放射性表面污染金属废物激光去污工艺研究

为实现激光去污技术在放射性表面污染金属废物清洁解控或循环再利用方面的应用,以350 W的纳秒脉冲光纤激光器为基础搭建了激光去污实验装置,针对激光功率、脉宽、频率、线间距、扫描速度等关键参数开展了一系列激光剥离去污工艺实验,根据实验结果分析得出激光去污工艺规律和不同去污深度的最佳工艺参数,并以某核电厂控制棒水池贮存搁架底板为对象开展验证试验。验证试验结果显示,采用激光去污技术,去污深度达到10 μm后,样品的β射线放射性表面污染水平已低于0.8 Bq/cm2,可达到清洁解控的表面污染水平要求。      

板式换热器放射性去污实践探索

核电站反应堆和乏燃料水池冷却处理系统(PTR)及设备冷却水系统(RRI)中使用的板式换热器,结构复杂,运行过程中放射性热点易淤积在导流函道及沟槽等死角位置,放射性去污难度较大,业内以作为放射性固体废物处理为主。板式换热器作为高价值专用设备,大量污染报废给成本管控和放射性废物最小化管理带来了较大的压力。针对上述问题,红沿河核电厂实践探索了“放射性污染板式换热器去污方法”,采用“化学去污+泡沫去污+可剥离膜去污”分段去污方法开展试验,经去污后的换热器表面污染水平均小于0.4 Bq/cm²,4个阶段累计去污率约为99.80%,193片放射性污染板式换热器全部实现复用。     

中子辐照导致线阵电荷耦合器件电荷转移效率退化实验研究

利用TRIGA型脉冲反应堆提供的快中子，对线阵电荷耦合器件进行中子辐照实验研究。研究结果表明：在10¹²～10¹³cm^-2中子注量范围内，该器件的电荷转移效率（CTE）随辐照中子注量的增加而线性下降；电荷转移效率的下降与电荷包在沟道中的转移时间及转移电荷包的电量有关。      

大面积表面污染测量仪主要性能测试与分析

针对用于测量α、β射线的便携式大面积表面污染测量仪,为确保其测量性能,本文根据JJG 478—2016和GB/T 5202—2008标准要求,对自主研发的便携式α、β大面积表面污染测量仪的均匀性、探测效率、探测限、响应时间和温度稳定性等主要性能指标进行了测试。结果表明:便携式大面积表面污染测量仪的表面发射率响应对²⁴¹Am α源为30.4%左右,对⁹⁰Sr-⁹⁰Y β源为48.1%,对³⁶Cl β源为45.5%左右;最低可探测下限对²⁴¹Am α源为0.15 Bq/cm²,对β源⁹⁰Sr-⁹⁰Y为0.07 Bq/cm²,对β源³⁶Cl为0.09 Bq/cm²;响应时间＜4 s;在-10 ℃～40 ℃时可正常工作。各项参数指标满足标准要求,能够有效达到防止污染扩散,保证工作人员安全的目的。     

废锆包壳的α去污北大核心CSCD

开展了利用Ag(Ⅱ)间接氧化溶解废锆包壳的α去污研究。结果表明:非放锆包壳不溶于Ag(Ⅱ)-硝酸溶液中,可能是包壳表面生成氧化锆,抑制了包壳的溶解;建立了废锆包壳电化学溶解工艺,开展了废锆包壳α去污工艺验证,溶解后废锆包壳残留的α比活度为2.0×10^(5)Bq/kg;分析了废锆包壳α污染源项,确定了废锆包壳α比活度的主要贡献是^(241)Am和^(238)Pu。     

涂硼电离室中子探测效率和灵敏度

从电离室工作原理导出了平板型涂硼中子电离室探测效率及灵敏度的计算公式,并求得其热中子探测效率和灵敏度。电离室对热中子探测效率饱和值为1.35%,灵敏度饱和值为9.65×10^-14A•cm^-2•s^-1,与已有公式所得结果8.43×10^-14A•cm^-2•s^-1相近。α粒子和Li离子对探测效率的贡献相差不大,但α粒子对灵敏度的贡献占主导地位。适当的硼膜厚度、慢化快中子、选用浓缩硼均有利于提高涂硼电离室探测效率和灵敏度。     

