GOLDSIM models of long-term radiation impact of conditionally cleared radioactive material

Management of waste materials from the decommissioning of nuclear installations offers opportunities for optimization. Large amounts of waste materials with low contents of various radionuclides could be recycled to save financial resources or repository capacity. The increase of the share of recycled materials compared to the current practice could be accomplished by applying the conditional clearance concept. Conditional clearance, as up to now is an unproven theoretical concept, demands utilization of the cleared material for the previously defined purpose (e.g., building construction). Safety studies needed for realization of this practice have to prove that conditionally cleared material will not cause radiation impact exceeding levels prescribed in health and safety regulations. Safety studies assess radiation impact during all manipulations with low level radioactive material (e.g., melting, component manufacturing, building of construction, etc.) as well as its impact on inhabitants living near the construction built using conditionally cleared material.The article is focused on modeling and calculation of long-term radiation impact on inhabitants living near the constructions. Models (scenarios) of various building applications were simulated using GOLDSIM software with Radionuclide Transport Module. Scenarios were selected according to information from the civil engineering business to cover the types of buildings most suitable for application of conditionally cleared material. The results of the calculations showed that conditional clearance represents no significant safety issue in the long-term. Calculated individual effective doses received by inhabitants did not exceed the given dose constraint (10 μSv/year) in case of any scenario evaluated. Detailed and transparent studies of the long-term impact of conditionally cleared materials are important especially for winning of public acceptance.     

Assessment of radioactive wastes from a DCLL fusion reactor: Disposal in El Cabril facility

Under the Spanish Breeding Blanket Technology Programme TECNO_FUS a conceptual design of a DCLL (Dual-Coolant Lithium–Lead) blanket-based reactor is being revised. The dually cooled breeding zone is composed of He/LiPb and SiC as material of the liquid metal flow channel inserts. Structural materials are ferritic-martensitic steel (Eurofer) for the blanket and austenitic steel (SS316LN) for the vacuum vessel (VV) and the cryostat.In this work, radioactive wastes are assessed in order to determine if they can be disposed as low and intermediate level radioactive waste (LILW) in the Spanish near surface disposal facility of El Cabril. Also, unconditional clearance and recycling waste management options are studied.The neutron transport calculations have been performed with MCNPX code, while the ACAB code is used for calculations of the inventory of activation products and for activation analysis, in terms of waste management ratings for the options considered.Results show that the total amount of the cryostat can be disposed in El Cabril joined to the outer layer of both VV and channel inserts, whereas only concrete-made biological shield can be managed through clearance and none of the steels can be recycled. Those results are compared with those corresponding to French regulation, showing similar conclusions.     

在短寿命放射性医疗废物中发现长寿命杂质核素

为了确定短寿命放射性医疗废物能否清洁解控及短期的衰变贮存是否有效,本研究使用γ谱仪分析已放置了10个半衰期以上的低水平含99Tcm的放射性废物中残留的放射性核素,通过核素的全能峰定性,通过感兴趣区的净峰面积定量.经过了衰变贮存,虽然所有核素的活度浓度均低于相应的清洁解控水平,但在99Tcm废物中检出了长寿命核素137Cs、155Eu、123Tem、154Eu,其中137Cs、155Eu分析为99Tcm的母体99Mo生产过程中产生的杂质核素进入99Tcm所致,123Tem、154Eu可能为99Mo靶中杂质核素衰变而成.半衰期最长的长寿命核素137Cs的半衰期为30a,短期的衰变贮存并不能使这些长寿命核素活度显著减少,可见控制放射性药物的核纯度具有重要意义,可防止后期产生的医疗废物处置复杂化.     

Shallow-Land Buriable PCA-Type Austenitic Stainless Steel for Fusion Application

Neutron-induced activity in the PCA austenitic stainless steel is examined, when used for first-wall components in a DEMO fusion reactor. Some low-activity definitions, based on different waste management and disposal concepts, are introduced.

