Spent fuel transport using cask loaded outside pool

Abstract

During the last year, Sogin (the Italian company in charge for decommissioning of Italian nuclear power plants) had to implement an accelerated decommissioning plan of a EUREX spent fuel pool due to finding a water leakage into the environment from the pool. EUREX is no longer operating a pilot reprocessing plant, which some years ago became the responsibility of Sogin. There were 52 spent fuel assemblies from the Trino Vercellese PWR nuclear power plant, 48 irradiated pins from a Garigliano BWR fuel assembly, and 10 plates from an irradiated MTR fuel assembly stored in the EUREX pool, so the first step of the accelerated decommissioning plan consisted in the evacuation of this spent fuel. Considering the necessity to start the evacuation as soon as possible, Sogin decided to use an already existing cask (AGN-1) used in the past for the transport of Trino and Garigliano fuel assemblies. This cask was requalified in order to obtain a transport licence for the fuel assemblies stored in the EUREX pool according to ADR 2005 regulation. The transport license for the AGN-1 cask loaded with EUREX fuel assemblies was released by APAT (the Italian Safety Authority) in the spring of 2007. Owing to the limited capacity of the EUREX pool crane (27 t for nuclear loads) and limited dimensions of pool operational area, it was not possible to transfer the AGN-1 cask (50 t) into the pool for fuel assemblies charging. The solution implemented to overcome this problem was the loading of the cask outside the pool. A special shielding shuttle was developed and used to allow safe spent fuel transfer between the pool and the cask. This procedure avoided also the problem of excessive contamination of cask surfaces that could have occurred due to very high level of contamination of EUREX pool water if the cask had been immersed in the pool. Additional shielding devices were developed and used to reduce dose rate during cask loading operations. Although the evacuation of spent fuel assemblies from the EUREX pool was a very challenging activity due to the short time available, unfavourable space conditions inside the pool building and handling tool limitations; all loading and transport operations were performed successfully and without particular problems. Ten transports were carried out to evacuate all of the spent fuel stored in the EUREX pool. Spent fuel was transferred to the Avogadro Deposit pool. The first loading sequence started on 2 May 2007 and the first transport was performed on 6 May 2007. The tenth and last transport was performed on 21 July 2007. A dose less than 50 μSv (neutron + gamma) was measured for the most exposed operator during a complete cask loading sequence.     

Guidance for Package Approvals in the United Kingdom

Abstract

Approval is required under the transport regulations for a wide range of package designs and operations, and applications for competent authority approval and validation are received from many sources, both in the UK and overseas. To assist package designers and applicants for approval, and to promote consistency in applications and their assessment, the UK Department for Transport issues guidance on the interpretation of the transport regulations and the requirements of an application for approval and its supporting safety case.The general guidance document, known as the Guide to an Application for UK Competent Authority Approval of Radioactive Material in Transport, has been issued for many years and updated to encompass the provisions of each successive edition of the IAEA transport regulations. The guide has been referred to in a number of international fora, including PATRAM, and was cited as a 'good practice' in the report of the IAEA TRANSAS appraisal of the UK in 2002. Specialist guides include the Guide to the Suitability of Elastomeric Seal Materials, and the Guide to the Approval of Freight Containers as Types IP-2 and IP-3 Packages. This paper discusses the guidance material and summarises the administrative and technical information required in support of applications for approval of package designs, special form and low-dispersible radioactive materials, shipments, special arrangements, modifications and validations.     

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.     

Mechanical design assessment approaches of actual spent fuel and HLW transport package designs

Abstract

In recent years, BAM Federal Institute for Materials Research and Testing finalised the competent authority assessment of the mechanical and thermal package design in several German approval procedures of new spent fuel and high level waste package designs. The combination of computational methods and experimental investigations in conjunction with materials and cask components testing is the most common approach to mechanical safety assessment. The methodology in the field of safety analysis, including associated assessment criteria and procedures, has evolved rapidly over the last years. The design safety analysis must be based on a clear and comprehensive safety evaluation concept, including defined assessment criteria and constructional safety goals. In general, for new package designs, the implementation of experimental package drop tests in the approval process should be obligatory. Additionally, pre- and post-test calculations as well as components or material testing could be important. The extent to which drop tests are necessary depends on the individual package construction, the materials used and identified safety margins in the design.     

