Critical Section Selection Methodology for the U.S. EPR™ Standard Nuclear Power Plant

2010 ◽  
Author(s):  
Se-Kwon Jung ◽  
Adam Goodman ◽  
Joe Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Critical Section ◽  
Selection Methodology ◽  
Critical Sections ◽  
The U.S

This paper presents a three-tier, critical section selection methodology that is used to identify critical sections for the U.S. EPR™ Standard Nuclear Power Plant (NPP). The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPR™ design; their structural design adequacy provides reasonable assurance of overall U.S. EPR™ structural design adequacy.

Critical Section Selection Methodology for the U.S. EPRTM Standard Nuclear Power Plant

2013 ◽  
Vol 135 (4) ◽  
Author(s):  
Se-Kwon Jung ◽  
Adam Goodman ◽  
Joseph Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Critical Section ◽  
Selection Methodology ◽  
Critical Sections ◽  
The U.S

A three-tier critical section selection methodology that is used to identify critical sections for the U.S. EPRTM standard nuclear power plant (NPP) is presented. The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPRTM design; their structural design adequacy provides reasonable assurance of overall U.S. EPRTM structural design adequacy.

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

Overview of the nuclear power plant structural integrity research in Finland

1993 ◽  
Vol 55 (1) ◽  
pp. 3-59 ◽  
Author(s):  
K. Törrönen ◽  
P. Aaltonen ◽  
H. Hänninen ◽  
K. Mäkelä ◽  
P. Karjalainen-Roikonen ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity

Seismic Response Analysis of Reactor Coolant Pump in Nuclear Power Plant

2008 ◽  
Author(s):  
Zhaohui Ren ◽  
Hui Ma ◽  
He Li ◽  
Guiqiu Song ◽  
Wenjian Zhou
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Response Spectrum ◽  
Response Analysis ◽  
Response Spectrum Analysis ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Modes Of Vibration

The reactor coolant pump in nuclear power plant is the only revolving equipment in the nuclear power plant. Its functional stability will directly affect the security of nuclear power plant. The coolant pump of a very nuclear plant is examined by using response spectrum analysis to analysis dynamic characteristics and responses aiming at finding the natural frequencies of vibration, modes of vibration and seismic responses, and any possible step which may cause damage of the whole system. The favorable spectrum and unfavorable one are investigated as well. The paper focuses on avoiding the detrimental effects caused by earthquakes, therefore may lay down a theoretical foundation for structural design and installation.

Validation and Benchmarking of Comprehensive Vibration Assessment Program for Prototype Reactor Internals

2012 ◽  
Author(s):  
Jianfeng Yang ◽  
Lixin Yu ◽  
Byounghoan Choi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Boiling Water ◽  
Pressurized Water ◽  
Water Reactor ◽  
Assessment Program ◽  
Water Reactors ◽  
The U.S ◽  
Reactor Internals

Reactor internals important to nuclear power plant safety shall be designed to accommodate steady-state and transient vibratory loads throughout the service life of the reactor. Operating experience has revealed failures of reactor internals in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) due to flow-induced vibrations (FIVs). U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) that the NRC staff considers acceptable for use in verifying the structural integrity of reactor internals for FIV prior to commercial operation. A CVAP supports the NRC reviews of applications for new nuclear reactor construction permits or operating licenses under 10 CFR Part 50, as well as design certifications and combined licenses that do not reference a standard design under 10 CFR Part 52. The overall CVAP should be implemented in conjunction with preoperational and initial startup testing. For prototype reactor internals, the comprehensive program should consist of a vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. Validation and benchmarking processes should be integrated into the CVAP throughout each individual program. Based on the authors’ experiences in Advanced Boiling Water Reactor and AP1000® CVAPs and based on detailed reviews of the U.S. Evolutionary Power Reactor and the U.S. Advanced Pressurized Water Reactor CVAPs, this article summarizes the essential CVAP validation and benchmarking processes with proper consideration of bias errors and random uncertainties. This article provides guidance to a successful CVAP that satisfies the NRC requirements and ensures the reliability of the evaluation of potential adverse flow effects on nuclear power plant components.

