Assessment of Applicability of Two-Fluid Model Code ACE-3D to Heat Transfer Test of Supercritical Water Flowing in an Annular Channel

A supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal hydraulics in rod bundles of the core. Experimental conditions of mockup tests, however, may be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique that can extrapolate experimental data to various design conditions of the reactor. Japan Atomic Energy Agency (JAEA) has improved the three-dimensional two-fluid model analysis code ACE-3D, which was originally developed for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water in the supercritical region. In the present study, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which were performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code is applicable to the prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of the SCWR core.     

Investigation of radiation-induced surface activation effect in austenitic stainless steel under ultraviolet and γ-ray irradiations

The radiation-induced surface activation (RISA) effect will be applied to the core design in supercritical light water reactor (SCWR) in order to achieve a high performance with excellent economy and safety. The purpose of the present study is to investigate the RISA effect in the candidate fuel cladding materials in SCWR such as PNC1520. The change of weldability due to RISA effect and the related microstructure analysis were performed in oxidized PNC1520 and 304 stainless steel with various oxidization periods. The phases contained in the surface oxide layer of the present specimen were identified as Fe₂CrO₄, γ-Fe₂O₃, and Fe₂O₃. The lifetime of 13.8 days for wettability improving factor was confirmed in the ultraviolet (UV) irradiation. Meanwhile, the long life of 13.8 days and short life of 0.8 days for wettability improving factors were identified in the γ-ray irradiation. Based on the fact that the band gap energies of Fe₂CrO₄, γ-Fe₂O₃, and Fe₂O₃ were, respectively, 2.1, 2.0, and 2.2 eV, and the photo energies of UV and γ-ray irradiation were 4.48 eV and 13.3 MeV, it is therefore clarified that the hydrophilization on the oxide layer is ascribed to the RISA effect.     

超临界水冷混合堆快谱区组件物理——热工分析

提出了超临界水冷混合堆快谱区多层燃料组件设计方案.应用MCNP程序为该组件建立计算模型,并进行了相应的物理计算;同时运用子通道分析程序STAFAS对多层燃料组件子通道进行了初步的稳态热工分析.计算结果表明:超临界水冷混合堆快谱区多层燃料组件燃料转换比超过1.0,并且获得负的冷却剂空泡反应性系数;燃料包壳表面最高温度约为595℃,低于设计准则规定的上限值,同时组件各子通道出口冷却剂温度均匀性较好.通过对燃料棒径敏感性分析可知,较大棒径组件燃料转换比较大,但也会导致热通道包壳表面温度峰值升高.     

超临界水冷堆与WWER1000型压水堆的安全特性比较分析

超临界水冷堆(SCWR)是在高于水的临界点(374℃,22.1 MPa)的温度和压力下运行的反应堆。它的设计为一次通过循环,其中没有再循环回路。这点是与现在运行的轻水堆的最大不同。在超临界水堆电站系统中,以控制棒、汽轮机控制阀与反应堆冷却剂泵为主要的控制方式。通过对比分析超临界水冷堆与田湾核电站WWER1000型压水堆主泵卡轴事故下的安全特性,得出超临界水堆给水流量的丧失会造成反应堆冷却剂流量的丧失,而WWER1000型压水堆给水流量的丧失并不会造成反应堆冷却剂流量的丧失;WW-ER1000型压水堆的安全系统有控制棒、蒸汽发生器的主蒸汽旁排阀、应急给水泵,这些安全配置与超临界水冷堆相似;相比WWER1000型压水堆,超临界水冷堆在压力较快达到稳定状态前提下,其最高包壳温度有个剧烈变化过程,但超临界水冷堆和WWER1000型压水堆在卡轴事故发生后,都能建立稳定的自然循环。     

SCWR-FQT回路的热工物理耦合分析

中欧核能合作研究项目超临界水堆燃料验证实验(SCWR-FQT)的主要研究内容为在超临界水环境下对一个小型燃料组件进行堆内性能分析和验证。本文应用修过后的系统程序ATHLET-SC对该实验回路进行建模,同时结合堆芯中子物理的计算结果,对由于压力管进口管破裂形成的失水事故进行热工水力和中子物理的耦合分析,并讨论了物理耦合中停堆棒的负反应性、冷却剂温度系数等参数对结果的影响。计算结果表明,进行了中子物理耦合的结果得到的最高包壳温度比未进行中子耦合的结果要低15℃,同时停堆棒引入的负反应性是该事故过程中影响燃料棒最高包壳温度的一个主要因素。     

