Safety and licensing of MHTGR

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant's exclusion area boundary (EAB). The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error.     

论解决核电安全问题的途径

从现有水冷反应堆核电厂存在堆芯熔化危险这一安全问题的焦点出发,分析了改进型反应堆AP-600、SIR、非能动安全反应堆PIUS和具有固有安全的模块高温气冷堆MHTGR等的安全特性.按照下一代水冷反应堆的设计要求和用户要求,提出了解决水堆核电厂安全问题的新概念——自安全铀氢锆反应堆,该堆型可能成为世界水堆核电发展的一个方问。中国核动力研究设计院正在探讨这种堆型。     

Thermal-hydraulic analysis on the effect of operator action in the loss of shutdown cooling system accident in a PWR

It is necessary to develop PSA methodology and integrated accident management technology during low power/shutdown operations. To develop this technology, thermal-hydraulic analysis is necessarily required to access the trend of plant process parameters and operator's grace time after initiation of the accident. In this study, the thermal-hydraulic behavior in the loss of shutdown cooling system accident during low power/shutdown operations at the Korean standard nuclear power plant was analyzed using the best-estimate thermal-hydraulic analysis code, MARS2.1. The effects of operator's action and initiation of accident mitigation system, such as safety injection and gravity feed on mitigation of the accident during shutdown operations are also analyzed.When steam generators are unavailable or vent paths with large cross-sectional area are open in the accident, the core damage occurs earlier than the cases of steam generators available or vent paths with small cross-sectional area. If an operator takes an action to mitigate the accident, the accident can be mitigated considerably. To mitigate the accident, high-pressure safety injection is more effective in POS4B and gravity feed is more effective in POS5. The results of this study can contribute to the plant safety improvement because those can provide the time for an operator to take an action to mitigate the accident by providing quantitative time of core damage. The results of this study can also provide information in developing operating procedure and accident management technology.     

A nodal dynamic model for control system design and simulation of an MHTGR core

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high-temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high-temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power-distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.     

压水堆核电站热工水力系统程序的研发现状与趋势

比较分析了目前世界上典型的压水堆核电站热工水力系统程序的研发历程、发展现状、应用范围,着重指出了最佳估算、程序耦合、程序评估在热工水力系统程序研发中的重要作用,阐述了各国热工水力系统程序研发模式对我国自主创新的借鉴意义。     

Integrated software safety analysis method for digital I&C systems

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide more powerful overall operation capability, and user friendly man-machine interface. The operator can obtain more information through digital I&C system. However, while I&C system being digitalized, three issues are encountered: (1) software common-cause failure, (2) the interaction failure between operator and digital instrumentation and control system interface, and (3) the non-detectability of software failure. These failures might defeat defense echelons, and make the Diversity and Defense-in-Depth (D3) analysis be more difficult. This work developed an integrated methodology to evaluate nuclear power plant safety effect by interactions between operator and digital I&C system, and then propose improvement recommendations. This integrated methodology includes component-level software fault tree, system-level sequence-tree method and nuclear power plant computer simulation analysis. Software fault tree can clarify the software failure structure in digital I&C systems. Sequence-tree method can identify the interaction process and relationship among operator and I&C systems in each D3 echelon in a design basis event. Nuclear power plant computer simulation analysis method can further analyze the available backup facilities and allowable manual action duration for the operator when the digital I&C fail to function. Applying this methodology to evaluate the performance of digital nuclear power plant D3 design, could promote the nuclear power plant operation safety. The operator can then trust the nuclear power plant than before, when operating the highly automatic digital I&C facilities.     

Design aspects of the Chinese modular high-temperature gas-cooled reactor HTR-PM

The modular high-temperature gas-cooled reactor (MHTGR) has distinct advantages in terms of inherent safety, economics potential, high efficiency, potential usage for hydrogen production, etc. The Chinese design of the MHTGR, named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM), based on the technology and experience of the HTR-10, is currently in the conceptual phase. The HTR-PM demonstration plant is planned to be finished by 2012. The main philosophy of the HTR-PM project can be pinned down as: (1) safety, (2) standardization, (3) economy, and (4) proven technology. The work in the categories of marketing, organization, project and technology is done in predefined order. The biggest challenge for the HTR-PM is to ensure its economical viability while maintaining its inherent safety. A design of a 450 MWth annular pebble bed core connected with steam turbine is aimed for and presented in this paper.     

