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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.
Shunsuke Uchida, Eishi Ibe, Kiyatomo Nakata, Motomasa Fuse, Katsumi Ohsumi, Yoshie Takashima
Nuclear Technology | Volume 110 | Number 2 | May 1995 | Pages 250-257
Technical Paper | Material | doi.org/10.13182/NT95-A35122
Articles are hosted by Taylor and Francis Online.
Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the reactor pressure vessel (RPV) and (b) moderating the crack growth rate, even if microcracks are present on the structural materials. Several disadvantages caused by HWC (e.g., turbine dose rate increase, 60Co radioactivity buildup, and effects on fuel cladding) are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated based on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants—e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom — are planned that would collect data to support the theoretical considerations.