An overview on thermal-hydraulic phenomena for water cooled nuclear reactors; part II: ALWRs and SCWRs

Validation of best-estimate thermal-hydraulic system codes is a necessary step to prove their applicability to calculate nuclear plant accident scenarios, including the course of events. It shall demonstrate that those physical thermal-hydraulic phenomena, which are important for a nuclear plant transient and accident scenarios, are calculated and analysed appropriately for the design and safety analysis purposes. Within the framework of international organizations, e. g., the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of The Organization for Economic Cooperation and Development (OECD/NEA-CSNI) and the International Atomic Energy Agency (IAEA), with the major support of institutions from different countries, work has been carried out for establishing validation matrices on Separate Effects Test (SET) facilities and Integral Test facilities of PWRs and BWRs. During this process, one of the major elements of the activity was to identify and characterize the thermal-hydraulic phenomena and, also test types. This process was based on expert knowledge of the scientists involved and, approved by the participating organizations and institutions in different countries. Additionally, identification and characterization of thermal-hydraulic phenomena were performed for VVERs, ALWRs (including SMRs) and, also SCWRs with emphasize to the design specifics of these nuclear plants. A common list of 67 thermal-hydraulic phenomena including basic phenomena is provided for Separate Effects Tests (SETs), and Integral Test facilities (ITEs) of PWRs, BWRs, and VVERs have about 50 thermal-hydraulic phenomena. Additionally, Advanced Water-Cooled Reactors having 15 identified phenomena, as well as SCWRs (specifically HPLWR) related 7 more phenomena identified. For all these internationally agreed thermal-hydraulic phenomena either some detailed descriptions for selected cases are given or needed references are provided in the paper. These identified phenomena also help to define research needs in the related field for the major number of water-cooled reactors. The knowledge of the thermal-hydraulic phenomena is the backbone and essential part for the development of nuclear thermal-hydraulics subject area. Thus, the subject will be presented in two parts: Part I of the paper will be devoted to thermal-hydraulic phenomena related to SETs, PWRs, BWRs, and VVER-400 plus VVER-1000; Part II will be dealing with additional thermal-hydraulic phenomena for ALWRs including VVER-1200, and SCWRs which are covered under GEN-III + and GEN-IV designs.