氚污染手套箱内壁及金属去污实验

对氚污染手套箱内壁及金属采用擦拭去污和可剥离膜去污。擦拭去污后,手套箱内壁残留氚污染水平降低到20Bq/cm2以下。对氚污染水平高处采用SO42－/TiO2固体超强酸掺杂可剥离膜或聚乙烯醇（PVA）可剥离膜去污,去污因子高,而氚污染水平低的金属通过可剥离膜去污后残留氚为20Bq/cm2。     

悬浮电解去污配方的去污能力及电化学性能测试

利用自制的悬浮电解去污装置对碳钢/不锈钢模拟污染样片进行电解去污。去污结果表明：H₂SO₄-Na₂SO₄体系的悬浮电解去污配方对碳钢/不锈钢模拟污染样片能有效去污。在电解电压6V时,碳钢模拟污染样片经过2h去污,表面放射性活度可降低到本底水平;不锈钢模拟污染样片经过3h去污,去污系数可以达到180。电化学性能测试表明,该电解去污配方能有效防止阳极极化,并且在电解去污过程中具有良好的稳定性。     

Pu(OH)4(am)与碳酸根的配位行为研究

为了解在惰气环境Pu(OH)_4(am)与碳酸盐溶液中HCO^-₃，CO^2-₃的配位行为，考察了放置时间对Pu总浓度的影响；同时也考察了pH值、碳酸根总浓度变化对碳酸盐溶液中Pu的主要存在形态及溶解总浓度的影响。实验结果表明，HCO^-₃离子与Pu(OH)_4(am)生成[Pu(OH)₄(HCO₃)₂]^2-（lg K=-2.61±0.18, lgβ=54.25±0.18）或[Pu(OH)₂(CO₃)₂]^2-（lgK=-2.61±0.18, lgβ=46.91±0.18）；CO^2-₃离子与Pu(OH)_4(am）生成[Pu(OH)₄(CO₃)₂]^4-（lgK=-3.52±0.11, lgβ=53.33±0.11）。可能的配位反应方程式为： Pu(OH)_4(am)+2HCO^-₃ = [Pu(OH)₄(HCO₃)₂]^2-, Pu(OH)_4(am)+2HCO^-₃ =[Pu(OH)₂(CO₃)₂]^2-+2H₂O, Pu(OH)_4(am)+2CO^2-₃=[Pu(OH)₄(CO₃)₂]^4-。     

2011—2015年秦山核电基地周围环境沉降物总α、总β放射性水平监测

报道2011—2015年浙江省辐射环境监测站对秦山核电基地周围环境沉降物总α、总β放射性水平的监测工作。监测结果表明, 2011—2015年, 秦山核电基地周围环境沉降物中总α活度浓度范围为0.07～0.92 Bq·m^-2d^-1, 平均值为0.28 Bq·m^-2d^-1;总β活度浓度范围为0.24～1.70 Bq·m^-2d^-1, 平均值为0.58 Bq·m^-2d^-1;总α/总β值范围为0.26～0.83;监测结果与对照点(杭州市)监测值处同一水平。     

Estimation of the amount of surface contamination of a water cooled nuclear reactor by cooling water analysis

Calculations, based upon on-the-spot measurements, were performed to estimate the contamination of NPP primary circuit and spent fuel storage pool solid surfaces via the composition of the cooling water in connection with a non-nuclear incident in the Paks NPP. Thirty partially burnt-up fuel element bundles were damaged during a cleaning process, an incident which resulted in the presence of fission products in the cooling water of the cleaning tank (CT) situated in a separate pool (P1). Since this medium was in contact for an extended period of time with undamaged fuel elements to be used later and also with other structural materials of the spent fuel storage pool (SP), it was imperative to assess the surface contamination of these latter ones with a particular view to the amount of fission material. In want of direct methods, one was restricted to indirect information which rested mainly on the chemical and radiochemical data of the cooling water. It was found that (i) the most important contaminants were uranium, plutonium, cesium and cerium; (ii) after the isolation of P1 and SP and an extended period of filtering the only important contaminants were uranium and plutonium; (iii) the surface contamination of the primary circuit (PC) was much lower than that of either SP or P1; (iv) some 99% of the contamination was removed from the water by the end of the filtering process.     

