A new scientific solution for preventing the misuse of reactor-grade plutonium as nuclear explosive

Yield analyses of nuclear explosions and thermal analyses of hypothetical nuclear explosive devices (HNEDs) based on reactor-grade plutonium are examined in a common approach. Three different levels of HNED technology are defined by criteria of geometric dimensions and thermodynamic characteristics of the chemical high-explosive implosion lenses. The results show the content of Pu-238 and the heat it generates in reactor-grade plutonium to be the key parameter. Low-technology HNEDs based on reactor-grade plutonium from spent low-enriched uranium (LEU) or MOX LWR fuel with burnups of 30 GWd/t or more are technically unfeasible. For medium technology, this limit rises to approximately 55 GWd/t burnup. Special cooling applied to such HNEDs would increase these burnup limits still further. Higher Pu-238 contents in reactor-grade plutonium are required to make such HNEDs technically unfeasible. Only for high-technology HNEDs, which could only be built by Nuclear Weapon States (NWSs), the limit to the Pu-238 content of reactor-grade plutonium would rise to approximately 9%.The paper discusses scientific lower limits of alpha-particle heat power or Pu-238 contents above which reactor-grade plutonium can be considered denatured or proliferation-resistant. However, eventually such limits could only be determined by IAEA in agreement with the countries concerned.Such denatured, proliferation-resistant reactor-grade plutonium, which makes reactor-grade plutonium HNEDs technically unfeasible, can be produced by various fuel cycle strategies employing enriched reprocessed uranium (ERU) or minor actinides (MAs). An interim phase of denatured proliferation-resistant plutonium production can be envisioned. A fully proliferation-resistant civil plutonium fuel cycle will become possible later. The use of MAs creates additional proliferation problems. While americium cannot be misused for weapon purposes, neptunium may well be. The neptunium actinide, therefore, must be avoided in an appropriate strategy of a future proliferation-resistant civil nuclear fuel cycle. A fuel cycle strategy of this type is proposed.     

A conceptual design study of proliferation-resistant PWR fuel using reprocessed uranium

A conceptual design study was carried out to enhance proliferation-resistant nature of current light water reactor fuels. Main features of the proliferation-resistant fuel design are adoption of alloy instead of oxide and utilization of enriched reprocessed uranium (10 wt% ²³⁵U). Major dimensions of the fuel assembly were not changed because of thermal-hydraulic considerations and back-fittability to current PWRs. Its smaller ²³⁸U inventory reduces generation of plutonium and ²³⁶U in the reprocessed uranium promotes generation of ²³⁸Pu that has large decay heat. The assembly calculation results of the fuel indicated that the fuel has good proliferation-resistant nature in the viewpoint of decreased plutonium generation, worse plutonium composition and increased decay heat. Neutronic analyses of an equilibrium core loaded with the proliferation-resistant fuels were carried out and calculation results indicate that variations of major core safety parameters are not very large. Therefore, basic feasibility of the proliferation-resistant fuel design using reprocessed uranium was confirmed in the course of this study.     

Hydrodynamic and mass transfer studies of 125 mm centrifugal extractor with 30% TBP/nitric acid system

The Indian nuclear power programme was conceived with a three-stage structure in order to utilize the resources optimally with at most importance to fuel reprocessing for closing the fuel cycle. The first stage of Indian nuclear power programme is based on natural uranium fuelled pressurized heavy water reactors to produce the plutonium (Pu) feed for the second stage. The second stage consists of plutonium fuelled fast breeder reactors to produce U-233 from thorium. The third stage envisages development and deployment of U-233 fuelled reactors. In the fuel cycle operations, solvent extraction is a major step for the recovery of uranium and plutonium. Centrifugal extractors plays a vital role in solvent extraction due to their compact size and high throughput. In the present work, the experimental studies for the hydraulic performance were reported for a single stage annular centrifugal contactor of ϕ125 mm rotor for two-phase flow. Maximum throughput with entrainment less than 1% of one phase to the other, of the centrifugal contactor was measured with the A/O ratio 1 to 6 for the three different bottom vane heights of 6, 8 and 10 mm and also for three different annular gaps of 12, 15 and 18 mm. The centrifugal extractor was operated at different speeds ranging from 1200 to 2200 rpm. In addition, mass transfer performance of the same unit was evaluated with 30% TBP/nitric acid biphasic systems.     

