模拟铯废物玻璃固化体组成对浸出性能的影响

按正交试验设计方法设计了熔制模拟铯废物钛硅酸盐玻璃固化体的配料组成 ,制备了相应的固化体样品。按静态浸出试验方法 (MCC- 1)对样品进行了浸出试验 ,条件为去离子水、90℃、7d、样品表面积与去离子水体积之比为 10 m-1。用电感耦合等离子体原子发射光谱 (ICP- AES)和原子吸收 (AAS)测定浸出液中各种离子的浓度。结果表明 ,固化体浸出液中主要含 Na+和 Si4+ ,其次是 Cs+。固化体中Zr O2 组分对浸出性能影响最大 ,其次是 Ti O2 / Si O2 (摩尔比 )。配料中 Zr O2 的摩尔分数选择在 1.5 %～4 .5 %之间、Ti O2 / Si O2 的摩尔比在 0 .4 0左右时 ,有利于减少 Cs+ 及 Na+ 、Si4+ 的浸出     

关于低中放废物陆地处置安全性研究——赴日考察技术简讯(2)

陆地处置的安全性评价是放射性废物处置的核心问题,因而也是世界各国共同关注的问题。日本低放废物的陆地处置目前正处在安全性研究阶段,已经做了比较多的工作。本文就赴日考察期间了解的情况简述如下:一、人工屏障安全性实证研究1.低放废物固化体长期浸出性试验日本原子能环境治理中心用“冷”试验进行模拟废物固化体长期浸出工程试验,他们以实际规模(1:1)的非放射性固化体(水泥、沥清固化体)在海水中和陆水(地面水)中进行为期三年时间的浸出试验,是从1982年开始准备的,1984年在进行试验装置调试,固化体模拟试制和预备试验的基础上开始了水泥固化体和沥青固化体在海水中的     

沥青-硝酸钠固化体长期浸出实验结果

放射性废物固化体的浸出性能研究是一项长期的综合性的课题.不仅要研究固化体中有害组分在长期浸泡过程中向介质(环境)中迁移的规律、机制、迁移速率和数量以及各种因素对浸出过程的控制和影响;而且还要研究固化体自身在与浸泡溶液(浸出剂)长期接触过程中的变化及其对废物处理和处置的直接影响或间接乃至潜在的影响,进而为废物管理提供必不可少的信息. 然而在上述逐项研究项目中至关重要的是实验基本数据,尤其是长期的试验数据.本文将叙述沥青-硝酸钠固化体的4.24年的浸出试验结果.     

模拟铯废物玻璃固化物组成对浸出性能的影响

按正交试验设计方法设计了熔制模拟铯废物钛硅酸盐玻璃固化体的配料组成，制备了相应的固化体样品，按静态浸出试验方法（MCC－1）对样品进行了浸出试验，条件为去离子水、90℃、7d、样品表面积与去离子水体积之比为10m^-1。用电感耦合等离子体原子发射光谱（ICP－AES）和原子吸收（AAS）测定浸出液中各种离子的浓度。结果表明，固化体浸出液中主要含Na^ 和Si^4 ，其次是Cs^ 。固化体中ZrO2组分对浸出性能影响最大，其次是TiO2/SiO2（摩尔比）。配料中ZrO2的摩尔分数选择在1.5%-4.5%之间、TiO2/SiO2的摩尔比在0.40左右时，有利于减少Cs^ 及Na^ 、Si^4 的浸出。     

废树脂苯乙烯固化研究(Ⅱ)

本文研究废离子交换树脂苯乙烯固化的工艺条件和固化产品性能鉴定。该法能包容62％(W)废树脂。固化产品均匀、坚韧、抗压强度239kg/cm~2。在去离子水中浸泡120d,浸出率对于~(137)Cs为10~(-6)cm/d,对于~(60)Co和~(85)Sr为10~(-7)-10~(-8)cm/d。长期水浸不溶胀。承受辐照剂量>10~8rad。闪点270℃,燃点290℃左右。加热到450℃不自燃,DTA曲线上235℃前无放热峰。     