Activity in the PCA is so high that any recycling of the irradiated material can be excluded. Disposal of PCA radioactive wastes in Shallow-Land Buriable (SLB) is prevented as well. Mo, Nb and some impurity elements have to be removed or limited, in order to reduce the radioactivity of the PCA. Possible low-activity versions of the PCA are introduced (PCA-la); they meet the requirements for SLB and may also be recycled under certain conditions.     

The role of pyro-processing in decreasing disposal cost in Korea

A comparative analysis regarding the disposal cost of HLW (High-Level Waste) from 20,000-ton PWR nuclear fuel, focusing on pyro-processing and direct disposal, was conducted in this study. A cost estimation of the major cost drivers in disposing of pyro-processed waste revealed that canisters would cost 67.32 MEUR and that the disposal holes and disposal tunnels would require about 11.2 MEUR for excavation. These estimates amount to 1/16 and 1/55 of the costs for direct disposal of PWR spent fuels, respectively. These significant disposal cost savings in pyro-processed radioactive waste result from a significant reduction in the amount of radioactive waste to be disposed of thanks to the recycling in a fast reactor.     

Some technological problems of fusion materials management

Within the framework of the International Energy Agency Program on Environmental, Safety and Economic Aspects of Fusion Power, an international collaborative study on management of fusion radioactive materials has been carried out to examine the back-end of the materials cycle. The strategy for handling fusion activated materials calls for three potential schemes: clearance, recycling and disposal. There is a growing international effort to avoid the underground disposal, for fusion in particular. Plasma facing components (divertor and blanket) normally contain high radioactivity and are not clearable. As clearance of sizeable components (such as biological shield, cryostat, vacuum vessel, and some constituents of magnets) is highly desirable, we identified the source of radioisotopes that hinder the clearance of these components and investigated the impact of impurity control. Another study assessed radioactivity build-up under repeated use of the divertor made of W–La₂O₃ alloy. Effect of impurities on activated materials management is illustrated by the examples of carbon-14 generation and impurities activation in concrete of biological shield. We think that consideration of activated materials management scenarios presented in this paper by example of blanket and divertor replacement is of interest as well.     

Lifecycle of palladium in Japan: for setting clearance levels of 107Pd

Research on the recycling of useful elements in spent nuclear fuels of light-water cooling reactors is progressing. Palladium (Pd), a fission product in high-level radioactive waste, is one of the elements that can be recovered and recycled. Even if the Pd recycling system is successfully established, ¹⁰⁷Pd contamination will be unavoidable. In this report, we review the lifecycle of Pd (applications, environmental levels, and human exposure) with respect to setting clearance levels for ¹⁰⁷Pd. The major applications of Pd in Japan are catalysts for automobile exhausts and chemical industries, followed by dental prostheses and electrical devices. The World Health Organization reported a daily uptake of Pd from drinking water and food of 0.03 and <2 μg person^?1, respectively. The uptake of a person with a dental prosthesis containing Pd might reach up to 15 μg day^?1. The Pd uptake by the public via inhalation of particulate matter from automobile exhaust catalysts is not as large, even in urban areas, although this industrial application is responsible for the largest portion of Pd released into the environment globally. The data presented here will be useful for setting clearance levels for ¹⁰⁷Pd.     