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.     

CANDU堆先进燃料循环的展望

介绍ＣＡＮＤＵ堆的天然铀燃料循环以及最近开发的适合未来近期的先进燃料循环。高中子经济性,不停堆换料以及简单的燃料束设计,使得ＣＡＮＤＵ堆具有非常优良的燃料循环灵活性和多样性。     

Impact on Geologic Repository Usage from Limited Actinide Recycle in Pressurized Light Water Reactors

A project has been conducted as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative to evaluate the impact of limited actinide recycling in light water reactors on the utilization of a geologic repository where loading of the repository is constrained by the decay heat of the emplaced materials. In this study, it was assumed that spent PWR fuel was processed, removing the uranium, plutonium, americium, and neptunium, along with the fission products cesium and strontium. Previous work had demonstrated that these elements were responsible for limiting loading in the repository based on thermal constraints. The plutonium, americium, and neptunium were recycled in a PWR, with process waste and spent recycled fuel being sent to the repository. The cesium and strontium were placed in separate storage for 100–300 years to allow for decay prior to disposal. The study examined the effect of single and mutliple recycles of the recovered plutonium, americium, and neptunium, as well as different processing delay times. The potential benefit to the repository was measured by the increase in utilization of repository space as indicated by the allowable linear loading in the repository drifts (tunnels). The results showed that limited recycling would provide only a small fraction of the benefit that could be achieved with repeated processing and recycling, as is possible in fast neutron reactors.     

Phase characteristics of a U-20Pu-3Am-2Np-15Zr metallic alloy containing rare earths

Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. The current irradiation test series design, designated AFC2, includes minor additions of rare earth elements to simulate expected fission product carry-over from the electrochemical molten salt reprocessing technique. The metal fuel alloys have been fabricated by an arc casting technique. The as-cast fuel alloys have been investigated for phase and thermal properties, specifically, enthalpies of transition, transition temperatures, and room temperature phase characteristics. Results and observations related to these characteristics for the “fresh” fuel alloys are provided. The alloy compositions are based on a U-20Pu-3Am-2Np-15Zr alloy, along with additions of 1 and 1.5 wt% RE (at the expense of U) where RE denotes rare earth alloy of cerium, lanthanum, praseodymium and neodymium. Phase behavior and associated transitions have been compared to available U-Pu-Zr ternary diagrams with acceptable agreement. Enthalpies of transition were deconvoluted from heating and cooling thermal traces for relatively reliable values. The rare earth additions to the base alloy have a minimal influence on the room temperature phases present, but the room temperature phases present slightly impacted the enthalpies of transition and transition temperatures.     

The Spallator and APEX Nuclear Fuel Cycle a New Option for Nuclear Power

A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high energy (1 to 2 GeV) protons on a heavy metal target. The neutrons are absorbed in a surrounding natural uranium or thorium blanket in which fissile Pu-239 or U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high beam current continuous wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of the short-lived fission products external to the fuel cycle eliminates the need for long-term geological age storage of fission product waste.     

钍铀核燃料循环研究

本文报道了中国科学院上海原子核研究所在开展钍铀燃料循环研究方面的进展和取得的成果。这些研究主要为克级量纯~(253)U的提取、钍基燃料后处理技术研究、新的铀钍萃取体系的研究、钍铀镤分离和分析方法研究、中子辐照ThO_2时产生有关核素的累积与中子积分通量和中子能谱的关系、钍的零功率试验等。本文还对钛的利用进行了评估和展望。     

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.     

