Comparative population dose risks from nuclear fuel cycle closure and renewal of the commercial nuclear energy alternative in the U.S.

The debate over a large expansion of commercial nuclear energy for electricity production in the U.S., termed a “nuclear renaissance,” has most recently focused on the issues of spent nuclear fuel transportation and the closing of the once-through nuclear fuel cycle through the licensing, construction, and operation of the national spent nuclear fuel repository at Yucca Mountain, Nevada. While such a commercial nuclear energy expansion is postulated to have environmental, climate, resource utilization, and economic benefits, the fundamental issue for typical U.S. citizens about nuclear energy concerns the potential for exposure to ionizing radiation. Two generations of U.S. citizens have experienced public and media “education” that has heightened their primal fears of ionizing radiation from commercial nuclear energy. In such an environment, comparing the risks of radiation doses from commercial nuclear energy fuel cycle closure and further nuclear energy expansion with ionizing radiation population doses experienced year after year, decade after decade from non-nuclear (conventional) industries seems worthwhile for use in achieving stakeholder education and concurrence. The U.S. National Academy of Sciences (NAS) has recently performed its own landmark risk assessment of spent fuel transport in the U.S., demonstrating the guiding principles and methods for use in comparative risk assessments involving radiation dose considerations. Using the NAS assessment approach, this paper broadens its application to the full consideration of the risk of nuclear fuel cycle closure and renewal of the commercial nuclear energy alternative in the U.S., to evaluate the ionizing radiation dose risks of such expansion compared to those routinely accepted for non-nuclear industries by policy makers and the public. The 50-year collective dose risk from the total commercial nuclear fuel cycle, even if the U.S. triples its installed nuclear capacity, transports spent fuel to Yucca Mountain, and operates the Yucca Mountain repository as planned, is shown to be in the range of 3.1-million person-cSv; for five selected non-nuclear industries, the corresponding 50-year collective dose risk exceeds 1 billion person-cSv, a more than 300 times greater risk. A key step towards renewing the commercial nuclear energy alternative, then, is to use this knowledge for education of various stakeholder parties.     

Spent fuel transportation risk assessment: routine transportation

Abstract

The RADTRAN model for calculating radiation doses is based on the well understood behaviour of ionising radiation. Absorption of ionising radiation depends on the energy and type of radiation and on the absorbing material. The casks that are used to transport spent nuclear fuel have walls that absorb most of the emitted ionising radiation and thereby shield the public and the workers. For routine transportation, RADTRAN models the cask as a sphere and assumes that the longest dimension of the trailer or railcar carrying the cask is the same as that of the cask. The dose rate in Sv/h at one metre from the cask is modelled as a virtual source at the centre of a sphere whose diameter is the longest dimension of the actual spent fuel cask. People who live along the cask’s route and the people in vehicles that share the route are exposed to external radiation from the cask. The dose to workers and the public from a cask during routine transportation depends on the time that the workers or public are exposed to the cask, the distance from the cask, and the cask’s external radiation. When the vehicle carrying the cask is travelling along the route, the faster the vehicle goes, the less exposure to anyone along the vehicle’s route. Therefore, an individual member of the public receives the largest dose from a moving vehicle when he or she is as close as possible to the vehicle, and the vehicle is travelling as slowly as possible. In the present analysis, these doses are in the range of four to seven nanosieverts. Collective doses along the route depend on the size of the exposed population. In this study, such doses were of the order of 0·1 person-millisieverts. The appropriate comparison between the collective dose from a shipment of spent fuel is not a comparison between the radiation dose from the shipment and zero dose, but between the background radiation dose in the presence and absence of a shipment, e.g. 8·810096 person-Sv if there is a shipment and 8·81000 person-Sv if there is no shipment.     

Operational experience of on-site transport of spent nuclear fuel in Korea

Abstract

With the rapid development of the nuclear power programme in Korea, the amount of accumulated spent nuclear fuel has inevitably increased year by year. The spent nuclear fuel is being stored in on-site storage pools at the nuclear power plants. As the current storage capacity for spent nuclear fuel is insufficient, at-reactor storage is being expanded at each site with regard to optimisation of technical and economic factors. On-site transport between neighbouring reactors has been necessary to secure sufficient storage capacity for pressurised water reactor spent nuclear fuel assemblies. A complete on-site transport system has been developed, and so far more than 800 spent nuclear fuel assemblies have been transported using two kinds of transport cask.     

