Exploring new coolants for nuclear breeder reactors

Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the cancelled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants.     

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

Actinide breeding and reactivity variation in a thermal spectrum ADSR – Part 2: Enhanced sustainability in the thermal spectrum

For more sustainable nuclear power to be realised improvements will be needed in the efficiency with which energy is derived from the remaining finite uranium reserves, while at the same time delivering reductions in the quantities of long-lived actinides contained within the spent fuel. The use of fast reactors to achieve this is the subject of renewed interest due to their beneficial capability to both burn and breed transuranic actinides. However, fast reactors have a mixed track record and have never been deployed in significant numbers despite considerable investment in the development of the technology over many years. In light of these difficulties, future advances in nuclear technology may be more readily realised through enhancements to the existing, well proven light water reactor (LWR) technology base.     

The evaluation of a reactivity insertion due to the group motion of FBR core subassemblies beyond Design Basis Ground Motion

The seismic probabilistic safety assessment (PSA) for fast breeder reactors (FBRs) has been carried out to confirm that the seismic safety is equivalent to that of light water reactors (LWRs). The seismic response on the reactor structure of FBRs causes seismic reactivity. The group motion of fuel assemblies is one of a typical seismic response. Therefore, much attention has been paid on the reactivity insertion mechanism due to the group motion of fuel assemblies and its consequence during the earthquake over the Design Basis Ground Motion (DBGM) condition. When the displacement of each subassembly is moving toward the same direction, each gap reduces coherently and the radial core compaction occurs, which results in positive reactivity insertion. We evaluate the gap reduction characteristics at the mid-plane of core by using a correlation coefficient. As a result, the fuel subassemblies are most concentrated when the input seismic motion of about 5 Hz frequency and 40 m/s² acceleration is applied. The amount of reactivity insertion is estimated approximately 1$ that corresponds to prompt criticality.     

Performance of natural uranium- and thorium-fueled fast breeder reactors (FBRs) for U fissile production

The performance of natural uranium and thorium-fueled fast breeder reactors (FBRs) for producing ²³³U fissile material, which does not exist in nature, is investigated. It is recognized that excess neutrons from FBRs with good neutron economic characteristics can be efficiently used for producing ²³³U. Two distinct metallic fuel pins, one with natural uranium and another with natural thorium, are loaded into a large sodium-cooled FBR. ²³³U and the associated-U isotopes are extracted from the thorium fuel pins. The FBR itself is self-sustained by plutonium produced in the uranium fuel pins. Under the equilibrium state, both uranium and thorium spent fuels are periodically discharged with a certain discharge rate and then separated. All discharged fission products are removed and all discharged actinides are returned to the FBRs except the discharged uranium utilized for fresh fuel of the other thorium-cycled reactors. ²³³U-production rate of the FBRs as a function of both the uranium–thorium fuel pins fraction in the core and the discharge fuel burnup is estimated. The result shows that larger fraction of uranium pins is better for the FBR criticality while larger fraction of thorium fuel pins and lower fuel burnup give higher ²³³U production rate.     

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.     

Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ? MFR ? 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant.     

聚变-裂变混合能源堆一维计算模型燃耗分析

聚变-裂变混合能源堆包括聚变中子源和以天然铀为燃料、水为冷却剂的次临界包层,主要目标是生产电力。利用输运燃耗耦合程序系统MCORGS计算了混合能源堆一维模型的燃耗,给出了中子有效增殖因数k_eff、能量放大倍数M、氚增殖比TBR等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。本文给出的结果可作为混合堆中子输运、燃耗分析程序校验的参考数据,为混合堆概念研究提供了基础数据。     

Application of candle burnup strategy for future nuclear energy utilization

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. If this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means about 40% of natural or depleted uranium can be utilized without either enrichment or reprocessing.

In the ideal nuclear energy utilization system, the radioactive toxicity in the environment should remain or decrease after the utilization. This requirement is very severe and difficult to be satisfied. It may take too much time for its realization. The CANDLE burnup may substitute this period. Though it is a once-through fuel cycle, the discharged fuel burnup is about ten times of the present value for light water reactors. The space necessary for final disposal can be drastically reduced. However, in order to realize such a high burnup of discharged fuels some innovative technologies should be developed. Either new material standing still for such a high burnup or intermediate recladding will be required. Especially new fuel development will take a lot of time. For the time being a small reactor with CANDLE burnup may be a good option for nuclear power generation. Even this kind of reactor requires some innovative technologies and a long period for their developments. For the first stage of CANDLE burnup the prismatic fuel high-temperature gas cooled reactor is preferable. Since the design of this reactor fits to the CANDLE burnup very well, only a little time is required for its research and development.     

