Natural Circulation Phenomena and Modeling for Advanced Water Cooled Reactors

As part of the IAEA’s overall effort to foster international collaborations that strive to improve the economics and safety of future water cooled nuclear power plants, an IAEA Coordinated Research Project (CRP) was started in early 2004. This CRP, entitled Natural Circulation Phenomena, Modelling and Reliability of Passive Safety Systems that Utilize Natural Circulation, focuses on the use of passive safety systems to help meet the safety and economic goals of a new generation of nuclear power plants. This CRP has been organized within the framework of the IAEA Department of Nuclear Energy’s Technical Working Groups for Advanced Technologies for Light Water Reactors and Heavy Water Reactors (the TWG-LWR and the TWG-HWR) and has provided an international cooperation on research work underway at the national level in several IAEA Member States. The use of passive safety systems was addressed in 1991 at the IAEA Conference on “The Safety of Nuclear Power: Strategy for the Future” [1-1]. Subsequently, experts in research institutes and nuclear plant design organizations from several IAEA Member States collaboratively presented their common views in a paper entitled “Balancing passive and active systems for evolutionary water cooled reactors” in Ref. [1-2]. The experts noted that a designer’s first consideration is to satisfy the required safety function with sufficient reliability, and the designer must also consider other aspects such as the impact on plant operation, design simplicity and costs. The use of passive safety systems such as accumulators, condensation and evaporative heat exchangers, and gravity driven safety injection systems eliminates the costs associated with the installation, maintenance and operation of active safety systems that require multiple pumps with independent and redundant electric power supplies. As a result, passive safety systems are being considered for numerous reactor concepts (including in Generation III and III+ concepts) and are expected to find applications in the Generation-IV reactor concepts, as identified by the Generation IV International Forum (GIF). Another motivation for the use of passive safety systems is the potential for enhanced safety through increased safety system reliability. The CRP benefits from earlier IAEA activities that include developing databases on physical processes of significant importance to water cooled reactor operations and safety [1-3, 1-4], technical information exchange meetings on recent technology advances [1-5–1-11], and status reports on advanced water cooled reactors [1-12, 1-13]. In the area of thermohydraulic phenomena in advanced water cooled reactors, recent IAEA activities have assimilated data internationally on heat transfer coefficients and pressure drop [1-3]; and have shared information on natural circulation data and analytical methods [1-5], and on experimental tests and qualification of analytical methods [1-6]. This CRP also benefits from a recent report issued by IAEA [1-14] on the status of innovative small and medium sized reactor designs. In order to establish the progress of work in this CRP, an Integrated Research Plan with description of the tasks addressing the objectives of the CRP was defined. These tasks are:  Establish the state of the art on natural circulation  Identify and describe reference systems  Identify and characterize phenomena that influence natural circulation  Examine application of data and codes to design and safety  Examine the reliability of passive systems that utilize natural circulation The results of the CRP work have been planned to be published in three consecutive IAEA-TECDOC reports. The activity under the first task is aimed at summarizing the current understanding of natural circulation system phenomena and the methods used experimentally to investigate and model such phenomena. In November 2005, the IAEA issued a technical document as first publication [1-15] in three TECDOC report series of this CRP on natural circulation, developed by the collaborative effort of the CRP participants and with major contributions from some selected experts in the CRP, aimed at documenting the present knowledge in six specific areas; advantages and challenges of natural circulation systems in advanced designs, local transport phenomena and models, integral system phenomena and models, natural circulation experiments, advanced computation methods, and reliability assessment methodology. The activity for the third task is aimed at identifying and categorizing the natural circulation phenomena of importance to advanced reactors and passive safety system operations and reliability. This task is the major link between the second and the fourth tasks. The activities related to the second task and the fourth task including the fifth task are agreed to be published in two different TECDOCs by the CRP participants. Since the third task is the backbone for both tasks, inclusion of this task in both TECDOCs in an appropriate form is a logical consequence. The aim of the second publication in this series of TECDOCs [1-16] was to describe passive safety systems in a wide range of advanced water cooled nuclear power plant designs with the goal of gaining insights into the system design, operation, and reliability without endorsement of the performance. This second publication has a unique feature which includes plant design descriptions with a strong emphasis on passive safety systems of the specific designs. These descriptions of the passive safety systems together with the phenomena identification (including the definitions of the phenomenon to describe in some detail the titles of the phenomenon considered) are also given in the Annexes and Appendix of the report [1-16], respectively. Based on the passive systems and phenomena, which are considered, a cross-reference matrix has been established and also presented in this second report. As basis for the phenomenon identification, earlier works performed within the OECD/NEA framework during 1983–1997 were considered. These are:  Code validation matrix of thermohydraulic codes for LWR loss of coolant accident (LOCA) and transients [1-17]  State of the art report (SOAR) on thermohydraulic of emergency core cooling in light water reactors [1-18]  Separate Effects Test (SET) Validation Matrix for Light Water Reactors [1-19]  Integral Facility Tests Validation Matrix for Light Water Reactors [1-20]  Status report on relevant thermohydraulic aspects of advanced reactor designs [1-21] Since the Generation III and III+ reactor designs contain technological features that are common to the current generation reactors, the phenomena identified during the work