高酸度电解质溶液中硅酸的形态及其聚合动力学研究

在核燃料后处理的水法工艺中,通常用萃取法和离子交换法回收铀、钚等放射性元素。为抑制萃取过程中的乳化和界面污物的形成以及离子交换中的树脂中毒,料液必需先经除硅处理。本工作结合实际任务,对高酸度硅含量较低的电解质溶液中硅的形态及其聚合动力学进行了研究。     

电势测量法研究单甲基肼和二甲基羟胺与Fe^3＋的氧化还原反应动力学

近年来，在核燃料后处理工艺的研究中，人们对有机无盐试剂的应用给予了极大关注。肼类衍生物及羟胺类衍生物与镎、钚的反应具有较好的选择性，不生成残留盐分，更适用于从动力堆核燃料中分离镎钚。在对其与镎钚反应动力学的研究过程中，观测到了一定量铁离子的存在会加速反应的进行这一现象。     

U、Np、Pu的X射线荧光激发参量比的应用研究

对于后处理厂的工艺控制分析,1AF料液中铀和钚浓度是至关重要的数据,同时,它也是核燃料后处理厂衡算分析的技术难点和重点。     

硝酸介质-荧光法测定亚硝酸根的分析方法

在核燃料后处理过程中，硝酸的光解和辐解以及元件的溶解会产生亚硝酸根，有时调料也需加入亚硝酸盐。亚硝酸根浓度影响着料液中镎、钚以及裂片元素的价态分布，因此，亚硝酸根浓度的测定是后处理工艺控制和后处理工艺研究中必要的分析项目。     

钚(Ⅲ)草酸盐溶解度

一、引言钚(III)草酸盐晶体具有适于钚转化过程的优良的物理化学性能。研究钚(III)草酸盐溶解度的文献所见不多,且未见系统的,特别是用硝酸羟胺(HAN)作为盥还原剂的研究。国外多数文献报道,尽管钚(III)草酸盐沉淀的各项性能均优于钚(IV),但因采用钚(III)草酸盐沉淀需要调价步骤雨未能用干工厂规模。如果核燃料后处理中钚的最后纯化循环采用羟胺还原反萃取,则所得钚产品液中主要为三价,此溶液可直接作为钚(III)     

武器用钚的控制及其核查问题

主要从核不扩散的角度来研究讨论武器用钚的控制和核查问题。(1)主要讨论了军用钚的生产及其特征,分析了几种估计军用钚库存量的方法,讨论了停止武器级钚的生产及其核查问题;(2)主要讨论了民用核燃料循环中钚的扩散及其安全保障。特别是对动力堆、后处理厂、钚燃料加工厂的安全保障问题进行了分析;(3)讨论了选择秘密生产武器级钚的可能性,给出了探测秘密生产活动的几种可能途径;(4)从核不扩散的角度,简要分析和评估了几种关于分离钚的处置方案。     

武器用钚的控制及其核查问题

主要从核不扩散的角度来研究讨论武器用钚的控制和核查问题。(1)主要讨论了军用钚的生产及其特征,分析了几种估计军用钚库存量的方法,讨论了停止武器级钚的生产及其核查问题;(2)主要讨论了民用核燃料循环中钚的扩散及其安全保障。特别是对动力堆、     

PMBP萃取法分析硝酸介质中钚的价态

在核燃料后处理工艺、锕系元素分离、钚的核素迁移和钚的化学研究中常常需要测定钚的价态组成。除仪器分析方法外,诸如沉淀载带法,溶剂萃取法,离子交换法和萃取色层及纸上色层法等化学分离法也被广泛研究和使用。Savage等曾使用TTA-二甲苯萃取法进行混合价态钚溶液中Pu(Ⅲ),Pu(Ⅳ)和Pu(Ⅵ)的测定。由于TTA萃取四价钚平衡很慢,不仅分析时间很长,而且也给被测体系中Pu(Ⅲ)的稳定带来困难。     

环境样品高灵敏Pu分析的优化程序设计

鉴于钚在核燃料循环中的重要性，同时，它具有极毒性，在环境中广为散布，因而，环境中的钚分析监测对保护环境生态和公众安全十分重要，一直受到学界的关注。总体来看，在经历了长期的方法学研究后，环境中钚的分析测定方法渐趋成熟，我国和日本等国家均颁布了环境样品中钚分析测定的国家标准方法，即采用硝酸介质中阴离子交换分离、电沉积制样、低本底α谱仪或计数器测定。     

多接收电感耦合等离子体质谱法精密测量铀基体中痕量钚同位素比值方法研究

准确测定经核燃料后处理得到的铀产品中的钚同位素比值，可大致判断铀的来源，这对防止核扩散、铀产品的质量控制具有重要意义。但核燃料后处理铀产品中钚的量非常低（小于1ng／g），精密测量其中痕量钚的同位素丰度比具有较大技术难度。     

Ion Exchange Technology in Spent Fuel Reprocessing

Solvent extraction is the major unit operation employed in spent nuclear fuel reprocessing. The operation yields three streams; fission product waste, uranium product and plutonium product. Ion exchange is primarily used in reprocessing as a tail-end method to concentrate and isolate the plutonium product stream. This review will describe the details of plutonium recovery and purification by both cation- and anion-exchange processing. A brief overview of miscellaneous uses of ion-exchange employed in reprocessing will also be given.     

