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Decommissioning of a Shielded αβγ Pie Facility at Harwell

[+] Author Affiliations
T. N. Chambers

United Kingdom Atomic Energy Authority

Paper No. ICEM2003-5022, pp. 1727-1734; 8 pages
doi:10.1115/ICEM2003-5022
From:
  • ASME 2003 9th International Conference on Radioactive Waste Management and Environmental Remediation
  • 9th ASME International Conference on Radioactive Waste Management and Environmental Remediation: Volumes 1, 2, and 3
  • Oxford, England, September 21–25, 2003
  • Conference Sponsors: Nuclear Engineering Division and Environmental Engineering Division
  • ISBN: 0-7918-3732-7 | eISBN: 0-7918-3731-9
  • Copyright © 2003 by ASME

abstract

Building 393.6 at Harwell was constructed and commissioned in the mid 1950’s as a remote handling facility used for the post irradiation examination (PIE) of a variety of radioactive materials. The facility contained a number of lead shielded cells, fume cupboards, gloveboxes and a block of three concrete shielded cells. A particular feature of the facility was a suite of ten lead shielded cells arranged as a cell line with interconnecting shielded tunnel sections, used for PIE of irradiated fuels. Operation of the plant continued until the mid-1990s supporting a variety of commercial programmes for the nuclear industry, including the assessment of fuel performance and metallurgy of a variety of cladding materials. Decommissioning of the facility commenced in 1995 with the objective of removing all radiological and toxic materials, dismantling of containments and removal of the building fabric, to enable release of the land for alternative uses. The key challenges encountered in decommissioning were: • B393.6 was the first Category 1 shielded facility to be decommissioned at the Harwell Site; • The hands on decommissioning team was divested from UKAEA to a contracting organisation part way through the decommissioning; • Development of remote decontamination techniques to ensure operator doses were ALARP; • Remote decontamination of heavily contaminated shielded enclosures to permit disposal of waste to Drigg as Low Level Waste; • Reducing the volume of Intermediate Level Waste into either Low Level Waste or free release material; • Decontamination of large volumes of lead to enable disposal of approximately 95% of lead arisings as free release. The paper will discuss the technical and managerial challenges encountered, and provide details of the solutions adopted in successfully reaching the radiological endpoint.

Copyright © 2003 by ASME

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