Impact on Geologic Repository Usage from Limited Actinide Recycle in Pressurized Light Water Reactors

A project has been conducted as part of the U.S. Department of Energy Advanced Fuel Cycle Initiative to evaluate the impact of limited actinide recycling in light water reactors on the utilization of a geologic repository where loading of the repository is constrained by the decay heat of the emplaced materials. In this study, it was assumed that spent PWR fuel was processed, removing the uranium, plutonium, americium, and neptunium, along with the fission products cesium and strontium. Previous work had demonstrated that these elements were responsible for limiting loading in the repository based on thermal constraints. The plutonium, americium, and neptunium were recycled in a PWR, with process waste and spent recycled fuel being sent to the repository. The cesium and strontium were placed in separate storage for 100–300 years to allow for decay prior to disposal. The study examined the effect of single and mutliple recycles of the recovered plutonium, americium, and neptunium, as well as different processing delay times. The potential benefit to the repository was measured by the increase in utilization of repository space as indicated by the allowable linear loading in the repository drifts (tunnels). The results showed that limited recycling would provide only a small fraction of the benefit that could be achieved with repeated processing and recycling, as is possible in fast neutron reactors.     

Final Status of the Decommissioning of Research Reactors in Korea

A decommissioning project for the Korean Research Reactors KRR-1 and KRR-2 was started in 1997 and had been carried out with the goal of completion by the end of 2008. All the facilities were dismantled and the building surfaces decontaminated. The radioactive waste was packed into 200 liter drums and 4m³ containers and temporarily stored on site until their final disposal at the national repository facility. Some of the releasable waste was freely released and utilized for non-nuclear industries. The assessment of the residual radioactivity was carried out according to the Multi Agency Radiation Site Survey and Investigation Manual (MARSSIM), and accordingly, the safety of the site release was verified. The site and buildings will be cleared for reuse for non-nuclear purposes after a review of the assessment. In this paper, the final status of the decommissioning of research reactors in Korea including dismantlement processes, waste management, and a final assessment for unrestricted use of the site and buildings as the final goal of the decommissioning project will be described.     

Shielded containers for radioactive waste using recycled contaminated metals

A family of shielded containers constructed of recycled contaminated lead and stainless steel from the US Deptartment of Energy (DOE) stockpile have been developed as a joint effort by the Idaho National Engineering and Environmental Laboratory and Lockheed Martin Energy Systems-Oak Ridge. The containers were designed primarily for the transportation and storage of 30- or 55-gallon drums of remote handled transuranic solid waste (RH-TRU). The use of recycled construction materials fulfills two purposes: (1) waste transportation and storage and (2) a reduction in the quantity of contaminated metals that require both storage and disposal. Currently, the DOE complex has millions of pounds of contaminated lead and stainless steel. Non-linear, dynamic finite element analyses were used to simulate the drop tests onto a rigid surface required for US Deptartment of Transportation (DOT) 7A Type A certification. During and after these tests, the container contents must remain within the container and shielding must not be reduced. Loads in the major connections and strains in the stainless steel and lead were monitored as a function of time during dynamic impact analyses for three simulated drop orientations. Results were used to optimize the final design. Stainless steel and lead were found to have maximum strains well below ultimate levels except at an impact corner where additional deformation is acceptable. The predicted loads in the connections indicate that some yielding will occur but the containment and shielding will remain intact. The results provide assurance that the containers will pass the required DOT certification tests. The methods used can be applied to other waste shipping containers allowing for optimization of designs without the expense of actual impact testing.     

Development of a treatment process and immobilization method for the volume reduction of uranium-bearing spent catalysts for final disposal

The uranium catalyst had been used in several industrial fields. The spent uranium catalyst became problematic radioactive waste awaiting a management strategy for the final disposal. This work studies a process to greatly reduce the volume of a spent uranium catalyst waste and the generation of a suitable waste form for final disposal. The process consists of several steps such as selective dissolution of the SiO₂ catalyst support, precipitation of dissolved silicon followed by its purification for release, treatment of uranium-laden wastewater generated during the process, and immobilization of the final uranium-bearing astes for disposal. Based on bench scale-level experiments, the process was confirmed to be effective to reduce the volume of the uranium catalyst waste. The final volume reduction yield obtained in this work was over 80% from the volume of the initial uranium catalyst waste. The radioactivity of the secondary wastes, namely, the recovered silica and effluent generated from the process, was confirmed to be sufficiently managed for clearance with meeting the discharge criteria in Korea. The process could achieve the maximum volume reduction of the uranium catalyst waste to be transferred to a disposal site, with the by-products from the process being released, meeting discharge criteria in view of both nuclear and non-nuclear environmental regulations.     

