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Complaint filed with FERC over Grand Gulf management
The Louisiana Public Service Commission (LPSC), the New Orleans City Council, and the Arkansas Public Service Commission on March 2 filed a complaint with the Federal Energy Regulatory Commission against Entergy Corporation, seeking damages of more than $360 million for what they term the utility’s “imprudent operation” of the Grand Gulf nuclear plant.
Located in Port Gibson, Miss., Grand Gulf is a single-unit plant with a 1,433-MWe boiling water reactor. The unit, which entered commercial operation in 1985, supplies power to customers of Entergy Louisiana, Entergy Mississippi, Entergy Arkansas, and Entergy New Orleans.
Marat Margulis, Erez Gilad
Nuclear Technology | Volume 196 | Number 2 | November 2016 | Pages 377-395
Technical Paper | dx.doi.org/10.13182/NT16-23
Articles are hosted by Taylor and Francis Online.
The application of best-estimate codes [coupled neutron kinetics (NK)/thermal hydraulics (TH)] for safety analyses of research reactors (RRs) has gained considerable momentum during the past decade. Application of these codes is largely facilitated by the high level of technological maturity and expertise that these codes allow as a safety technology in nuclear power plants, and it is largely driven by International Atomic Energy Agency activities. The present study belongs in this framework and presents the development and application of the coupled NK and TH code THERMO-T to the analysis of protected reactivity insertion accidents and loss-of-flow accidents in a typical RR with standard Materials Testing Reactor plate-type fuel elements. The coupling is realized by considering the neutronic reactivity feedbacks of the fuel and coolant temperatures and a heat generation model for the reactor power. The neutron flux in the reactor core is solved by applying point reactor kinetic equations and employing radial and axial power distributions calculated from a three-dimensional full-core model by the continuous-energy Monte Carlo reactor physics code Serpent. The evolution of temporal and spatial distributions of the fuel, cladding, and coolant temperatures is calculated for all fuel channels by using a finite volume time implicit numerical scheme for solving a three-conservation equation model. In this study, additional features, such as critical heat flux ratio prediction and decay heat model, are implemented for both highly enriched uranium and low-enriched uranium cores, and a comprehensive comparison of THERMO-T results is performed against other codes.