Developments in Structural Integrity Safety Cases for Highest Reliability Components in the UK

2016 ◽  
Author(s):  
T. Jelfs ◽  
M. Hayashi ◽  
A. Toft
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Reliability ◽  
Structural Integrity ◽  
Lessons Learned ◽  
Advisory Group ◽  
Design Assessment ◽  
Safety Case ◽  
The Uk

Gross failure of certain components in nuclear power plant has the potential to lead to intolerable radiological consequences. For these components, UK regulatory expectations require that the probability of gross failure must be shown to be so low that it can be discounted, i.e. that it is incredible. For prospective vendors of nuclear power plant in the UK, with established designs, the demonstration of “incredibility of failure” can be an onerous requirement carrying a high burden of proof. Requesting parties may need to commit to supplementary manufacturing inspection, augmented material testing requirements, enhanced defect tolerance assessment, enhanced material specifications or even changes to design and manufacturing processes. A key part of this demonstration is the presentation of the structural integrity safety case argument. UK practice is to develop a safety case that incorporates the notion of ‘conceptual defence-in-depth’ to demonstrate the highest structural reliability. In support of recent Generic Design Assessment (GDA) submissions, significant experience has been gained in the development of so called “incredibility of failure” arguments. This paper presents an overview of some of the lessons learned relating to the identification of the highest reliability components, the development of the structural integrity safety arguments in the context of current GDA projects, and considers how the UK Technical Advisory Group on Structural Integrity (TAGSI) recommendations continue to be applied almost 15 years after their work was first published. The paper also reports the approach adopted by Horizon Nuclear Power and their partners to develop the structural integrity safety case in support of the GDA process to build the UK’s first commercial Boiling Water Reactor design.

Countermeasures Derived From the Lessons of the Fukushima Daiichi Nuclear Power Plant Accident

2013 ◽  
Author(s):  
Tadashi Narabayashi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pacific Ocean ◽  
Electric Power ◽  
Nuclear Power ◽  
Safety Evaluation ◽  
Lessons Learned ◽  
The Other ◽  
Nuclear Accidents ◽  
Fukushima Daiichi

On March 11, 2011, Tokyo Electric Power Company’s Fukushima Daiichi Nuclear Power Plant (NPP) was hit by a tsunami caused by the Tohoku-Pacific Ocean Earthquake, resulting in nuclear accidents in Units #1 to #4. With the aim of improving the safety of NPPs worldwide, we summarize the lessons that have been learned following a thorough analysis of the event and make specific proposals for improving the safety of such facilities. The author has been involved in investigating the causes of the accidents and developing countermeasures for other NPPs in Japan as a member of the Committee for the Investigation of Nuclear Safety of the Atomic Energy Society of Japan [1], an advisory meeting member of NISA with regard to technical lessons learned from the Fukushima Daiichi NPP accidents, and a Safety Evaluation Member of NISA for the other NPPs in Japan [2].

CFD Analysis of Thermal Stratification Phenomena in the RNS Suction Line of a Gen-III Nuclear Power Plant

2013 ◽  
Author(s):  
Yan Li ◽  
Daogang Lu ◽  
Zhigang Wang ◽  
Jian Wu ◽  
Fengyun Yu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Heat Removal ◽  
Radial Temperature ◽  
Piping Systems ◽  
Suction Line ◽  
Power Operation

Thermal stratification phenomena in piping systems of nuclear power plant would threaten the structural integrity of pipes, which are caused by the significant change of water density with temperature. To provide temperature gradients for the stress analysis of Normal heat Removal System (RNS) suction line of a Gen-III nuclear power plant, the relevant thermal stratification phenomena are analyzed by CFD in this paper. Cases without leakage (normal power operation) and with leakage are both studied. The results show that the first portion of pipe (one meter or so) near the hot leg is isothermal for normal power operation due to the penetrating flow. In the remaining portion, the radial temperature drops are of the order of 20∼27 K for no leakage case. For the leakage case, the radial temperature drops are 23 K or less, which are relatively smaller than those for the no leakage case due to the net hot flow from the hot leg to the valve.

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