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.     

核燃料水法后处理现状和展望

本文评论了世界各国乏燃料后处理技术。后处理能力在目前和不久将来不能满足核能的发展需要,它在燃料循环中占有重要的地位。普雷克斯流程不仅对于轻水堆燃料,而且对快中子增殖堆燃料后处理然是一种主要流程。近期后处理研究和发展的重点在于使流程最佳化,并引入新技术,尤其是U(Ⅳ),电解氧化还原,硝酸羟胺还原和亚硝气氧化等无盐过程。当处理高燃耗的乏燃料时,应采取特别措施,以避免造成溶剂的严重辐解和钚的临界问题。氚在流程中必须控制,并限定在一定区域,使其废液体积尽量减小。     

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.     

乏燃料后处理厂主工艺中的主要化学安全问题

水法后处理工艺过程涉及很多化学反应,反应条件和反应产物不同,需要关注的化学安全问题也不同。描述了后处理主工艺不同阶段的化学反应,分析了各阶段应关注的主要化学安全问题,为商用后处理厂的设计和事故分析提供参考。     

Effective Utilization of Weapon-Grade Plutonium to Upgrade Repeatedly-Reprocessed Mixed-Oxide Fuel for Use in Pressurized Water Reactors

A “Multiple Recycling” mode of fuel management is proposed for effectively utilizing weapon-grade plutonium from discarded military material to compensate plutonium degradation in repeatedly-reprocessed mixed-oxide (MOX) fuel. Comparative calculations on core performance are undertaken for comparison between the proposed fuel management mode of Multiple Recycling—using recovered depleted plutonium upgraded by admixture with weapon-grade plutonium while retaining unincreased the total plutonium” content—and a reference mode of using repeatedly reprocessed spent MOX fuel with plutonium upgraded through increase of the plutonium content. Multiple Recycling proves all calculated safety parameters to be retained unimpaired through multiple cycles of MOX fuel reprocessing, whereas in the reference mode of refueling with spent MOX fuel reprocessed without upgrading with weapon-grade plutonium, many of the calculated safety parameters come to exceed stipulated limits with repetition of fuel cycles. Moreover, Multiple Recycling mode can be implemented with application solely of techniques already practiced in the fabrication of MOX fuel.     

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     

草酸铈和草酸铀的热分解机理及煅烧遗传性北大核心CSCD

在核燃料后处理过程钚尾端处理中,通常利用草酸制备草酸钚沉淀然后进一步煅烧获得氧化钚颗粒、用于MOX燃料的制造,因此掌握草酸钚热分解机制、控制氧化钚产品颗粒的形貌和粒度具有实际意义。针对草酸盐的热分解机理和沉淀在煅烧后颗粒形貌遗传性问题,本工作选用草酸铈和草酸铀作为研究对象,系统研究了草酸铈、草酸铀的热分解反应,结合同步热分析仪(TG/DSC)与X射线衍射仪(XRD)的表征,获得了草酸铈和草酸铀的热分解数据;制备了不同粒度的草酸铈和草酸铀,用激光粒度仪考察了煅烧前后颗粒的粒度,并用扫描电镜(SEM)观察颗粒聚集状态、粒度和形貌,结果表明煅烧分解会导致粒径有规律地下降,但形貌得以保留。     

铀钚萃取洗涤-共反萃工艺Ⅰ.串级工艺优化

快堆燃料后处理是实现快堆燃料闭式循环的关键环节之一,快堆乏燃料中裂变产物含量高,进行后处理需要多个铀钚萃取洗涤-共反萃循环才能达到去污效果。本研究针对快堆乏燃料高钚浓度和需要多个萃取洗涤 共反萃循环净化裂变产物的特点,采用模拟料液通过多次串级实验,确定了满足铀钚收率及避免钚聚合的铀钚萃取洗涤-共反萃工艺,实验结果表明,1A铀、钚萃取收率分别为99.995%和99.996%,1B铀、钚反萃收率分别为99.936%和99.996%。     

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.