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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.
Ian C. Gauld, Georgeta Radulescu, Germina Ilas, Brian D. Murphy, Mark L. Williams, Dorothea Wiarda
Nuclear Technology | Volume 174 | Number 2 | May 2011 | Pages 169-195
Technical Paper | Special Issue on the SCALE Nuclear Analysis Code System / Fuel Cycle and Management | doi.org/10.13182/NT11-3
Articles are hosted by Taylor and Francis Online.
The calculation of fuel isotopic compositions is essential to support design, safety analysis, and licensing of many components of the nuclear fuel cycle - from reactor physics and severe accident analysis to back-end fuel cycle issues, including spent-fuel storage and transportation, reprocessing, and radioactive waste management. Versions of the ORIGEN code, developed by Oak Ridge National Laboratory, have been used worldwide for isotopic depletion and decay analysis for more than three decades. The supported version of ORIGEN, maintained as the depletion analysis module for SCALE 6, performs detailed time-dependent isotopic generation and depletion for 1946 nuclides for reactor fuel and activation analysis. Stand-alone ORIGEN calculations can be performed using cross-section libraries developed for a wide range of reactor types and fuel designs used worldwide, including light water reactors UO2 and MOX, CANDU, VVER 440 and 1000, RBMK, and graphite reactors. Alternatively, within SCALE 6, ORIGEN can be automatically coupled to two-dimensional discrete ordinates or three-dimensional Monte Carlo transport solvers that provide problem-dependent cross sections for use in the ORIGEN depletion calculation. The hybrid ability to function as either a stand-alone or coupled depletion code provides ORIGEN advanced capabilities to simulate a broad range of applications for various reactor systems. The nuclear data libraries in ORIGEN have been significantly improved recently, using modern ENDF/B nuclear data evaluations. The most recent developments in SCALE 6.1 include the addition of ENDF/B-VII decay data, energy-dependent fission yields, and fine-group ORIGEN neutron cross sections based on the JEFF-3.0/A special purpose activation files. Advanced methods and data for neutron and gamma source energy spectral analysis are also available in the current version of the code. The ORIGEN code and associated nuclear data libraries have been extensively validated against experimental data that include spent nuclear fuel isotopic assay data for actinides and fission products, radiation source spectra, and decay heat measurements.
