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.     

SVBR-75/100 Multipurpose Modular Low-Power Fast Reactor with Lead–Bismuth Coolant

The results of development work on the SVBR-75/100 modular fast reactor cooled by lead–bismuth coolant are presented. The design is based on the experience gained operating reactors in nuclear-powered submarines. The SVBR-75/100 reactor is inherently safe because of its low power, the chemical inertness and high boiling point of the lead–bismuth coolant, and the integrated basin-type arrangement of the first-loop equipment. This eliminates the conflict between the safety and economic requirements that is characteristic of conventional-type reactors. The lower power makes it possible to build the reactor in a factory and deliver the reactor using any form of transportation to the nuclear power plant site. All safety systems operate passively. Most of these systems are used as normal-operating systems. This creates conditions for multipurpose applications of such reactor modules in developed and developing countries.     

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.     

Mononitride Fuel for Fast Reactors

Substantiation is given for the development of nuclear power based on inherently-safe fast reactors with a mononitride core. Fundamental studies and design work on the development of such reactors with lead (BREST-OD-300), lead–bismuth (SVBR-75/100), and sodium coolant (BN-800) are being performed. The development of nuclear power in our country is based on organizing a closed fuel cycle. The results of experimental investigations of the properties of mononitride fuel are correlated. Mononitride fuel meets all requirements for fast-reactor fuel.     

Development of Fast Helium Reactors in Russia

The need for fast reactors in nuclear power is discussed. The advantages of fast helium reactors over fast liquid-metal reactors are shown. Fast helium reactor concepts and the possibilities of improving breeding by using fuel of higher density than oxide fuel, specifically, carbides and nitrides, are examined. It is indicated that the development and use of high-temperature helium-cooled thermal- and fast-neutron reactors are promising as part of an international collaboration in nuclear power in the 21st century.     

Fast reactors in the energy security for the stable development of Russia

New aspects of the development of a future nuclear power system based on the advanced technologies of a closed nuclear fuel cycle with fast-neutron reactors are discussed on the basis of an analysis of systems problems pertaining to present-day nuclear power. The systems requirements ensuring adequate fuel for nuclear power with any installed capacity and maximum use of natural uranium and thorium brought into the fuel cycle are formulated. Sodium-cooled fast reactors, which thus far possess the highest level of technological readiness for commercialization, are given a special role in the formation of a new technological platform for large-scale nuclear power of the 21st century.     

BN-800 as a New Stage in the Development of Fast Sodium-Cooled Reactors

The role of fast reactors in a strategy for developing nuclear power in Russia because of the inevitable exhaustion of natural uranium deposits in the foreseeable future is discussed. The BN-800 reactor, which is under construction and incorporates unique solutions – greatly enhancing the safety of the reactor – to technical and constructional problems, is examined. Cost assessments taking account of the complete life cycle show that fast reactors could be no more expensive than the most widely reactors in the world – water-moderated water-cooled reactors.Closing the BN-800 nuclear fuel cycle will make it possible to solve the problem of utilizing plutonium and actinides. This makes fast reactors safer for the environment.     

SVBR-100 module-type fast reactor of the IV generation for regional power industry

In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used.The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.     

The closed on-site fuel cycle of the brest reactors

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10⁻² – 10⁻³ of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature.

The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000.

The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.     

Self-consistent model of nuclear power and nuclear fuel cycle

Under discussion are such major aspects of the nuclear energy sector as cost effectiveness, nuclear and environmental safety of reactors and nuclear fuel cycle facilities, sustained fuel supply, and proven feasibility of a proliferation-resistant technology. These requirements can be met, for instance, by a two-circuit nuclear facility with an inherently safe fast reactor of the BREST type which is expected to produce electricity at a cost not higher than that at modern LWRs. Fuel supply to such facilities and to a relatively small number of thermal reactors with BR<1, could be provided by fast reactors using depleted uranium as makeup fuel and having a small breeding gain in the core (CBR≈1.05) and bottom blanket (full BR≈1.1). Use of a high-boiling metallic coolant (lead) affords deterministic nuclear, technical and environmental safety of the plants in design-basis and hypothetical accidents. Introduction of a transmutational NFC is viewed as one of the avenues to global environmental safety, when the equivalent activity of long-lived high-level waste is made lower or close to the activity of the source material going into energy production. With such a balance in place, nuclear power could be regarded, in a sense, as waste-free.     

