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.     

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。     

加速增产核燃料的天然安全“核热泉”快中子增殖堆

为保证21世纪中国经济的持续稳定地高速增长,必须充分发挥核能的巨大潜力,使之配合其他可再生能源同步增长,及早大规模替代煤炭等化石能源。由于目前国内大量兴建的核电站以压水堆为主,需要消费大量天然铀资源,倚靠廉价铀供应难于维持长期增长,必须依靠快中子增殖生产人造裂变燃料——钚,才能摆脱天然铀原料短缺的束缚。然而,传统的快中子增殖堆的核燃料增产速度较慢,难于配合中国核电的高速增长。本文介绍一种先进快中子增殖堆(AFBR)方案,其中利用在线连续换料的空心球形燃料元件,依靠载热剂的出入口之间的温度差实现满功率自然循环,可以成倍地提高燃料比功率与核燃料增殖速度。本快中子增殖堆改进了俄罗斯称为"天然安全"的BREST铅冷快堆设计方案,成为无须人为控制的"核热泉",它能在不设置加压泵及高位铅池的情况下,自动按外部负荷需要供应必要的热量,完全依靠自然循环将全部裂变热能及停堆后堆芯余热散出,不至对环境产生放射性污染。     

热表面电离质谱法测定钚氧化物中的钚同位素组成

采用热表面电离质谱法对钚氧化物中钚同位素丰度进行了测定。通过对钚氧化物样品预处理、离子源和分析器的真空控制、法拉第杯接收效率检测、测量过程中的信号强度大小控制、信号强度稳定性控制以及测量时间的控制等条件进行优化,确定了最佳预处理条件和测量条件,实现了钚氧化物中钚同位素组分的准确测定。在选定的条件下,测定了钚标准样品中的钚同位素丰度,主同位素239Pu和242Pu测量精密度(sr)均优于0.05%(n=6)。     

钚年龄评估技术

在军控核查技术中应用钚的年龄属性对核弹头认证和“禁产公约”监督具有非常重要的意义。根据^241Pu的不同衰变模式，对母核和子核衰变伴随的γ辐射随时间的变化、各种情况下子核与母核的原子比随时间的变化规律等进行了描述。研究表明，通过某种手段获得^241Pu和^241Am的原子比或测量γ辐射可确定钚的年龄。     

气体采样滤膜中钚的激光诱导击穿光谱定量分析研究

在核设施排放物的监测中,为确保环境安全及实现核材料衡算,需针对通风管气体采样滤膜中钚的含量进行分析。为探索利用激光诱导击穿光谱（LIBS）技术定量分析滤膜中钚元素含量的可行性,开展了某核设施热室内含钚气体采样滤膜的LIBS分析研究。通过分析含钚气体采样滤膜中钚的LIBS特征光谱信号,筛选出钚的有效发射谱线,并优化LIBS测量延迟时间参数,最终得到了浓度范围较低情况下的定量标准曲线,并进行了技术验证分析。研究结果显示,针对滤膜中的钚元素可采用有效特征谱线进行LIBS定量分析,能用于定量分析钚的可分析有效特征谱线为PuⅡ 443.298 nm、PuⅠ 460.721 nm、PuⅡ 466.389 nm。分析结果表明,LIBS分析结果的精度好于3%。     

乏燃料后处理厂草酸钚沉淀器临界安全分析研究

本文对乏燃料后处理厂中钚尾端工艺环节的关键设备草酸钚沉淀器进行了临界控制方法和参数的详细分析。针对连续沉淀器的工艺和结构特点,对易裂变物质的状态进行了一系列分析,比较了均匀溶液和悬浮颗粒溶液反应性的差别。对单个沉淀器和多个沉淀器并行工作的情况分别进行了临界安全分析,并分别研究了不含中子毒物、布置中子毒物层以及布置中子毒物棒等情况下能达到的最大处理能力。选取了临界安全基准实验国际评价中的相似实验方案进行了验证计算,分析了所用程序计算此类问题的不确定度。本文开展的临界安全分析研究总结了连续沉淀器临界安全控制的规律性结论,可为后续连续沉淀器的工艺设计及今后的工程应用提供参考。     

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.     

