercritical pressure, is one of the next generation reactors for the purpose of improving economic efficiency and safety. In the SCWR, water pressure passes the critical pressure during startup, shutdown or in case of loss of coolant accident (LOCA). In the near-critical pressure region which the pressure slightly below the critical pressure, critical heat flux (CHF) condition tends to occur at relatively low heat flux and then there is a risk of serious damage of fuel rod due to abrupt rise of surface temperature. In this study, experiments on CHF at near-critical pressure in vertical upward flow inside a tube were conducted to clarify characteristics of CHF at near-critical pressure.

The thermal circuit of a nuclear power station’s unit built around a supercritical-pressure water-cooled reactor

Main results obtained from calculations of the steam generator and thermal circuit of the steam turbine unit for a nuclear power unit with supercritical-pressure water coolant and integral layout are presented. The obtained characteristics point to the advisability of carrying out further developments of this promising nuclear power technology.

Coupled neutronics and thermohydraulics calculations with burn-up for HPLWRs

The HPLWR (High-Performance Light Water Reactor) is the European version of the SCWR (Supercritical-pressure Water Cooled Reactor) concept, which is one of the Generation IV reactors. In this reactor the primary water enters the core of the HPLWR under supercritical-pressure condition (25 MPa) at a temperature of 280 °C and leaves it at a temperature of up to 510 °C. Due to the significant changes in the physical properties of water at supercritical-pressure the system is susceptible to local temperature, density and power oscillations. This inclination is increased by the pseudocritical transformation of the supercritical-pressure water used as primary coolant. At the Budapest University of Technology a coupled neutronics–thermohydraulics program system has been developed which is capable of determining the steady-state equilibrium parameters and calculating the related power, temperature, etc. distributions for the above-mentioned reactor. The program system can be used to optimize the axial enrichment profile of the fuel rods (e.g. in order to obtain a uniform power distribution). Since the power distribution will change with burn-up, which causes a change in the temperature and density distributions, a coupled neutronics – thermohydraulics – burn-up calculation is required. Therefore, the program system has been extended with the ORIGEN-S burn-up calculator. In the paper the developed program system and its features are presented. Parametric studies for different operating conditions were carried out and the obtained results are discussed.

RESEARCH ACTIVITIES ON A SUPERCRITICAL PRESSURE WATER REACTOR IN KOREA

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.