Waste disposal aspects of the long term cooling characteristics of irradiated nuclear fuels

A study has been made of the long term cooling characteristics of nuclear fuels irradiated in commercial reactor designs of interest within the U.K. In the case of thermal reactors, Magnox, AGR, SGHWR, LWR and HTR systems fuelled with either natural or enriched uranium are considered, together with a fast reactor fuelled with plutonium derived either from the Magnox or the AGR programmes. Alternative uses for Magnox plutonium are considered by simulating a plutonium fuelled HTR thermal system and the development of a $\text{Th}{}^{233}\text{U}$ fuel cycle has been anticipated for both a fast reactor and an HTR.For each system the activities as a function of cooling time are considered on the assumption of U/Pu recovery from the fuel during reprocessing within a year of discharge from the reactor and for the alternative case of no U/Pu extraction. The reprocessing waste products associated with the various fuel cycles have then been compared both on the basis of decay heating and radiological hazard per GW(e) yr. Finally, recycling of transplutonium elements is also considered with a view to reducing the long term heating commitment from the higher actinides.     

Pu的利用和处置

从核弹头拆卸出的武器级Ｐｕ以及核电动力堆产生的反应堆级Ｐｕ的贮存、利用和处置是全世界关注的问题。已提出的处理方案可归纳为两类：一类是将Ｐｕ作为能源利用；另一类是将其作为废物永久性处置。从能源利用、环境影响、经济效益、技术成熟程度和防止核扩散等方面对处理方案进行了综合比较，初步认为，全堆芯混合氧化物燃料的先进轻水堆是利用Ｐｕ的最佳方案，而与高放废物一起玻璃固化后永久性贮存是处置Ｐｕ的最有希望的措施。     

Evaluation of uranium thorium and plutonium thorium fuel cycles in a very high temperature hybrid system

In recent times, there is a renewed and additional interest in thorium because of its interesting benefits. Thorium fuel cycle is an attractive way to produce long term nuclear energy with low radiotoxicity waste. In addition, the transition to thorium could be done through the incineration of weapons grade plutonium or civilian plutonium. Th-based fuel cycles have intrinsic proliferation-resistance and thorium is 3–4 times more abundant than uranium. Therefore, thorium fuels can complement uranium fuels and ensure long term sustainability of nuclear power.In this paper, the main advantages of the use of fuel cycles based on uranium-thorium and plutonium-thorium fuel mixtures are evaluated in a hybrid system to reach the deep burn of the fuel. To reach this goal, the preliminary conceptual design of a hybrid system composed of a critical reactor and two Accelerated Driven Systems, of the type of very high temperature pebble-bed systems, moderated by graphite and cooled by gas, is analyzed.Uranium-thorium and plutonium-thorium once-through and two stages fuel cycles are evaluated. Several parameters describing fuel behaviour and minor actinide stockpile are compared for the analyzed cycles.     

Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO₂. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.     

Inherent Protection of Plutonium by Doping Minor Actinide in Thermal Neutron Spectra

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% ²³⁵U and 20% ²³⁵U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, ²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% ²³⁵U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties.     

Categorisation and priorities for future research on inert matrix fuels: An extended synthesis of the panel discussions

There are three categories of basic fuel cycle needs, which are being addressed by the different types of inert matrix fuel (IMF) concepts currently under development. These are: plutonium burning in existing LWRs, plutonium burning in fast reactors and minor actinide transmutation — corresponding to three distinct timescales for perceived IMF implementation, viz. short, medium and long term, respectively. The current paper, based partly on the two panel discussions organised at the 6^th IMF workshop, presents viewpoints and priorities for each of the three categories of IMF applications, both in terms of the fuel concepts to be pursued and the corresponding R&D requirements.     

