Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

J. K. Wright ; R. N. Wright

关键词: ACCIDENTS;    AFTER-HEAT REMOVAL;    COMMERCIAL SECTOR;    CONFIGURATION;    CONSTRUCTION;    DEMONSTRATION PLANTS;    ELECTRICITY;    GAS COOLED REACTORS;    GRAPHITE;    HYDROGEN PRODUCTION;    MARTENSITIC STEELS;    NUCLEAR POWER;    PEBBLE BED REACTORS;    PRESSURE VESSELS;    RADIOACTIVE MATERIALS;    SERVICE LIFE;    STEELS;    STRESSES;    URANIUM;    WATER High Temperature Gas-cooled Reactor;    Next Generation Nuclear Plant;    Reactor Pressure Vessel;    Research and Development Plan;   

DOI  :  10.2172/952022
RP-ID  :  INL/EXT-08-14108
PID  :  OSTI ID: 952022
Others  :  TRN: US0902411

学科分类：核能源与工程

美国|英语

来源: SciTech Connect

【 摘 要 】

The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2ŸCr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.