Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the ²³¹Pa–²³²U–²³³U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of ²³²U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.     

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium–plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.     

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.     

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and ²³⁸U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with ²³⁸U alone should be reprocessed in the usual way.     

Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing <Superscript>235</Superscript>U, <Superscript>233</Superscript>U,and <Superscript>232</Superscript>Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of ²³⁵U. It operates in the open-cycle mode involving ²³³U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.     

Physical particularities of nuclear reactors using heavy moderators of neutrons

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using ²³³U as a fissile nuclide and ²³²Th and ²³¹Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.     

Operation of CANDU power reactor in thorium self-sufficient fuel cycle

This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the same as for standard operation in uranium cycle. Two modes of operations are discussed in the paper: mode of preliminary accumulation of ²³³U and mode of operation in self-sufficient cycle. For the mode of accumulation of ²³³U it was assumed for calculations that plutonium can be used as additional fissile material to provide neutrons for ²³³U production. Plutonium was placed in fuel channels, while ²³²Th was located in target channels. Maximum content of ²³³U in target channels was estimated to be ∼13 kg/t of ThO₂. This was achieved by irradiation for six years. The start of the reactor operation in the self-sufficient mode requires ²³³U content to be not less than 12 kg/t. For the mode of operation in self-sufficient cycle, it was assumed that all channels were loaded with identical fuel assemblies containing ThO₂ and certain amount of ²³³U. It is shown that nonuniform distribution of ²³³U in fuel assembly is preferable.      

The concept of the use of recycled uranium for increasing the degree of security of export deliveries of fuel for light-water reactors

The present paper deals with investigation of the possibilities for reducing the risk of proliferation of fissionable materials by means of increasing the degree of protection of fresh fuel intended for light-water reactors against unsanctioned use in the case of withdrawal of a recipient country of deliveries from IAEA safeguards. It is shown that the use of recycled uranium for manufacturing export nuclear fuel makes transfer of nuclear material removed from the fuel assemblies for weapons purposes difficult because of the presence of isotope ²³²U, whose content increases when one attempts to enrich uranium extracted from fresh fuel. In combination with restricted access to technologies for isotope separation by means of establishing international centers for uranium enrichment, this technical measure can significantly reduce the risk of proliferation associated with export deliveries of fuel made of low-enriched uranium. The assessment of a maximum level of contamination of nuclear material being transferred by isotope ²³²U for the given isotope composition of the initial fuel is obtained. The concept of further investigations of the degree of security of export deliveries of fuel assemblies with recycled uranium intended for light-water reactors is suggested.     

The thorium cycle for fast breeder reactors

The role that could be played by liquid metal-cooled fast breeder reactors (LMFBRs) in the utilization of India’s considerable thorium resources is reviewed in this article. Distinct advantages of thorium-based fuels over plutonium-uranium fuels in LMFBRs pertain to a more favourable coolant voiding reactivity coefficient and better fuel element irradiation stability. The poorer breeding capability of thorium-fuelled fast reactors can in principle be overcome by improved core design and development of advanced fuel concepts. The technical feasibility of such advanced thorium fuels and core designs must be established by sustained research and development. It is also necessary to efficiently close the thorium fuel cycle of fast breeder reactors by appropriate development of the fuel reprocessing and refabrication stages. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is expected to be an important tool for development of thorium fuel and fuel cycle technology. A quick look at the economics of the thorium cycle for fast reactors, vis-a-vis the more conventional uranium cycle indicates only a small and acceptable cost disadvantage on account of the need for remote fabrication of recycled thorium fuel. The authors felicitate Prof. D S Kothari on his eightieth birthday and dedicate this paper to him on this occasion.     

Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed

A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the ²³⁶U isotope is compensated by re-enrichment in the ²³⁵U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the ²³²U isotope are obtained.     

Closed Thorium-Uranium-Plutonium Cycle Concept in Nuclear Power Engineering

The use of 232Th instead of 238U as a fertile isotope, 233U instead of 239Pu as the main fissile isotope, heavy water instead of light water as a coolant, and its dilution with light water in the VVER reactor campaign make possible self-enrichment of fuel with fissile isotopes, including the time upon achieving the balanced isotopic abundance ratio of actinides, and also provide conditions for closing the Th-U-Pu fuel cycle. This allows increasing the fuel lifetime by around two orders of magnitude, making it much easier to handle radioactive waste, reducing the nuclear hazard of PWE reactors, and providing a technological barrier to prevent the distribution of fissile materials and nuclear technologies.     

