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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.
Te-Chuan Wang, Shih-Jen Wang, Jyh-Tong Teng
Nuclear Technology | Volume 156 | Number 3 | December 2006 | Pages 347-359
Technical Note | Thermal Hydraulics | doi.org/10.13182/NT06-A3796
Articles are hosted by Taylor and Francis Online.
Chinshan is a Mark-I boiling water reactor nuclear power plant (NPP) located in north Taiwan. It incorporates several severe-accident-mitigating features, especially two raw-water tanks in the mountain. According to a probabilistic risk assessment (PRA) of Chinshan NPP, station blackout (SBO) sequences are the most dominant sequences in internal core damage frequency. No credit is taken for the raw-water system in the development of a Chinshan PRA. Therefore, two dominant sequences (T3UTERDGX and T3UTERDG) of the SBO in the Chinshan PRA are cited as reference cases to evaluate the capacity of the raw-water system in the PRA and severe accident. The T3UTERDGX sequence is initiated by loss of off-site power (T3) followed by failure of both diesel generators (DG), failure of gas turbine generators, and failure to recover alternating current (ac) power (ER). That results in loss of all on- and off-site ac power. The high-pressure injection systems fail (UT) initially and timely reactor depressurization fails (X). The T3UTERDG sequence is the same as the T3UTERDGX sequence, except for failure of timely reactor depressurization (X). The MAAP4 code is used as a tool to evaluate the effectiveness of the raw-water system. Based on MAAP4 analysis, the raw-water system cannot cool down the core in the T3UTERDG sequence after introducing severe-accident-management guidelines. The raw-water system cannot flood dry-well water level above minimum debris submerge level (MDSL) in the T3UTERDGX sequence after reactor pressure vessel (RPV) breach. Sensitivity studies show that raw-water injection before the vessel water level reaches level 2 (L-2) can keep core coolability in the T3UTERDG sequence. Three times the raw-water injection rate is the minimum flow rate to flood the dry-well water level above MDSL and cool down the corium on the dry-well floor in the T3UTERDGX sequence. A raw-water system can be used as a mitigating measure, especially in an SBO. The RPV should be depressurized as quickly as possible if a raw-water system is the only mitigation measure in the accident. It is worthwhile to increase the raw-water flow rate to cool down the debris in the dry well after RPV breach.