Economic feasibility of heat supply from simple and safe nuclear plants

Use of nuclear energy as a heating source is greatly challenged by the economic factor since the nuclear heating reactors have relative small size and often the lower plant load factor. However, use of very simple reactor could be a possible way to economically supply heat. A deep pool reactor (DPR) has been designed for this purpose. The DPR is a novel design of pool type reactor for heat only supply. The reactor core is put in a deep pool. By only putting light static water pressure on the core coolant, the DPR will be able to meet the temperature requirements of heat supply for district heating. The feature of simplicity and safety of DPR makes a decrease of investment cost compared to other reactors for heating only purposes. According to the economical assessments, the capital investment to build a DPR plant is much less than that of a pressurized reactor with pressure vessels. For the DPR with 120 or 200 MW output, it can bear the economical comparison with a usual coal-fired heating plant. Some special means taken in DPR design make an increase of the burn-up level of spent fuel and a decrease of fuel cost. The feasibility studies of DPR in some cities in China show that heating cost using nuclear energy is only one third of that by coal and only one tenth of that by nature gas. Therefore, the DPR nuclear heating system provides an economically attractive solution to satisfy the demands of district heating without contributing to increasing greenhouse gas emissions.     

Analysis of decay heat removal from RBMK-1500 reactor in decommissioning phase by natural circulation of water and air

The main problem in nuclear energy is providing of safety at all stages of lifetime of nuclear installations in conditions of normal operation, accidents and at shutdown. Ignalina NPP, located in Lithuania, is one of the latest with RBMK reactors at highest capacity. Ignalina NPP has two units, both are closed for decommissioning now (in 2004 and 2009). Both units are equipped with RBMK-1500 reactors, the thermal power output is 4200 MW, the electrical power capacity is 1500 MW for each. In RBMK-1500 reactor the fuel assemblies remain for long time inside reactor core after the final shutdown. The paper discusses possibility of heat removal from the RBMK-1500 core at shutdown condition by natural circulation of water (1) and air (2) inside the fuel channels. In first case the decay heat from fuel assemblies is removed due to natural circulation of water and the piping above reactor core should be cooled by means of ventilation in the drum separator compartments. To warrant free access of air in to fuel channels (in the second case) the reactor cooling system should be completely dry out and the pressure headers and the steam discharge valves in steam lines should be opened. If mentioned conditions will be fulfilled, the reactor core will be cooled by natural circulation of water or air and fuel rods remain intact.     

Nuclear Design Study on a Super High-burnup Fuel Assembly for Pressurized Water Reactors Using Transuranium

A conceptual design study was carried out on a super high-burnup mixed-oxide (MOX) fuel assembly (SHB FA) for pressurized water reactors (PWRs) using transuranium (TRU). This study aims to avoid the surplus plutonium (Pu) accumulation and to reduce the accumulation of long-lived radioactive minor actinides (MAS) by utilizing the currently existing PWRs under the condition that the Japanese program to develop fast breeder reactors (FBRs) is tend to delay. For this purpose, an SHB FA with discharged burnup of ?80 GWd/t was investigated by utilizing MAS positively as both burnable absorbers and fissile suppliers and loading high-content Pu. It is possible to load the SHB FAs in a current PWR together with UO₂ FAs and to use 2.5 times as much amount of Pu as that in a standard 1/3 MOX core. Moreover, it is found to be possible to reduce the total number of fresh FAs further from that of a high-burnup (55 GWd/t in maximum) UO₂ (4.9 wt%) core and also to reduce the accumulation of MAS in the nuclear fuel cycle significantly.     