Specific transport and storage solutions for residues return programmes

Abstract

The treatment of used nuclear fuel, performed at AREVA's La Hague plant, allows recovering uranium 95% and plutonium 1% for recycling, the remaining 4% being considered as ultimate waste that can be sorted into two categories: high level activity waste (HLW) which is vitrified, and long-lived intermediate level waste (ILW) composed of structural elements of used nuclear fuel which is compacted. Whether vitrified or compacted, the waste is conditioned in the same universal and multipurpose container, named the Universal Canister. The resulting residue is named CSD-V for vitrified waste and CSD-C for compacted waste; they both remain property of the utilities and must be returned to countries of origin. In order to transport Universal Canisters in the best technical and economical conditions, TN International designs two kinds of cask solutions for its customers, either for transport only or for dual purpose, storage and transport, depending on the facility. Since the mid-1990s, TN International has transported CSD-V residues to Belgium, the Netherlands, Switzerland, Germany and Japan and is now starting the CSD-C return program. The purpose of this paper is to explain how the experience gained during the CSD-V return program has been used to optimize the CSD-C return program, in terms of cask design and licensing and of transport logistics. In some cases, casks initially developed for CSD-V transports have been adapted and in other cases, new casks are being designed specifically for CSD-C transport to increase the cask capacity and reduce the number of shipments.     

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

Japanese perspectives and research on packaging,transport and storage of spent fuel

Abstract

The present paper gives an overview of Japanese experimental studies of dual-purpose metal casks. The studies included: cask drop without impact limiters, drop of a heavy weight onto a cask due to building collapse, burial of a cask in debris from building collapse, tipping over of a cask during an earthquake, long-term containment of metal gaskets and transportability of casks after long-term storage. Most of the studies employed full-scale casks for the experiments.     

Use of Alcan's Al-B4C metal matrix composites as neutron absorber material in TN International's transportation and storage casks

Abstract

Major issues in the area of transportation and/or storage of radioactive materials are reliability and safety of engineering components. Among the functions to be undertaken, transportation and storage systems shall allow the criticality control of the transported matter, the control of its temperature, as well as the capacity to withstand the mechanical stresses due to normal, incidental and accidental conditions of use. In most cases, criticality control requires the use of an internal arrangement made of a neutron absorber material, which must also have high thermal conductivity properties to ensure the temperature control. When, as in many AREVA-TN International applications, the design takes credit of the neutron absorber material as a structural component, it must show high mechanical performance. Alcan's Al-B₄C metal matrix composites (Al-B₄C MMCs) meet all the above mentioned requirements, due to their special capability of capturing neutrons, their light weight, and their superior thermal conductivity and mechanical properties. The significant advantage of Alcan's technology is its flexibility with regards to a wide range of boron carbide contents and matrix alloys (from AA1XXX to AA6XXX). This enables the adjustment of the properties to the exact needs of the design. TN International presently uses extruded and/or rolled Al-B₄C MMC parts in several of its internal arrangements. The present paper gives an overview of the manufacture processes of Alcan's Al-B₄C MMCs, from the mixing of B₄C into liquid aluminium to the extrusion and rolling operations. It describes the methods and results for the qualification tests in terms of the neutron absorption, thermal, physical and mechanical properties of the material. Finally, details are given on the use of Alcan's MMCs as a neutron absorber with enough credit for structural material in TN International's TN24 designs.     

Analysis and design of spent fuel transport cask impact limiters

Abstract

Three Latin American countries which operate research reactors, Argentina, Brazil and Chile, have joined efforts to improve the capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half scale model for materials test reactor fuel was constructed in Argentina and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions.

In this paper both the numerical modelling and mechanical tests to select adequate shock absorbers materials are presented. Results of these tasks are used to improve the cask design.     

Advancing US public acceptance of spent fuel storage and transport: proposed outreach services for ionising radiation education support

Abstract

Expansion of commercial nuclear energy could be one of the future US sources for clean, safe, reliable and economic electricity. However, no federal policy has effectively achieved wide acceptance of nuclear energy, with such policies having fallen victim to the politics of public radiation fears from nuclear energy usage and from spent fuel storage and transport. Many experts have described the foundation of public fear as not so much nuclear technology, but the ionising radiation to which people fear they might be exposed, and this issue has been talked and written about, yet gone substantially unaddressed with respect to public education for more than three decades. In the USA, the Blue Ribbon Commission Final Report is just the latest of clear statements where such an educational need is firmly asserted. The lamentable fact is that no one has made that substantive and concerted effort to do anything about it. Indeed, the only effort seems to have been talk about ‘better communication’, with a focus on risk based communication. Any rejuvenation of public acceptance of commercial nuclear energy in the USA, including spent fuel storage and transport, can only be sustained using a different strategy from that of earlier decades. This paper highlights professional opinion on the radiation fear issue and why current industry efforts in risk based information for and communication with the public have not achieved the desired success. Education to expand the public’s understanding of comparative radiation sources and exposures while ameliorating concern about radiation from nuclear energy is the proposed alternative. In addition, here, the clear linkage between education supporting nuclear energy and facilitating necessary spent fuel storage and transport is unmistakable. The paper summarises a concept for outreach services for ionising radiation education support for application in the US, as well as key elements of such a process: its basis for success, its education content and potential implementation approaches. Comparative radiation education of the public can prove effective using current research, which has been effective in other industries. Additionally, while this discussion addresses the US situation, much of the content is likely applicable to many of the world’s nuclear energy producing countries.     

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.     