Parametric Survey for Benefit of Partitioning and Transmutation Technology in Terms of High-level Radioactive Waste Disposal

Benefit of implementing Partitioning and Transmutation (P&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. First, the amount and characteristics of HLWs caused from UO₂ and MOX spent fuels of light-water reactors (LWR) were evaluated for various reprocessing schemes and cooling periods. The emplacement area in the repository site required for the disposal of these HLWs was then estimated with considering the temperature constrain in the repository. The results showed that, by recycling minor actinides (MA), the emplacement area could be reduced by 17–29% in the case of UO₂-LWR and by 63–85% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of 241Am which was a long-term heat source. Further 70–80% reduction of the emplacement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types. To achieve this benefit of P&T, however, it is necessary to confirm the engineering feasibility of these unconventional disposal concepts.     

Status of the FRG activities on direct disposal of spent LWR-fuel

As stipulated by the German Atomic Energy Act, reprocessing is the reference waste management route for LWR's in the Federal Republic of Germany (FRG).Spent fuel disposal without reprocessing is being developed to technical maturity for those fuel elements for which reprocessing is either technically not feasible or economically not justifiable. The reference concept for direct disposal is the emplacement of large and heavily-shielded casks in drifts of a repository mine located in a salt dome. Moreover, a back-up solution is being pursued which results in smaller canisters which are emplaced in boreholes.The mining authorities have pointed out that the feasibility of direct disposal is to be demonstrated before a license for industrial scale deployment could be granted. Demonstration tests are necessary in the following areas: shaft transport of large and heavily shielded casks, handling of the casks in the repository and thermal and rock mechanics investigations with respect to the drift emplacement concept.The results of the demonstrations tests as well as the results from layout and optimization studies for a common repository for both reprocessing waste and spent fuel will be available early enough to be incorporated into the licensing procedure for the FRG's first repository for heat-generating nuclear wastes. This means that direct disposal of spent fuel not suitable for reprocessing could be introduced in the future in addition to the reprocessing and recycling waste management concept.     

Radioactive waste management plan for the PBMR (Pty) Ltd fuel plant

The Pebble Bed Modular Reactor (Pty) Ltd Fuel Plant (PFP) radioactive waste management plan caters for waste from generation, processing through storage and possible disposal. Generally, the amount of waste that will be generated from the PFP is Low and Intermediate Level Waste. The waste management plan outlines all waste streams and the management options for each stream. It also discusses how the Plant has been designed to ensure radioactive waste minimisation through recycling, recovery, reuse, treatment before considering disposal. Compliance to the proposed plan will ensure compliance with national legislative requirements and international good practice.The national and the overall waste management objective is to ensure that all PFP wastes are managed appropriately by utilising processes that minimize, reduce, recover and recycle without exposing employees, the public and the environment to unacceptable impacts. Both International Atomic Energy Agency (IAEA) and Department of Minerals and Energy (DME) principles act as a guide in the development of the strategy in order to ensure international best practice, legal compliance and ensuring that the impact of waste on employees, environment and the public is as low as reasonably achievable. The radioactive waste classification system stipulated in the Radioactive Waste Management Policy and Strategy 2005 will play an important role in classifying radioactive waste and ensuring that effective management is implemented for all waste streams, for example gaseous, liquid or solid wastes.     

低放废物近地表岩洞处置实践

利用辽宁兴城退役铀矿山废矿井的＋43m中段处置退役的镭厂低放废物.对该处置设施基本情况、需处置的废物来源、整备设施及运输、废物处理处置方法及环境影响进行了分析.该处置设施已接受8 747m3废物、尚有2 000m3的容量空间,待剩余空间接受其他来源的废物后实施处置场封闭.     

