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Warrington, United Kingdom

Egarr D.A.,MMI Engineering Ltd | Burt D.,MMI Engineering Ltd | Cory A.,International Nuclear Services Ltd
Packaging, Transport, Storage and Security of Radioactive Material

The Pacific Grebe is one of Pacific Nuclear Transport Ltd's fleet of ships and is managed by International Nuclear Services Ltd. She is an Irradiated Nuclear Fuel (INF) 3 certified vessel under the INF code of the International Maritime Organization. This means the Pacific Grebe is certified to carry irradiated nuclear fuel, vitrified residues or plutonium with no restriction of the maximum aggregate activity of the materials. The Pacific Grebe entered full service in 2011, commencing a maiden operational voyage to Japan on August 3 with a shipment of vitrified residue from spent fuel reprocessing. This was the first shipment of vitrified residue following the tsunami that struck Japan's northeast coast in March 2011. The residue product was transported inside TN28VT flasks within the Grebe's cargo holds. There are four holds, each capable of transporting a number of flasks. To support the safety case, thermal assessments were undertaken to determine the following: resin temperatures in the TN28VT flask to demonstrate whether the resin is maintained below the limiting temperature when transporting the maximum heat load of 40.88 kW; the hold air temperature as measured by two thermocouples, which would be used as a control in the Shipment Approval Certificate; and the average hold air temperature to determine whether it was within the limit of 55°C specified by the INF code. The paper describes the work undertaken, which required computational fluid dynamics with conjugate heat transfer and a bespoke onedimensional heat transfer calculation to determine the thermal boundary conditions at the hold walls. © 2013 W. S. Maney & Son Ltd. Source

Nuttall M.,Sellafield Ltd. | Cory A.R.,International Nuclear Services Ltd
Packaging, Transport, Storage and Security of Radioactive Material

This paper discusses aspects of legacy breeder fuels, which can be important to the transport criticality safety case and which are different from fuels from thermal reactors. The issues that are addressed include inventory determination and uncertainty, fuel damage from impact, and corrosion and validation of the criticality calculations. The paper is based on a real project. The paper also describes the iterative nature of the design process, showing how it benefited from a large computer cluster and a parameterised criticality code. © W. S. Maney & Son Ltd 2013. Source

Nuttall M.,Sellafield Ltd. | Purcell P.C.,International Nuclear Services Ltd
Packaging, Transport, Storage and Security of Radioactive Material

Many fissile material transport packages incorporate boron as a neutron poison [predominantly as a boron-metal matrix composite (MMC)] to maintain criticality safety. The MMC contains a prespecified proportion of Boron 'homogeneously' distributed throughout the metal matrix. Uncertainty arises as to the meaning of 'homogeneous', in the context of providing the neutron absorbing properties assumed in the safety case for a package and the potential effect on criticality. During criticality analyses, it is usual for homogeneous materials to be assumed as uniform, without irregularities, with equal properties in all directions, at the atomic level. Since the boronated materials are 'alloys', the constituents are not chemically combined but finely mixed, with the boron particles, of various sizes, viewable via a microscope. This signifies that at the atomic level, the material is not homogeneous, but a heterogeneous mixture with size and distribution of boron within the MMC being not strictly uniform. Depending on variation in boron size and distribution, the neutron absorption capability of the MMC could be reduced, with consequential reduction in criticality safety margins. During the recent manufacture of a MOX fuel transport package, which included a boron carbide (B4C)-aluminium alloy, material quality tests were performed to examine the structure of the material. Although the tests confirmed the size and distribution of B4C within the MMC to be such that it could be classified as 'homogeneous', supporting calculations were completed to determine potential effects on criticality of a heterogeneous versus homogeneous neutron absorbing material. To determine the change in neutron absorbing properties of the MMC due to atomic versus microscopic assumptions, the Monte Carlo neutronics code MONK was used to extensively examine the effects of a heterogeneous MMC with a boron particulate of various sizes and proportions compared to a homogeneous boron distribution within the material. The paper presents, from a criticality viewpoint, the effects of heterogeneity versus homogeneity for boronated poisons in a particular fissile material transport package. It also emphasises the benefits of the utilisation of software/computing developments during the calculational process, which enable wide ranging surveys over many variables to be completed quickly and efficiently. © W. S. Maney & Son Ltd 2013. Source

Competent authorities worldwide have the power to issue an extension to an existing package design certificate with the process appropriately termed timely renewal in the USA. An extension or timely renewal allows competent authorities sufficient time to challenge and review the safety of the package design, applicants to respond to questions raised and user's to continue to use the package. However, when requests are made to validate a design certificate of foreign origin, no similar processes or conventions exist, giving little flexibility to competent authorities. Several issues arose during recent European validation applications that had the potential to delay validation issue. If any significant delay had occurred, it had the potential to adversely impact transports. The present paper aims to encourage discussion surrounding the certificate validation process legal framework and promote the development of working practices and processes that readily support the safe, efficient and effective extension of existing validations to facilitate continuing transports. Several approaches are considered, with the proposal that the US timely renewal (or a combination of timely renewal with overlap provision) would seem be the ideal model to resolve this issue. It is also recommended that further discussion on this subject takes place via the invited contribution of members of the World Nuclear Transport Institute and the World Nuclear Association, with the purpose to, if deemed necessary, develop a UK proposal in connection with the UK regulator for change during the next International Atomic Energy Agency TS-R-1 or TS-G-1?1 review cycle. INS makes no representations or warranties or any kind concerning this article, express or implied, statutory or otherwise, including without limitation, warranties of accuracy or the absence of errors. © W. S. Maney & Son Ltd 2013. Source

Egarr D.,MMI Engineering Ltd | Duda A.,MMI Engineering Ltd | Ganeshalingam J.,MMI Engineering Ltd | Cory A.,International Nuclear Services Ltd | Acker B.,International Nuclear Services Ltd
Packaging, Transport, Storage and Security of Radioactive Material

Over the next few years, the engineering support required to meet the UK Nuclear Decommissioning Authority's strategic targets for redistribution of materials will ramp up significantly. To carry out these activities efficiently requires innovative solutions to be applied. This paper describes an approach for the transport of certain categories of heat generating materials, which offers operational and payload benefits. For the transport of certain materials, the use of a relatively small package is required for handling purposes, carrying internal product cans of material containing fissile product. It is proposed that 'INS3578' packages could be used. The packages are to be contained within an ISO container. Owing to the heat generating nature of some of the material to be transported, consideration must be given to the evacuation of heat from the container. A passively cooled container has the advantage of not requiring the complication of a forced ventilation system and refrigeration plant. This has operational, licensing and security benefits. A concept study undertaken to investigate a passively cooled container is described, including options that have been considered and the results from calculations undertaken to determine the package temperature. The results from the concept study suggest that with further work and further consideration of the heat load to be placed inside the container, the concept of a passively cooled ISO container for the transport of material might be a viable option. © W. S. Maney & Son Ltd 2015. Source

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