Handling Irradiated RBMK-1000 and VVÉR-1000 Fuel during the Growth of Nuclear Power

On-site storage facilities, consisting of ponds with water, for irradiated RBMK-1000 fuel are now close to being filled. To continue operating nuclear power plants with RBMK reactors, it is necessary to select one possible method for handling irradiated fuel.A variant of long-term storage followed by reprocessing is examined and considerations are presented for future use of reprocessed irradiated RBMK and VVÉR fuel as fuel for an initial load for naturally-safe fast reactor. Important points in handling irradiated RBMK-1000 fuel include economic assessments and requirements for a strategy for development of nuclear power in Russia based on closure of the nuclear fuel cycle with radiation-equivalent burial of wastes and utilization of accumulated plutonium for fast reactors. 3 figures, 2 tables, 8 references.     

金属核燃料快中子脉冲堆核-热-力耦合计算方法研究

为确保快中子脉冲堆的运行安全,防止超临界脉冲对材料造成物理损伤,需要对快中子脉冲堆脉冲工况进行模拟分析。本研究针对金属核燃料快中子脉冲堆,基于点堆动力学方法、蒙特卡罗方法和有限元力学方法,对Godiva-I脉冲堆开展了核热力耦合计算分析研究。计算结果表明,反应性温度系数和裂变率与实验值吻合良好,反应性、温升、表面位移、表面应力与实际情况相符合。因此,本文建立的“核-热-力”耦合计算方法可应用于金属核燃料快中子脉冲堆的分析计算,具有一定的可靠性。      

Radiation Characteristics of Fuel and Wastes in the Uranium–Plutonium and Thorium–Uranium Fuel Cycle

The radiation characteristics of fuel cycles of various reactors – replacement candidates in the future nuclear power – are compared. Proceeding from the basic requirements (safety, fuel supply, and nonproliferation of fissioning materials), inherently safe fast reactors of the BREST type can be used as the basis for large-scale nuclear power. Thermal reactors, which can burn enriched uranium, thorium–uranium fuel, or mixed uranium–plutonium fuel with makeup with fissioning materials from fast reactors, will operate for a long time simultaneously with fast reactors in the future nuclear power. VVÉR-1000 and CANDU reactors are examined as representatives of thermal reactors; for each of these reactors the operation in various variants of the fuel cycle is simulated. It is shown that with respect to radiation characteristics of the fuel and wastes the thorium–uranium fuel cycle has no great advantages over the uranium–plutonium cycle.     

A “Generation III+” nuclear reactor for space needs

Nuclear power has been extensively used in space applications but in very few cases it entailed the use of nuclear fission reactors. The paper deals with a feasibility study of a space nuclear reactor for planet settlements, based on a “Generation III+” technology, specifically the integral type PWR one, currently under study in several projects (e.g. IRIS, CAREM, IMR, SMART) for its simplicity, compactness and cost reduction, but mainly based on already proven technology.This approach was suggested by the following constraints, typical in research activities for space applications, especially when nuclear is considered: to obtain a solution which has to be extremely reliable, moderately expensive in the R&D program, deployable in a reasonable period of time, safe and operable/controllable with no human intervention for several years. These requirements have been coped with by means of a reactor design based mainly on the well proven PWR technology of present terrestrial reactors.The fuel selection, the core configuration and related burnup, the primary and conversion systems, the reactivity and system control as well as preliminary considerations on safety are presented.The space nuclear reactor is envisaged to be suited mainly for stationary needs on a planet, by producing electrical energy of the order of 100 kW for nearly 4000 days. The obtained results are fully satisfactory and suggest to continue the study to reach a more detailed design.     