Nuclear mechanism for TRU burning oxide fueled core in Advanced Recycling Reactor

This paper presents an approximation approach to predict the core characteristics based on parametric survey and an analysis of nuclear mechanism in a conceptual nuclear design for enhanced transuranics (TRU) burning mixed oxide fueled and sodium cooled fast reactor which can be realized in the near future. The design study of Advanced Recycling Reactor was conducted in the context of the program for the industry in Global Nuclear Energy Partnership initiatives, including a core in the first plant for demonstration and cores with enhanced TRU burning capability for the future plants. Both concepts for the first plant; low core height and large volume fraction of structure are deployed, seeking small TRU conversion ratio (CR)^* and small void reactivity which are crucial in the design, but different approaches. In this paper, the TRU CR and the sodium void reactivity have been approximated with a single equation in these concepts, based on the theoretical formula related to the chain reaction in the reactor and the calculation results. Shortening the core height and increasing the structure volume fraction will enhance TRU enrichment through increased neutron leakage and capture, which will reduce a ratio of U-238 to sum of Pu-239 and Pu-241 so that TRU CR decreases from 0.78 to 0.57. A small ratio of sodium loss to plutonium fissile will be effective also in the reduction of positive reactivity effect by spectral hardening. On the other hand, when this ratio and geometrical buckling of flux reduce, negative contribution by the neutron leakage becomes small. Theses relations related to the positive void reactivity have been formularized by the approximation with few parameters within several percent respectively as well as the TRU CR, indicating that one of dominating parameters is the ratio of sodium loss to plutonium fissile in the void reactivity at large fast reactors. * = (1 − net loss of TRU/loss of heavy metal).     

Proposal of a plutonium burner system based on HTGR with high proliferation resistance

An innovative plutonium burner concept based on high temperature gas cooled reactor (HTGR) technology, “Clean Burn”, is proposed by Japan Atomic Energy Agency (JAEA). That is expected to be as an effective and safe method to consume surplus plutonium accumulated in Japan. A similar concept proposed by General Atomics (GA), Deep Burn, cannot be introduced to Japan because of its adopting highly enriched plutonium, which shall infringe on a Japanese nuclear nonproliferation policy according to Japan–US reprocessing negotiation. The Clean Burn concept can avoid this problem by employing an inert matrix fuel (IMF) and a tightly coupled fuel reprocessing and fabrication plants. Both features make it impossible to extract plutonium alone out of the fabrication process and its outcomes. As a result, the Clean Burn can use surplus plutonium as a fuel without mixing it with uranium matrix. Thus, surplus plutonium alone will be incinerated effectively, while generation of plutonium from the uranium matrix is avoided. High neutronic performance, i.e., achievement of burn-up of about 500 GWd/t and consumption ratio of plutonium-239 reaching to about 95%, is also assessed. Furthermore, reactivity defect caused by the inert matrix is found to be negligible. It is concluded that the Clean Burn concept is a useful option to incinerate plutonium with high proliferation resistance.     

Demonstration Experiment of 3 BN-600 MOX Vibropac FAs Irradiation for the Excess Weapons Plutonium Disposal

The paper presents results of a demonstration experiment on conversion of 50 kg of weapon-grade plutonium in the form of metal ingots into granulated MOX-fuel to be used for manufacturing fuel pins and 3 fuel assemblies (FAs) for the fast power-generating reactor BN-600, irradiation parameters of these FAs and the data from post-irradiation examinations. It can be concluded from the PIE results that the 3FAs were successfully irradiated in BN-600 without any fuel pin failures. Therefore, disposition of weapongrade plutonium with a weight of about 20 kg was successfully done. This represents the first disposition of Russian surplus weapon-grade plutonium as an international cooperation (this experiment was performed in collaboration between RIAR and JNC). The possibility of using MOX vipac fuel as a method for weapon plutonium disposition is clearly shown.     

Variations in Neutronic Characteristics Accompanying Burnup in a Large Fast Breeder

Burnup calculations based on one-dimensional slab model approximation have been performed to determine the variations occurring in the neutronic characteristics of a pancake-shaped 1,000 MWe sodium-cooled fast breeder when operated under a specified refueling scheme.

The fissile plutonium enrichment [fiss.Pu/(Pu+U)] of the initially loaded fuel is 18.0%. It proves from the present calculations that the partial replacement of spent fuel according to the specified scheme requires use of fresh fuel containing 21.6% fissile plutonium in order to prevent the decline of reactivity with progress of refueling cycles. The above enrichment of the replacement fuel will assure equilibrium of the neutronic characteristics after about 5 years of operation. Thus, unless the refueling charges are provided with fissile plutonium enrichment higher than that of the initially loaded fuel, the state of the reactor will soon fall below criticality.

When the refueling cycle is repeated with fuel of the above specified enrichment, the breeding ratio will decline with progress of operation, from 1.24 in the initial state to 1.14 in the equilibrium state.

