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Colin Judge: Testing structural materials in Idaho’s newest hot cell facility
Idaho National Laboratory’s newest facility—the Sample Preparation Laboratory (SPL)—sits across the road from the Hot Fuel Examination Facility (HFEF), which started operating in 1975. SPL will host the first new hot cells at INL’s Materials and Fuels Complex (MFC) in 50 years, giving INL researchers and partners new flexibility to test the structural properties of irradiated materials fresh from the Advanced Test Reactor (ATR) or from a partner’s facility.
Materials meant to withstand extreme conditions in fission or fusion power plants must be tested under similar conditions and pushed past their breaking points so performance and limitations can be understood and improved. Once irradiated, materials samples can be cut down to size in SPL and packaged for testing in other facilities at INL or other national laboratories, commercial labs, or universities. But they can also be subjected to extreme thermal or corrosive conditions and mechanical testing right in SPL, explains Colin Judge, who, as INL’s division director for nuclear materials performance, oversees SPL and other facilities at the MFC.
SPL won’t go “hot” until January 2026, but Judge spoke with NN staff writer Susan Gallier about its capabilities as his team was moving instruments into the new facility.
Clay M. Davis
Nuclear Technology | Volume 87 | Number 4 | December 1989 | Pages 778-785
Technical Paper | TMI-2: Decontamination and Waste Management / Nuclear Safety | doi.org/10.13182/NT89-A27671
Articles are hosted by Taylor and Francis Online.
The March 28, 1979, loss-of-coolant accident at Three Mile Island Unit 2 resulted in the exposure of ∼3000 m2 of reactor building (RB) internal concrete surfaces to both liquid- and vapor-borne contaminants. The period of contact between these surfaces and aqueous solutions of mixed fission products ranged from a few days to several years. At the completion of gross decontamination of the accessible elevations of the RB in 1982 by water flushing, high-pressure spraying, strippable coating application, and hands-on wiping, dose rates remained above expected levels. Surveys and limited surface sampling indicated that contaminants had penetrated the protective coatings on the structural concrete, creating a substantial fixed source. To assess the depth of contaminant penetration into the concrete, a sampling program was conducted in September 1983. Based on the results of this program, it was determined that where coatings were intact, penetration past the matrix of the coating was insignificant. Where the coatings had been damaged prior to the accident, however, penetrations into the concrete were observed up to 20 mm. Subsequent modeling using the ISOSHLD II code using these values indicated that between 23 and 40% of the 1983 observed dose rates could be attributed to this source. Coatings removal tests conducted on the samples demonstrated that removal of the coatings could result in the removal of between 50 and 98% of the activity. Subsequent to this work, coatings and concrete removal on the accessible upper elevations of the RB resulted in dose reductions of 15 to 38%. These data, and that of subsequent work in the RB basement, indicate that protective coatings applied to structural concrete substantially reduce the degree to which the substrate will be penetrated by aqueous contaminant solutions. Relative to the ability to successfully remove absorbed contamination, the coating matrix containing the major fraction of the radionuclide load may be removed with less aggressive effort than that required to remove a potentially larger volume of contaminated concrete.