Analysis of the Feed & Bleed procedure for the Ascó NPP: First approach study for operation support

The Asociación Nuclear Ascó-Vandellòs (ANAV) is the consortium responsible for the Ascó and Vandellòs-II reactors. The reactors are Westinghouse-design 3-loop PWRs with an approximate electrical power of 1000 MW. In order to operate the reactors ANAV prepared an Integral Plant Model for each plant. The model is used for supporting plant operation and control from the point of view of safety and competitiveness. The Technical University of Catalonia (UPC) has been working with ANAV in order to establish, qualify and use these Best Estimate (BE) models. This paper develops a specific use of the Ascó plant model for operation support and tries to clarify a group of scenarios related to the total loss of feed water (FW) and the corresponding Feed & Bleed (F&B) recovery procedure. The study characterizes the procedure with different calculations and provides results that have become helpful in establishing the impact of partial availability of Power Operated Relief Valves (PORVs) and HPIS trains, the maximum time for starting the procedure and some considerations on heat sink recovery. The use of the BE model provides a consistent overview of the final problem. The results of the analysis are a first approach for operation support.     

Advanced qualification process of ANAV NPP integral dynamic models for supporting plant operation and control

Extensive technical literature exists aimed at establishing the requirements needed to qualify a Nuclear Power Plant model. Most of this literature is focused on qualifying a model for licensing uses. Less documentation is available nowadays on the requirements needed when an Integral Plant Model is used for supporting plant operation and control of an actual commercial facility, while fulfilling its goals of safety and competitiveness. For the last 15 years the Technical University of Catalonia (UPC) has been working in this field along with Asociación Nuclear Ascó–Vandellòs (ANAV), which is a utility that presently runs three operating PWRs. The paper develops an advanced qualification process (AQP) of plant models for operation support, introduces the concept of plant configuration and explains how this activity complements other usual validation tasks.     

Analysis of an actual reactor trip operating event due to a high variation of neutron flux occurring in the Vandellòs-II nuclear power plant

The Asociación Nuclear Ascó-Vandellòs-II (ANAV) is a utility that runs three operating reactors. All of them are Westinghouse-design 3-loop Pressurized Water Reactor (PWR), with an approximate electrical power of 1000 MW. In the past ANAV prepared an Integral Plant Model for each plant using Relap5. The model is used to support plant operation and control from the point of view of safety and operation. The Technical University of Catalonia (UPC) has been working with ANAV since 1991 in order to establish, qualify and use these Best Estimate (BE) models. Among the different uses of such models, simulation of actual transients has special significance, since it is usually extremely helpful in order to understand their dynamic behaviour. This paper shows an example of how the analysis helped to clarify some questions about the sequence of events and the influence of manual actions. As some parameter time trends are not measurable by plant instrumentation, the model helped to provide a complete explanation of a transient for which the initiating event was a loss of off-site power.     

Licensing requirements for initial commissioning programs in Spain: Application to different PWR designs

This paper describes the overall licensing process in Spain, focusing on the initial commissioning requirements. The significance of this part of the regulatory work is evident both from the licensing and the licensee's points of view.Licensing in Spain is ruled by different laws which determine the general requirements and fix the licensing frame. Being a nuclear technology importer country, the base of the regulatory work lies on the rules and regulations of the country of origin of the plant, with the addition of case specific requirements. The application of this methodology to plants designed in different countries produces licensing processes which are similar in the overall, but very different in its development. It also means a special technical effort on the part of the regulatory body to cope with the problems arising from the use of different technologies and safety standards.The start-up programs from fuel loading to full power of a Westinghouse plant (Vandellós 2) and a Siemens-KWU plant (Trillo 1) are compared from the technical point of view, enhancing the differences that can be relevant for the regulatory work. The difficulties arising from the application of both the German and US concepts are discussed.     

Water chemistry technology – one of the key technologies for safe and reliable nuclear power plant operation

Water chemistry control is one of the key technologies to establish safe and reliable operation of nuclear power plants. Continuous and collaborative efforts of plant manufacturers and plant operator utilities have been focused on optimal water chemistry control, for which, a trio of requirements for water chemistry should be simultaneously satisfied: (1) better reliability of reactor structures and fuel rods; (2) lower occupational exposure and (3) fewer radwaste sources. Various groups in academia have carried out basic research to support the technical bases of water chemistry in plants. The Research Committee on Water Chemistry of the Atomic Energy Society of Japan (AESJ), which has now been reorganized as the Division of Water Chemistry (DWC) of AESJ, has played important roles to promote improvements in water chemistry control, to share knowledge about and experiences with water chemistry control among plant operators and manufacturers and to establish common technological bases for plant water chemistry and then to transfer them to the next generation of plant workers engaged in water chemistry. Furthermore, the DWC has tried and succeeded arranging R&D proposals for further improvement in water chemistry control through roadmap planning. In the paper, major achievements in plant technologies and in basic research studies of water chemistry in Japan are reviewed. The contributions of the DWC to the long-term safe management of the damaged reactors at the Fukushima Daiichi Nuclear Power Plant until their decommissioning are introduced.     