Plutonium management and Generation IV systems

This paper discusses the potential role of Generation IV nuclear energy systems in managing plutonium. It briefly reviews the Generation IV goals and their relevance to plutonium management. Each of the six selected Generation IV systems [very high temperature reactor (VHTR), gas-cooled fast reactor (GFR), sodium-cooled fast reactor (SFR), super-critical-water-cooled reactor (SCWR), lead-cooled fast reactor (LFR), molten salt reactor (MSR)] is briefly discussed. The main characteristics of each system are summarised and the capability for plutonium management indicated. The potential for the management of plutonium using Generation IV systems is briefly reviewed from a complete fuel cycle perspective to illustrate the issues in the context of a fleet of reactor and fuel cycle facilities.     

Proliferation-resistant fuel options for thermal and fast reactors avoiding neptunium production

Sustainable nuclear energy production requires reuse of spent nuclear fuel while avoiding its misuse. In the paper we assume that plutonium with sufficiently high content of the Pu-238 isotope (about 6% or more) and americium from spent nuclear fuel are proliferation-resistant. On the other hand, neptunium should be considered as material that is fissionable in a fast neutron spectrum and could be misused.We also assume that plutonium denatured by Pu-238 can be produced in nuclear reactors of, e.g. nuclear weapon states and used for fuel fabrication there or in multilateral reprocessing and re-fabrication centers as suggested by IAEA. Then the fabricated fuel can be utilized in nuclear reactors everywhere provided that the reactors may operate safely and the fuel remains proliferation-resistant after utilization. Options to meet these criteria are investigated in the paper for two reactor types: pressurized water reactors (PWRs) and fast reactors (FRs).In PWRs, the investigated fresh fuel compositions include denatured plutonium and depleted uranium mixed with a small amount of U-233, thorium and, optionally, with americium, presence of U-233 making the coolant void effect negative. In FRs, use of americium makes plutonium denatured, both for the burner (without fertile blanket) and breeder options. It is shown that the proposed design and fuel options are proliferation-resistant, the generation of neptunium being very low. Safety parameters are acceptable. Advanced aqueous or pyrochemical reprocessing for plutonium/thorium/uranium fuel and related fuel re-fabrication technology applying remote handling may become necessary to realize the considered fuel cycles.     

Basic Analysis on Isotopic Barrier of Material Attractiveness Based on Plutonium Composition of FBR

Fuel behaviors of the large fast breeder reactor have been investigated, as well as material attractiveness based on isotopic plutonium composition for evaluating proliferation resistance with regards to a combined evaluation of decay heat and spontaneous fission neutron barrier as key parameters of isotopic material barrier. Trans-uranium fuel (TRU) (MA + U-Pu) in the core regions and MA doping (MA + natural U) in the blanket regions as options of MA loading produce a higher Pu-238 composition for denaturing plutonium, which mainly comes from converted Np-237. The isotopic plutonium composition of TRU fuel is relatively less than the Pu composition of MOX fuel except for the Pu-238 composition that is higher than that of MOX fuel. MA in the core or blanket regions, which produces a higher Pu-238 composition, plays a key role in obtaining a high-level material barrier of decay heat and spontaneous fission neutron compositions. The material attractiveness level of plutonium composition in the core regions can be categorized as practically unusable and its level becomes less by adopting TRU fuel. In addition, the material attractiveness level in the blanket regions as being practically unusable can be reached from weapon grade by loading MA at a 2% doping rate.     

A view of strategies for transmutation of actinides

Developments in the field of recycling and transmutation of actinides are discussed. Three general strategies are discriminated: (i) an evolutionary strategy based on gradual implementation of partitioning and transmutation techniques in the fuel cycle; (ii) a radical strategy based on implementation of partitioning and transmutation in the fuel cycle, once all steps of this technology are proven; (iii) plutonium incineration, based on the conversion, with existing reactor types, of separated plutonium into a spent fuel form that is suited for direct storage.     

Advanced fast reactor development requirements: Is there any need for HEU?

