Design and licensing strategies for the fluoride-salt-cooled,high-temperature reactor (FHR) technology

Fluoride-salt-cooled, high-temperature reactor (FHR) technology combines the robust coated-particle fuel of high-temperature, gas-cooled reactors with the single phase, high volumetric heat capacity coolant of molten salt reactors and the low-pressure pool-type reactor configuration of sodium fast reactors. FHRs have the capacity to deliver heat at high average temperature, and thus to achieve higher thermal efficiency than light water reactors. Licensing of the passive safety systems used in FHRs can use the same framework applied successfully to passive advanced light water reactors, and earlier work by the NGNP and PBMR projects provide an appropriate framework to guide the design of safety-relevant FHR systems. This paper provides a historical review of the development of FHR technology, describes ongoing development efforts, and presents design and licensing strategies for FHRs. A companion review article describes the phenomenology, methods and experimental program in support of FHR.     

Sensitivity studies of modular high-temperature gas-cooled reactor postulated accidents

The results of various accident scenario simulations for the two major modular high temperature gas-cooled reactor (HTGR) variants (prismatic and pebble bed cores) are presented. Sensitivity studies can help to quantify the uncertainty ranges of the predicted outcomes for variations in some of the more crucial system parameters, as well as for occurrences of equipment and/or operator failures or errors. In addition, sensitivity studies can guide further efforts in improving the design and determining where more (or less) R&D is appropriate. Both of the modular HTGR designs studied – the 400-MW(t) pebble bed modular reactor (PBMR, pebble) and the 600-MW(t) gas-turbine modular helium reactor (GT-MHR, prismatic) – show excellent accident prevention and mitigation capabilities because of their inherent passive safety features. The large thermal margins between operating and “potential damage” temperatures, along with the typically very slow accident response times (approximate days to reach peak temperatures), tend to reduce concerns about uncertainties in the simulation models, the initiating events, and the equipment and operator responses.     

Neutronic and thermo-hydraulic analyses of a small,long-life HTGR for passive decay-heat removal

Since the accident at Fukushima Daiichi Nuclear Power Plant in 2011, design concepts for nuclear reactors have been reconsidered with much greater emphasis placed upon passive systems for decay-heat removal. By considering this issue, the design parameter conditions for high temperature gas-cooled reactors (HTGRs) with passive safety features of decay-heat removal were obtained by residual-heat transfer calculation using equations for fundamental heat transfer mechanisms in our previous works. In the present study, the appropriate size of reactor core for a 100 MW_t reactor operating at 1123 K of the initial core temperature was found using the conditions. Consequently, neutronics and thermo-hydraulic analyses for the proposed reactor core were performed and the proper optimizations to control the excess reactivity and flatten the change in power peaking factor during operation were done successfully. By the systematic method to decide the core design which satisfies the condition for passive decay-heat removal, a long-life small HTGR concept whose excess reactivity was small during the operation was shown. The small excess reactivity is a significant advantage from the view point of safety in reactivity accident.     

Innovative nuclear technology based on modular multi-purpose lead–bismuth cooled fast reactors

Today's nuclear power is in the state of an intrinsic conflict between economic and safety requirements. This fact makes difficult its sustainable development.

One of the ways of finding the solution to the problem can be the use of modular fast reactors SVBR-75/100 cooled by lead–bismuth coolant that has been mastered in conditions of operating reactors of Russian nuclear submarines.

The inherent self-protection and passive safety properties are peculiar to that reactor due to physical features of small power fast reactors (100 MWe), chemical inertness and high boiling point of lead–bismuth coolant, integral design of the pool type primary circuit equipment.

Due to small power of the reactor, it is possible to fabricate the whole reactor at the factory and to deliver it to the NPP site in practical readiness by using any kind of transport including the railway.

Substantiation of the high level of reactor safety, adaptability of the SVBR-75/100 reactor relative to the fuel type and fuel cycle, issues of non-proliferation of nuclear fissile materials, opportunities of multi-purpose usage of the standard SVBR-75/100 reactors have been viewed in the paper.     

Reactor systems developed by OKBM in the development of the atomic energy industry, nuclear power, and the nuclear-powered fleet in Russia

The various types of reactor systems which OKBM designed for different purposes over the 60 years of its existence are reviewed: commercial uranium-graphite and heavy-water reactors, reactors and steam-generating systems for naval and for civilian (icebreakers) ships, fast reactors, low-and medium-capacity reactor systems for regional power generation, and high-temperature gas-cooled reactors. The results of work performed by OKBM on the development of the cores of propulsion reactors and alternative fuel for VVéR-1000 reactors are also presented. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 3–21, January, 2007.     

Safety features of the gas-cooled fast breeder reactor

The safety features of the gas-cooled fast breeder reactor (GCFR) are described in the context of the 300-MW(e) demonstration plant design. They are of two general types, inherent and design-related. The inherent features are principally associated with the helium coolant and the nuclear coefficients. Design-related features influencing safety include shutdown systems, residual heat removal systems, method of core support, and the prestressed concrete reactor vessel (PCRV). This paper discusses the safety-related aspects of each of these. Recently completed residual heat removal system reliability studies are also discussed. The probability of residual heat removal system failure in the GCFR is found to be lower than that described for light water reactors. The safety characteristics of larger plants are examined, and increases in size are found to improve GCFR safety margins.     

Nuclear power development in market conditions with use of multi-purpose modular fast reactors SVBR-75/100

This article presents an innovative nuclear power technology, based on the use of modular type fast-neutron reactors SVBR-75/100 having heavy liquid-metal coolant, i.e. eutectic lead–bismuth alloy, which was mastered in Russia for the nuclear submarines’ reactors. Reactor SVBR-75/100 possesses inherent self-protection and passive safety properties that allow excluding of many safety systems necessary for traditional type reactors. Use of this nuclear power technology makes it possible to eliminate conflicting requirements among safety needs and economic factors, which is particularly found in traditional reactors, to increase considerably the investment attractiveness of nuclear power based on the use of fast-neutron reactors for the near future, when the cost of natural uranium is low and to assure development of nuclear power in market conditions. On the basis of the factory-fabricated “standard” reactor modules, it is possible to construct the nuclear power plants with different power and purposes. Without changing the design, it is possible for reactor SVBR-75/100 to use different kinds of fuel and operate in different fuel cycles with meeting the safety requirements.     

Design of the exhaust system in the confinement of the HTR-10

The modular high temperature gas-cooled reactor has a vented confinement instead of a gastight pressurized containment due to its passive safety features. The safety class negative pressure exhaust system is used in the heating, ventilation and air conditioning system to fulfill all kinds of safety-related functions at the normal operation and during accidents. This paper introduces and reviews the design of safety class negative pressure exhaust systems of the 10 MW high temperature gas-cooled reactor-test module.     

IAEA activities on passive safety systems and overview of international development

Although a major focus of nuclear power reactor development efforts in industrialised countries is on large evolutionary units and design modifications that take maximum advantage of successful proven features and components, consideration is also given to utilisation of passive safety systems and inherent safety features. The various advanced reactor designs: evolutionary, large water-cooled reactor designs; evolutionary, medium size water-cooled reactor designs; concepts requiring substantial development; gas-cooled reactor concepts; and liquid metal-cooled fast reactors, incorporate a wide variety of passive safety features for initiation of safety systems, for residual heat removal and for containment heat removal. Organizations have established testing programmes to confirm their performance. A number of IAEA activities have lead to the conclusion that the use of passive safety features can be a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions. However, care should be taken to evaluate possible new failure mechanisms, and both passive and active systems should be assessed from the standpoint of reliability and economics. Key technical issues include: the quantification of reliability over a wide range of conditions: economics, speed of action, plant ageing, demonstration of technical feasibility, in-service testing, ease of maintenance and minimization of personnel radiation exposure. Many member states conduct substantial work on the design, modelling, development and reliability assessments of passive safety systems. Continued information exchange can benefit the involved member states, and the IAEA is providing a forum for review of programmes, project directions, and the results achieved.     

