Isotope separation methods for self-consistent nuclear energy system

It is known that for transmutation of fission products(FPs) in the concept of self-consistent nuclear energy system(SCNES) based on fast neutron reactor it is necessary to apply isotope separation of some FPs to keep neutron balance (to decrease parasitic capture of neutrons by stable isotopes). It is a question whether such FPs isotope separation can be feasible or not within amount of nuclear fission energy production. So it is necessary to consider isotopic content of FPs after fast reactor and to choose energetically appropriate isotope separation method for each radioactive FPs taking into account safe radioactivity level of FPs. In this paper we discuss about isotope separation method for SCNES. Isotopic composition of FPs was calculated using tables of fission yields from ²³⁹Pu fission. It isshown that concentrations of radioactive isotope in the main FPs to be isotopically separated are significant and vary from 2% in ruthenium up to 74% in iodine. We consider new isotope separation methods developed recently such as plasma separation process (PSP) based on selective ion cyclotron resonance heating and atomic vapor laser isotope separation (AVLIS) as a possible candidates. It seems to be energetically profitable to combine various methods to achieve desired separation characteristics. Since the most of FPs have a high initial concentration of radioactive isotope, PSP method seems to be a good candidate for first stages of separation process. We consider the main parts of energy expenditure in one PSP module and its separation characteristics. Estimations of energy consumption in multistage isotope separation process of FPs give maximum value 100keV/fiss. using PSP only and 3MeV/fiss. using AVLIS only. We can significantly decrease these values using AVLIS after PSP when concentration of target isotope in separation cascade will become sufficiently low. We can affirm that energy consumption in isotope separation of FPs is less than 60 MeV of electricity per one fission in nuclear reactor.     

Feasibility study of nitride fuel core and recycle system toward self-consistent nuclear energy system

A concept of nitride core and recycling system in relation to SCNES⁽¹⁾ has been investigated and it has been found that the safety of intact core is enhanced by zero burnup reactivity and negative feedback at ULOF (Unprotected Loss of Flow) and ULOHS (Unprotected Loss of Heat Sink). The self-actuating modules are examined so as to eliminate recriticality of degraded core. The nitride core can employ transuranics-added fuels and burn long life radioactive fission products after isotope-separation of FPs. The enrichment of ¹⁵N and the laser isotope separation seem to need about 2.4% of produced electric energy in the preliminary evaluation.     

Applicability evaluation to a MOX fueled fast breeder reactor for a self-consistent nuclear energy system

The potential for a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in a self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long-lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a two layer radial blanket region without a significant impact on core nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 10² years is as small as that of a typical uranium ore. Regarding self-controllability in the system's safety, the proposed FBR core concept has an inherent negative reactivity feedback with a gas expansion module, sodium plenum above the core and burnup reactivity compensation module. So sodium boiling and fuel melting will be avoided in anticipated transient without scram events. Regarding self-terminability, even if the MOX fuel melting should cause a core compaction process, re-criticality of the core can be avoided by a fuel dilution and relocation module.     

A metallic fuel cycle for Self-Consistent Nuclear Energy System (SCNES)

A fast reactor core and fuel cycle concept has been discussed for Self-Consistent Nuclear Energy System (SCNES) concept. This paper discussed loading material candidates for long-lived fission products (LLFPs) and removal of stable nuclides from radioactive nuclides with isotope separation using tunable laser. Some of LLFPs were possible to be loaded in metal of the generated form. The potential for LLFP-confinement in the reactor system is discussed along with a metallic fuel cycle concept. The proposed fuel cycle scheme is a successful candidate for SCNES concept.     

HYPER (Hybrid Power Extraction Reactor): A system for clean nuclear energy

The Korea Atomic Energy Research Institute (KAERI) has been performing accelerator driven system related research and development (RID) called HYPER (HYbrid Power Extraction Reactor) for the transmutation of nuclear waste and energy production through the transmutation process. HYPER program is within the frame work of the national mid and long-term nuclear research plan. KAERI is aiming to develop the elemental technologies for the subcritical transmutation system by the year of 2001 and build a small bench scale test facility (5 MW) by the year of 2006. Some major features of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with a minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Lead–bismuth (Pb–Bi) is adopted as a coolant and spallation target material. 1 GeV 16 mA proton beam is designed to be provided for HYPER. HYPER is to transmute about 380 kg of TRU a year and produce 1000 MW of power. The support ratio of HYPER for LWR units producing the same power is believed to be 56.     