ENUN 24P乏燃料运输容器排气排水孔盖泄漏检测方法的改进北大核心CSCD

乏燃料运输容器内盖上的排气/排水孔盖作为容器包容边界之一,采用双○型金属密封圈,在容器装载乏燃料组件后需对排气/排水孔盖进行氦泄漏检测。ENUN 24P乏燃料运输容器调试过程中,发现原泄漏检测工具存在孔盖与密封面对中困难、操作复杂、易损坏密封面、增加操作人员受照风险和检测方法未考虑本底值等问题。针对以上问题,提出了改进检测工具和增加本底测量的检测改进措施,经过试验验证改进后的检测工具能有效地加快泄漏检测时间,操作简便,并减少操作人员受照剂量。改进后的检测工具也可应用于国内已有的NAC-STC型乏燃料运输容器排气/排水孔盖泄漏检测。     

Computational and experimental validation of protective features of transport packages for long-term storage of spent nuclear fuel from nuclear power plants

Abstract

The purpose of this paper is to describe the work of our Institute (RFNC-VNIITF) on the development of a unique long transport container using new manufacturing technologies. The container is designed for transport and long-term storage of spent fuel elements more than 10 m long. The disposal of such elements usually implies their cutting. This requires considerable expense in building special cutting plants. Another way is to use long protective shells to contain and store the entire fuel cell. Development of such containers has been made possible by the use of rolled-strip technology ensuring maximum strength of containers. The paper contains certain results of calculations and experimental research justifying and confirming the protective properties of the long package. To confirm reliability with minimum material consumption an available package model, manufactured by new technology, is used. This package was enhanced in the flange area to simulate strictly a flange joint of a full-scale transport package. Calculations and experimental work were performed according to IAEA regulations. During investigations the full-scale cask was considered as both a model and a prototype as it simulated completely the bottom and the flange joint 'lid-cask body' (i.e. being a model) while being shorter and thinner than the cylindrical part of the body (i.e. being a prototype). Here it is called simply a prototype. The complete simulation was not set as an objective. At the same time the cask body loading for tests simulating an air crash was done with regard to the full-scale design. The paper contains results obtained analytically on the basis of experimental data to justify the properties of a full-scale package without experimentally testing it: therefore only some photographs and experimental results for the prototype are presented here. The title of the paper indicates that the analytical study was conducted with use of numerical constants obtained experimentally. It is impossible to present all the information, including experimental research, here because of the large scope of the work performed and the limited space available.     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.     

Discrete ordinates method-like transport computation with equivalent group condensation and angle-collapsing for local/global iteration

The group condensation of the transport equation is studied in this paper so that the computational burden can be reduced. The group condensation procedure leads to equivalent total cross section that becomes angle dependent. The difficulty of angle dependency has been traditionally treated by consistent P or extended transport approximation in multigroup transport computation. However, in this study, the angle dependency of the total cross section is applied directly to the discrete ordinates equation, and the solution procedure is validated numerically on a test problem. In addition, an angle-collapsing concept is proposed for the purpose of further simplifying the group condensed problem. A local/global iteration framework is also described, in which fine-group discrete ordinates calculation is used in local problems while few-group angle collapsed transport calculation is used in the global problem, with excellent test results in the k_eff and flux estimation.     

快速响应PNIPA／Fe3O4复合凝胶辐射制备与性能研究

采用化学共沉淀法制备了四氧化三铁粒子,以N-异丙基丙烯酰胺（NIPA）、N,N-亚甲基双丙烯酰胺（MBA）和聚乙二醇（PEG）为原料,以60Coγ,射线为放射源,辐照聚合制备了多孔PNIPA／Fe3O4复合水凝胶,并对其温度敏感性、平衡溶胀率进行了表征。研究发现：磁性四氧化三铁纳米粒子在凝胶中分散均匀;凝胶具有明显的温度敏感性;致孔剂的添加提高了水凝胶的平衡溶胀率,多孔复合水凝胶失水率达96％,比普通磁性水凝胶失水率提高了约76％;致孔剂的添加使复合凝胶的最低临界相转变温度由34℃升高至37℃左右。     

Thermoluminescence properties of nanocrystalline K2Ca2(SO4)3:Eu irradiated with gamma rays and proton beam

Thermoluminescence properties of nanocrystalline K₂Ca₂(SO₄)₃:Eu prepared by ball milling technique have been studied and the nanophosphor’s suitability as an effective gamma radiation and proton beam dosimeter material has been examined. It is found that the nanophosphor is suitable for dosimetry over a very wide range of doses ∼1 Gy to 1 kGy for gamma radiation. And for proton beam the same nanophosphor shows a more or less linear response for the dose range 0.1-100 Gy. A comparative study of this nanophosphor with its corresponding microcrystalline form (prepared by solid-state diffusion method) as well as the nanocrystalline form prepared by (the more conventional) co-precipitation technique has shown that the nanophosphor prepared by the ball milling technique is in almost all respects better than the other two forms reported earlier.