放射性化学沉淀泥浆水泥固化研究

本文探讨了用水泥同化放射性化学沉淀泥浆的可行性,研究了同化工艺配方和固化物的浸出性能。固化试验结果表明,固化体中添加20%斜发沸石可提高固化体的抗压强度,降低固化体中~(137)Cs 的浸出牢;固化体上表面复盖3mm 厚的沥青浸泡1066d,铯累积浸出分数仅6.87×10~(-2)cm;固化体外表面包复薄层沥青后,铯累积浸出分致更低,仅9.12×10~(-5)cm。     

模拟高放废物玻璃固化体在处置条件下的浸出行为研究（Ⅱ）

模拟高放废物玻璃固化体在处置条件下的浸出行为研究（Ⅱ）吴兆广，罗上庚，于承泽，盛嘉伟（中国原子能科学研究院）柳得橹（北京科技大学）关键词高放废物，玻璃固化体，浸出试验，表面分析１引言为了预测高放废物长期处置的安全性，需要研究玻璃固化体的浸出过程和浸出...     

粉煤灰基沸石固化体中Sr~(2+)、Cs~+的浸出行为

以电厂粉煤灰为原料水热制备粉煤灰基沸石,利用粉煤灰基沸石对模拟放射性废液中Sr~(2+)、Cs~+进行分离富集,在碱激发剂的作用下,以粉煤灰、粉煤灰基沸石制备地聚合物固化体,测试了所得固化体的抗压强度和抗浸出性能,并采用X射线衍射法(XRD)和扫描电镜(SEM)技术对浸出机理进行了初步探讨。结果表明,不同沸石掺量对固化体的抗压强度和抗浸出性能有很大的影响,当沸石掺量(质量分数)为20%~30%时,其抗压性能达到国家标准,浸出率和累积浸出分数均远低于国家标准限值。同时,固化体对Sr~(2+)、Cs~+阻滞效果不同,其中对Sr~(2+)的固化效果更加优异,42d浸出率最低为1.87×10~(-6)cm/d,累积浸出分数为3.3×10~(-4)cm。实验得出,粉煤灰基沸石固化体对Sr~(2+)、Cs~+具有较优异的固化效果。     

放射性废物水泥固化体浸出试验的比对

本文总结了1983年6月至7月进行的放射性废物水泥固化体浸出试验的比对工作。这次比对有6个单位7个实验室参加,比对试验条件基本上按照“放射性废物固化体长期浸出试验”(标准讨论稿)中的规定。比对结果表明,各实验室第42天宋总β累积浸出分数的平均值之间的最大偏差为±20%,同时也表明“标准讨论稿”基本上是可行的。     

Sr、Cs和Np在北山花岗岩上的吸附行为研究

<正>~(90)Sr、~(137)Cs和~(237) Np是高放废物深地质处置研究中需要考虑的关键核素。由于地下水的侵入和浸蚀,高放玻璃固化体中的~(90)Sr、~(137)Cs和~(237) Np最终被浸出并进入地下水,最终在花岗岩中迁移。~(90)Sr、~(137)Cs和~(237) Np在北山花岗岩上的吸附行为决定其在处置库远场的迁移程度,也是处置库能否安全处置高放废物的决定性因素。因此本文研究~(90)Sr、~(137)Cs和~(237) Np在北山新场、算井子、沙枣园花     

On the pH dependence of leaching of nuclear waste glasses

Two simulated nuclear waste glasses were leached for periods up to one month at 90°C in high-purity water, following standard MCC-1 test procedures. The changes in composition of the surface layers were determined using ESCA and SIMS, which analyses layers of different depths. The results are discussed with reference to the different pH values in the tests performed.     

