Advanced Water-cooled Reactors Research Articles

Comparison of sensitivity measures in probabilistic fracture mechanics

Probabilistic fracture mechanics (PFM) simulates the behavior of cracked structures and propagates uncertainties from input parameters to a failure probability or its uncertain estimate. In nuclear technology, this approach supports the assessment of the rupture probability of highly reliable pipes, which is an important parameter for the safety analysis of a nuclear power plant. For the appropriate probabilistic modelling of a structure with consideration of uncertainties, but also for the analysis of PFM application cases, the question arises, which input parameter of a probabilistic model has a higher impact on the estimate of computed failure probability, and which has a minor impact. This question is associated with the sensitivity measures or importance factors of the input parameters and their ranking concerning their influence.In this paper, six different approaches for the quantification of the sensitivity of parameters PFM evaluations are investigated: the amplification ratio, the direction cosine, the degree of separation, the analysis of the most probable failure point, the separation of uncertainty method, and the simple sample-based sensitivity study. Each method is described, visualized, applied to a common test case, and compared. The application case and the comparison are part of the Coordinated Research Project (CRP), “Methodology for Assessing Pipe Failure Rates in Advanced Water-Cooled Reactors (AWCRs)” by the International Atomic Energy Agency (IAEA), which is dedicated to the development of failure rates of piping in AWCRs. The participants used different PFM computer codes to analyze the test case and individual sensitivity methods to rank the input parameters, which motivated the comprehensive survey.The predicted parameter ranking of the approaches is consistent between the methods and between different PFM codes, but the approaches differ in the scope and the required effort. A conclusion is drawn and recommendations for the six different approaches are given.

An overview on thermal-hydraulic phenomena for water cooled nuclear reactors; part I: SETs, and ITFs of PWRs, BWRs, VVERs

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 (ITFs) 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 6 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 SMRs, and SCWRs which are covered under Gen-III+ and GEN-IV designs.

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 (ITFs) 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.

Passive Safety Systems in Advanced PWRs

Passive safety systems have been widely applied to advanced water-cooled reactors (WCRs), to enhance the safety of nuclear power plants. For the near term and medium term, the Chinese government decided for advanced pressurized water reactors with an extensive usage of passive safety systems. The International Workshop on Passive Safety Systems of Advanced PWR (IPASS’08) was held in Shanghai on April 28–30, 2008, coorganized by the Shanghai Jiao Tong University and the Pisa University. The main purposes of the Workshop IPASS’08 are (a) to provide a platform to the international nuclear community for exchanging research results and design experience on passive safety systems applied to advancedWCR and (b) to enhance the contact and collaboration between the Chinese research institutions and international partners. More than 100 scientists and nuclear engineers from eight countries, that is, Canada, China, France, Germany, Italy, Korea, Pakistan, and USA, were participating in this workshop. More than 60 technical papers were presented covering the following topics:

Role of Passive Safety Systems in Chinese Nuclear Power Development

Passive safety systems have been widely applied to advanced water‐cooled reactors, to enhance the safety of nuclear power plants. The ambitious program of the nuclear power development in China requires reactor concepts with high safety level. For the near‐term and medium‐term, the Chinese government decided for advanced pressurized water reactors with an extensive usage of passive safety systems. This paper describes some important criteria and the development program of the Chinese large‐scale pressurized water reactors. An overview on representative research activities and results achieved so far on passive safety systems in various institutions is presented.

Studies on advanced water-cooled reactors beyond generation III for power generation

China’s ambitious nuclear power program motivates the country’s nuclear community to develop advanced reactor concepts beyond generation III to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics, sustainability and technology availability. It is a logical extension of the generation III PWR technology in China. The status of international R&D work is reviewed. A new supercritical water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydraulics method is carried out. It shows good feasibility for the new design proposal.

Conceptual designing of reduced-moderation water reactor with heavy water coolant

The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06–1.11.

B106 低減速スペクトル炉の概念設計

Based on the experienced light water reactor technology, conceptual design studies on advanced water-cooled reactors have been performed. They are named "Reduced-Moderation Water Reactor" (RMWR) with the high conversion ratio more than 1.0 and the negative void reactivity coefficients. Several concepts have been successfully established for them based on the neutronics calculations. Also the thermal hydraulic investigation has been performed for each concept, including the basic critical heat flux experiments at 15.5 MPa in the triangular tight-lattice core configuration. Through this thermal hydraulic investigation, the feasibility of each design is being confirmed.

Assessment of a tube-based bundle CHF prediction method using a subchannel code

At the conceptual design stage for advanced water-cooled reactors (AWCRs), a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. A basic idea for general CHF prediction is to utilize the tube-based CHF models covering wide applicable ranges with the help of supplementary terms for bundle effects. In this study, feasibility assessments have been performed with the bundle CHF data relevant to pressurized water reactors. A subchannel analysis is adopted in order to be able to consider the geometrical variations properly even in untested fuel bundle geometries. As a result, a CHF look-up table method (i.e., the use of a round tube CHF table with appropriate bundle effect factors) turns out to be a promising way to fulfill the needs in many aspects among some selected correlations and theoretical models. Though improvements of the supplementary factors, especially for the cold wall and the bundle heated length effects, are desirable to make better predictions, the tube-based bundle CHF prediction method clearly shows a potential as a general CHF predictor.