Implementation of an LBE spallation target in an accelerator-driven molten salt subcritical reactor

An accelerator-driven system (ADS) combined with a subcritical molten salt reactor (MSR) is a type of hybrid reactor originally designed to use Th/U (or U/Pu ) fuel cycles. In most accelerator-driven molten salt reactor (AD-MSR) concepts, the salt material is also used as a target for inducing spallation neutrons. Although a neutron source is an important component in the design of ADS, only a few studies have addressed the effects of the neutron spallation source in the AD-MSR. Incidentally, there is no quantitative study on how much the beam power can be reduced by installing a spallation target in a sodium chloride-based fast reactor. We studied the proton and the neutron source efficiencies of an AD-MSR with chloride fuels by considering an Lead Bismuth Eutectic (LBE) spallation target. This LBE target is found to increase the proton source efficiency significantly. The required beam power for an AD-MSR can be reduced by 33 % and 16 % for NaCl-Th/²³³U and NaCl-U/Pu fuels, respectively, relative to the AD-MSR without the LBE spallation target by keeping the same k_eff. The energy gain can be increased up to 1.5 times and 1.2 times for NaCl-Th/²³³U and NaCl-U/Pu fuels, respectively. Thus, incorporating a spallation target module in an AD-MSR can significantly reduce the burden on the accelerator.     

Application of fluoride volatility method to the spent fuel reprocessing

ABSTRACT

An advanced reprocessing system has been developed to treat various SF (spent fuels): spent UO₂ and MOX (mixed oxide) fuels from LWR (light water reactor) and MOX fuel from FR (fast reactor). The system consists of SF fluorination to separate most U (uranium) as volatile UF₆, dissolution of solid residue containing Pu (plutonium), FP (fission products), MA (minor actinides) and partial U by nitric acid, and Pu+U separation from FP and MA by conventional solvent extraction. Gaseous UF₆ is purified by the thermal decomposition and the adsorption of volatile PuF₆ and adsorption of other impurities. This system is a hybrid process of fluoride volatility and solvent extraction and called FLUOREX. Fluorination of most U in the early stage of the reprocessing process is aimed at sharply reducing the amount of SF to be treated in the downstream aqueous steps and directly providing purified UF₆ for the enrichment process without conversion. The FLUOREX can flexibly adjust the Pu/U ratio, rapidly separate UF₆ and economically treat aqueous Pu+U. These features are especially suitable for the transition period fuel cycle from LWR to FR. This paper summarizes the feasibility confirmation results of FLUOREX.     

快堆中不同核燃料类型的长寿期性能评价

建立了典型的快堆六角形栅元堆芯模型,研究了多种类型的燃料在快中子能谱辐照环境下经过较长时间辐照后的性能,对不同燃料堆芯在运行寿期末的乏燃料组成成分进行了分析.结果表明,在栅元结构完全一样且初始剩余反应性基本相同的情况下,燃料反应性损失从小到大的顺序是:金属燃料＜氮化物燃料＜碳化物燃料<氧化物燃料;在整个寿期中,使用Pu驱动的燃料比使用235U驱动的燃料反应性下降得慢;金属燃料寿期末乏燃料中按初始装载燃料质量平均后的超铀核素的质量最小,其他依次为氧化物<氮化物<碳化物;由于初始装载量的增多,使用Pu驱动的燃料寿期末乏燃料超铀核素的总量比使用235U驱动的燃料多,同时,乏燃料Pu中的易裂变同位素的份额比235U驱动燃料的少.从中子学角度考虑,UZr燃料是比较理想的长寿命快堆堆芯燃料类型.     

Toward Innovative Actinide Separation Processes: Sequential Reduction Scheme of Uranium,Neptunium, and Plutonium in 3 M HNO3 by External Ultrasound Irradiation

An innovative remote valency control technique for actinide ions induced by external ultrasound irradiation was reported in the present study. It is known that ultrasound irradiation to water causes the oxidation and/or reduction of the solute by H? or OH? There were some reports on the redox behavior of actinide elements under ultrasound irradiation. Nevertheless, they showed that the ability of ultrasound is not sufficient for the use of the actinide separation processes. However, very recently we found that a noble metal catalyst drastically enhances the sonochemical effect and that even highly stable U(VI) is reduced to U(IV) by ultrasound irradiation. Employing this catalytic reaction, we are developing low-emission actinide ion separation schemes driven by external ultrasound irradiation. In the present work, U(VI), Np(VI) and Pu(VI) in 3 M HNO₃ medium were chosen as target ions. Their valency was first adjusted to U(VI)/Np(V)/Pu(IV) by external ultrasound irradiation and, then, further sonochemical reduction to U(IV)/Np(IV)/Pu(III) was carried out. The present results confirmed the possibility to design the low-emission U/Np/Pu mutual separation scheme driven by external ultrasound irradiation.     

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.     