核电厂乏燃料贮存格架水下去污装置研制

针对核反应堆乏燃料贮存格架去污的必要性与功能要求，研制了一种二代核电厂乏燃料贮存格架水下冲洗去污装置。详细介绍了该装置的结构组成、功能原理及控制系统设计。经某核电厂现场使用验证，该装置操作简便，具有良好的冲洗去污能力，可大幅降低乏燃料贮存格架的辐射剂量水平。      

高放废液处理与处置不同技术方案的放射性健康风险比较

本文估计和比较了乏燃料后处理高放废液不同处理和处置方案的风险。结果表明：两种方案之间风险只有很小的差别；如果分离流程中对风险贡献最大的９９Ｔｃ核素有足够的去污因子，高放废液就能降级成为中、低放废液。本文还指出只有分离和嬗变相结合的技术方案才能真正降低高放废液的处置风险。     

乏燃料处置的必要性及其处置库环境化学行为

为了确保核燃料循环的安全性，不宜处理的乏燃料也应该同玻璃固化体一样作为高放废物进行深地质处置。本文综述了一些前期工作，归纳了空气侵入和水的辐解产生氧化性产物是导致乏燃料UO2基体氧化溶解的主要因素; 核燃料浸出实验结果显示铀和锕系镧系元素每天的浸出量是相应核素总量的1/107，比裂变产物的浸出速率小一个数量级。铁金属被各国选为高放废物处置容器材料的原因是其低价格、高强度和优秀的还原能力。在最不利的地下水侵入深地质处置库、近场处置容器防腐层破损的情景下，铁容器材料表面与地下水反应产生氢气，氢气通过还原反应消耗辐解产生的氧化性自由基和分子, 并能还原乏燃料表面的U（Ⅳ），大幅度减缓乏燃料的腐蚀和溶解；乏燃料中裂变产物贵金属合金颗粒对氢气有催化作用；处置容器表面铁金属能还原沉积溶解的多价态核素U（Ⅵ）、Np（Ⅴ）、Tc（Ⅶ）、Se（Ⅳ）和Se（Ⅵ）。希望本文对我国确立以铁基金属为处置容器材料的包括乏燃料在内的高放废物深地质处置概念有参考作用。     

Environmentally acceptable nuclear fuel cycle development of a new reprocessing system

The aim of the present study is to establish a new reprocessing system for spent nuclear fuel, which would overcome the environmental problems in the nuclear fuel cycle. In order to achieve this, the following subjects have to be conquered: recoveries of high ratios of uranium and trans uranium elements from spent nuclear fuel, separations of strong radioactive elements, such as Sr and Cs, and assurance of the extreme safety during operation. The last subjects might be of particular importance in order to avoid any potential danger. Therefore, in the present system all processes were performed under mild aqueous conditions. Experiments were carried out for a simulated spent fuel solution, which was calculated from the ORIGEN CODE containing uranium and 17 major elements. The system consists of the following processes: 1. dissolution of spent UO₂ fuel involving off-gas treatment of I and Ru; 2. neutralization of the dissolved fuel solution with NaHCO₃---Na₂CO₃ mixed solution to be slightly basic at pH about 9 followed by the separation of precipitated fission products by centrifugation; 3. separation of Cs by a precipitation method using tetraphenylborate ion; 4. recovery of U, Np and Pu as precipitates of hydrolyzed compounds from alkaline solution; 5. separation of Am and Cm from lanthanide elements. The concentration of residual uranium in the final solution was measured to be ppm order, indicating that the decontamination factor of U was as large as 10⁴. Other hexa-valent actinide ions, NpO₂²⁺ and PuO₂²⁺, also have extremely large stability constants for the complex formation with carbonate ion, and are expected to behave similarly with UO₂²⁺. In conclusion, the present reprocessing system enables us to recover U, Pu and Np from spent nuclear fuel by means of a simple precipitation method in much higher ratios compared with other reprocessing methods, to separate hazardous Cs and Sr from high-level waste, and to exclude any potential danger owing to chemical processes under mild aqueous conditions.     