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.     

Modeling of condensation in horizontal tubes

Condensation in horizontal tubes plays an important role for example in the determination of the operation mode of horizontal steam generators of VVER reactors or passive safety systems for the next generation of nuclear power plants. Two different approaches (HOTKON and KONWAR) for modeling this process have been undertaken by Forschungszentrum Jülich (FZJ) and the University for Applied Sciences Zittau/Görlitz (HTWS) and implemented into the 1D-thermohydraulic code ATHLET, which is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH for the analysis of anticipated and abnormal transients in light water reactors. Although the improvements of the condensation models are developed for different applications between generators and emergency (VVER steam generators-emergency condenser of the SWR1000) with very different operation conditions (e.g. the temperature difference over the tube wall in HORUS is up to 30 K and in NOKO up to 250 K, and the heat flux density in HORUS is up to 40 kW/m² while that in NOKO is up to 1 GW/m²) both models are now compared and assessed by Forschungszentrum Rossendorf FZR e.V. Therefore post-test calculations of four selected HORUS experiments were performed with ATHLET/KONWAR. It can be seen that the calculations with the extension KONWAR as well as HOTKON significantly improve the agreement between computational and experimental data.     

Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during cooldown to cold shutdown, and in the validity of a two-tier calculational method. The results have been directly used in updating the plant shutdown PSA, by changing the success criteria for core cooling during cooldown of the plant and showing a reduction in overall risk.     

基于顺序控制技术的反应堆自动启动系统

在深入分析反应堆启动过程各项技术操作的基础上,提出利用顺序控制技术实现反应堆自动启动的构想.对基于顺序控制技术的反应堆自动启动系统的设计思路、层次结构、系统组成和系统设计的关键技术进行了研究,并初步进行了系统设计.所设计的顺序控制系统能够实现反应堆的自动启动,减轻运行人员工作负担,从而提高了反应堆启动运行的安全性和经济性.     

Research and development on nuclear hydrogen production using HTGR at JAERI

JAERI has been conducting R&D on HTGR and on hydrogen production using HTGR. The reactor technology has been developed using HTTR installed at Oarai site of JAERI. HTTR reached its full power operation of 30MW in 2001 and demonstrated reactor outlet helium temperature of 950°C in April 2004. As for the hydrogen production technology, the thermo-chemical IS process is under study. The process control method for continuous hydrogen production has been examined using a bench-scale apparatus. Also, studies are underway on process improvement and on materials of construction to be used in the corrosive environment. As for the system integration of HTGR and the hydrogen production plant, R&D is underway aiming to develop technologies for safe and economical connection. It covers safety technology against explosion, safety technology against radioactive materials release, control technology to prevent the thermal disturbance from hydrogen production plant to reactor, etc.     

Progress in the studies of the advanced plutonium fuel assembly

A new assembly concept, designated APA (for dvanced lutonium fuel ssembly), should make it possible to multi-recycle plutonium in pressurized water reactors. The basic idea is founded on the manufacture of a large plutonium thin annular fuel rod with an inert support, cooled on both faces. The absence of plutonium generation, combined with moderate fuel temperature should make it possible to achieve substantial burn-up fractions in these rods. The assembly is compatible with the internals of a Pressurized Water Reactor (PWR), and provides for permanent reversibility. Neutronic studies showed a compliance with actual safety/control criteria. A multi-recycling scenario was simulated for 84 years' operation with a 65 GW electrical power installed capacity, comprising forty-five 1450 MW electrical power PWRs, 32 of which are loaded with UO₂ and 13 with APAs. It showed that the plutonium inventory is controlled. Thermal-hydraulic studies showed one can find an annular rod geometry allowing one to respect both margins to Critical Heat Flux (CHF) during normal and accidental operations and void fraction limitations.     