空间核反应堆电源用核燃料研制进展

核燃料是空间核反应堆电源的主要材料之一,由于空间核反应堆电源的运行条件明显有别于地面反应堆,空间核反应堆电源用核燃料的类型和技术要求也明显不同于地面反应堆。国际上空间核反应堆电源用核燃料研制取得了长足的进展,多种核燃料材料在工程应用中得到了检验,并在持续开发新型核燃料。我国在亚化学计量二氧化铀芯块、铀钼合金、铀氢锆合金、碳化铀芯块、氮化铀芯块等多种具备在空间堆中应用的燃料材料上开展了一定的研究,并掌握了部分材料性能数据。本文就上述内容展开论述,同时针对与国际相应领域明显落后的实际情况,提出了我国后续核燃料研究的初步设想。     

商用压水堆核燃料研发进展与应用展望

压水堆(PWR)是目前核电厂反应堆的主力堆型,而核燃料是反应堆的能量源泉和放射性裂变物质的主要来源,关乎核电厂的经济性和安全性。本文对当前国际上面向商用PWR应用研发的掺杂UO₂燃料、高铀密度燃料、微封装燃料和金属燃料的性能特点、技术状态及前景进行了归纳和评价。在掺杂UO₂燃料中,大晶粒燃料具有较高的技术成熟度,将在PWR实现大规模商用;高铀密度燃料和金属燃料在高温水腐蚀氧化问题以及事故下的行为仍待研究解决;具有极致安全的微封装燃料更适合特殊用途的小型反应堆。应协同开展先进燃料组件设计、建立设计准则以及研发高保真的性能分析技术等,以充分发挥新型燃料的可靠性及高燃耗优势。     

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.     

星球表面用核反应堆临界安全分析

核反应堆电源具有寿命长、可全天候工作等特点,可作为星球表面及其他深空探测任务的电源。针对星球表面用核反应堆电源在发射过程中重返地面的临界安全问题,提出了星球表面用核反应堆的临界安全分析要求、分析假设与模型,并对反应堆临界安全特性及采取的临界安全措施进行了计算分析。计算结果表明,不同假设掉落环境下的星球表面用核反应堆的有效增殖因数均小于0.98,满足临界安全要求。反应堆通过采用Mo-14%Re合金结构材料、设置相对较厚的堆芯反射层以及在反射层包壳和堆芯外围涂覆Gd₂O₃涂层等措施有利于确保反应堆在事故时处于次临界状态。     

Requirements for 21st centry nuclear power plants

The development of energy production in the 21st century will be subject to more uniform per capita and regional consumption. Among the competing sources of energy, the positive qualities of nuclear power-unlimited fuel resources, high energy intensiveness, and ecological compatibility with the possibility of the wastes being highly concentrated—predetermine the development of large-scale nuclear power. The conditions for the development of such nuclear power are its ecological effectiveness and safety (of the reactors and the fuel cycle with the production of wastes), nuclear fuel breeding with adequate characteristics, and guarantees of nonproliferation of fissioning materials. Continuity in the development of nuclear power dictates the requirements for reactor systems in the near and distant future. The acceptable level of safety is closely related to the scales of nuclear power and the applications of nuclear energy sources. However, progress in decreasing the potential danger of reactors and decreasing the cost of protective systems is unavoidable. In choosing new directions, it is important to demonstrate the new qualities in the solution of the problems facing nuclear power in the future. An adequate diversity of reactor technologies could exist in the future. The requirements that will face nuclear power plants in the future stages of development and the expected stages of this development are discussed. The jourmal variant of this report at the 10th annual conference of the Nuclear Society “From the first nuclear power plant in the world to power engineering of the twenty-first century” (June 28–July 2, 1999, Obninsk) Russian Science Center “Kurchatov Institute”. Translated from Atomnaya énergiya, Vol. 88, No. 1, pp. 3–14, January, 2000.