VALMOX: validation of nuclear data for high burn-up MOX fuels

VALMOX, an acronym for validation of nuclear data for high burn-up MOX fuels, is one of the projects of the cluster evolutionary fuel concepts: high burn-up and MOX fuels (EVOL). It covers 30 months, from October 2001 to March 2004.It considers the evaluation of the actinide inventory of MOX fuel at high burn-up (typically 60 GWd/t) in light water reactors, with special attention to the helium production. Calculated values for the spent fuel isotopic masses are compared to the measured ones, with sensitivity analyses made in support. The JEF 2.2 nuclear data file is taken as a basis for calculation. The resulting recommendations on nuclear data should be employed in the preparation and testing of the next JEFF3 file.So far, the major effort was placed on the evaluation of MOX fuel irradiations in pressurised water reactors, and first results will be presented and compared.     

Study on fuel handling system suitable for MA loading of commercial FBR

On the commercial stage of FBR, the improvement of the fuel handling system which directly removes spent fuels from the reactor without in-vessel storage, will be useful to enhance an effective utilization of nuclear resources.

Minor actinides (MA) loading measures such as the heterogeneous loading (two kinds of MA contents-fuel assemblies are loaded to the same reactor) and the homogeneous loading were extracted, and their suitability for the commercial FBR was estimated from standpoints of the fuel handling system cost and the contribution to the smooth fuel recycle flow.

And it was clarified that the heterogeneous MA loading was useful for the transition period from LWRs to FBRs due to the promotion for the Pu utilization and the MA transmutation, and the homogeneous MA loading for the period of only FBRs due to the reduction for the TRU (Pu and MA) inventory outside the reactor.     

Development of next generation PWR in Japan

The demand for energy in Japan is expected to increase steadily into the future, and it seems that the importance of nuclear power generation will be heightened more when the situation of our country which is not rich in energy resources is taken into account.

Furthermore, when we consider the present situation that the light water reactors have become common, recent outlook for the supply and demand for uranium resources, trends in the development of the fast breeder reactor technology, etc., the light water reactors are expected to remain dominant in the nuclear power generation of our country until at least the second half of the 21st century.

Based on such a background five PWR utilities in Japan (Hokkaido, Kansai, Shikoku, Kyushu, and the Japan Atomic Power), and Mitsubishi Heavy Industries, Ltd. have jointly started researching the Next Generation PWR which is expected to be the leading nuclear power plant taking place of APWR.     

Small PWRs using coated particle fuel for district heating, PFPWR50

It has been said that nuclear energy is an important option for especially developing countries to satisfy their increasing energy demand. However, it will be difficult to deploy first of a kind nuclear power plant in developing countries because extensive safety demonstration has to be conducted in industrialized countries. On the other hand, it will be essential to present rigid proof of reliable operational experience to develop proper understanding of the safety features of new reactor systems among the people around the demonstration plant sites. One of the ways to solve the issue is to integrate existing technologies supported by a great deal of data and experience into a new reactor design. Based on the consideration, a small-sized district heating reactor system based on the pressurized water reactor (PWR) technologies combined with the fuel concept of high temperature gas cooled reactors (HTGRs) has been studied. The purpose of the combination of these two existing concepts is to take the best advantages of both excellent operational experience of PWRs and the integrity of HTGR fuel, coated particle fuel, against fission products release even at high temperature. We expect that this approach will help create a breakthrough to the current stagnation of nuclear power deployment.     

Economic viability of small to medium-sized reactors deployed in future European energy markets

Future plans for energy production in the European Union as well as other locations call for a high penetration of renewable technologies (20% by 2020, and higher after 2020). The remaining energy requirements will be met by fossil fuels and nuclear energy. Smaller, less-capital intensive nuclear reactors are emerging as an alternative to fossil fuel and large nuclear systems. Approximately 50 small (<300 MWe) to medium-sized (<700 MWe) reactors (SMRs) concepts are being pursued for use in electricity and cogeneration (combined heat and power) markets. However, many of the SMRs are at the early design stage and full data needed for economic analysis or market assessment is not yet available. Therefore, the purpose of this study is to develop “target cost” estimates for reactors deployed in a range of competitive market situations (electricity prices ranging from 45-150 €/MWh). Parametric analysis was used to develop a cost breakdown for reactors that can compete against future natural gas and coal (with/without carbon capture) and large nuclear systems. Sensitivity analysis was performed to understand the impacts on competitiveness from key cost variables. This study suggests that SMRs may effectively compete in future electricity markets if their capital costs are controlled, favorable financing is obtained, and reactor capacity factors match those of current light water reactors. This methodology can be extended to cogeneration markets supporting a range of process heat applications.     

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.