Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

Technical Report · Sat Jan 01 00:00:00 EST 2005

DOI:https://doi.org/10.2172/911007· OSTI ID:911007

MacDonald, Philip E

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.

• Research Organization: Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: DOE - NE
• DOE Contract Number: DE-AC07-99ID-13727
• OSTI ID: 911007
• Report Number(s): INEEL/EXT-04-02530; TRN: US0704338
• Country of Publication: United States
• Language: English

Related Subjects

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
29 - ENERGY PLANNING
POLICY AND ECONOMY
BUILDING MATERIALS
CRITICAL PRESSURE
DRILLING EQUIPMENT
ELECTRIC POWER
FUEL CHANNELS
FUEL CYCLE
NUCLEAR ENERGY
PRESSURE VESSELS
PRODUCTION
REACTOR CORES
REACTOR SAFETY
STEAM GENERATORS
STEAM SEPARATORS
STRESS CORROSION
THERMAL EFFICIENCY
WATER COOLED REACTORS
Generation IV
SCWR