Radiation effect of neutrons produced by D–D side reactions on a D–3He fusion reactor

One of the most important characteristics in D–³He fusion reactors is neutron production via D–D side reactions. The neutrons can activate structural material, degrading them and ultimately converting them into high-level radioactive waste, while it is really costly and difficult to remove them. The neutrons from a fusion reactor could also be used to make weapons-grade nuclear material, rendering such types of fusion reactors a serious proliferation hazard. A related problem is the presence of radioactive elements such as tritium in D–³He plasma, either as fuel for or as products of the nuclear reactions; substantial quantities of radioactive elements would not only pose a general health risk, but tritium in particular would also be another proliferation hazard. The problems of neutron radiation and radioactive element production are especially interconnected because both would result from the D–D side reaction. Therefore, the presentation approach for reducing neutrons via D–D nuclear side reactions in a D–³He fusion reactor is very important. For doing this research, energy losses and neutron power fraction in D–³He fusion reactors are investigated. Calculations show neutrons produced by the D–D nuclear side reaction could be reduced by changing to a more ³He-rich fuel mixture, but then the bremsstrahlung power loss fraction would increase in the D–³He fusion reactor.     

Neutronic simulation of a research reactor core of (232Th, 235U)O 2 fuel using MCNPX2.6 code

The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can be adopted because a high fissile production rate of ²³³U converted from ²³²Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2% enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core.     

Fission decay properties of ultra neutron-rich uranium isotopes

The fission decay of highly neutron-rich uranium isotopes is investigated which shows interesting new features in the barrier properties and neutron emission characteristics in the fission process. ²³³U and ²³⁵U are the nuclei in the actinide region in the beta stability valley which are thermally fissile and have been mainly used in reactors for power generation. The possibility of occurrence of thermally fissile members in the chain of neutron-rich uranium isotopes is examined here. The neutron number N = 162 or 164 has been predicted to be magic in numerous theoretical studies carried out over the years. The series of uranium isotopes around it with N = 154–172 are identified to be thermally fissile on the basis of the fission barrier and neutron separation energy systematics; a manifestation of the close shell nature of N = 162 (or 164). We consider here the thermal neutron fission of a typical representative ²⁴⁹U nucleus in the highly neutron-rich region. Semiempirical study of fission barrier height and width shows that ²⁵⁰U nucleus is stable against spontaneous fission due to increase in barrier width arising out of excess neutrons. On the basis of the calculation of the probability of fragment mass yields and the microscopic study in relativistic mean field theory, this nucleus is shown to undergo exotic decay mode of thermal neutron fission (multi-fragmentation fission) whereby a number of prompt scission neutrons are expected to be simultaneously released along with the two heavy fission fragments. Such properties will have important implications in stellar evolution involving r-process nucleosynthesis.      

核爆聚变电站概念设想

提出了一种新的核爆聚变电站概念设想，并对其可行性作了初步的分析讨论，利用核爆炸实现聚变放能已为氢弹研制的成功所证明，关键是如何把核爆炸能安全地转化成熟能和电能，概念设想通过爆洞，喷钠，选择核装置，铀－钍循环，核燃料回收等措施，合理地解决了能量安全转化，爆洞建造运行、核燃料循环供给等技术困难，使核爆聚变站的设想有可能成为现实。     

铀-钍混合燃料反应堆的可行性分析

分析了以铀为燃料的核电系统的弊端、钍燃料反应堆的理论技术依据和世界范围内钍燃料反应堆的研究状况。提出在我国开发利用钍资源,建立铀．钍混合燃料反应堆具有的独特优势,建议应加大钍资源开发人力物力投入,改变我国核电利用水平落后和钍资源流失之现状。Nuclear energy is a preferred option for electric power generation. The disadvantages of the current uranium-dioxide （UO2 ） fuel in nuclear power were presented and the reactor using the mixed thorium dioxide and uranium dioxide fuel （ ThO2-UO2 ） in the near future was foretold. A proposal to strengthen the research cooperation on the use of the thorium mineral resources in china was put forward.     

Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle

Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.     

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium–Uranium–Plutonium Fuel Cycle

It is shown for a closed thorium–uranium–plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain ~54 kg of fission products, ~0.8 kg of thorium, ~0.10 kg of uranium isotopes, ~0.005 kg of plutonium isotopes, ~0.002 kg of neptunium, and “trace” amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.     

Total reflection X-ray fluorescence and energy-dispersive X-ray fluorescence characterizations of nuclear materials

Nuclear energy is one of the clean options of electricity generation for the betterment of human life. India has an ambitious program for such electricity generation using different types of nuclear reactors. The safe and efficient generation of electricity from these reactors requires quality control of different nuclear materials, e.g. nuclear fuel, structural materials, coolant, moderators etc. These nuclear materials have to undergo strict quality control and should have different specified parameters for their use in nuclear reactors. The concentration of major and trace elements present in these materials should be within specified limits. For such chemical quality control of these materials, major and trace elemental analytical techniques are required. Since some of these materials are radioactive, the ideal chemical characterization techniques should have multielement analytical capability, should require very less sample (micrograms level) for analysis so that the radioactive waste generated, and radiation exposure to the detector and operator are minimum. Total reflection X-ray fluorescence (TXRF) and energy dispersive X-ray fluorescence (EDXRF) with improved features, e.g. application of filters, secondary target and instrumental geometry require very small amount of sample and thus can be suitably used for the characterization of nuclear materials mainly for the determination of elements at trace and major concentration levels. In Fuel Chemistry Division, TXRF analytical methods have been developed for trace element determinations in uranium and thorium oxides, chlorine determination in nuclear fuel and cladding materials, sulphur in uranium, uranium in sea water etc. Similarly, EDXRF analytical methods with radiation filters (to reduce background) and improved sample preapartion techniques, e.g. fusion bead and taking samples in the form of solution on filter papers have been used for developing analytical methods for the determination of U and Th in their mixed matrices, Cd in uranium etc. Some of these studies have been reported in this paper.     

小型模块化熔盐堆钍利用方案

钍基熔盐堆（Thorium Molten Salt Reactor,TMSR）核能系统先导专项的研究目标是研发第四代裂变反应堆核能系统（即钍基熔盐堆）。为充分利用液态燃料熔盐堆的在线添料与在线燃料处理的优势,同时考虑熔盐堆的快速部署,TMSR先导专项部署了小型模块化熔盐堆。考虑燃料处理技术现状及其可能的发展方向,小型模块化熔盐堆钍利用方案采用"三步走"战略。第一阶段采用在线加料与离线处理,实现钍的成规模利用;第二阶段采用在线加料和在线处理（U）与离线处理（MA）的结合,实现钍的高效利用;第三阶段采用在线加料及在线处理全部重金属,实现钍的自持增殖利用。随着"三步走"战略的逐步实施,钍铀燃料循环模式及后处理性能稳步提高,重金属利用率得到明显改善,同时有效降低了卸料毒性。考虑燃料许可容易度和建堆时间,首先为钍利用方案第一阶段布置了三种可能的启堆燃料,分别为低富集铀、低富集铀加钍和233U加钍。计算结果显示:以低富集铀启堆时,燃料循环性能与水堆相当;以233U启堆时,燃料利用率明显高于水堆,且其放射性毒性比水堆低约2个数量级。The missions of the Thorium Molten Salt Reactor (TMSR) Nuclear Energy System are to research and develop the thorium based molten salt reactors (MSR) belonging to the fourth generation of nuclear fission reactor system. A Small modular Molten Salt Reactor (SmMSR) is deployed to make full use of the advantages of online refueling and online reprocessing and to consider the rapid deployment of MSR. An innovative "three-stage" strategy of thorium utilization based on SmMSR is proposed to take the current condition of fuel reprocessing and its future evolution. The first stage can realize the thorium utilization at a large scale with online refueling and off-line processing. The second stage can obtain efficient thorium utilization with online refueling, online processing of uranium and off-line processing of minor actinides (MAs). The third stage is implemented with self-sustaining or breeding mode with online refueling and online processing of all heavy metals. Along with the development of three stages, the utilization of heavy metals will be obviously improved and the radio-toxicity will be significantly reduced. A SmMSR is designed to achieve the goals of the first stage of thorium utilization. And three kinds of nuclear fuel cycles with different startup fuel types (i.e., low enriched uranium (LEU), thorium mixed with LEU (LEU+Th) and thorium mixed with 233U (233U+Th)) are implemented. The results show that the performance for fuel cycle containing LEU is comparable to the pressurized-water reactor (PWR). Meanwhile, the nuclear utilization for that containing 233U is much higher than PWR, and the radio-toxicity for which is lower by ~2 magnitudes than that for PWR.