芬兰开发新的钻孔技术

[欧洲核学会《核新闻网》1994年2月7日报道] 在芬兰的奥尔基洛托VLJ低、中放废物处置库,人们正在试验可用于乏燃料(高放废物)的地下最终处置的钻孔技术。 试验内容包括在地面下约60米深的研究隧道内,穿透基岩打三个孔。这项工作是由运营该处置库的芬兰国家工业动力公司     

sp地下实验室在瑞典高放废物深部处置库计划中的作用

瑞典核电站的乏燃料将在500m深的基岩坑道系统之中进行处置。瑞典政府于1984年原则上接受了代号为KBS—3的处置方案,其基本设想是将乏燃料包容在多重屏障之中,利用这些屏障在甚长时间内把乏燃料与环境隔离。在置于深部处置库之前,乏燃料组件被包容于带有钢质内衬的圆柱形铜罐之中。在处置坑道的底板上开挖一些专门的处置孔用于放置废物罐,而在每个废物罐周围放置压实的膨润土。深部处置库将分阶段建造。预计2008年将建成处置库的首期工程,其容量约为乏燃料总数的5％～10％。当所有乏燃料放置完毕后,处置库的所有巷道将用膨润土和砂回填、封闭。     

(A)sp(o)地下实验室在瑞典高放废物深部处置库计划中的作用

瑞典核电站的乏燃料将在500m深的基岩坑道系统之中进行处置.瑞典政府于1984年原则上接受了代号为KBS-3的处置方案,其基本设想是将乏燃料包容在多重屏障之中,利用这些屏障在甚长时间内把乏燃料与环境隔离.在置于深部处置库之前,乏燃料组件被包容于带有钢质内衬的圆柱形铜罐之中.在处置坑道的底板上开挖一些专门的处置孔用于放置废物罐,而在每个废物罐周围放置压实的膨润土.深部处置库将分阶段建造.预计2008年将建成处置库的首期工程,其容量约为乏燃料总数的5%～10%.当所有乏燃料放置完毕后,处置库的所有巷道将用膨润土和砂回填、封闭.     

Aospoe地下实验室在瑞典高放废物深部处置库计划中的作用

瑞典核电站的乏燃料将在500m深的基岩坑道系统之中进行处置。瑞典政府于1984年原则上接受了代号为KBS一3的处置方案，其基本设想是将乏燃料包容在多重屏障之中，利用这些屏障在甚长时间内把乏燃料与环境隔离。在置于深部处置库之前，乏燃料组件被包容于带有钢质内衬的圆柱形铜罐之中。在处置坑道的底板上开挖一些专门的处置孔用于放置废物罐，而在每个废物罐周围放置压实的膨润土。深部处置库将分阶段建造。预计2008年将建成处置库的首期工程，其容量约为乏燃料总数的5％～10％。当所有乏燃料放置完毕后，处置库的所有巷道将用膨润土和砂回填、封闭。     

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。     

乏燃料与高放废物的深层钻孔处置

<正>【本刊2015年6月综合报道】美国核废物技术评审委员会(NWTRB)2015年4月17日公布了一份题为《乏燃料与高放废物的深层钻孔处置》的研究报告,介绍了乏燃料与高放废物深层钻孔地质处置技术的潜在优势及其面临的技术挑战。对核电厂乏燃料和乏燃料后处理厂产生的固体高放废物进行深层钻孔处置,这一技术方案最早可追溯至20世纪70年代,在20世纪90年代和21世纪初又被再次提起(主要是在瑞典和英国)。最近,这一技术又     

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。     

低放固体废物核素活度浓度上限值的推导

介绍了近地表处置设施在300 a监护期前及其以后的任何时间,公众个人及闯入者通过各种途径的受照剂量分别小于剂量限值时所要求的低放固体废物核素活度浓度上限值的推导方法及过程。以我国放射性废物近地表处置的基本安全要求为前提,并以遥田处置场和北龙处置场为对象,分析处置设施关闭后各景象的核素迁移过程和照射途径,建立各景象核素迁移的概念模型、数学模型,并计算各景象对人类产生的照射剂量。假设核素活度浓度与剂量之间呈线性关系,推导满足剂量准则下各景象各放射性核素的活度浓度上限值,选择最小的上限值,从而确定出低放固体废物各核素活度浓度上限值的量级。     