The present paper reviews R&D needs foreseen for advanced fast reactor development. The focus is on physics, safety and fuel validation needs using experimental facilities, i.e. critical zero power facilities or irradiation reactors. This review underlines the crucial need for a continuous availability of such facilities in the next decades. As for the fuel to be used in the experiments the need is for plutonium fuelled experiments and no specific need for HEU seems to be foreseen. However, the already existing stocks of HEU at some critical facilities like MASURCA in France will be certainly needed and should be kept in order to allow the required flexibility to perform critical experiments of large size, representative of most future advanced systems.     

球床式高温气冷堆MOX燃料循环优化

与压水堆相比,球床式高温气冷堆能在堆芯结构不做明显改变的情况下采用全堆芯装载混合氧化物（MOX）燃料元件。基于250 MW球床模块式高温气冷堆堆芯结构,设计了4种球床式高温气冷堆下MOX燃料循环方式,包括铀钚混合的燃料球和独立的钚球与铀球混合装载的等效方式,采用高温气冷堆设计程序VSOP进行分析,比较了初装堆的有效增殖因数、燃料元件在堆芯内滞留时间、卸料燃耗、温度系数等主要物理特性。结果表明：采用纯铀和纯钚两种分离燃料球且铀燃料球循环时间更长的方案,平均卸料燃耗较高,总体性能较其他循环方式优越。     

Non-proliferation attributes of molten salt reactors

Molten salt reactors (MSR) have many non-proliferation attributes. They can operate on the thorium-uranium fuel cycle which protects the fissile material by the daughter products of the inseparable U-232. MSRs can completely fission all plutonium and HEU, and as desired, ‘convert’ them to U-233. This also results in high, and efficient resource utilization, while diminishing the plutonium stock. On line processing, when applied, could free the waste from all fissile material. The fuel in the reactor stays protected by the intense radiation of the fission products. Fuel can also be protected in the reactor as well as outside the reactor by denaturing with natural uranium. A wide variety of MSRs are available, from ‘once through’ minimum processing reactors to ones with fuel processing which can breed fuel for converters. MSRs are extremely safe and simple reactors with good economic potential.     

Plutonium breeding of light water cooled fast reactors

Light water cooled fast reactor with new fuel assemblies (FA) has been studied for high breeding of fissile plutonium. It achieves fissile plutonium surviving ratio (FPSR) of 1.342 (discharge/loading), 1.013 end and beginning of equilibrium cycle (EOEC/BOEC), and compound system doubling time (CSDT) of 95.9 years at the average coolant density of pressurized water reactor (PWR). It is further improved for reduced moderation boiling water reactor (BWR) (RMWR) coolant density. Fissile plutonium surviving ratio reaches 1.397 (discharge/loading), 1.030 (EOEC/BOEC) and CSDT is 37 years. The present study has shown the possibility of breeding at the PWR coolant density and meeting the growth rate of energy demand of advanced countries at the RMWR and Super FR coolant density for the first time. The new FA consist of closely packed fuel rods. The integrity of welding of fuel rods at the top and bottom ends is maintained as the conventional fuel rods. The coolant to fuel volume fraction is reduced to 0.085, one-sixth of that of RMWR. The volume fraction remains unchanged with the diameter of the fuel rod. The thermal hydraulic design of the cores remains for the future study.     

Burning of Reactor Grade Plutonium Mixed with Thorium in a Hybrid Reactor

Significant amount of plutonium have been discharged and accumulated from the conventional LWRs and CANDU reactors. Reducing this reactor grade (RG) plutonium is very important because it may be misused and/or released accidentally into the environment. Fusion-fission (hybrid) reactors have strong potential on burning plutonium effectively. This study presents the burning of RG plutonium mixed with thorium in a hybrid reactor for an operation period of 24 months. The effect of various fuel mixtures (98% ThO₂ + 2% RG-PuO₂, 94% ThO₂ + 6% RG-PuO₂ and 90% ThO₂ + 10% RG-PuO₂) and coolants (Flinabe, natural lithium and Li₂₀Sn₈₀) on the reactor’s performance was investigated. Numerical results showed that utilization of RG plutonium in the mixed fuel in such a hybrid reactor not only enhanced the reactor’s performance but also reduced its ²³⁹Pu content significantly.     