DVI管小破口失水事故实验研究

在模块化小型反应堆非能动安全系统综合模拟实验装置上进行了压力容器直接注入（DVI）管小破口失水事故实验,研究了DVI管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。研究结果表明：模块化小型反应堆DVI管小破口失水事故中,非能动安全系统可对堆芯进行注水,有效导出堆芯衰变热量,保护堆芯安全。     

Update of the French R&D strategy on gas-cooled reactors

Gas-cooled reactors take up a strong second role in France's R&D strategy on future nuclear energy systems as priority was given in 2005 to fast neutron reactors with multiple-recycle for their potential to optimally use uranium resource and minimize the long term burden of radioactive waste. Owing to the European past experience on sodium-cooled fast reactors (SFRs), this reactor type was logically selected as reference for a new generation fast neutron reactor intended to be tested as a prototype in the 2020s and be ready for industrial deployment around 2040. At the same time, the potential merits of a gas fast reactor (GFR) with ceramic clad fuel for a safe management of cooling accident are acknowledged for the potential of this reactor type to resolve critical issues of liquid fast reactors (safety, operability and reparability). A pre-feasibility report on a first concept of GFR was issued in 2007 that summed-up results of a 5-year international R&D effort on GFR fuel technology, reactor design and operating transient analyses. This report established a global confidence in the feasibility of this concept and its potential for attractive performances. Furthermore, it suggested directions of R&D to generate by 2012 an updated concept with improved performances and taking better benefit from GFR specific technologies.A second activity on gas-cooled reactors originates from the current interest of CEA's industrial partner AREVA in high or very high temperature reactors (V/HTR) for supplying hydrogen, synthetic hydrocarbon fuels and process heat for the industry. This activity currently encompasses R&D on V/HTR key technologies such as particle fuel fabrication, high temperature compact heat exchangers and coupling technologies to various power conversion systems. R&D on V/HTR and GFR are synergistic in various respects. The GFR can be viewed as a more sustainable version of the VHTR and synergies exist in research on heat resisting materials, helium system technology and power conversion systems. Both reactors require active research in materials and spur developments of new metallic alloys and ceramics applicable to other advanced nuclear systems.     

Analysis of emergency regimes of modular HTGR

The designs of the modular high-temperature gas-cooled reactors GT-MHR and MHR-T, developed in our country, are based on world experience in the development and operation of HTGR with prismatic fuel assemblies and ceramic microfuel as well as on innovative solutions for the energy conversion system. The high safety level of nuclear power plants with modular HGTRs is reached by developing internal self-protection properties and using passive safety systems. Approaches to safety analysis, the computer codes used, and the verification of these codes are presented. The results of an analysis of the most serious accidents from the standpoint of core damage and radiation after effects are examined. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 52–57, January, 2007.     

铅铋冷却氮化物燃料小型模块化快中子反应堆堆芯物理特性分析

国际原子能机构(International Atomic Energy Agency,IAEA)认为小型模块化反应堆具有很好提高核能安全性、经济性和防止核扩散的能力,是未来核能最具发展前景的堆型之一。为适应未来核能发展的需求,提出了一种铅铋冷却氮化物燃料小型模块化反应堆(Small Modular Pb-Bi Cooled Reactor with Nitride Nuclear Fuel,SMPBN)设计方案,并利用PIJ组件计算程序和CITATION堆芯计算程序对SMPBN的物理特性和安全特性,包括反应性系数及其随燃耗变化、卸料燃耗、功率峰因子、燃料转换比和停堆余量等进行了深入分析。通过分析,认为SMPBN在20年寿期内,具有很好的燃料转换能力,不需要换料,反应性波动很小,反应性系数均为负值,具有固有安全性,符合国际上第四代反应堆的要求。     

Nuclear power technologies at the stage of sustainable nuclear power development

It is not simple to solve the problem of competitiveness of nuclear power technologies in evolutionary upgrading the conventional nuclear power plants (NPP) such as light water reactors (LWR), which requires high expenditure for safety. Moreover, the existing LWRs cannot provide nuclear power (NP) for a long time (hundreds of years) because the efficiency of use of natural uranium is low and closing the nuclear fuel cycle (NFC) for those reactors is not expedient.The highlighted problem can be solved in the way of use of innovative nuclear power technology in which natural uranium power potential is used effectively and the intrinsic conflict between economic and safety requirements has been essentially mitigated.The technology that is most available and practically demonstrated is the use of reactors SVBR-100 — small power multi-purpose modular fast reactors (100 MWe) cooled by lead-bismuth coolant (LBC). This technology has been mastered for nuclear submarines’ reactors in Russia.High technical and economical parameters of the NPP based on RF SVBR-100 are determined from the fact that the potential energy stored in LBC per a volume unit is the lowest.The compactness of the reactor facility SVBR-100 that results from integral arrangement of the primary circuit equipment allows realizing renovation of power-units LWRs, the vessels’ lifetime of which has been expired. So due to this fact, high economical efficiency can be obtained.The paper also validates the economical advantage of launching the uranium-fueled fast reactors with further changeover to the closed NFC with use of plutonium extracted from the own spent nuclear fuel in comparison with launching fast reactors directly with on uranium-plutonium fuel on the basis of plutonium extraction from spent nuclear fuel of LWRs.     

