乏燃料后处理溶液中U,Pu的^57Co原激发K—XRF分析

自制一套＾３７Ｃｏ源激发与Ｋ系Ｘ射线荧光（Ｋ－ＸＲＦ）分析系统，用＾３７Ｃｏ的ｋｅＶγ射线激发工艺溶液中Ｕ，Ｐｕ的Ｋ系Ｘ射线荧光，用ＨＰＧｅ探测器－多道微机分析系统进行测量，并以１２２ｋｅＶγ射线康昔顿散射线为内标，建立强度比－浓度校正曲线，快速同地测定了ＰＷＲ乏燃料后处理工艺溶液中Ｕ，Ｐｕ浓度，测定范围为０．５－２００ｇ／Ｌ，精密度为５．０％－１．５％。方法适于ＰＷＲ乏燃料后处理工艺中，Ｕ，Ｐｕ     

碘化汞探测器的研制进展

使用片状法生长ＨｇＩ２单晶片制砀２０ｍｍ２×０．４ｍｍ探测器在常温下对５９．５ｋｅＶγ射线和５．９ｋｅＶＸ射线的能量分辨率分别为１．９ｋｅＶ和９００ｅＶ。已用于Ｃｕ、Ｐｂ、Ｚｎ等矿样分析，取得了新结果。     

^75Se和^113Sn的KX射线发射几率测量

用＾５４Ｍｎ、＾５７Ｃｏ、＾６５Ｚｎ、＾８５Ｓｒ、＾１０９Ｃｄ、＾１３７Ｃｓ等核素发射的ＫＸ射线和Ｉ（ＫＸ）／Ｉ（γ）比值法在５￣４０ｋｅＶ范围内对Ｓｉ（Ｌｉ）探测器的效率进行了刻度（不确定度小于２％）。在此基础上,用Ｓｉ（Ｌｉ）Ｘ射线谱仪系统和ＨＰＧｅγ射线谱仪系统对核素＾７５Ｓｅ和＾１１３Ｓｎ的主γ射线和ＫＸ射线进行了测量,获得了＾７５Ｓｅ和＾１１３Ｓｎ的ＫＸ射线发射几率,并与文献报道值进行了     

X,γ系列标准源的研制

本文主要介绍用于Ｘ、γ谱仪能量及效率刻度的一套系列标准源的研制方法和检验结果。它是发射射线能量范围为５．９－１８３６．１ｋｅＶ的￣（５５）Ｆｅ、￣（１０９）Ｃｄ、￣（２４１）Ａｍ、￣（５７）Ｃｏ、￣（１３３）Ｂａ、￣（１３７）Ｃｓ、￣（５４）Ｍｎ、￣（６０）Ｃｏ、￣（２２）Ｎａ和￣（８８）Ｙ１０种核素标准源。采用的高纯放射性核素标准溶液，杂质的相对强度小于０．１％；由４π（ＬＳ）和４π（ＰＣ）β－γ符合方法确定的放射性浓度均在总不确定度范围（＜２．０％，置信水平为９９．７％）内相符。放射性核素密封于质量厚度为１５ｍｇ／ｃｍ￣２（除￣（５５）Ｆｅ源用７ｍｇ／ｃｍ￣２）的圆形聚脂膜中，活性区直径≤３ｍｍ。对６个￣（５７）Ｃｏ源活度用ＮｐＧｅ谱仪测定检验，偏差在０．５１％－１．０２％之间。经擦试检验表明，未见表面污染和泄漏。     

混合式K边界技术及其应用的实验室研究（I）

描述了１台用于乏燃料后处理厂首端溶解液中铀，钚浓度同时直接测量的混合式Ｋ边界／Ｘ荧光装置的研制和实验结果，包括装置调试软件开发，系统刻度，粗度试验以及标准铀溶液样品的分析，系统对铀浓度在１００－３５０ｇ／Ｌ范围内Ｋ边办刻度系数的平均值△μ＝（３．３１７６±０．００１０）ｃｍ＾２／ｇ，对ρ（Ｕ）〉１００ｇ／Ｌ样品的分析精度在０．２２％－０．４０％之间，用该装置对Ｕ３Ｏ８标准样品的分析结果与化学测量值     

探测器校准用的6130keV单能γ源

通过混合＾２３８Ｐｕα辐射体和＾１３Ｃ制得单能６１３０ｋｅＶ高能γ射线校准源，测量了ＨＰＧｅ探测器效率、标准源的中子发射率、γ射线能谱及其发射率，讨论了不同投料比和工艺。     