加速器驱动先进核能系统的乏燃料循环再生研究北大核心CSCD

乏燃料后处理是核燃料循环的关键环节,制约核电的可持续发展。借助于加速器驱动先进核能系统(ADANES)提供的高通量、硬能谱的外源中子,其乏燃料后处理只需除去乏燃料中的挥发性裂变产物和影响次锕系元素嬗变的中子毒物,长寿命的次锕系元素Np、Am、Cm可与二氧化铀一起转化为新的燃料元件在加速器驱动燃烧器中燃烧、嬗变、增殖和产能。基于此,本课题组提出了加速器驱动的乏燃料后处理及再生制备的技术路线,包括高温氧化粉化与挥发、选择性溶解分离和燃料再生制备。本文主要介绍了近几年本课题组在这三方面所取得的一些成就,希望能为加速器驱动先进核能系统的乏燃料后处理提供基础数据。     

基于IQS/MC方法的ADS次临界反应堆中子时空动力学模拟分析

采用改进准静态近似与蒙特卡罗中子输运程序相结合(IQS/MC)的方法实现了加速器驱动的次临界系统(ADS)中子时空动力学模拟计算。以加速器驱动嬗变研究装置的靶堆耦合参考方案物理模型为例,通过对束流瞬变引入和燃料组件提升两种工况进行动态模拟,计算得到了堆芯总的相对功率、分能群相对中子注量率及相对功率三维网格分布随时间的变化。将IQS/MC方法计算结果与点堆计算结果进行了对比分析,模拟结果符合物理规律,两种方法对比结果与国外相关文献一致,表明IQS/MC方法适用于ADS次临界反应堆中子时空动力学过程的瞬态安全分析。     

Light Water Breeding with Nitride Fuel

This scoping study proposes using mixed nitride fuel in Pu-based high conversion LWR designs in order to increase the breeding ratio. The higher density fuel reduces the hydrogen-to-heavy metal ratio in the reactor which results in a harder spectrum in which breeding is more effective. A Resource-renewable Boiling Water Reactor (RBWR) assembly was modeled in MCNP to demonstrate this effect in a typical high conversion LWR design. It was determined that changing the fuel from (U,TRU)O₂ to (U,TRU)N in the assembly can increase its fissile inventory ratio (fissile Pu mass divided by initial fissile Pu mass) from 1.04 to up to 1.17.     

Electrorefining of metallic fuel with burn-up of ~7 at% in a LiCl-KCl melt

Electrorefining of irradiated metallic fuels (burn-up ~ 7 at%) in a LiCl-KCl melt at 773 K was successfully demonstrated: Actinides in the fuels were anodically dissolved in the melt. Both a selective U metal deposition on a solid cathode and a grouped recovery of actinides, U, Pu, Np, Am, and Cm, in a liquid Cd cathode were confirmed. The behavior of fission products, such as lanthanides, alkali metals, alkaline earth metals, and noble metals, were also investigated. It was found that the behaviors of actinides and fission products in the electrorefining of the fuels with ~ 7 at% burn-up were similar to those in electrorefining of fuels with ~ 2.5 at% burn-up.     

MgO-pyrochlore composite as an inert matrix fuel: Neutronic and thermal characteristics

Inert matrix fuels are an important component of advanced nuclear fuel cycles, as they provide a means of utilizing plutonium and reducing the inventory of ‘minor’ actinides. We examine the neutronic and thermal characteristics of MgO-pyrochlore (A₂B₂O₇: La₂Zr₂O₇, Nd₂Zr₂O₇ and Y₂Sn₂O₇) composites as inert matrix fuels in boiling water reactors. By incorporating plutonium with resonance nuclides, such as Am, Np and Er, in the A-site of pyrochlore, the k_infvs. burn-up curves are shown to be similar to those of UO₂, although the Doppler coefficients are less negative than UO₂. The Pu depletion rates are 88-90% (²³⁹Pu) and 54-58% (total Pu) when the inert matrix fuels experience a burn-up equivalent of 45 GWd/tHM UO₂. Because of the high thermal conductivity of MgO, the center-line temperatures of the MgO-pyrochlore composites at 44.0 kW/m are lower than those of UO₂ pellets. After burn-up, the A-site cation composition is 15-35 at.% lower than that of the B-site cations in pyrochlore (e.g., A_1.84B_2.17O_7.00) due to the fission of Pu in the A-site and the presence of fission product elements in the A- and B-sites of the pyrochlore structure.     

跳源法在启明星1#次临界装置中的应用

简要介绍了跳源法在启明星1#次临界装置上测量次临界度的原理、外源驱动的次临界中子学实验装置、堆芯布置及中子源驱动系统。主要研究了中子源在堆芯轴向中心位置、不同装载情况下的反应性变化,并给出不同的有效倍增系数keff。实验测量结果与理论计算结果进行了比较,结果符合较好。     

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO₂ assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO₂ sample of 72 GWd/t.     