Future potential of nuclear heat utilization in energy, economy and environment

In this study quantitative analyses are made to clarify the possible roles of S-HTGRs (Small-sized modular High Temperature Gas-cooled Reactors) in our future energy systems. The results obtained show the good possibility of S-HTGRs to compete economically with L-HTGRs (Large-sized HTGR) taking into account the effects of modularization, learning, mass production, and simplification of safety systems. In the electricity market, S-HTGRs can well compete with coal steam electric power and LWR electric power if they are located close to demand areas. In addition the high temperature nuclear heat from small-sized modular gas-cooled reactors has the potential of contributing to reduce the amount of imported fossil fuels and also SO₂, NO_x, and CO₂ emissions.     

Simulation of the small modular reactor severe accident scenario response to SBO using MELCOR code

Advanced small modular reactors (SMRs) use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In present work, the SMRs severe accident caused by the station blackout (SBO) was modeled and analyzed using MELCOR code, and the simulation of the accident scenario response to SBO was conducted. Based on the steady state calculation, which agrees well with designed values, we introduced the SBO accident for transient calculation. First, the case of the SBO accident without the passive core cooling system (PXS) was calculated. The progression and scenario in the reactor pressure vessel (RPV) and the containment were simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. The station black-out transient in the SMR can be simulated accurately, and the main failure model in the accident process can be concluded. Then three other cases of the SBO accident with different passive safety systems (core makeup tank (CMT), accumulator (ACC), passive residual heat removal system heat exchanger (PRHR HX), automatic depressurization system (ADS)) of the PXS were calculated respectively, and the results for different passive safety systems were compared. The passive core cooling system can not only provide water to the primary coolant system, but also take away the reactor decay residual heat. So in a station black-out transient, we can get more time for restoring AC power, and effectively prevent the accidents such as Fukushima.     

Sensitivity studies of air ingress accidents in modular HTGRs

Postulated air ingress accidents, while of very low probability in a modular high-temperature gas-cooled reactor (HTGR), are of considerable interest to the plant designer, operator, and regulator because of the possibility that the core could sustain significant damage under some circumstances. Sensitivity analyses are described that cover a wide spectrum of conditions affecting outcomes of the postulated accident sequences, for both prismatic and pebble-bed core designs. The major factors affecting potential core damage are the size and location of primary system leaks, flow path resistances, the core temperature distribution, and the long-term availability of oxygen in the incoming gas from a confinement building. Typically, all the incoming oxygen entering the core area is consumed within the reactor vessel, so it is more a matter of where, not whether, oxidation occurs. An air ingress model with example scenarios and means for mitigating damage are described. Representative designs of modular HTGRs included here are a 400-MW(th) pebble-bed reactor (PBR), and a 600-MW(th) prismatic-core modular reactor (PMR) design such as the gas-turbine modular helium reactor (GT-MHR).     

Materials and design concepts of gas-cooled reactor systems

The family of gas-cooled reactors being developed in the United States by Gulf General Atomic consists of the steam-raising and direct cycle versions of the high temperature gas-cooled reactor (HTGR) for electric power generation, the hydrogen-producing HTGR for chemical process applications, and the gas-cooled fast reactor (GCFR), a high gain breeder. The aim of this paper is to describe the underlying design concepts that are common to all of these reactors and relate these design concepts to the choice of both structural and fuel materials for the wide variety of environmental conditions encountered throughout the world. Interwoven with this discussion are typical examples of the interaction of design activities and materials selection required to give a reactor system of maximum safety and reliability, favourable environmental features, and minimum cost.     

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants.Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.     

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。