Temperature Effect on Plutonium Leach Rate of Nuclear Waste Glass

The two kinds of nuclear waste glass with similar composition, a ²³⁸Pu-doped and nonradioactive waste glass, were leached under the ISO-test conditions at temperature between 23 and 90°C. An activation energy of 22±10kJ/mole was obtained from the initial leach rates of Pu, which was much lower than the 78±9kJ/mole obtained from those of Si, Na, Sr and Cs, It is suggested that in the initial stages of leaching. Pu is not released from the waste glass with the same mechanisms as the releases of Si, Na, Sr and Cs, but the dissolution of hydrous plutonium dioxide PuO₂·_xH₂O formed on the glass surface becomes predominant. In the long duration tests (<32d), the release of Pu appears to be affected by the solubility of PuO₂·_xH₂O remaining in the leached surface layers.     

Immobilization of simulated evaporator concentrate waste in low density polyethylene

Simulated evaporator concentrate was prepared by pre-treating sodium borate with calcium hydroxide to produce an insoluble, ulexite-like borate salt. The resultant solid waste was blended by extrusion with virgin and recycled low density polyethylene (LDPE) at 30 and 50 wt% load. Samples were evaluate to homogeneity by density determinations (ASTM D792-91), compressive strength (ASTM D695-91) and leaching behavior by accelerated leaching tests (ASTM C1308-95). Coefficient of variation of density was lower than 5% for all evaluated waste forms, indicating a satisfactory homogeneity. Compressive strengths have complied to U.S.NRC standard and were above CNEN standard limit for cement waste products if 5% strain could be considered a reasonable limit to assure the structural integrity of the material. Cumulated fraction leached after 11 days of accelerated leach test was found to be lower than 10%, and diffusion coefficients of boron have varied between 3.86 × 10^?10 and 9.06 × 10^?10 cm²/s. Boron concentrations for all materials have shown a tendency to reach an asymptotic value at the end of the test (around 0.25 and 0.7 mg/L for 30 and 50% waste load, respectively). Measured cumulated fraction leached (CFL) data with time have shown a tendency of becoming asymptotic sooner than it was predicted by the diffusion model.     

Stainless steel containers for the storage of low and medium level radioactive waste

Corrosion investigations have been performed on the austenitic structural steel AISI 304L, in comparison with the structural steel AISI 316L, in an aerated and a de-aerated solution, which was leached from low and medium level radioactive waste. On the basis of measured potentio-dynamic anodic polarization curves and the results of cyclic polarization tests, it was found that both types of steel, as well as the corresponding welds, had a high pitting potential and a high protective potential, which means that they have a strong tendency to form a compact and corrosion-resistant passive film. The repassivation capability of both types of steel prevents the occurrence of stress corrosion cracking at the level of concentration of chloride ions which corresponds to the described type of waste, whereas absorbed atomic hydrogen does not reduce toughness or cause hydrogen embrittlement. The results of the research work confirmed that it is possible to use AISI 304L structural steel for the construction of containers for the temporary, 30-year storage of low and medium level radioactive waste.     

乏燃料棒中关键核素径向分布特点及对废包壳测量准确度影响

废包壳是水法乏燃料后处理工艺首端高放固态废物的主要来源,通常采用非破坏性测量方法进行整体测量并分析其中残余的U、Pu等感兴趣关键核素的量,传统方法中多引用组件的平均燃耗作为分析计算的输入参数。但根据反应堆运行经验,乏燃料组件和乏燃料棒中燃烧生成的核素及残余U的浓度呈非均匀空间分布状态,这一特性增大了废包壳非破坏性测量分析结果的不确定度。本文采用模拟计算的方法重建乏燃料棒中感兴趣关键核素的径向分布特征,数据表明废包壳中感兴趣核素的质量浓度比与采用燃料棒平均燃耗计算的结果相差可达100%,由此建立了采用非均匀分布特性修正废包壳中感兴趣核素浓度测量分析方法。     

Leaching behavior of boric acid and cobalt from paraffin waste forms

Ninety-day leaching test was performed to investigate the leaching characteristics of paraffin waste forms that had been recently generated in large quantities at Korean nuclear power plants(KNPPs). In the case where mixing weight ratio of boric acid to paraffin was 78/22, which was adopted in concentrate waste drying system(CWDS) of KNPPs, the cumulative fraction leached(CFL) of boric acid and cobalt was about 51% and 61%, respectively. The compressive strengths of waste form before and after the leaching test exhibited 666psi(4.53MPa) and 232psi(1.58MPa), respectively. The CFLs of paraffin waste form were well expressed by diffusion-controlled dissolution model such as Gintsling-Brounshtein kinetics. The internal cross-sectional view of specimen after the test demonstrated the applicability of this unreacted shrinking core model to the leaching analysis of paraffin waste form.     

