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

Journal Article · Wed Mar 01 04:00:00 EST 2017 · Journal of Nuclear Engineering and Radiation Science

DOI:https://doi.org/10.1115/1.4035851· OSTI ID:1470321

Alfonsi, Andrea ^[1]; Mesina, George L. ^[1]; Zoino, Angelo ^[2]; Anderson, Nolan ^[1]; Rabiti, Cristian ^[1]

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1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
2. Univ. La Sapienza, Rome (Italy)

The Nuclear Regulatory Commission (NRC) has considered revision of 10 CFR 50.46C rule to account for the effects of burn-up rate in future analysis of reactor accident scenarios so that safety margins may evolve with the operation and reloading of the reactor as dynamic limits. In order to find these limiting conditions, both cladding oxidation and maximum temperature must be cast as functions of fuel exposure. It is computationally intensive to run a plant model through a long duration operational transient to the time of fuel reload, and this must be repeated for each reload until the time of the accident scenario. Moreover, fuel reload patterns may differ in many ways so that calculations with many different core configurations must be considered. To perform such new analyses in a reasonable amount of computational time with good accuracy, Idaho National Laboratory (INL) has developed new multi-physics tools by combining existing codes and adding new capabilities. The Parallel Highly Innovative Simulation INL Code System (PHISICS) toolkit for neutronic and reactor physics is coupled with the Reactor Excursion and Leak Analysis Program (RELAP5-3D) for the Loss of Coolant Accident (LOCA) analysis and RAVEN for the Probabilistic Risk Assessment (PRA) and margin characterization analysis. Expansion of code capability was also necessary. For RELAP5-3D to process a single sequence of cores in a continuous run required modification of RELAP5-3D to restart a sequence of input decks, each with a slightly modified neutronics model or thermal-hydraulic flow region and culminating in an accident scenario. A new multi-deck input processing capability was developed and verified for this analysis. In conclusion, the combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical Pressurized Water Reactor (PWR).

• Research Organization: Idaho National Laboratory (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• Grant/Contract Number: AC07-05ID14517
• OSTI ID: 1470321
• Report Number(s): INL/JOU--16-40110-Rev000
• Journal Information: Journal of Nuclear Engineering and Radiation Science, Journal Name: Journal of Nuclear Engineering and Radiation Science Journal Issue: 2 Vol. 3; ISSN 2332-8983
• Publisher: ASME
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
22 GENERAL STUDIES OF NUCLEAR REACTORS
97 MATHEMATICS AND COMPUTING
Equilibrium core configuration
PHISICS
Probabilistic Risk Assessment
RAVEN
RELAP5-3D
peak clad temperature