Consideration of aging mechanism influence on transport safety of dual purpose casks for spent nuclear fuel or HLW

When storage of spent nuclear fuel or high level waste is carried out in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable and can be justified and certified permanently throughout that period. The effects of aging mechanisms (e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. Consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components that cannot be directly inspected or changed without opening the cask cavity, like the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not the subject of technical aspects only but also of ‘intellectual’ aspects, like changing standards, scientific/technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of license holders and in appropriate design approval update processes. The paper addresses issues that are subject of an actual International Atomic Energy Agency TECDOC draft ‘Preparation of a safety case for a dual purpose cask containing spent nuclear fuel’.     

Assessment of integrity of a dual-purpose metal cask after long-term interim storage: seal performance under transport conditions

Abstract

Spent fuels generated in nuclear power plants (NPPs) must be stored until they can be reprocessed into new energy sources in Japan. The quantity of spent fuel stored at each NPP site is increasing, and early realisation of a method of interim storage is expected. Dual-purpose metal casks will be used which will not be reopened until they are delivered to a reprocessing plant, in order to minimize radiation exposure to personnel. The Japan Nuclear Energy Safety Organization (JNES) was established on 1 October 2003 with the mission of ensuring public safety from the potential hazards of nuclear energy. JNES have also been conducting a study of dry storage technology for interim storage. The study of verification of metal cask storage technology, the results of which from a storage and subsequent transport safety point of view are presented in this paper, was originally initiated with government funds in 1999 at the Nuclear Power Engineering Corporation, before being transferred to and conducted by JNES up to the end of 2003.     

Similarity of high-temperature gas-cooled reactor technologies and designs in Russia and USA

Historical information concerning the development of high-temperature gas-cooled reactors in the USA and Russia is presented. The reactor facilities MHTGR (USA), VG-400 (Russia), VGM (Russia), GT-MGR (Russia, USA), and at the Fort St. Vrain nuclear power plant (USA) are described. The US programs for developing innovative high-temperature nuclear reactor technologies are examined. It is shown that the Russian and US technological developments for the fuel, reactor system, energy conversion system, and fission-product transport are similar.     

我国乏燃料运输现状探讨

随着我国经济的持续发展,核能作为安全、清洁能源在我国能源战略中地位日益突出。在保证安全的前提下,我国核电机组按照国家规划合理增加,乏燃料的产量也将逐步增加。根据我国核电站乏燃料贮存及外运规则,以及我国核电站主要位于东部沿海,而乏燃料后处理厂处在西北腹地这一国情,必将面临乏燃料的大量、长距离及安全运输的问题。乏燃料运输作为联接核电站与后处理厂或最终处置场的纽带,在维持核燃料循环体系的正常运行上发挥至关重要的作用。对国内外乏燃料运输涉及的运输方式、运输容器、运输安全监管及事故应急体系等问题进行了分析和讨论,对我国乏燃料运输中存在问题的解决提出了建议。     

Aging management and post-storage transport of dual purpose cask for spent fuel

Aging management of spent fuel storage facility may follow lessons learned from literature for nuclear power plant and a review for spent fuel dry cask storage system by US NRC, DOE, by German BAM, that by Japan NISA, etc. Namely, the essence of systematic approach to aging management includes Understanding aging, Plan (Development and optimisation of activities for aging management), Do (Managing aging mechanisms), Check (Monitoring, inspection and assessment), and Act (Maintenance). The PDCA cycle will optimise the systematic approach to the aging management. An aging management programme (AMP) for the storage system over the period of extended storage will address uncertainties in the safety relevant functions of the system that may otherwise be impaired by aging mechanisms. The AMP identifies system, structure and components (SSCs) that need specific actions to mitigate aging and ensures that no aging effects result in a loss of their intended function of the SSCs, during an intended licensed period. AMPs generally include Prevention, Mitigation, Monitoring, Inspection, and Maintenance programmes. Aging management plans should ensure compliance with transportation requirements after extended storage. Potential issue would be a significant change of the transport regulations in the future. If the regulations changed significantly, a gap analysis should be performed to identify any impact to the cask safety. Compensating arrangements, if necessary, should be proposed at that time. Assuming that the regulations will not change significantly after long term storage, we will be able to renew the license both for transport and storage of the cask during the storage period. For example, in Japan, a holistic approach was established for the license of a 50 year storage and transport. In this approach, we can evaluate integrity of spent fuel, basket, etc. with respect to chemical, thermal, mechanical, and radiation factors. With this approach we will not have to open the cask lid for visual inspection of the spent fuel, basket, etc. prior to the post-storage transport.     

