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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.
Michael L. Fensin, John S. Hendricks, Samim Anghaie
Nuclear Technology | Volume 164 | Number 1 | October 2008 | Pages 3-12
Technical Paper | Icapp '06 | doi.org/10.13182/NT08-A4003
Articles are hosted by Taylor and Francis Online.
As advanced reactor concepts challenge the accuracy of current modeling technologies, a higher-fidelity depletion calculation is necessary to model time-dependent core reactivity properly for accurate cycle length and safety margin determinations. The recent integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a completely self-contained Monte Carlo-linked depletion capability. Two advances have been made in the latest MCNPX capability based on problems observed in prereleased versions: continuous-energy collision density tracking and adequate fission yield selection.Prereleased versions of the MCNPX depletion code calculated the reaction rates for (n,2n), (n,3n), (n,p), and (n,) by matching the MCNPX steady-state 63-group flux with 63-group cross sections inherent in the CINDER90 library and then collapsing to one-group collision densities for the depletion calculation. The accuracy of this procedure is therefore dictated by the adequacy of the 63-group energy structure of the cross-section set to accurately capture the spectrum of a specific model. Different types of models would therefore require different types of cross-section energy group structure. MCNPX 2.6.A eliminates this dependency by using the continuous-energy reaction rates determined during the MCNPX steady-state calculation to calculate energy-integrated collision rates to be used by CINDER90.MCNPX 2.6.A now also determines the proper fission yield to be used by the CINDER90 code for the depletion calculation. The CINDER90 code offers a thermal, fast, and high-energy fission yield for each fissile isotope contained in the CINDER90 data file. MCNPX 2.6.A determines which fission yield to use for a specified problem by calculating the integral fission rate for the defined energy boundaries (thermal, fast, and high energy), determining which energy range contains the majority of fissions, and then selecting the appropriate fission yield for the energy range containing the majority of fissions.The MCNPX depletion capability enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. This study focuses on the methodology development of the two improvements described here. Further improvements are under development to enhance the usefulness of this new capability.
