Validation of Ultrasonic Inspections for Sizewell ‘B’ Pressurized Water Reactor

1989 ◽  
Vol 203 (1) ◽  
pp. 59-64
Author(s):  
D K Cartwright
Keyword(s):  
United Kingdom ◽  
Structural Integrity ◽  
High Standard ◽  
Atomic Energy ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Atomic Energy Authority ◽  
Circuit Components

To assist in ensuring a high standard of structural integrity of the Sizewell ‘B’ pressure circuit, components are inspected both during fabrication and in service. The more critical inspections are to be validated. This paper describes the validations of ultrasonic inspections to be carried out at the Inspection Validation Centre, United Kingdom Atomic Energy Authority, Risley.

Probabilistic Structural Integrity Evaluation on a Pressurized Water Reactor Pressure Vessel Under Pressure–Temperature Limit Operations

2015 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Temperature Limit ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

The normal reactor startup (heat-up) and shut-down (cool-down) operation limits are defined by the ASME Code Section XI-Appendix G, to ensure the structural integrity of the embrittled nuclear reactor pressure vessels (RPVs). In the paper, the failure risks of a Taiwan domestic pressurized water reactor (PWR) pressure vessel under various pressure-temperature limit operations are analyzed. Three types of pressure-temperature limit curves established by different methodologies, which are the current operation limits of the domestic RPV based on the KIa fracture toughness curve in 1998 or earlier editions of ASME Section XI-Appendix G, the recently proposed limits according to the KIC fracture toughness curve after the 2001 edition of ASME Section XI-Appendix G, and the risk-informed revision method proposed in MRP-250 report that provides more operational flexibility, are considered. The ORNL’s probabilistic fracture mechanics code, FAVOR, is employed to perform a series of fracture probability analyses for the RPV at multiple levels of embrittlement under such pressure-temperature limit transients. The analysis results indicate that the pressure-temperature operation limits associated with more operational flexibility will result in higher failure risks to the RPV. The shallow inner surface breaking flaw due to the clad fabrication defect is the most critical factor and dominates the failure risk of the RPV under pressure-temperature limit operations. Present work can provide a risk-informed reference for the safe operation and regulation of PWRs in Taiwan.

Comparison of Pressure-Temperature Limits for a Pressurized Water Reactor Pressure Vessel Considering Beltline and Extended Beltline Regions

2018 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Yu-Yu Shen ◽  
Chin-Cheng Huang
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Pressure Vessels ◽  
Reference Temperature ◽  
Regulatory Body ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

To ensure the structural integrity of the embrittled reactor pressure vessels (RPVs) during startup or shutdown operation, the pressure-temperature (P-T) limits are mainly determined by the fracture toughness of beltline region material with the highest level of neutron embrittlement. However, other vessel parts such as nozzles with structural discontinuities may affect the limits due to the higher stress concentration, even though the neutron embrittlement is insignificant. Therefore, not only beltline material with the highest reference temperature, but also other components with structural discontinuities have to be considered for the development of P-T limits of RPV. In the paper, the pressure-temperature operational limits of a Taiwan domestic pressurized water reactor (PWR) pressure vessel considering beltline and extended beltline regions are established per the procedure of ASME Code Section XI-Appendix G. The three-dimensional finite element models of PWR inlet and outlet nozzles above the beltline region are also built to analyze the pressure and thermal stress distributions for P-T limits calculation. The analysis results indicate that the cool-down P-T limit of the domestic PWR vessel is still dominated by the beltline region, but the heat-up limit is partially controlled by the extended beltline region. On the other hand, the relations of reference temperature between nozzles and beltline region on the P-T limits are also discussed. Present work could be a reference for the regulatory body and is also helpful for safe operation of PWRs in Taiwan.

ASME Section XI Appendix L Flaw Tolerance Evaluation of Pressurized Water Reactor Piping Systems to Support Second License Renewal (80-Years Operation)

2018 ◽  
Author(s):  
Anees Udyawar ◽  
Charles Tomes ◽  
Alexandria Carolan ◽  
Steve Marlette ◽  
Thomas Meikle ◽  
...  
Keyword(s):  
Crack Growth ◽  
Structural Integrity ◽  
Operating Conditions ◽  
Pressurized Water Reactor ◽  
Safe Operation ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flaw Tolerance ◽  
Piping Systems ◽  
Operating Plant

One of the goals of ASME Section XI is to ensure that systems and components remain in safe operation throughout the service life, which can include plant license extensions and renewals. This goal is maintained through requirements on periodic inspections and operating plant criteria as contained in Section XI IWB-2500 and IWB-3700, respectively. Operating plant fatigue concerns can be caused from operating conditions or specific transients not considered in the original design transients. ASME Section XI IWB-3740, Operating Plant Fatigue Assessments, provides guidance on analytical evaluation procedures that can be used when the calculated fatigue usage exceeds the fatigue usage limit defined in the original Construction Code. One of the options provided in Section XI Appendix L is through the use of a flaw tolerance analysis. The flaw tolerance evaluation involves postulation of a flaw and predicting its future growth, and thereby establishing the period of service for which it would remain acceptable to the structural integrity requirements of Section XI. The flaw tolerance approach has the advantage of not requiring knowledge of the cyclic service history, tracking future cycles, or installing systems to monitor transients and cycles. Furthermore, the flaw tolerance can also justify an inservice inspection period of 10 years, which would match a plant’s typical Section XI in-service inspection interval. The goal of this paper is to demonstrate a flaw tolerance evaluation based on ASME Section XI Appendix L for several auxiliary piping systems for a typical PWR (Pressurized Water Reactor) nuclear power plant. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth mechanism, such as fatigue crack growth, and the inspection detection capabilities. The purpose of the Section XI Appendix L evaluation is to demonstrate that a reactor coolant piping system continues to maintain its structural integrity and ensures safe operation of the plant.

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