Transport of nuclear heat by means of chemical energy (nuclear long-distance energy)

Nuclear long-distance energy, i.e. the transportation of chemically bound energy, represents a potential application for process heat plants in which the endothermic reaction takes place at the heat source (high temperature reactor) whereas the exothermic back reaction occurs at the region of heat utilization (consumer). Due to the following criteria, i.e. reversibility of the chemical reaction, sufficiently large reaction enthalpy, favourable temperature region for the forward and back reactions, and the available technology, a combination of the methods of endothermic steam reforming of methane and exothermic methanation is chosen. As well as supplying household and industrial consumers with heating, process steam and electrical energy, an interconnected system with synthesis gas consumers (e.g. methanol production and iron ore reduction plants) is possible. It is shown that the amount of reactor heat which is convertible into long-distance energy depends considerably on the helium temperatures in the high temperature reactor and lies between 60 and 73% of the reactor power. Conceivable circuit schemes for the nuclear steam-reforming plants and the methanation plants are described. Finally, it is demonstrated, with the help of a simple model for cost estimations, that the nuclear long-distance energy system can make heating for households available in competition with oil heating and that due to the lower specific transport costs, for distances larger than 50 km it is also more economical than the hot water supply from the thermal power coupling of steam turbine plants using light water reactors (LWRs) or high temperature reactors (HTRs).     

The feasibility study on perfect burning reactor system

The system of 100% natural uranium burning with once-through fuel cycle is defined as the Perfect Burning Reactor System (PBRS). This kind of nuclear system can be expected to have some good characteristic such as resource efficiency, radiotoxicity reduction, proliferation and nuclear safety. Therefore, the feasibility of the concept is studied in this paper. The preliminary results show that the system of 100% natural uranium burning with once-through fuel cycle is physically possible with a plenty supply of external neutron, and that the system demands no activities concerning with fuel cycle such as uranium enrichment, fuel fabrication, spent fuel reprocessing and radioactive waste treatment. The study also quantitatively clarifies the external neutron source strength, the nuclear criticality safety, the demanded accelerator performance and the energy balance. In addition, the more precise analysis is requested for well understanding and improving the characteristic and economical rationality of the system.     

The fusion-supported decentralized nuclear energy system

A decentralized nuclear energy system is proposed comprising mass-produced pressurized water reactors in the size range 10 to 300 MW (thermal), to be used for the production of process heat, space heat, and electricity in applications where petroleum and natural gas are presently used. Special attention is given to maximizing the refueling interval with no interim batch shuffling in order to minimize fuel transport, reactor downtime, and opportunity for fissile diversion. The smallest reactors could be deployed as nuclear batteries, kept in the equivalent of spent-fuel shipping casks and returned to nuclear fuel centers for refueling. These objectives demand a substantial fissile enrichment (7 to 15%). The preferred fissile fuel is U-233, which offers an order of magnitude savings in ore requirements (compared with U-235 fuel), and whose higher conversion ratio in thermal reactors serves to extend the period of useful reactivity and relieve demand on the fissile breeding plants (compared with Pu-239 fuel). Application of the neutral-beam-driven tokamak fusion-neutron source to a U-233 breeding pilot plant is examined. This scheme can be extended in part to a decentralized fusion energy system, wherein remotely located large fusion reactors supply excess tritium to a distributed system of relatively smallnonbreeding D-T reactors.     

A study on physicochemical properties of epoxy coating system for nuclear power plants

In this work, the thermal properties of epoxy coating system on the liner plate in the containment structure have been investigated by irradiation dose rate and design basis accident (DBA) conditions. Also, the effect of immersion in hot water on adhesion strengths of epoxy coating system has been studied. The glass transition temperature (T_g) and thermal stability of the epoxy coating system after DBA tests were measured by differential scanning calorimeter (DSC) and thermogravimetric analyser (TGA) analyses, respectively. Contact angle measurements were used to determine the effect of immersion on the surface energetics of epoxy coating system, including surface free energy and work of adhesion. Adhesion tests were also executed to evaluate the adhesion strength at interfaces between carbon steel plate and epoxy resins. As a result, the DBA test led to the improvement of the internal structure in cured epoxy systems, resulting in significantly increasing the thermal stability, as well as the T_g. Also, the immersion in hot water had a role in the post curing of epoxy resins and increased the mechanical interlocking of the network system, resulting in increasing the adhesion strengths of the epoxy coating system.     

An approach for evaluating fuel system applied in innovative nuclear energy system

One of the important issues in the study of Innovative Nuclear Energy Systems (INES) is the integrity of the fuel system applied. An approach of evaluating fuel system integrity is discussed here based on the procedure currently used in the integrity evaluation of fast reactor fuel. The fuel failure modes controlling fuel life were reviewed and fuel integrity was analyzed and compared with the failure criteria.Metal and nitride fuels with austenitic and ferritic cladding tubes were examined in this study. For the purpose of representative irradiation behavior analyses of the fuel for INES, the correlations of the cladding characteristics were modeled based on well-known characteristics of austenitic modified 316 SS (PNC316), ferritic-martensitic steel (PNC-FMS) and oxide dispersion strengthen steel (PNC-ODS).The analytical result showed that fuel lifetime is limited by channel fracture which is a nonductile type (brittle) failure associated with a high level of irradiation-induced swelling in the case of austenitic steel cladding. In the case of ferritic steel, on the other hand, the fuel life is controlled by cladding creep rupture. The lifetime evaluated here is no more than 200 GWd/t, which is still lower than the target value 400 GWd/t burnup. Possible measures to extend metal fuel lifetime may be reducing fuel smear density and ventilating fission gas in the plenum.     

