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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.
Theron D. Marshall, Robert W. Hockenbury, John A. Honey, Lee C. CadWallader
Nuclear Technology | Volume 114 | Number 1 | April 1996 | Pages 84-96
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT96-A35225
Articles are hosted by Taylor and Francis Online.
Probabilistic risk assessment methodology is applied to generate an evaluation of the relative likelihood of safe recovery following selected pressurized water reactor (PWR) design basis accidents for a Russian V213 nuclear power reactor. U.S.-designed PWRs similar to the V213 are used for reference and comparison. This V213 risk assessment is based on comparison analyses of the following aspects: accident progression event tree success paths for typical PWR accident initiating events, safety aspects in reactor design, and perceived performance of reactor safety systems. The four initiating events considered here were taken from a U.S. Nuclear Regulatory Commission summary report on severe accident risk: loss of offsite power with station blackout, large-break loss-of-coolant accident (LOCA), medium-break LOCA, arid small-break LOCA. The success probabilities for the V213 reaching a non-core-damage state after the onset of the selected initiating events are calculated for two scenarios: (a) using actual component reliability datafrom U. S. PWRs and (b) assuming common component reliability data. U.S. PWR component reliability data are used because of the unavailability of such data for the V213 at the time of the analyses. While the use of U.S. PWR data in this risk assessment of the V213 does strongly infer V213 comparability to U.S. plants, the risk assessment using common component reliability does not have such a stringent limitation and is thus a separate scoping assessment of the V213 engineered safety systems. The results of the analyses suggest that the V213 has certain design features that significantly improve the reactor’s safety margin for the selected initiating events and that the V213 design has a relative risk of core damage for selected initiating events that is at least comparable to U.S. PWRs. It is important to realize that these analyses are of a scoping nature and may be significantly influenced by important risk factors such as V213 operator training, quality control, and maintenance procedures. Additionally, the analyses make no implications as to the effects of the selected initiating events on the health and safety of the public.