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Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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ANS Student Conference 2025
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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.
R. J. Sheu, Y. F. Chen, S. H. Jiang, J. N. Wang, U. T. Lin
Nuclear Technology | Volume 175 | Number 1 | July 2011 | Pages 335-342
Technical Paper | Special Issue on the 16th Biennial Topical Meeting of the Radiation Protection and Shielding Division / Radiation and Protection | doi.org/10.13182/NT11-A12305
Articles are hosted by Taylor and Francis Online.
This study reevaluates the dose rates at the site boundary of an independent spent-fuel storage installation (ISFSI) using the MAVRIC computational sequence in the SCALE6 code package. Based on advanced variance-reduction techniques and powerful geometry modeling capabilities, MAVRIC can tackle this large ISFSI shielding problem by directly simulating the radiation transport in a full-scale model. This study started with a benchmark calculation of a single storage cask and then investigated the impact of a fully loaded ISFSI on the dose rates at the site boundary. Because of the short distance to the nearest site boundary, additional shielding to the cask itself or the site is necessary to meet the stringent design dose limit. Compared to the two-step cask-by-cask approach adopted in the original safety analysis report, this method of analyzing the site boundary doses is straightforward and efficient enough to allow us to evaluate the effect of the cask design modification and to test various options for further improvement.