乏燃料的干法分离技术

【日本《能源》2000年3月刊第48～50页报道】 分离转化为短半衰期核素 核电站(轻水堆)使用的核燃料(乏燃料)以被称为普雷克斯法的工艺进行后处理。通过这种方法就可从乏燃料中回收铀(U)、钚(Pu),这些回收物质又可再次在燃钚轻水堆和快堆中作为燃料利用。另一方面,把产生于后处理工序的高放废液做成玻璃固化体,然后对其进行深层处置。 高放废液中含有镎(Np)、镅(Am)、锔(Cm)等半衰期非常长的超铀元素,它们和半衰期短的其他核素一并被处置。现正在对高放玻璃固化体处置的安全性进行评估,这将成为今年处置工作的主体。如果能够除去来自高放废…     

加完成处置库候选场址的首个试验性钻孔

正【英国《国际核工程》网站2018年1月23日报道】加拿大核废物管理组织(NWMO)2018年1月4日在伊格纳斯(Ignace)地区完成首个1000米深试验性钻孔的钻探。该钻孔位于伊格纳斯以西约35千米的Revell Batholith岩层中,其所在场址是加5个乏燃料深层地质处置库候选场址之一。废物管理组织表示,由地球科学、环境、工程和处置库安全专家组成的工作组将分析岩石样本,未来可能需要进一步钻孔。     

Effects of α-radiation on a direct disposal system for spent nuclear fuel – (2) review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North America

The Japanese geological disposal programme has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter “direct disposal of SF”) as an alternative management option other than reprocessing followed by vitrification and deep geological disposal of high-level radioactive waste (HLW). In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of α-radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.     

Effects of α-radiation on a direct disposal system for spent nuclear fuel – (1) review of research into the effects of α-radiation on the spent nuclear fuel,canisters and outside canisters

The Japanese geological disposal programme has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter “direct disposal of SF”) as an alternative management option other reprocessing followed by vitrification and geological disposal of high-level radioactive waste. In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with decomposition of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate of SF and the solubility of radionuclides. Focusing especially on the effects of α-radiation in safety assessment, this study has reviewed research into the effects of α-radiation on the SF, canisters and environment outside the canisters.     

钻孔处置简介

介绍了钻孔处置概念及3个国家的钻孔处置设施。对于废放射源和少量放射性废物,钻孔处置将是一个安全、经济和有效的最终处置方案。     

The effect of organic retarders on grout thickening and setting during deep borehole disposal of high-level radioactive waste

Deep borehole disposal (DBD) is being increasingly seen as a viable and potentially superior alternative to comparatively shallow mined repository concepts for disposal of some high-level radioactive wastes. We report here details of proof-of-concept investigations into the use of cementitious grouts as sealing/support matrices for use in low temperature DBD scenarios. Using the cementitious grout to fill annular space within the disposal zone will not only support waste packages during placement, but will also provide a low permeability layer around them which will ultimately enhance the safety case for DBD. Grouts based on Class G oil well cement are being developed. The use of retarders to delay the accelerated onset of thickening and setting (caused by the high temperature and pressure in the borehole) is being investigated experimentally. Sodium gluconate and a polycarboxylate additive each provide sufficient retardation over the range 90–140 °C in order to be considered for this application. Phosphonate and sulphonate additives provide desirable retardation at 90 °C. The additives did not affect grout composition at 14 days curing and the phases formed are durable at elevated temperature and pressure.     

化学种态分析软件CHEMSPEC最新进展

我国高放废物地质处置研发工作已经步入建造地下实验室阶段。地下实验室建造安全评价和未来的处置库性能评价中均需要关键放射性核素在相应深部地质条件下的扩散和迁移参数,而关键核素的扩散和迁移参数与核素在相应水岩体系中的化学种态密切相关。为满足我国核设施退役治理工作的需要,尤其是我国高放废物地质处置相关安全评价的需要,北京大学核环境化学课题组于2008年开始编写具有完全著作权的化学种态分析软件CHEMSPEC。经过多次修改和完善,目前已经具备了较好的计算功能。本文介绍该软件在表面配合模型、数据库补充和程序优化方面的最新进展,以实例形式介绍该软件的新性能,以期为我国相关实验研究者使用该软件提供参考。     

对法国高放废物处置安全审评的分析研究

介绍了法国高放废物处置研究现状和规划,对法国高放处置场的审评技术单位法国核与辐射安全研究院（IRSN）所开展的高放处置安全研究和审评工作及其提出的审评原则和审评要点进行了分析研究,并对我国的高放处置安全审评工作提出了建议。     