乏燃料公海铁联运核应急体系建设研究

公海铁联运作为解决大宗乏燃料远距离运输的最佳方案,在国际上是一种较为普遍的运输模式,如果未来我国采用该运输模式,需探索相关核应急工作思路。本文调研梳理了国内外乏燃料公海铁联运核应急相关法规标准,参考借鉴国外乏燃料运输相关实践,提出我国乏燃料公海铁联运核应急体系建设相关工作建议。     

信源可信度对核电公众接受影响的研究

公众对核电的态度受各种信息的影响,信息源可信度是说服力的重要决定因素.在我国,核安全监管机构、核电企业和专家是主要的核信息发布源,研究不同信息源可信度对公众核电态度的影响对核电的可持续发展十分重要.本研究基于源可信度理论,考察了公众对不同信源的可信度认知,构建了核电公众接受回归模型并开展实证研究.研究发现,不同源的可信...     

核电厂乏燃料组件厂内转运关键设备研制

乏燃料组件厂内转运是解决核电厂燃料水池贮存空间不足问题的方法之一。本文分析了乏燃料组件厂内转运的设计准则、安全风险,介绍了用于运输容器内破损组件检测和运输容器内组件冷却用设备的工作原理及其应用情况。应用结果表明:破损检测设备可以快速有效地检测乏燃料运输容器内是否存在破损组件;乏燃料组件冷却设备可以较为安全地冷却装有乏燃料组件的运输容器。      

Twenty Years of Experience in Spent Fuel Shipment from German Nuclear Power Plants—A View of the Competent Authority

Abstract

A survey of the transport of spent fuel in and from Germany during the last 20 years is presented. The spent fuel is now transported from the German nuclear power facilities to the reprocessing plants in France and the United Kingdom. In the past, there were also shipments to the former reprocessing plant WAK Karlsruhe (Germany), to the long-term storage facility CLAB (Sweden) and also from the former German Democratic Republic to the USSR. The transport of the spent fuel is carried out in specially built flasks requiring an extensive quality assurance programme. Due to the heavy weight of these packages, the shipments are mostly carried out by rail, but also by road and sea. An overview is given of the following matters: (i) quantities of spent fuel transport, (ii) organisation of transport, (iii) licensing matters, and (iv) reported incidents. In addition, an analysis is included of the radiation exposure for normal conditions of transport, especially of the transport workers. Difficulties and hindrances during transport are also reported.     

Criticality risks during transportation of spent nuclear fuel

Abstract

A probabilistic risk assessment (PRA) quantifies the frequency of criticality accidents during railroad transport of spent nuclear fuel casks (SFCs) in the USA. It evaluates the likelihood that undetected errors in fuel selection and/or fuel handling could result in a misloaded SFC susceptible to a criticality event following an accident during rail transport of the cask. The PRA shows that existing fuel burnup records and formal procedures for loading a SFC make the likelihood of shipping a misloaded SFC on the order of 2·6 × 10^–6 per SFC. When combined with historical evidence regarding train accidents and an estimate of the likelihood that an accident could breach and submerge a SFC, the calculated frequency of criticality is below 2 × 10^–12 over the 11 000 shipments that would be required to ship the spent fuel inventory generated by the current US fleet of nuclear reactors, assuming that they each operate for 60 years.     