performed for first item to fourth item can be used as base knowledge. The fifth item provides the important and relevant thermohydraulic phenomena for advanced reactor designs in addition to the relevant thermohydraulic phenomena identified for the current generation of light water reactors (LWR). The list of relevant phenomena established in Ref. 1-21 has been taken as basis for the CRP work and has been modified according to the reactor types and passive safety systems considered in TECDOC-1624 [1-16]. It is to be noted that in identifying the relevant thermohydraulic phenomena in the list which is provided in this report, expert judgement is the main contributor. The two TECDOCs (the present TECDOC and TECDOC-1624) have been produced as consequent volumes independent of each other in their use and contents. The present TECDOC is the third publication in this series of TECDOC reports and its contents are described below. The identification and definition of the phenomena for advanced water cooled reactors with emphasis to passive safety systems is also presented in the second chapter of this TECDOC. Phenomena have been classified into two categories. These are phenomena occurring during interaction between primary system and containment; and phenomena caused by the presence of new components and systems or special reactor configurations. These descriptions are supplement to the definitions of phenomena which are developed for the current operational water cooled nuclear power plants [1-19, 1-20]. Twelve phenomena have been identified as key outcome of the process in Chapter 2 and the characterization of these additional phenomena for advanced water cooled reactors are provided in Chapter 3. The characterization of these additional phenomena includes exhaustive description of phenomena considering a comprehensive picture of the transient performance of the class of innovative reactors which are described in Ref. 1-16. Each of these descriptions includes outlines of models and an overview of the experimental data base that support the characterization of phenomena. The capabilities of thermohydraulics transient system codes or computational fluid dynamics (CFD) codes in predicting the same phenomena are evaluated and the results from the application of the codes to the analysis of experimental data are provided, as applicable to the case. After providing an overview on the phenomena and their characterization, Chapter 4 includes example cases for integral test facilities which simulate the prototypical plant design, some analysis of the experimental data obtained, use of these experimental data to help assess the predictive capabilities of the computer codes to model the phenomena that are occurring in the experimental test facilities, and application to the nuclear power plant analysis. In this chapter, a wide variety of prototypical plant designs, integral test facilities and thermohydraulics transient system codes, which are used in the analysis of plants and integral test facilities, are included. The collected and summarized information makes this chapter an excellent resource for providing understanding on the general capabilities of the integral test facilities and computer codes used for the examination of the natural circulation phenomena. As it has already been mentioned, the passive safety systems in their designs rely on natural forces to perform their accident prevention and mitigation functions once actuated and started. These driving forces are not generated by external power sources (e.g., pumped systems), as is the case in operating and some evolutionary reactor designs. Because the magnitude of the natural forces, which drive the operation of passive systems, is relatively small, counterforces (e.g. friction) can be of comparable magnitude and cannot be ignored as it is generally the case of systems including pumps. Moreover, there are considerable uncertainties associated with factors on which the magnitude of these forces and counter forces depends (e.g. values of heat transfer coefficients and pressure losses). In addition, the magnitude of such natural driving forces depends on specific plant conditions and configurations which could exist at the time a system is called upon to perform its safety function. All these uncertainties affect the thermohydraulic performance of the passive system. This particular aspect, inherent to the passive systems, has been dealt with in Chapter 5 and the methodology, which was developed within the framework of a project called Reliability Methods for Passive Safety Functions (RMPS) and performed under the auspices of the European Commission’s 5th Framework Programme, has been presented in some detail in that chapter. To assess the impact of uncertainties on the predicted performance of the passive system, a large number of calculations with best estimate thermohydraulic codes are needed. If all the sequences where the passive system studied is involved are considered, the number of calculations can be prohibitive. For all these reasons, it is necessary to create a specific methodology to assess the reliability of passive systems. The methodology addresses the following problems:  Identification and quantification of the sources of uncertainties and determination of the important variables;  Propagation of the uncertainties through thermohydraulic models and assessment of thermohydraulic passive system unreliability;  Introduction of passive system unreliability in the accident sequence analysis. In Section 1 of Chapter 5 (Section 5.1), each step of the methodology is described and commented and a diagram of the methodology is presented. Some improvements of this methodology, proposed after the end of the RMPS project (in early 2004) are highlighted in Section 5.2. Alternative methodologies, which have been developed by other institutions in parallel, are presented in Section 5.3. These methodologies are the ENEA methodology (developed by ENEA, Bologna, Italy) and Assessment of Passive System Reliability (ASPRA) (developed by BARC, India). Finally the application of the RMPS methodology, as an example, to a Passive Residual Heat Removal System (PRHRS) of a “CAREM-like” reactor has been performed within the present IAEA CRP and this application is described in Section 5.4. The major aim of this work is to quantify the failure probability of the passive safety function associated with the system under analysis. The last chapter of this TECDOC (Chapter 6) is devoted to summarize the detailed conclusions and recommendations, which were already provided in the contents of the various chapters. These conclusions and recommendations are given in a concise form for use of further needs in research, development, and technology areas.