低本底反康普顿HPGe γ谱仪的调试

本文报道的低本底反康普顿ＨＰＧｅγ谱仪．ＨＰＧｅ探测器对￣（６０）Ｃｏ的１３３２ｋｅｖγ射线的相对探测效率为３８．３％．能量分辨率为１．７７ｋｅＶ。在阱型反符合屏蔽下．对放在探测器端面的￣（１３７）Ｃｓ点状薄膜源的峰康比可达６８５．８：１；测量时间１００ｍｉｎ．置信度９５％时．￣（１３７）Ｃｓ点源的最小判断限为１．１２ｘ１Ｏ￣（－４）Ｂｑ。在物质屏蔽和阶型反符合屏蔽下，在５０～２１５２．８ｋｅＶ能区的积分本底为０．３４３ｓ￣（－１）。与无反符合屏蔽时相比，压缩系数大于４．５．对￣（１５２）Ｅｕ体源，谱仪积分非线性为０．０２７％。     

压水动力反应堆燃料元件安全性的监测与分析

为了对１座压水型动力反应堆作燃料元件破损的现场监测，计算了一些裂变产物的主要γ光子用７６．２ｍｍ×７６．２ｍｍＮａｌ探测器测量时产生的光电峰相对计数率随反应堆启动不同时间的变化，并作了监测中的干扰因素分析。计算和分析结果表明：在元件安全性监测中，最适合选择的γ光子能量是２２０．９ｋｅＶ（８９Ｋｒ）、４０２．７ｋｅＶ（８７Ｋｒ）、１９６．３ｋｅＶ（８８Ｋｒ）、５２９．８ｋｅＶ（１３３Ｉ）和８１ｋｅＶ（１３３ｘｅ）。在监测中存在的主要干扰因素是高能γ射线产生的湮没辐射、ｗａｌ探测器周围ｐｂ屏蔽上产生的７５ｋｅＶＸ射线及由１９Ｏ和１６Ｎ产生的γ射线。在１座反应堆２次事故排除的元件安全性监测中，分析方法成功地得到了应用。     

电离辐射作用后粒－巨噬细胞修复的特性

运用粒－巨噬细胞集落（ＣＦＵ－ＧＭ）体外培养技术，观察了受次致死剂量６０Ｃｏγ射线照射的小鼠骨髓ＣＦＵ－ＧＭ的修复能力。实验测得正常小鼠每根股骨中ＣＦＵ－ＧＭ的总数和骨髓有核细胞总数分别为４．８９×１０４和２．４×１０７，经６Ｇｙ剂量照射后第３ｄ，两计数值分别下降到正常值的０．８％和４％，照射后第５ｄ出现回升。骨髓有核细胞计数和外周血白细胞数与ＣＦＵ－ＧＭ的变化规律相一致。     

硝酸溶液中U（Ⅳ）,U（Ⅵ）和Pu（Ⅲ）的二次微分光谱及其同时测定

研究了ＨＮＯ３介质中Ｕ（Ⅳ），Ｕ（Ⅵ）和Ｐｕ（Ⅲ）的微分光谱和吸收光谱，讨论了Ｈ＾＋，ＮＯ＾－３和ＨＮＯ３浓度变化对这三种离子二次微分光谱的影响，并在此基础上，对Ｕ、Ｐｕ浓度进行了测定。在硝酸介质中同时测定Ｕ（Ⅳ），Ｕ（Ⅵ）和Ｐｕ（Ⅲ）的适宜硝酸浓度范围为２．５－３．０ｍｏｌ／ｌ，相对偏差〈５．０％，对Ｕ（Ⅳ），Ｕ（Ⅵ）和Ｐｕ（Ⅲ）的检测下限分别为０．１、１和０．５ｍｍｏｌ／ｌ。     

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of ²³⁹Pu and ²⁴¹Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.     