Application of a Chemical ion Exchange Model to Transport Cask Surface Decontamination

Abstract

Radionuclide contamination of stainless steel surfaces occurs during submersion in a spent fuel storage pool. Subsequent release or desorption of these contaminants from a nuclear fuel transportation cask surface under varying environmental conditions occasionally results in the phenomenon known as contamination ‘weeping’. Experiments have been conducted to determine the applicability of a chemical ion exchange model to characterise the problem of cask contamination and release. Surface charge characteristics of Cr₂O₃ and stainless steel (304) powders have been measured to determine the potential for ion exchange at metal oxide-aqueous interfaces. The solubility of Co and Cs electrolytes at varying pH and the adsorption characteristics of these ions on Cr₂O₃ and stainless steel powders in aqueous slurries have been studied. Experiments show that Co ions do reversibly adsorb on these powder surfaces and, more specifically, that adsorption occurs in the nominal pH range (pH=4–6) of a boric acid moderated spent fuel pool. Desorption has been demonstrated to occur at pH≤3. Cs⁺ ions also have been shown to have an affinity for these surfaces although the reversibility of Cs⁺ bonding by H⁺ ion exchange has not been fully demonstrated. These results have significant implications for effective decontamination and coating processes used on nuclear fuel transportation casks.     

研究堆破损乏燃料元件快速排查技术研究

研究堆燃料元件在安全转移至乏燃料贮存水池前,需对其进行破损检测。目前的检测方法耗时长,难以对具体的破损对象快速判定。本文提出一种破损乏燃料元件快速排查法,该方法能在短期内实现对破损乏燃料筛查,提高后续待运乏燃料破损检测通过率。      

The interim fuel storage facility of the PBMR

The PBMR’s spent fuel and partially burnt fuel are stored in the sphere storage system (SSS), which acts as the interim fuel storage facility of the plant. It is unique in the world since the fuel is stored in bulk containers (called storage tanks), each capable of holding more than 500,000 spheres for a period of about 80 years. The SSS has the ability to transfer the contents of one tank to another tank, and to return partially burnt fuel back to the reactor core for re-fuelling.The storage tanks are individually sealed carbon steel pressure vessels. They form the final barrier of any fission products that have diffused from the fuel spheres. Sub-criticality is achieved through the geometric cross-section of the tank, and by taking credit for fuel burn-up. Protection from the corrosive environment where the PBMR Demonstration plant will be built is achieved by actively cooling the tank with clean dry air. In the event of an active cooling failure, louvres open automatically and cooling is done passively via natural convection making use of the chimney-effect. Sufficient radiation protection is provided around each tank to facilitate maintenance and inspection operations where needed.The design of the SSS is nearing the end of its basic design phase, and for some components, detail design work has already commenced. The design complies with all spent fuel storage requirements and is seen as a cost-effective solution for the interim storage of PBMR spent fuel.     

Simulations of H2O2 concentration profiles in the water surrounding spent nuclear fuel

A simple mathematical model describing the hydrogen peroxide concentration profile in water surrounding a spent nuclear fuel pellet as a function of time has been developed. The water volume is divided into smaller elements, and the processes that affect hydrogen peroxide concentration are applied to each volume element. The model includes production of H₂O₂ from α-radiolysis, surface reaction between H₂O₂ and UO₂ and diffusion. Simulations show that the surface concentration of H₂O₂ increases fairly rapidly and approaches the steady-state concentration. The time to reach steady-state is sufficiently short to be neglected compared to the times of interest when simulating spent fuel dissolution under deep repository conditions. Consequently, the steady-state approach can be used to estimate the rate for radiation-induced spent nuclear fuel dissolution.     

模拟放射性污染土壤样品中~(147)Pm的分离及液闪测量

本工作针对受乏燃料污染的环境土壤样品,研究了Pm的分离纯化方法,建立了~(147)Pm的分析流程,流程收率大于70%,对主要干扰核素的去污因子大于10~3。同时,通过~(147)Pm标准源对效率示踪法测量结果进行验证,测量值与推荐值相对偏差小于0.5%。将该方法应用于模拟样品分析,其结果与~(147)Nd测量推荐值的相对偏差小于2%。