废放射源长期贮存方案初步研究

由于我国国内近期无法对多数长寿命废放射源和高活度废放射源实施地质处置或钻孔处置,可行的管理措施应该是将其整备后进行长期贮存。根据国内外废放射源贮存实践经验,结合对国内废放射源贮存现状及存在问题分析,建议首先对贮存前的废放射源实施整备,整备方案考虑安全原则、回收原则和废物最小化原则,整备后的废放射源以深井贮存方案为主,辅以地坑贮存,并建议运输容器和长期贮存容器应实现标准化和系列化。     

Am(Ⅲ)在地下水中的溶解度

溶解度对于验证地质化学程序的有效性非常重要,而地质化学程序是迁移模型的一部分。241 Am和243 Am是高放废物深地质处置研究中须重点考虑的核素,Am溶解度的准确测定将为Am的深地质处置安全评价提供可靠的数据。本文采用过饱和法测定了低氧高纯氩气氛中,不同恒定温度下,Am(Ⅲ)在甘肃北山花岗岩地下水中的溶解度,并探讨了温度、硫酸根浓度、碳酸氢根浓度及pH值对溶解度的影响。结果显示,Am(Ⅲ)的溶解度随温度和pH值的升高而减小,随起始硫酸根浓度和碳酸氢根浓度的增大而增大。虽然由于地下水中阴离子的配合作用使Am(Ⅲ)在地下水中的溶解度有增大趋势,但由于处置库近场环境中的温度较高,偏碱性地下水中Am(Ⅲ)的溶解度在温度和pH值升高的影响下大幅减小,最终有利于处置安全。     

Development of the Nirex Generic Transport Safety Assessment to Assist in the Provision Advice of Waste Packaging

Abstract

The current Nirex mission is to provide the United Kingdom with safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials. As part of this role, Nirex has developed a phased deep geological disposal concept which is defined by six ‘generic documents’ that describe systems, processes and safety assessments that are not specific to anyone location or geology. These generic documents give access to detailed information about the ideas and approaches that underpin the phased disposal concept, and have been published with an invitation to enter into dialogue with Nirex regarding these issues. The generic documents identify the requirements for an integrated transport system that would be necessary for the management of the intermediate-level (ILW) and low-level (LLW) wastes within Nirex's remit — the so-called reference case volume. This has involved Nirex in the development of transport hardware and associated safety reports and modelling and assessment tools for transport system logistics and system safety. Although the phased disposal concept is only one option for the long-term management of waste, the integrated transport system and associated modelling tools are likely to be of equal relevance to other options. The safety assessment of the generic transport operation for the movement of ILW and LLW waste from waste producers' sites to a future radioactive waste disposal facility is described in one of the generic documents — the generic transport safety assessment (GTSA). The GTSA demonstrates that the transport operation is compliant with Nirex safety principles, and that the nuclear and non-nuclear risks to the public and workers from routine transport and from accidents are acceptable. This paper describes the types of risk that are calculated, and discusses the data requirements and calculation methodology. The verification and validation methodology is outlined, together with a discussion of the results and a comparison of the risks with the Nirex dose and risk targets. In addition, this paper also describes how the methodology of the GTSA has been developed into an innovative software tool, TranSAT, which is routinely used as part of the packaging waste advice service offered by Nirex.     

Concept of a Korean Reference Disposal System for Spent Fuels

A deep geologic disposal system for the spent fuels from nuclear power plants has been developed since this program was launched in 1997 in Korea. In this paper, the concept of a Korean reference high-level waste (HLW) vertical disposal system (KRS-V1) is described. Though no site for the underground repository has yet been specified in Korea, a generic site with a granitic rock is considered for a reference spent fuel repository design. The depth of the repository is assumed to be 500 m. The repository consists of a disposal area, a controlled area, and an uncontrolled area. The disposal area consists of disposal tunnels, panel tunnels, and a central tunnel. In the controlled area and the uncontrolled area, there are technical rooms and tunnels and/or shafts to connect them to the ground level, respectively. The repository will be excavated, operated, and backfilled in several phases including an underground research laboratory (URL) phase. The result of this concept development will be used for an evaluation of its feasibility, analyses of its long-term safety, information for public communication, and a cost estimation, among others.