At the same time, both sodium void effect and Doppler coefficient will tend toward more unfavorable values, the former from ?0.083Δk/k% to ?0.068Δk/k% (calculated for cases of complete sodium removal from core), and the latter from ?0.0075T-dk/dT to ?0.0051T-dkldT (with sodium). Thus, reactor safety is foreseen to be gradually encroached as the operation progresses.     

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.     

Evaluation of Proliferation Resistance of Plutonium Based on Decay Heat

The present paper is prepared for the peaceful use of nuclear energy. Proliferation resistance of plutonium can be enhanced by increasing the decay heat of plutonium. For example, it can be enhanced by increasing the isotopic fraction of ²³⁸Pu, which has the largest decay heat among plutonium isotopes. In the present paper, the proliferation resistance of plutonium was evaluated based on decay heat with a physical assessment model. New criteria were proposed to evaluate the proliferation resistance based on isotopic compositions of plutonium from the viewpoint of decay heat. The present criteria were applied to evaluate the proliferation resistance of plutonium produced in typical Light Water Reactor and Fast Breeder Reactor based on an evaluation function “Attractiveness” as case studies.     

Build-Up and Decay of Actinide Nuclides in Fuel Cycle of Nuclear Reactors

In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1,000MW. The following items were examined;

1. quantities of actinide nuclides build-up in the reactor

2. build-up and decay of activities of actinides in the spent fuel

3. build-up and decay of activities of actinides after reprocessing, and

4. variation of isotopie composition of plutonium with high burn-up.

It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.     

Uranium-based TRU multi-recycling with thermal neutron HTGR to reduce environmental burden and threat of nuclear proliferation

To reduce environmental burden and threat of nuclear proliferation, multi-recycling fuel cycle with high temperature gas-cooled reactor has been investigated. Those problems are solved by incinerating trans-uranium (TRU) nuclides, which is composed of plutonium and minor actinoid, and there is concept to realize TRU incineration by multi-recycling with fast breeder reactor. In this study, multi-recycling is realized even with a thermal reactor by feeding fissile uranium from outside of the fuel cycle instead of breeding fissile nuclide. In this fuel cycle, recovered uranium and natural uranium are enriched and mixed with recovered TRU to fabricate fresh fuels.

The fuel cycle was designed for a gas turbine high temperature reactor (GTHTR300). Reprocessing is assumed as existing reprocessing with four-group partitioning technology.

As a result, the TRU nuclides excluding neptunium can be recycled by the proposed cycle. The duration of potential toxicity decaying to natural uranium level can be reduced to approximately 300 years, and the footprint of repository for high-level waste can be reduced by 99.7% compared with the standard case. Surplus plutonium is not generated by this cycle. Moreover, incineration of TRU from light water reactor cycle can be performed in this cycle.     

钚的利用与核裂变能的可持续发展

简要分析了当今世界的能源结构 ,指出以化石燃料为主的能源供应不可持续。概述了乏燃料后处理与钚的循环对充分利用铀资源的贡献 ,指出钚和其他锕系元素的彻底焚烧 ,有可能最大限度地减少放射性废物量及其毒性 ,从而实现核裂变能的可持续发展     

Decoding of Thermal Data in Fleischmann & Pons Paper

A uranium-free fast reactor was simulated at FCA in order to examine the prediction accuracy for sodium void effect of plutonium burning fast reactors. Material sample worth for plutonium and B4C was also measured to compare its prediction accuracy with that for the sodium void reactivity worth. It was found that an axial distribution of plutonium sample worth and the central B₄C sample worth for various kinds of ¹⁰B enrichment were precisely calculated by the conventional calculation method for fast reactors with 70-group structure. The sodium void reactivity worth was, however, poorly predicted especially for the non-leakage term. This discrepancy seems to be caused by the peculiar energy breakdown of the non-leakage term in the uranium-free fast reactor.     

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.     

核数据测量用钚靶的制备

钚的核数据测量是核科学领域一项重要的研究内容,要获得钚的相关核数据,需将钚制备成钚靶置于加速器或反应堆中进行辐照,因此在测量过程中钚靶必不可少。根据测量相关需求,需要将m≥4μg的钚制备成活性区直径D≤5 mm的钚靶。本工作对于小面积Pu靶的制备进行了大量的实验工作,通过对异丙醇、硫酸铵、硫酸钠三种不同电沉积体系的研究,确定了以硫酸钠为电沉积液的电沉积体系,并对装置进行改进以及对相关工艺条件进行研究,确定了Pu靶制备工艺:在pH=2.0～3.0的硫酸钠电沉积液中、750 mA/cm²的电流密度、65℃下,电沉积150 min,并通过液闪谱仪对电沉积率进行了检测。结果表明,该方法对于Pu的电沉积率可达98%以上。