双堆布置核电厂公用设施对双堆超设计基准事故缓解的影响和改进

福岛事故后,同一厂址多台机组同时发生超设计基准事故(包括严重事故)的后果开始受到关注,为此需要从设计上保证核电厂事故应对措施的独立性。我国运行和在建的大部分核电厂为双堆布置的二代改进型核电厂。分析表明,水压试验泵和安全壳过滤排放系统(EUF)为双堆公用,对双堆超设计基准事故的应对能力存在影响。进而研究了这两个公用设施在现有电厂中的潜在改进选项,从尽量减少硬件改动的目的出发提出了最可能的改进方案。其中EUF交替排放仅仅通过操作规程的变化,凭借一套公用系统即可实现双堆的卸压目的。进一步计算也证明,合理选取交替排放的时间窗口,EUF交替排放在最保守及最现实的事故情况下均能确保双堆安全壳的安全。     

Computer Control System for Sodium Test Loops

A digital computer system was developed for automatic operation of the Toshiba Sodium Test Loops, intended for studying the control system of liquid metal fast breeder reactors.

The test installation is controlled by a system of two TOSBAC-40C central processing units, which take care of plant monitoring, data logging, information display, sequence control and closed loop control. The operator console linked to the computers can handle more than 3,000 process input/output signals. The state of the plant and measured data are displayed on two color cathode-ray tube installed on the console. Normal operation of the four loops constituting the installation can be supervised by one single operator.

Safety in the case of power failure and other emergencies is assured by a safety protection system that functions independently of the computer system.     

我国核电厂数字化仪控系统相关运行事件分析

从事件原因和事件后果两个方面分析了我国核电厂发生的25起数字化仪控系统相关运行事件,找出了安全级和非安全级数字化仪控系统存在的问题,总结了我国核电厂数字化仪控系统的运行情况,并提出了需要关注和进一步改进的方面,为后续新建核电厂提供借鉴。     

Phenomena-based thermal-hydraulic modeling requirements for systems analysis of a Modular High Temperature Gas-Cooled Reactor

The Modular High Temperature Gas-Cooled Reactor (MHTGR) is a candidate design for new production and commercial power nuclear reactors. The MHTGR has inherent safety features including: (1) passive shutdown and decay heat removal, (2) reduced requirements for operator intervention, thereby reducing sensitivity to operator error, and (3) long time intervals for corrective action. In support of the Department of Energy's (DOE) initial development of the design, the authors have completed an evaluation of the thermal-hydraulic phenomena that will dominate the plant response during representative normal, off-normal and accident conditions. Phenomena having significance to the plant behavior have been identified, and ranked with respect to their relative importance in satisfying operational, investment and/or safety criteria. The resulting information provides the basis for evaluating the applicability of existing computer codes, and defines the requirements for the development of new codes, for thermal-hydraulic systems analysis. The phenomena-based requirements also support the quality assurance related verification and validation of these codes. This paper briefly describes the methodology employed, and gives illustrative examples of the resulting requirements. References are cited for reports that document the full body of requirements and provide additional information for the methodology.     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

微型核反应堆技术发展趋势研究

微型核反应堆采用四代非轻水堆、热管堆以及三代轻水堆等固有安全性高的堆型,可以为偏远海岛和矿区、边防哨所和基地、应急救灾、太空和深海探索等创新场景提供长续航高可靠能源,具有广阔的应用前景,是实现国家战略的重要技术支撑之一。本研究总结了微型核反应堆的定义和主要研发堆型,描述了微型核反应堆固有安全性高、易于模块化和扩展、可运输性、便于部署、自主运行等创新技术特征,分析了微型核反应堆新型燃料、主回路一体化、新型热电转换装置、非能动安全系统、智能运维以及核能与其他能源耦合等关键技术的发展趋势,可为制定适用于我国的微型核反应堆发展技术路线提供支撑。      

AP1000钢制安全壳厚度对传热性能的影响

AP1000是目前世界上安全性最高的第三代大型压水堆之一,相比于二代压水堆,其重要特征是将预应力混凝土的安全壳改为钢制安全壳,在整个冷却过程中钢制安全壳起着重要的作用。本文利用WGOTHIC程序建立AP1000整体长期空气冷却模型,对安全壳厚度进行研究,得到了传热性能与安全壳厚度的关系。结果表明,在一定范围内随安全壳厚度的增加,总体安全性得到较大提升,这为采用钢制安全壳的核电站设计提供了理论参考。     

Safe operation and life extension of RBMK plants

The article presents an overview of the issues for ensuring safe operation of the Russian nuclear power plants with RBMK-1000 reactors.The operating data for such NPPs are included, the upgrading activities are described. The examples of the major implemented activities and their contribution to safety enhancement are given.The activities aimed at the justification of safe operation of NPP with RMBK-1000 reactors are described and some results are provided.The basic conclusion is that today it can be claimed with good reason that operation of the nuclear power plants with RBMK reactors is as safe as operation of NPP with domestic and foreign reactors of other types.     

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.