Next Generation Nuclear Plant GAP Analysis Report

As a follow-up to the phenomena identification and ranking table (PIRT) studies conducted recently by NRC on next generation nuclear plant (NGNP) safety, a study was conducted to identify the significant 'gaps' between what is needed and what is already available to adequately assess NGNP safety characteristics. The PIRT studies focused on identifying important phenomena affecting NGNP plant behavior, while the gap study gives more attention to off-normal behavior, uncertainties, and event probabilities under both normal operation and postulated accident conditions. Hence, this process also involved incorporating more detailed evaluations of accident sequences and risk assessments. This study considers thermal-fluid and neutronic behavior under both normal and postulated accident conditions, fission product transport (FPT), high-temperature metals, and graphite behavior and their effects on safety. In addition, safety issues related to coupling process heat (hydrogen production) systems to the reactor are addressed, given the limited design information currently available. Recommendations for further study, including analytical methods development and experimental needs, are presented as appropriate in each of these areas.

[1]  Sydney J Ball,et al.  Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 1: Main Report , 2008 .

[2]  R. C. Martin,et al.  Compilation of Fuel Performance and Fission Product Transport Models and Database for MHTGR Design , 1993 .

[3]  D. S. Griffin,et al.  Elevated-temperature structural design evaluation issues in LMFBR licensing , 1985 .

[4]  R. Fuge,et al.  Method for determining the differential reactivities of power reactors , 1973 .

[5]  Timothy O. Lenz.  The New ‘U’ , 1997, PS: Political Science & Politics.

[6]  Michael Parkin.  Flow Induced Vibration Problems in Gas Cooled Reactors , 1980 .

[7]  G. Jones.  The modular high temperature gas cooled reactor , 1988 .

[8]  R. Ballinger,et al.  TIMCOAT: An Integrated Fuel Performance Model for Coated Particle Fuel , 2004 .

[9]  Robert I. Jetter,et al.  DOE-ASME Generation IV Materials Tasks , 2006 .

[10]  P. Zweifel,et al.  Reactor physics , 1973 .

[11]  B. T. Kelly.  Irradiation creep in graphite—some new considerations and observations , 1992 .

[12]  R. A. Ainsworth,et al.  An Assessment Procedure for the High Temperature Response of Structures , 1991 .

[13]  Ugur Emre Sikik.  TINTE analysis of PBMR reactivity insertion transients , 2008 .

[14]  Timothy D. Burchell,et al.  The analysis of irradiation creep experiments on nuclear reactor graphite , 1994 .

[15]  W. J. Gray,et al.  Constant stress irradiation-induced compressive creep of graphite at high fluences , 1973 .

[16]  Gennady Ziskind,et al.  Resuspension of particulates from surfaces to turbulent flows : review and analysis , 1995 .

[17]  D. R. Vondy,et al.  Solving the uncommon nuclear reactor core neutronics problems , 1983 .

[18]  Barry Marsden,et al.  Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 5: Graphite PIRTs , 2008 .

[19]  F. A. Silady,et al.  A vented low pressure containment strategy for the Modular High Temperature Gas-Cooled Reactor (MHTGR) , 1994 .

[20]  R. Ainsworth.  R5 procedures for assessing structural integrity of components under creep and creep–fatigue conditions , 2006 .

[21]  Y. Takahashi,et al.  7.10 – Leak before Break , 2003 .

[22]  E. Hontañón,et al.  THE CÆSAR CODE FOR AEROSOL RESUSPENSION IN TURBULENT PIPE FLOWS. ASSESSMENT AGAINST THE STORM EXPERIMENTS , 2000 .

[23]  A. K. Postma,et al.  Evaluation of air cleaning system concepts for emergency use in LMFBR plants , 1976 .