A cost-effective methodology to internalize nuclear safety in nuclear reactor conceptual design

A new methodology to perform nuclear reactor design, balancing safety and economics at the conceptual engineering stage, is presented in this work. The goal of this integral methodology is to take into account safety aspects in an optimization design process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behavior during accidents (safety performance indicators), are synthesized on Design Maps. These maps allow one to compare the safety indicator with limits, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimization process, by means of additional rules to the neutronic, thermal-hydraulic, and mechanical calculations.An application of the methodology, implemented in Integrated Reactor Evaluation Program 3 (IREP3) code, to optimize safety systems of CAREM prototype is presented. It consists in balancing the designs of the Emergency Injection System (EIS), the Residual Heat Removal System (RHRS), the primary circuit water inventory and the containment height, to cope with loss of coolant and loss of heat sink (LOHS) accidental sequences, taking into account cost and reactor performance.This methodology turns out to be promising to internalize cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels.     

“玲龙一号”反应堆研发关键技术

模块式小型反应堆（SMR）是一种新型的核能系统。“玲龙一号”反应堆（ACP100）是我国完全自主创新的多用途模块化小型压水反应堆。本文介绍了ACP100的研发过程、堆芯设计和安全设计的主要特点,主要包括堆芯核设计、热工水力设计、安全设计理念、固有安全设计、事故应对策略等关键技术。ACP100反应堆通过基于全非能动的设计理念以及确定论与概率安全评价相结合的设计方法,极大地提高了安全性,超过了三代核电安全标准要求。      

Outlook for MHTGR project development

The MHTGR design and technical basis is sufficiently advanced in the U.S. to warrant the initiation of Lead Project development activities. Two efforts are currently underway. A variant of the civilian MHTGR design is being advanced as a New Production Reactor to supply materials for the Nation's nuclear defense requirements. In parallel, a private-sector initiative is underway through a Lead Project Feasibility Study to evaluate whether and how to deploy the first civilian MHTGR. This paper reviews the status of these efforts.     

Investigation of large-scale gas mixing and stratification phenomena related to LWR containment studies in the PANDA facility

The thermal-hydraulic processes governing the containment response to postulated accidents and the mixing and distribution of hydrogen following a severe accident are relevant issues identified by several international expert groups as ‘research needs’ for current and advanced LWRs. The development and validation of modern computational codes that will accurately predict gas distribution in LWR containments is required for analyses related to safety, design and operational issues in current and advanced reactors. This objective requires the availability of separate-effect test data collected in facilities where the 3D distribution of the relevant variables is measured with sufficient resolution and accuracy and tests are performed under well-controlled initial and boundary conditions. Within the scope of the OECD project ‘SETH’, a series of 25 tests has been initiated in the large-scale thermal-hydraulic facility PANDA, in order to investigate mixing and stratification phenomena in a large multi-compartment gas volume approaching the dimensions of actual containment compartments. This experimental programme investigates jets and plumes and the resulting propagation of stratification fronts. The presentation of the first results from this test series demonstrates the value of these new data for code validation purposes.     

模块式高温气冷堆超临界循环一次再热方案研究

基于模块式高温气冷堆先进技术和超临界蒸汽动力循环先进技术,研究了高温气冷堆模块与超临界蒸汽动力循环耦合配置方案。结合超临界热力循环理论及模块化高温气冷堆的特性,研究了超临界热力循环方案及相应的循环参数。针对标准一次再热循环,研究了反应堆模块与汽轮机组匹配模式;计算了循环可能达到的效率,并与先进压水堆效率进行了比较。结果表明：模块化高温气冷堆超临界循环效率比压水堆电厂约高30%。本研究结果可作为高温气冷堆超临界循环电站概念设计的理论基础,为进一步的技术研究与方案设计提供依据。     

TRETA and TIZONA fast running thermal-hydraulic codes

TRETA and TIZONA codes have been developed to help analysts with the understanding of probabilistic safety analysis (PSA) event trees involving complex transients taking place in nuclear power plants. The two-phase thermal-hydraulic sections of the TIZONA code convey an original model that hinges on the assumption that one of the phases is in the saturation condition. The simulation extends to virtually all plant systems, including control, protection and balance of plant. However, the codes are not tightly bound to a particular technology and can be used to perform simulations of other physical systems. The design of the codes is modular, their inputs being built as a block diagram in which each block is an instance of a more general entity called module provided with particular data. Other particular-purpose codes can be connected to TRETA and TIZONA to perform a joint simulation in which several tasks may be run in parallel if the problem so allows.     

IR-360 nuclear power plant safety functions and component classification

The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.     