Numerical Simulation of Fluid Flow and Heat Transfer of Supercritical Fluids in Fuel Bundles

A supercritical water-cooled reactor (SCWR) was proposed as a kind of generation IV reactor in order to improve the efficiency of nuclear reactors. Although investigations on the thermal-hydraulic behavior in SCWR have attracted much attention, there is still a lack of CFD study on the heat transfer of supercritical water in fuel channels. In order to understand the thermal-hydraulic behavior of supercritical fluids in nuclear reactors, the local fluid flow and heat transfer of supercritical water in a 37-element fuel bundle has been studied numerically in this work. Results show that secondary flow appears and the cladding surface temperature (CST) is very nonuniform in the fuel bundle. The maximum cladding surface temperature (MaxCST), which is an important design parameter for SCWR, can be predicted and analyzed using the CFD method. Due to a very large circumferential temperature gradient in cladding surfaces of the fuel bundle, the precise cladding temperature distributions using the CFD method is highly recommended.     

堆芯中子学程序系统SARCS-4.0的开发及初步验证

开发了堆芯中子学程序系统SARCS-4.0,该程序系统能处理由任意方形燃料组件组成的堆芯;能计算铀钚、钍铀燃料循环;能计算硼、钆、铒、铪、银、铟、铕、钐等各类可燃毒物和含硼、铪、银-铟-铬、铕、镝等各类控制棒;具备堆芯核设计的基本功能.使用SARCS-4.0系统对超临界水冷反应堆(SCWR)堆芯进行计算以验证程序系统的计算准确性,结果表明,SARCS-4.0系统具有较高计算精度,该系统从功能上、精度上均适用于新型反应堆堆芯选型研究.     

超临界水冷堆燃料性能验证实验回路的冷却剂丧失事故分析

超临界水冷堆燃料性能验证实验（SCWR-FQT）将对1个小型燃料组件在超临界水环境下进行堆内性能测试。本工作应用修改过的ATHLET程序对包含该燃料组件的超临界水冷实验回路进行建模,并对其冷却剂管道破口导致的失水事故进行分析计算。计算结果表明,现有安全系统设计基本能保证在这些事故情况下维持燃料棒实验段的有效冷却。结果显示,修改过的ATHLET程序对超临界水冷系统的模拟具有良好的适用性。     

Discussion on the stability of heated channels with different fluids at supercritical pressures

The paper extends previous work on the stability of heated channels with fluids at supercritical pressure as predicted by available models.A set of dimensionless numbers proposed to predict the threshold of instabilities is further discussed to highlight their capabilities and possible improvements. In particular, it is shown that the choice made for the reference value of the derivative of specific volume with respect to specific enthalpy is justified as an extension of classical formulations adopted for boiling channels. Moreover, the degree of universality to be expected by the use of these dimensionless numbers while using different fluids is clarified; in this aim four different fluids are considered: water, carbon dioxide, ammonia and refrigerant R23.In order to provide a clear perspective of the usefulness of the proposed dimensionless numbers for dealing with different fluids, linear stability maps generated by a previously developed in-house code, making use of balance equations in dimensionless form, are then compared with the results obtained by computations performed in dimensional terms. In this aim, both an in-house code and RELAP5 are used. The reference considered system is a long circular channel with uniform heating and no singular pressure drops, already addressed in previous analyses, here assumed both in vertical upward and in horizontal flow conditions. The comparison of the predictions obtained for the different fluids allows to ascertain the level of applicability of the dimensionless numbers and, as an interesting by-product, confirms the possibility to encounter static instabilities also in systems at supercritical pressure.     

Experimental study of critical flow of water at supercritical pressure

Experimental studies of the critical flow of water were conducted under steady-state conditions with a nozzle 1.41 mm in diameter and 4.35 mm in length, covering the inlet pressure range of 22.1–26.8 MPa and inlet temperature range of 38–474°C. The parametric trend of the flow rate was investigated, and the experimental data were compared with the predictions of the homogeneous equilibrium model, the Bernoulli correlation, and the models used in the reactor safety analysis code RELAP5/MOD3.3. It is concluded that in the near or beyond pseudo-critical region, thermal-dynamic equilibrium is dominant, and at a lower temperature, choking does not occur. The onset of the choking condition is not predicted reasonably by the RELAP5 code.