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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.
H. Y. Khater, M. E. Sawan, I. N. Sviatoslavsky, L. J. Wittenberg, W. R. Meier
Fusion Science and Technology | Volume 21 | Number 3 | May 1992 | Pages 2138-2144
Blanket Shield and Neutronic | doi.org/10.13182/FST92-A30037
Articles are hosted by Taylor and Francis Online.
A detailed safety analysis was performed for the inertial confinement fusion reactor OSIRIS. The radioactivity induced in the carbon fabric chamber concrete shield and Flibe breeder is very low allowing for their disposal at the end of the reactor life as Class A low level waste (LLW). The biological dose rate after shutdown behind the reactor biological shield shield is very low (0.11 µmrem/hr) allowing only for hands-on maintenance. A total of 91.5 Ci/day are routinely released to the environment producing an off-site dose to the maximally exposed individual (MEI) of 2.43 mrem/yr at the reactor site boundary. Only a small fraction (0.2%) of the reactor first wall would be mobilized during a loss of coolant/loss of flow accident. The decay heat generated in the concrete shield is very low such that its temperature would only increase by less than 2 degrees during such an accident OSIRIS contains 660 tonnes of liquid Flibe as a coolant and breeder. A severe accident including a breach of the reactor building and chamber is estimated to release less than 0.5 kg of the activated Flibe to the environment. The total whole body (WB) early dose at the reactor site boundary during a pessimistic accident that includes the potential release of radioactive products from the chamber, shield and Flibe coolant would amount to 13.5 mrem. In addition, a 100% release of all the tritium steady state inventory (12.7 gm) inside the reactor building during operation would result in an additional 114 mrem of off-site dose. The total of 128 mrem off-site dose produced from OSIRIS eliminates the need for using N-stamp nuclear grade components in the reactor.