燃耗信任制下燃耗计算对临界计算的偏差及不确定度的研究

为量化燃耗信任制中燃耗计算传递给临界计算的不确定度,本文基于参数统计法对燃耗计算的核素偏差及偏差不确定度展开分析,并以蒙特卡罗（MC）抽样方法计算的kinf不确定度为基准,比较不同抽样方法对临界计算不确定度的影响。结果表明,核素偏差与偏差不确定度是随样品燃耗变化的分段函数。对于临界计算,拉丁超立方抽样（LHS）方法与MC抽样方法的kinf不确定度计算结果吻合较好,且LHS方法可考虑参数间的相关性,计算结果更真实,可进一步提升电厂的经济性。     

干法后处理技术典型流程综述

虽然基于溶剂萃取的Purex流程在乏燃料后处理几十年的应用中取得的成功,使得水法后处理至今没有发展出可以取代这一流程的新萃取剂,但干法后处理却有了两种可供进一步发展的流程:氟化物挥发法和高温电化学法。氟化物挥发法存在的最大问题是热力学上PuF6必须在有大量F2过剩的条件下才稳定。高温电化学法适合于处理合金元件,以及氧化物和碳化物元件。首先,将核燃料熔解在熔盐中,然后,电解使铀钚在阴极上沉积,再对阴极上沉积出来的铀钚进行精制而得到铀钚产品。但该方法存在熔盐对MOX的熔解能力和对过程设备的腐蚀问题。     

Separation and Determination of Pu and Np in Existence of Large Quantityof Uranium：Establishment and Applicationof Argon Pressurized Anion-Exchange Evacuated Column System

Np and Pu in the uranium product must strictly be controlled in spent fuel reprocessing process.Determination of Np, Pu in uranium product needs accuracy, stability and efficiency for uranium product quality control.     

俄罗斯氧化物乏燃料电沉积流程研究进展

高温熔盐干法后处理以熔盐作为电解质，通过电解精炼和电沉积回收核燃料中的铀和钚。目前，俄罗斯、美国、日本、韩国和欧盟等国均在积极发展乏燃料高温熔盐干法后处理技术的研究，其中俄罗斯的金属氧化物核燃料电沉积流程是经典的流程之一。本文对俄罗斯原子反应堆研究所（Research Institute of Atomic Reactors,RIAR）发展的氧化物乏燃料高温熔盐电沉积干法后处理的发展现状、流程及特点进行了综述。     

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.     

Thermochemical modelling of electrotransport of uranium and plutonium in an electrorefiner

Electrotransport behaviour of U and Pu in a molten salt electrorefining cell has been numerically simulated with an improved thermochemical model. Depending upon saturated or unsaturated states of the liquid Cd electrodes with respect to U or Pu or with both U and Pu, 16 conditions of electrorefiner cell operation have been categorised and electrotransport simulated for all the realistic conditions. Algebraic equations for determining the compositions of the salt phase and the two electrodes under each condition of electrotransport are derived. Fractional mass transport coefficients and relative fractional mass transport coefficients are derived for each condition to illustrate the electrotransport behaviour. Comparison is made between modeling with concentration dependent and concentration independent activity coefficients for U and Pu in liquid Cd. The electrotransport to a solid cathode and anodic dissolution have also been simulated. Application of the model to reprocessing of spent metallic fuel is discussed with respect to U recovery, Pu enrichment and reconstitution of the spent fuel with desired fuel composition.     

后处理有机相料液中Pu(Ⅳ)和硝酸的同时快速分析

Pu(Ⅳ)和硝酸的测定在核燃料后处理厂工艺控制分析中占有重要的地位。采用自行研制的分析装置,利用Pu(Ⅳ)和硝酸的近红外吸收光谱,结合偏最小二乘回归（PLS）法,建立了后处理工艺有机相料液中Pu(Ⅳ)和硝酸含量的同时快速分析方法。Pu(Ⅳ)及硝酸的浓度测量范围分别为0.15～15 g/L、0.05～0.80 mol/L,测量范围覆盖了后处理流程大部分的工艺点。料液中硝酸测量的相对标准偏差小于5%,Pu(Ⅳ)测量的相对标准偏差小于2%。模拟样品的分析结果通过t检验,Pu(Ⅳ)和硝酸的重加回收率均为95%～103%。     

基于拟Newton算法的Purex流程计算机模拟程序

在多级逆流萃取MESH方程基础上，基于拟Newton算法编写了用于模拟Purex萃取流程的计算机程序，使用SEPHIS分配比模型，分别以1A、2D槽工艺条件为例，计算了其各级出口的U、Pu酸浓度。程序计算结果与实验浓度剖面符合很好。相对于已有的一些计算机模拟程序，该程序具有适用性广、收敛性强的特点。     

Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR／CANDU synergism

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexi-bility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nu-clear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (-22.5%), increase the energy output (-41%), decrease the quantity of spent fuels to be disposed (-2/3) and lower the cost of nuclear poower, Because of the inherent flexibility of nuclearfuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modifica-tion of the reactor core structure and operation mode.It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.