Gas-cooled reactor thermal hydraulic analyses with MELCOR

Pursuant to the Energy Policy Act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the reference design for the Next Generation Nuclear Plant (NGNP). Stemming from a U.S. Nuclear Regulatory Commission (NRC) HTGR research initiative, a need was identified for validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform licensing analyses. Because the NRC has used MELCOR for light water reactors (LWR) in the past and because MELCOR was recently updated to include gas-cooled reactor (GCR) physics models, MELCOR is among the system codes of interest to the NRC. This paper describes MELCOR modeling of the General Atomics' Modular High Temperature Gas-Cooled Reactor (MHTGR). The MHGTR is a suitable design for demonstration of MELCOR GCR modeling competency for two reasons: 1) the MHTGR is a predecessor to the more advanced General Atomics’ Gas-Turbine Modular High Temperature Reactor (GTMHR), and 2) experimental data useful for benchmark calculations may soon become available. Using the most complete literature references available for the MHTGR design, researchers at Texas A&M University (TAMU) constructed a MELCOR input deck for the MHTGR to partially validate MELCOR GCR modeling capabilities. Normal and off-normal system operating conditions were modeled with appropriate boundary and initial conditions. MELCOR predictions of system response were obtained for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) scenarios. Code results were checked against nominal MHTGR design parameters, physical intuition, and anticipated GCR thermal hydraulic response. No inherent deficiencies in MELCOR modeling capability were observed, suggesting that the newly-implemented GCR models are adequate for systems-level analysis. If and when experimental benchmark data becomes available, further validation activities may proceed given the modeling efforts discussed herein.     

RELAP5铅铋快堆模型拓展及验证

为研究铅铋快堆瞬态热工水力特性,对RELAP5程序进行二次开发,添加铅铋合金（LBE）物性模型和液态金属流动换热模型,并与NACIE-UP和CIRCE-ICE台架的实验结果进行对比。计算结果表明：NACIE-UP台架稳态流量和温度相对误差在2%以内,瞬态相对误差不超过5%,与其他系统程序CATHARE、ATHLET、RELAP5-3D、RELAP5/MOD3.3(modified)相比,本文程序的相对偏差不超过10%;CIRCE-ICE台架稳态流量和温度相对误差在2%以内,瞬态相对误差不超过10%。本文程序满足反应堆系统热工水力分析程序精度要求,可作为铅铋快堆安全分析的有效工具。     

模块式高温气冷堆核电站设计特点

模块式高温气冷堆具有安全、灵活、可靠、经济性好的优点,受到核技术先进国家的重视。本文着重介绍了美国新近推出的模块式高温气冷堆核电站的设计特点和安全特性。     

Comprehensive uncertainty and sensitivity analysis for coupled code calculations of VVER plant transients

The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics codes, is an important step to perform best-estimate calculations for plant transients of nuclear power plants. For applications in safety analysis, these coupled codes should be validated by benchmark calculations and, preferably, by comparison with plant transient data from operating plants. In addition, the results should be supplemented by applying uncertainty and sensitivity analysis methods, which allow to identify relevant parameters of models and solution procedures affecting the results and to quantify their relative importance. Both objectives were part of the VALCO project. The aspect of validation is presented in [S. Mittag, et al., 2004. Neutron-Kinetic Code Validation against Measurements in the Moscow V-1000 Zero-Power Facility, in press; T. Vanttola et al., 2004. Validation of coupled codes using VVER plant measurements, in press], the aspect of a comprehensive uncertainty and sensitivity analysis for coupled code calculations is the topic of this contribution. The results and experiences obtained by the analysis for two plant transients in a VVER-440 and a VVER-1000, respectively, are presented and discussed.     

Trends and needs in experimentation and numerical simulation for LWR safety

The very complex phenomena that need to be considered in safety analyses require use of sophisticated analytical tools. Basically, one-dimensional (1D) system codes have been used for a long time and have reached a degree of maturity. There are, however, limits to their capabilities and further developments are underway; these are outlined. The development of new generations of tools and methods can profit from the availability of increasingly powerful computers and advances in multiphase flow, information technology and numerical techniques. Three-dimensional (3D) situations need also to be addressed more frequently now. Certain developments in these directions that are already taking place in various EURATOM research programs and elsewhere are briefly reviewed; case studies of applications are discussed and lessons drawn.Future safety analyses for nuclear power plants may include use of Computational Fluid Dynamics (CFD) for parts of the primary system and the containment. First applications in this direction have already been made. Although 3D, single-phase CFD computations are commonplace, the size of the systems considered make these quite challenging. The real challenges lie, however, in two-phase flow CFD applications that are still at their very infancy. Coupling of neutronic and thermal-hydraulic codes is also necessary for certain problems.