Concept of innovative water reactor for flexible fuel cycle (FLWR)

In order to ensure sustainable energy supply in the future based on the matured light water reactor (LWR) and coming mixed oxide (MOX)-LWR technologies, a concept of innovative water reactor for flexible fuel cycle (FLWR) has been investigated in Japan Atomic Energy Research Agency (JAEA). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming MOX-LWR technologies without significant technical gaps. The second part represents the reduced-moderation water reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-developed LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the future fuel cycle circumstances during the reactor operation period around 60 years.At present, reprocessed plutonium from the LWR spent fuel is to be utilized in MOX-LWR. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or MOX-LWR. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the MOX-LWR utilization. When the fuel cycle for plutonium multiple recycling with MOX fuel reprocessing is realized, the fuel assembly will be replaced with another type of the tight-lattice one for RMWR with different rod diameter, rod gap width and so forth even in the same reactor system, being flexibly corresponding to the fuel cycle circumstances.Investigation on the core for both the parts of the FLWR concepts has been performed, including the core conceptual design, the core characteristics under Pu multiple recycling, the thermal hydraulic investigation in the tight-lattice core, and so forth. Up to the present, promising results have been obtained.     

Status of the small modular fluidized bed light water nuclear reactor concept

The state of the art of a small modular reactor concept with a suspended core is presented. The reactor design is based on a fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident conditions. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another option is to modify the shape of the reactor core to produce a non-fluctuating bed and, consequently, guarantee the dynamic stability of the reactor. The mixing of tantalum in the fuel is also proposed as an additional inhibition to the power excursion. The spent fuel pellets may not be considered nuclear waste, since they are of a shape and size that can easily be used as a source of radiation for food irradiation and industrial applications. The reactor can easily operate as a plutonium burner or can operate with a thorium fuel cycle.     

Optimum utilization of nuclear fuel with gas and vapor core reactors

Gas and Vapor Core Reactors (G/VCR) are externally reflected and moderated nuclear energy systems fueled by stable uranium compound in gaseous or vapor phase. In G/VCR systems the functions of fuel and coolant are combined and the reactor outlet temperature is not constrained by solid fuel-cladding temperature limitations. G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Furthermore, G/VCR systems feature a low inventory and fully integrated fuel cycle with exceptional sustainability and safety characteristics. With respect to fuel utilization, there is practically no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from almost solely a burner to a breeder is achievable. The continuous recycling of fuel in G/VCR systems allows for continuous burning and transmutation of actinides without removing and reprocessing of the fuel. The only waste product at the backend of the gas core reactors' fuel cycle is fission fragments that are continuously separated from the fuel. As a result the G/VCR systems do not require spent fuel storage or reprocessing.

G/VCR systems also feature outstanding proliferation resistance characteristics and minimum vulnerability to external threats. Even for comparable spectral characteristic, gas core reactors produce fissile plutonium two orders of magnitude less than Light Water Reactors (LWRs). In addition, the continuous transmutation and burning of actinides further reduces the quality of the fissile plutonium inventory. The low fuel inventory (about two orders of magnitude lower than LWRs for the same power generation level) combined with continuous burning of actinides, significantly reduces the need for emergency planning and the vulnerability to external threats. Low fuel inventory, low fuel heat content, and online separation of fission fragments are among the key constituent safety features of G/VCR systems.     

The status of development of small and medium sized reactors

Several IAEA Member States have shown their interest in reactor designs, having a smaller power rating [100–500 MW(e) range] than those generally available on the international market. These small and medium sized power reactors are of interest either for domestic applications or for export into countries with less developed infrastructure. There are different developments undertaken for these power reactors to be ready for offering in the nineties and beyond.The paper gives an overview about the status and different trends in IAEA Member States in the development of small and medium sized reactors for the 90's and provides an outlook for very new reactor designs as a long term option for nuclear power.     

Pu-breeding feasibility in PWR

This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO₂ or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived.