September-October 2017 NPJ

TRACG- LOCA Methodology By James Tusar, Exelon Generation. James Tusar James Tusar is a graduate of the Pennsylvania State University with a BS in Nuclear Engineering. He has an MS in Environmental Engineering from Drexel University, and has a Professional Engineer’s License in Nuclear Engineering. He is currently the Senior Manager of Spent Fuel for Exelon Generation which includes responsibility for spent fuel management for 23 nuclear reactors across five states. James has been recognized for his nuclear industry accomplishments with eleven (11) Nuclear Energy Institute (NEI) Top Innovative Practice (TIP) Awards. Nuclear Energy Institute’s Top Innovative Practice Process Awards highlight the nuclear industry’s most innovative techniques and ideas. This innovation won the 2017 GE Hitachi Nuclear Energy Vendor Award. The team members who participated included James Tusar, Exelon Generation, Nuclear Fuels; John Massari, Exelon Generation, Nuclear Fuels; David Knepper, Exelon Generation, Nuclear Fuels; Travis Bement, Exelon Generation, Nuclear Fuels; Ronnie Reynolds, Exelon Generation, Licensing & Regulatory Affairs; George Inch, Exelon Generation, Nine Mile Point, Design Engineering; Heather Czernicki, Oyster Creek, Design Engineering; Baris Sarikaya, GE- Hitachi, LOCA & Containment Analysis; Joe Fricano, GE-Hitachi, LOCA & Containment Analysis. Innovation Sometimes a “game-changer” is needed. For Exelon’s and the United States’ only BWR/2 reactors, Nine Mile Point Unit 1 and Oyster Creek, that game-changer is the GE Hitachi (GEH) TRACG-LOCA (Transient Reactor Analysis Code GEH) methodology. This innovative methodology will provide ad- ditional thermal limit margin and signifi- cantly reduce fuel costs. Unfortunately, in the last four years, companies shutdown – or announced their intent to shutdown – several nuclear power reactors, in part, due to economic chal- lenges. This innova- tion results in a sus- tained fuel cost reduc- tion that improves the economic viability of nuclear power plants. Nine Mile Point Unit 1 was the first plant in the Unites States to implement this methodology. Enhancing the methods used for Loss of Coolant Accident (LOCA) analyses provides reasonable assurance that nuclear reactors operate safely. In the past, analytical tools for LOCA assessments provided conservative calculations aimed at overcoming the limited understanding of the complex physical phenomena associated with a LOCA and the resulting physical model uncertainties, as well as the limitations in numerical analysis methods and restrictions in computational resources. Additional regulatory assumptions are applied to increase assurance of public safety. GEH’s realistic TRACG-LOCA methodology is an alternative to the traditional 10 CFR 50 Appendix K, ECCS Evaluation Models , methods. GEH has developed and licensed a realistic methodology for Emergency Core Cooling System (ECCS) performance evaluation based on the TRACG methodology. This methodology was implemented at Nine Mile Point Unit 1 in March 2017 and will be at Oyster Creek in September 2018. TRACG uses improved physical models and uncertainty analysis capabilities, and it implements numerical advancements beyond legacy computational tools. TRACG has been extensively validated and used for other types of analyses including AOO (Anticipated Operational Occurrences)/ transient, stability, and Anticipated Transients Without SCRAM (ATWS) events. It is also used for licensing of the ESBWR. TRACG-LOCA methodology is a best-estimate analysis tool, coupled with an uncertainties quantification technique, which differs from the traditional 10 CFR 50 Appendix K methods. The methodology is structured to follow the Code Scaling, Applicability and Uncertainty (CSAU) Evaluation Methodology approach. In a study chartered by the U.S. Nuclear Regulatory Commission (NRC), LOCA analysis is the first example of a best-estimate method application of CSAU. Additional guidance and principal criteria for such applications are provided by the NRC in Regulatory Guide 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance . GEH adhered to the principles provided in that guidance and continue to develop and maintain the TRACG code per Regulatory Guide 1.203, Transient and Accident Analysis Methods . Both Exelon and GEH have been extensively engaged with the NRC throughout the review and approval process. The model uncertainties are evaluated by comparing code predictions to available and applicable data. Improvements are obtained by refining model uncertainties in calculations when a best-estimate methodology is used. Among some of the conservative features that are considered to compensate for the model uncertainties in the traditional 10 CFR 50 Appendix K methods, is the decay heat requirement, that is the most dominant factor causing high Peak Clad Temperature (PCT) and oxidation results, which produces the restrictive Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits. In realistic calculations, there are no mandatory restrictions for models that can be used, as long as adequate validation and 46 NuclearPlantJournal.com Nuclear Plant Journal, September-October 2017 (Continued on page 48)