秦山第三核电厂乏燃料干式贮存设施竣工环保验收监测与评价

介绍了秦山第三核电厂乏燃料干式贮存设施项目的组成及设计。参照相关法规、规范,提出了该项目竣工环保验收监测方案,给出了竣工验收监测结果,并对监测结果进行了分析评价。在分析评价的基础上,对项目是否满足竣工验收条件给出了明确结论,对存在的问题提出了改正建议。     

Spent fuel transportation risk assessment: conclusions

Abstract

The US Nuclear Regulatory Commission (NRC) has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. The study reached the following findings. First, the collective dose risks from routine transportation are vanishingly small. These doses are about four to five orders of magnitude less than collective background radiation doses. Second, the routes selected for this study adequately represent the routes for spent nuclear fuel transport, and there was relatively little variation in the risks per kilometre over these routes. Third, radioactive material would not be released in an accident if the fuel is contained in an inner welded canister inside the cask. Fourth, only rail casks without inner welded canisters would release radioactive material, and only then in exceptionally severe accidents. Fifth, if there were an accident during a spent fuel shipment, there is less than one in a billion chance the accident would result in a release of radioactive material. Sixth, if there were a release of radioactive material in a spent fuel shipment accident, the dose to the maximally exposed individual would be <2 Sv (200 rem) and would not cause an acute fatality. Seventh, the collective dose risks for the two types of extraregulatory accidents (accidents involving a release of radioactive material and loss of lead shielding) are negligible compared to the risk from a no release, no loss of shielding accident. Eight, the risk of loss of shielding from a fire is negligible. Ninth, none of the fire accidents investigated in this study resulted in a release of radioactive material. Based on these findings, this study reconfirms that radiological impacts from spent fuel transportation conducted in compliance with NRC regulations are low. In fact, this study’s radiological impact estimates are generally less than the already low estimates reported in earlier studies. Accordingly, with respect to spent fuel transportation, this study reconfirms the previous NRC conclusion that the regulations for transportation of radioactive material are adequate to protect the public against unreasonable risk.     

Microbial biofilm growth on irradiated, spent nuclear fuel cladding

A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 10³ Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.     

External radiation of a container used for dry storage of spent VVER-1000 nuclear fuel from the Zaporozhie nuclear power plant

The program MCNP (Monte Carlo Modeling of radiation transfer) is used to calculate the characteristics of external neutron and γ radiation from a ventilated dry-storage container for spent nuclear fuel. Data are obtained on the spatial, energy, and angular distribution of the neutron and γ-ray flux outside the container and the dependence of the dose rate on the storage time of the spent fuel is determined. It is shown that γ-rays make the main contribution to the dose rate on the side surface of the container and neutrons do on the cover. The computed dose rate is 1.4 times higher than the measured value on an individual loaded container at the Zaporozhie nuclear power plant.     

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。      

Residual energy release and actinide and fission product radiotoxicities during long-term storage of high burnup spent VVÉR fuel

The residual energy release and radiotoxicity of spent high burnup VVéR-1000 fuel during long-term storage is investigated as a function of time. The contributions of α, β, γradiation and radiotoxicity-the maximum admissible activity of nuclides in air and water-are taken into account in the calculations of the energy release. The data presented can be used to develop methods for handling spent nuclear fuel from prospective power reactors. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 292–296, May, 2007.     

A Screening Model for Calculating the Risks of Transporting High Level Radioactive Waste

Abstract

In transporting high level radioactive waste (including spent nuclear fuel), shippers (and sometimes carriers) need to evaluate the risks of potential radiation exposure to the public and transport workers. A simple model is presented that can be applied to nuclear waste transport risk assessments. The model considers radiation risks arising from incident free exposure, accidental release-caused exposure to on-link population, off-link population, crew, transport workers, etc. Important parameters and factors that affect the radiation dose level are grouped using the physics of the different exposure phenomena. The total radiation risk (in person-rems) is given by a linear combination of the groups of these factors, each representing a different type of exposure. The radiation exposure risk assessment is reduced to the evaluation of a single linear algebraic equation containing five distinct terms and each term containing the groups of parameters and a constant coefficient. The estimation of the values of the constant coefficients was accomplished by selecting a sample of 65 origin-destination (0-D) pairs and simulating the shipment of high level nuclear waste or spent nuclear fuel between each O-D pair, and evaluating the radiation dose risks to each group of population. RADTRAN 4 was used in the detailed assessments. The coefficient values were tested for statistical robustness using a sampling hypothesis and t-statistics. These values are presented. The simplified model presented here represents a viable and economical option as a radiological risk assessment tool, to be used in mode or route options screening.