Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria

Abstract

The Nuclear Regulatory Commission (NRC) has considered revision of 10-CFR-50.46C rule (Borchard and Johnson, 2013, “10 CFR 50.46c Rulemaking: Request to Defer Draft Guidance and Extension Request for Final Rule and Final Guidance,” U.S. Nuclear Regulatory Commission, Washington, DC.) to account for burn-up rate effects in future analysis of reactor accident scenarios so that safety margins may evolve as dynamic limits with reactor operation and reloading. To find these limiting conditions, both cladding oxidation and maximum temperature must be cast as functions of fuel exposure. To run a plant model through a long operational transient to fuel reload is computationally intensive, and this must be repeated for each reload until the time of the accident scenario. Moreover for probabilistic risk assessment (PRA), this must be done for many different fuel reload patterns. To perform such new analyses in a reasonable amount of computational time with good accuracy, Idaho National Laboratory (INL) has developed new multiphysics tools by combining existing codes and adding new capabilities. The parallel highly innovative simulation INL code system (PHISICS) toolkit (Rabiti et al., 2016, “New Simulation Schemes and Capabilities for the PHISICS/RELAP5-3D Coupled Suite,” Nucl. Sci. Eng., 182(1), pp. 104–118; Alfonsi et al., 2012, “PHISICS Toolkit: Multi-Reactor Transmutation Analysis Utility—MRTAU,” PHYSOR 2012 Advances in Reactor Physics Linking Research, Industry, and Education, Knoxville, TN, Apr. 15–20.) for neutronic and reactor physics is coupled with the reactor excursion and leak analysis program—three-dimensional (RELAP5-3D) (The RELAP5-3D© Code Development Team, 2014, “RELAP5-3D© Code Manual Volume I: Code Structure, System Models, and Solution Methods,” Rev. 4.2, Idaho National Laboratory, Idaho Falls, ID, Technical Report No. INEEL-EXT-98-00834.) for the loss of coolant accident (LOCA) analysis and reactor analysis and virtual-control environment (RAVEN) (Alfonsi et al., 2013, “RAVEN as a Tool for Dynamic Probabilistic Risk Assessment: Software Overview,” 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Sun Valley, ID, May 5–9, pp. 1247–1261.) for the probabilistic risk assessment (PRA) and margin characterization analysis. For RELAP5-3D to process a single sequence of cores in a continuous run required a sequence of restarting input decks, each with different neutronics or thermal-hydraulic (TH) flow region and culminating in an accident scenario. A new multideck input processing capability was developed and verified for this analysis. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical pressurized water reactor (PWR).