ITER中国液态锂铅实验包层模块活化特性分析与废料处理

使用中子学程序系统VisualBUS和活化数据库EAF-99对DFLL-TBM的高级子模块DLL-TBM的活化特性进行了计算和分析,包括DLL-TBM各部件在不同停堆时间的活度、衰变余热和剂量率.活化计算所需要的三维中子能谱通过MCNP/4C中子/光子输运程序和国际原子能机构发布的FEND1.0数据库计算得到.在活化计算分析的基础上,参照欧洲聚变堆安全和环境评估(SEAFP)策略中有关核废料的处理标准评估了TBM各区材料在退役后的废料处理工作,包括核废料应该采用何种适当的方式进行处理及其被完全清除干净的可行性.     

美国放射性废物最终处置政策及技术路线

反应堆安全和核废物安全处置被认为是影响今后核能事业发展的两大障碍。自１９８０年以来，美国颁布了３个关于核废物处置的政策法令，对放射性废物的安全处置的要求及责任做出了明确规定。文章介绍了美国放射性废物处置的政策、技术路线及现状。对乏燃料及高放废物的处置，美国采取了比较慎重的态度，进行了各种方案的比较，虽然已有基本轮廓，但仍在探索之中。文章还介绍了高放废物处置中存在的一些有争议的重大问题和倾向性意见。     

重水研究堆退役废物再利用研究

实现废物再利用是废物最小化的重要措施之一,从废物流中将有潜在利用价值的物料分离出来实现再利用可大幅减少对环境的影响。本文以中国原子能科学研究院重水研究堆退役为实例研究了放射性废物再利用问题。通过全面分析和计算重水研究堆在退役期间产生的各类废物,得出具有一定数量的物料有潜在的利用价值,可直接或经适当处理后再利用在其他行业领域中。研究表明,通过采取废物最小化控制措施（如废物分类和废物流分离等）,采用适当的去污技术和执行清洁解控要求,至少可使重水研究堆退役过程中产生的几十吨钢铁、10 t铝材和5 t重水实现再利用。     

Nuclear waste management in the US: The Nuclear Waste Technical Review Board's perspective

The US program for the management and disposal of commercial spent nuclear fuel and high level waste is in a period of potential programmatic, regulatory, and legislative change. Proposals currently being considered by the US Congress would authorize the development of a storage facility as soon as possible adjacent to the potential repository site at Yucca Mountain. The legislation also would establish regulatory requirements for a permanent repository at an individual dose limit of 1 mSv year⁻¹ (100 mrem year⁻¹) for the average person living near the repository. Concurrently, the fiscal year 1996 appropriation to characterize the Yucca Mountain site has been reduced by approximately 40%. These initiatives portend possible changes in the focus of the US program, including a fundamental shift in priority from permanent disposal to temporary storage, and a change in the approach to licensing a potential repository at the Yucca Mountain site. This paper provides the perspective of the members of the Nuclear Waste Technical Review Board on the impact these developments could have on the future of the US program. It discusses the Board's opinion on how to address the issues these and other developments raise in a way which moves the US civilian radioactive waste management program forward.     

Utilization of Crushed Radioactive Concrete for Mortar to Fill Waste Container Void Space

Minimizing the volume of radioactive waste generated during dismantling of nuclear power plants is a matter of great importance. In Japan waste forms buried in a shallow burial disposal facility as low level radioactive waste must be solidified by cement or other materials with adequate strength and must provide no harmful opening. The authors have developed an improved method to minimize radioactive waste volume by utilizing radioactive concrete for fine aggregate for mortars to fill void space in waste containers. Tests were performed with pre-placed concrete waste and with filling mortar using recycled fine aggregate produced from concrete. It was estimated that the improved method substantially increases the waste fill ratio in waste containers, thereby decreasing the total volume of disposal waste.