Cyclic mode of neptunium, americium and curium transmutation in heavy-water reactor

Characteristics of process of transmutation of neptunium, americium and curium from spent nuclear fuel in heavy-water reactor during first 10 lifetimes and at transition to equilibrium mode are calculated. During transmutation, dangerous nuclides, first of all, ²⁴⁴Cm and ²³⁸Pu are accumulated. They cause an increase of radiotoxicity. At first 10 cycles of transmutation, the radiotoxicity is increased by 8.7 times in comparison with radiotoxicity of initial load of transmuted actinides. Heavy-water reactor with thermal power of 1000 MW can transmute neptunium, americium and curium extracted from 3.7 VVER-1000 type reactors. It means, that the required power of transmutation reactor makes about 8% of thermal power of VVER-1000 type reactors.     

Comparison of thorium-based fuels with different fissile components in existing boiling water reactors

With the aim of investigating the technical feasibility of fuelling a conventional BWR (Boiling Water Reactor) with thorium-based fuel, computer simulations were carried out in a 2D infinite lattice model using CASMO-5. Four different fissile components were each homogenously combined with thorium to form mixed oxide pellets: Uranium enriched to 20% U-235 (LEU), plutonium recovered from spent LWR fuel (RGPu), pure U-233 and a mixture of RGPu and uranium recovered from spent thorium-based fuel. Based on these fuel types, four BWR nuclear fuel assembly designs were formed, using a conventional assembly geometry (GE14-N). The fissile content was chosen to give a total energy release equivalent to that of a UOX fuel bundle reaching a discharge burnup of about 55 MWd/kgHM. The radial distribution of fissile material was optimized to achieve low bundle internal radial power peaking. Reactor physical parameters were computed, and the results were compared to those of reference LEU and MOX bundle designs. It was concluded that a viable thorium-based BWR nuclear fuel assembly design, based on any of the fissile components, can be achieved. Neutronic parameters that are essential for reactor safety, like reactivity coefficients and control rod worths, are in most cases similar to those of LEU and MOX fuel. This is also true for the decay heat produced in irradiated fuel. However when Th is mixed with U-233, the void coefficient (calculated in 2D) can be positive under some conditions. It was concluded that it is very difficult to make savings of natural uranium by mixing LEU (20% U-235) homogenously with thorium and that mixing RGPu with thorium leads to more efficient consumption of Pu compared to MOX fuel.     

Radiological and environmental aspects of Th-U fuel cycle facilities

The comparatively higher level of thorium reserves and the absence of long lived actinides of environmental concern offer real advantages for utilization of thorium in nuclear reactors. While use of uranium is likely to continue for some more time in view of investments already made, a shift to thorium eventually is an imperative necessity. It is in fact inevitable for a country like India. The paper presents a detailed comparative analysis of occupational radiation exposures as well as environmental releases. Different stages such as mining, fuel fabrication, reactor operation, spent fuel storage and reprocessing are considered. The factors that need to be taken into account include among others, the relatively lower occupational exposures and environmental releases in sodium cooled fast reactors compared to LWRs, the occurrence of thorium as surface deposits obviating the need for deep mining as in the case of uranium and the special dose reduction measures that need to be devised to minimize occupational exposures due to daughter products of ²³²U present in ²³³U during fuel fabrication operations. If once through mode of fuel cycle is to be adopted, thorium oxide materials are likely to be more enduring than would be the case with uranium.     

What is the potential use of thorium in the future energy production technology?

Today's nuclear technology has principally been based on the use of fissile U-235 and Pu-239. While the natural thorium isotope Th-232 can finally be transformed to a fissile U-233 nucleus following a thermal neutron capture reaction, the existence of thorium in the nature and its potential use in the nuclear technology were not unfortunately into account with a sufficient importance. This was probably because of the geological availability of natural resources of thorium and uranium. Global distributions of thorium and uranium reserves clearly indicate that in general some developed countries such as the USA, Canada, Australia have considerable uranium reserves and contrarily only some developing countries such as Brazil, Turkey, India, Egypt have considerable thorium reserves as being totally about 70 % of the global reserve. All technical parameters obtained from the studies on thorium fuel cycle during the last 50 years indicate that thorium fuel cycle can be used in most of reactor types already operated. In addition, accelerated-driven hybrid systems promise to use the thorium based nuclear fuels. So, thorium will probably be a nuclear material much more valuable than uranium in the future. For this reason, all developing countries having thorium reserves should focus their technological attentions to the evaluation of their national thorium resources like in the case of India. In this paper a brief story on the studies of thorium and its potential use in the future energy production technology have been summarized.     

Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(BB) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/~(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PBB) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.     

Use of weapons plutonium in VVÉR-1000 reactors

One scenario for using excess Russian weapons plutonium is to load it into VVéR-1000 reactors. It is proposed that up to 40% of the fuel assemblies with uranium fuel be replaced with structurally similar fuel assemblies with mixed uranium-plutonium fuel. The stationary regime for burning fuel has the following characteristics: the run time is about 300 or 450 eff. days, the yearly plutonium consumption reaches 450 kg, the neutron-physical characteristics are close to the corresponding regimes with uranium fuel. The nuclear safety criteria and the irradiation dose for workers handling fresh and spent mixed fuel remain within the limits of the normative values. The use of mixed fuel makes it necessary to upgrade certain systems at nuclear power plants. A substantial quantity of weapons plutonium can be loaded every year into VVéR-1000 reactors, effectively using the energy potential of this plutonium. __________ Translated from Atomnaya énergiya, Vol. 103, No. 4, pp. 215–222, October, 2007.     

钍基柱状高温气冷堆不同启动燃料特性初步分析

钍是一种可转换材料,将其转换成²³³U能极大提高现有核燃料资源的储量。为实现对钍的合理利用,以模块式柱状高温气冷堆GT-MHR的燃料组件作为研究对象,选取低浓缩铀、武器级钚、核反应堆级钚等作为其启动燃料。利用栅格输运计算程序DRAGON对这3种启动燃料下的钍基柱状燃料组件的寿期初中子能谱、无限增殖系数、燃耗、转换比以及²³³U和²³²Th的含量等参数进行了分析。结果表明,在易裂变物质初装量约为9%时,与低浓缩铀和武器级钚相比,核反应堆级钚作为启动燃料时组件寿期初中子能谱较硬、转换比较高;其燃耗达90 GW•d/tHM;其无限增殖系数在寿期内的波动最小;燃耗为75 GW•d/tHM时组件中²³³U存余量与²³²Th消耗量之比达0.566。     

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.     

Purex co-processing of spent LWR fuels: flow sheet

Purex co-processing of spent LWR fuel is investigated. In purex co-processing, uranium and plutonium in spent fuel are processed and recovered together as a single stream, while in standard purex reprocessing uranium and plutonium are obtained as separate streams. A two-step (co-decontamination and co-stripping) flow sheet for purex co-processing is devised; concentrations, recoveries and decontamination factors are calculated; and methods to co-convert uranium–plutonium nitrate to mixed oxide are reviewed. A closed nuclear fuel cycle in which at no point uranium and plutonium are separated from each other is reached.     

Burnup simulations and spent fuel characteristics of ZrO2 based inert matrix fuels

Reducing the inventory of long lived isotopes that are contained in spent nuclear fuel is essential for maximizing repository capacity and extending the lifetime of related storage. Because of their non-fertile matrices, inert matrix fuels (IMF’s) could be an ideal vehicle for using light-water reactors to help decrease the inventory of plutonium and other transuranics (neptunium, americium, curium) that are contained within spent uranium oxide fuel (UOX). Quantifying the characteristics of spent IMF is therefore of fundamental importance to determining its effect on repository design and capacity. We consider six ZrO₂ based IMF formulations with different transuranic loadings in a 1-8 IMF to UOX pin-cell arrangement. Burnup calculations are performed using a collision probability model where transport of neutrons through space is modeled using fuel to moderator transport and escape probabilities. The lethargy dependent neutron flux is treated with a high resolution multigroup thermalization method. The results of the reactor physics model are compared to a benchmark case performed with Montebruns and indicate that the approach yields reliable results applicable to high-level analyses of spent fuel isotopics. The data generated show that a fourfold reduction in the radiological and integrated thermal output is achievable in single recycle using IMF, as compared to direct disposal of an energy equivalent spent UOX.