ECRH for JET: A feasibility study

For JET to fulfil its mission in preparing ITER operation, the installation of an electron cyclotron resonance heating system on JET would be desirable. The study described in this paper has investigated the feasibility of installing such a system on JET. The principal goals of such a system are: current drive over a range of radii for NTM stabilization, sawtooth control and current profile tailoring and central electron heating to equilibrate electron and ion temperatures in high performance discharges. The study concluded that a 12 gyrotron, 10 MW, system at the ITER frequency (170 GHz) adapted for fields of 2.7–3.3 T would be appropriate for the operation planned in JET. An antenna allowing toroidal and poloidal steering over a wide range is being designed, using the ITER upper launcher steering mechanism. The use of ITER diamond windows and transmission line technology is suggested while power supply solutions partially reusing existing JET power supplies are proposed. Detailed planning shows that such a system can be operational in about 5 years from the time that the decision to proceed is taken. The cost and required manpower associated with implementing such a system on JET has also been estimated.     

A design study on Pb-208 cooled compact CANDLE burning reactor for future nuclear energy supply

A compact pool-type Pb-208 cooled CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) with a thermal power rating of 125 MWth is considered for the future nuclear energy supply. Natural Pb consists of Pb-204, Pb-206, Pb-207 and Pb-208. Pb-208 has a small capture and inelastic-scattering cross-section, which makes it possible to reduce neutron capture by coolant and to make neutron spectrum harder. In case of Pb-208 coolant instead of natural Pb, the core height and radius are reduced to 1.5 m and 1 m, respectively. The effective multiplication factor of the core, k_eff, could be increased from k_eff = 0.984 of natural Pb up to k_eff = 1.006. For increasing natural circulation head, coolant velocities in each core zone are adjusted by orifice at the core inlet position. The reactor vessel height is equal to that of a typical loop-type demonstration FBR vessel to obtain natural circulation head.     

模糊控制在核子秤系统中的应用研究

讨论了模糊控制在核子秤系统中的应用,表明它在非线性过程、弛豫时间长以及无法用数学方法精确描述的过程控制中具有很大优点,并给出了模糊变量、隶属度函数以及模糊控制规则。最后比较了模糊控制与PID控制的噪声扰动。     

Challenge of nuclear data for innovative nuclear energy

The article devoted to assessment of present-day demand to nuclear data for transmutation problem, including the discussion of required accuracies, status and perspectives of nuclear data evaluation and development of nuclear models. The effect of nuclear data uncertainties on radiation damage of structural materials is discussed. An analysis of ISTC projects related to nuclear data measurement and evaluation is presented. The recommendations for differential, integral experiments and recommendations on the evaluated data preparation are presented.     

Requirements for nuclear energy in the 21st century nuclear energy as a sustainable energy source

Nuclear energy must compete against other energy technologies in the 21st century. It must be economical and it must be proven that it fulfills the conditions for sutainability. This means that the requirements of — no short term depletion of resources — extremely low emission of noxious or radioactive substances to the environment — extremely low release of radioactivity from a nuclear plant in case of the most severe accidents and — the present very long term problem of high active waste must be transformed into a few hundred years problem through destruction of plutonium, transmutation of the minor actinides and the most important very long lived fission products.     

AMBIDEXTER nuclear energy complex: a practicable approach for rekindling nuclear energy application

Aiming at one of the decisive alternatives for long-term perspectives of the nuclear power, an integral and closed nuclear energy system concept is proposed; namely, the Advanced Molten-salt Break-even Inherently-safe Dual-missioning Experimental and Test Reactor (AMBIDEXTER) nuclear energy complex. This essentially comprises two mutually independent circuits of the radiation/material transport and the heat/energy conversion, centered at the integral reactor assembly, which enables one to utilize maximum benefits of nuclear energy under minimum risks of nuclear radiation. The entire reactor system resides in a thin and large Hastelloy vessel, the internal part of which is divided into a number of equipment compartments with neither connection pipings nor active valves necessary. As the reactor operates at very low FP inventory throughout its designed lifetime and there is no primary heat transport pipings outside the reactor vessel, significant release of radioactive materials due to any equipment failure should be incredible. The nuclear-thermalhydraulic characteristics of the molten ThF₄–²³³UF₄ fuel salt extend the self-sustainability of the AMBIDEXTER fuel cycle to enhance the resource security and safeguard transparency. While maintaining the break-even conversion ratio criterion, a flexible fuel management strategy using a certain choice of denaturants should improve its own proliferation-resistance characteristics. As the core technologies associated with developing the AMBIDEXTER concept are mostly available in commercialized forms at present, investigating the integral performance of the concept should be the prime research topic in ongoing 250 MW_th prototype design studies.     

聚变中子源驱动的次临界清洁核能系统─聚变能技术的早期应用途径

提出作为聚变能技术早期应用途径的聚变中子源驱动的清洁核能系统概念,并从国家的能源需求、国内外核电发展状况论述开发这种系统的必要性和意义,根据国内外聚变驱动器技术及次临界包层技术进展和国内多年的可行性研究结果,说明开发这种系统的现实性和基础。文中也给出了建议的发展进程。