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

Aspects of 238Pu production in the experimental fast reactor JOYO

Experimental determination of ²³⁸Pu in ²³⁷Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of ²³⁸Pu production from ²³⁷Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the ²³⁸Pu isotopic ratio. ²³⁸Pu production amount in the irradiated ²³⁷Np samples was determined by a radioanalytical technique. Aspects of ²³⁸Pu production were examined on the basis of the present radioanalysis. The ²³⁸Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu.     

Theoretical cross sections of Bi,Th, U and U on deuteron-induced reactions

There are many application fields for fast neutrons. The main application fields of the fast neutrons are accelerator-driven sub-critical systems (ADS) and fusion–fission (hybrid) reactor systems for fission energy production. Thorium (Th) and uranium (U) are nuclear fuels in fusion–fission (hybrid) reactor systems and bismuth (Bi) is also the target nucleus in the ADS reactor systems. In this study, neutron production cross sections produced by (d, xn) reactions for spallation targets such as ²⁰⁹Bi, ²³²Th, ²³⁵U and ²³⁸U have been investigated. New evaluated hybrid model and geometry dependent hybrid model have been used to calculate the pre-equilibrium neutron production cross sections. For the reaction equilibrium component, Weisskopf–Ewing model calculations have been preferred. The obtained results have been discussed and compared with the available experimental data and found in agreement with each other.     

加速器驱动的次临界系统快堆次锕系核素非均匀布置堆芯的中子学研究

运用MCNP与ORIGEN2耦合计算程序COUPLE，对加速器驱动的次临界系统（ADS）钠冷金属燃料快堆堆芯进行稳态与燃耗计算，比较分析次锕系核素（MA）非均匀布置堆芯与均匀布置堆芯在MA嬗变效果与反应性参数方面的差异。计算结果表明，对比均匀布置，非均匀布置具有更高的MA嬗变率与嬗变支持比，在反应性参数方面导致多普勒效应与有效缓发中子分额降低，钠空泡效应增大，在堆芯功率分布与加速器束流功率方面没有明显变化。     

Accelerator driven sub-critical systems with enhanced neutron multiplication

Utilisation of thorium, by way of the U-Th cycle, is of particular interest to the Indian Nuclear Power Programme because of large thorium deposits and limited Uranium reserves. Several schemes, such as fast and advanced heavy water reactors, leading to thorium utilisation, are under study at this centre. The present paper discusses a scheme for evolving a practical accelerator driven sub-critical U-Th system with increased neutron multiplication and consequentually a reduced requirement of the accelerator current. It is shown that the requirement of the accelerator current is considerably reduced if a sub-critical assembly with a given K_eff is composed of two partially coupled regions.     

TRU burning by dual tier system of LWR-SFR

This paper presents a concept of the dual tier system consisting of the existing light water reactor (LWR) plants and sodium-cooled fast reactor (SFR) for transuranics (TRU) burning for the purpose of downsizing the required SFR. In this system, Pu is combusted by the LWR at first and then the remaining Np, Am, and Pu are destructed by the SFR. The iteration number of Pu combustion by the LWR is chosen to be twice owing to the sodium void reactivity limitation of $6. As a result of combustion calculation, the twice Pu burning of LWR lessens the TRU amount by 27% and changes the composition significantly. Moderator pins of zirconium hydride are deployed to the SFR fuel subassembly so as to enhance TRU burning and reduce the sodium void reactivity. The nuclear calculation found that the core characteristics become similar to the conventional SFR due to the moderator: the sodium void reactivity remains still $4 and the Doppler coefficient becomes −6 × 10⁻³ Tdk/dT. This study concludes that this dual tier strategy can downsize the required SFR to approximately 40% of the single tier system of SFR with TRU conversion ratio of 0.6.     

Melting behavior of MgO-based inert matrix fuels containing (Pu,Am)O2−x

The melting behavior of MgO-based inert matrix fuels containing (Pu,Am)O_2−x ((Pu,Am)O_2−x-MgO fuels) was experimentally investigated. Heat-treatment tests were carried out at 2173 K, 2373 K and 2573 K each. The fuel melted at about 2573 K in the eutectic reaction of the Pu-Am-Mg-O system. The (Pu,Am)O_2−x grains, MgO grains and pores grew with increasing temperature. In addition, Am-rich oxide phases were formed in the (Pu,Am)O_2−x phase by heat-treatment at high temperatures. The melting behavior was compared with behaviors of PuO_2−x-MgO and AmO_2−x-MgO fuels.     

ADS次临界反应堆的中子共轭方程

与临界反应堆相比,ADS次临界反应堆的外源中子和裂变中子的空间分布具有严重的不均匀性,对应的中子价值也不同。本工作对次临界反应堆的稳态输运方程作分群扩散近似,得到了多群方程,进一步推导出按堆芯功率归一化的中子共轭方程表达式和与功率相关的中子价值函数表达式,给出了次临界反应堆中子价值的物理意义。由稳态中子共轭方程组出发,给出了两种带外加中子源的次临界反应堆增殖因数的表达式。