放射性废物固化体抗浸出性快速测定方法探讨

固化体的抗浸出性是放射性废物安全管理的一重要参数。目前，国内采用国标GB7023—86中的标准浸出试验方法测试固化体的抗浸出性，试验周期长。并且，国标GB14569.1—93仅对核素第42d的浸出率作了规定。这一规定不能很好反映不同固化基材、不同配方固化体间抗浸出性的差异。美国国家标准ANSI/ANS-16.1—2003采用快速浸出试验方法，并用浸出因子来表征核素的抗浸出性。本工作参照美国标准对试验结果的处理方法，对以往获得的真实或模拟放射性废物水泥固化体的浸出试验数据进行重新计算。计算结果表明，当核素累积浸出百分数小于20%时，核素的浸出率与浸出因子间存在一定的换算关系。据此，可考虑建立快速浸出试验方法和新的试验结果表述式，以较全面地判定放射性废物固化体的抗浸出性能。     

Three Year Experimental Study of Leaching Phenomena from Low Level Radioactive Cement-Based Waste Forms

Abstract

This study was carried out in order to demonstrate the safety of homogeneous cementbased waste forms (hereinafter called cement forms) for BWR's low level radioactive wastes as engineered barriers. Eighteen full scale simulated cement forms were manufactured with the addition of ¹³⁷Cs, ⁶⁶Co and ⁹⁰Sr.

Leaching tests on these forms were carried out for approximately three years. In order to study the relationship of leachability to environments at disposal sites, this Three Year Leaching Test was conducted for three kinds of environmental conditions, sea water, land water and soil. After the tests, all of these forms were cut to measure the distribution of the radionuclide's density within them. In case of the soil tests, the distribution of radionuclide in the soil was also measured.

The radionuclide leachability results reveal that ⁶⁰Co was almost completely retained in the cement forms and that ¹³⁷Cs leached from cement forms was mostly adsorbed by the soil. On the other hand, ⁹⁰Sr was not trapped in the forms and leaked through the soil around them in retard. This study also showed that simulated cement forms buried in the soil were more physically and chemically stable, and had longer term stable radionuclide containment capability than those which were submerged in sea or land water.     

Modelling of radioactive material release from waste packages in case of a fire event

Assumed incidents in the operational phase of the planned German repository Konrad for radioactive waste with negligible heat production were investigated in order to assess their possible radiological consequences. Release fractions of the radioactive substances contained in waste packages were assessed from experimental data obtained under thermal impact. They are given for halogens, tritium, ‘4C and other radionuclides and are classified according to the waste form groups and waste container classes.     

Release of Neptunium from Neptunium-Doped Borosilicate Waste Glass

The release of neptunium from a neptunium-doped borosilicate waste glass was studied at 90°C in deionized water and silicate water. The standard MCC-1 static leach method was applied to the tests for durations up to 91 days with the SA/V ratio of 10 m^?1.

The normalized elemental mass loss obtained for neptunium was about 5 g/m² for both the deionized and the silicate water leachates. This value is similar to those for currently typical borosilicate waste glasses. That is, the studied glass is comparable with the typical glasses in terms of the ability to immobilize neptunium.

The time dependence of the release of neptunium from the glass was different from those of soluble glass components such as sodium, boron and cesium, but similar to that of strontium. A part of neptunium, like strontium, probably remained in the surface layer formed on the leached glass. The neptunium species in the surface layer was predicted to be NpO₂.xH₂O(am) based on available solubility data.