Evaluation of Fast Neutron Inserted to RPV Materials and Management of RPV Lifetime for PWR Power Plant

2011 ◽  
Vol 413 ◽  
pp. 514-519
Author(s):  
Shou Hai Yang ◽  
Yi Xue Chen ◽  
Ye Hu ◽  
Ye Shuai Sun ◽  
Guo Ping Quan ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Neutron Fluence ◽  
Daya Bay ◽  
Low Leakage ◽  
Inner Surface ◽  
Reactor Pressure

For the PWR power plant in service, the lifetime of RPV is mainly determined by fast neutron (E>0.1MeV and E>1.0MeV) fluence. A new Low-Leakage Core Loading (LLCL) pattern is considered to prolong the lifetime of RPV to 60 years for Daya Bay nuclear power plant. Fast neutron (E>0.1MeV and E>1.0MeV) flux inserted to Reactor Pressure Vessel (RPV) under LLCL pattern is calculate by MCNP and compared with that under Long-Short Alternating Core Loading (LSACL) pattern. The results show that fast neutron fluence over inner surface of RPV under LLCL pattern decrease 28.50% (E>1.0MeV) and 28.22% (E>0.1MeV) compared with LSACL pattern and it is feasible expand the life of LWR-RPV to 60 years by LLCL pattern.

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.

Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

2014 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

Comparison of Allowable Bending Moments for Circumferentially Flawed Pipes in ASME Boiler and Pressure Vessel Code Section XI Nonmandatory Appendices C and H

2019 ◽  
Author(s):  
Hune-Tae Kim ◽  
Ji-Su Kim ◽  
Jun-Min Seo ◽  
Yun-Jae Kim ◽  
Kuk-Hee Lee ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Bending Moments ◽  
Analytical Evaluation ◽  
Evaluation Procedures ◽  
Failure Assessment ◽  
Inner Surface ◽  
Plant Components

Abstract In this paper, allowable bending moments for circumferential inner surface cracked pipes are evaluated. ASME Boiler and Pressure Vessel code Section XI, “Rules for Inspection of Nuclear Power Plant Components” provides analytical evaluation procedures. Analytical evaluation methods based on the failure mechanism are provided in nonmandatory Appendix C and those based on failure assessment diagram are given in nonmandatory Appendix H. Allowable bending moments are evaluated using both appendices and compared with experiments. Conservativeness is compared quantitatively between both methods by normalizing allowable bending moments with experimental maximum moments.

Welding Residual Stress in a Large Diameter Nuclear Reactor Pressure Vessel Nozzle

2013 ◽  
Vol 135 (2) ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Large Diameter ◽  
Reactor Pressure Vessel ◽  
Welding Residual Stress ◽  
Stress Fields ◽  
Reactor Pressure

The Atucha II nuclear power plant is a pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant was designed by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. Welding residual stresses in nuclear power plant piping can lead to cracking concerns later in the life of the plant, especially for stress-corrosion cracking. Hence, understanding the residual stress distribution from welding is important to evaluate the reliability of pipe and nozzle joints with welds. In this paper, a large-diameter reactor pressure vessel (RPV) hot-leg nozzle was analyzed. This is a nozzle from Atucha II nuclear power plant in Argentina. The main piping material is 20MnMoNi55 with Tenacito 65R weld metal, and inner diameter (ID) welded cladding at the girth weld locations is made of 309L. The special materials and weld geometry will lead interesting welding residual stress fields. In addition, postweld heat treatment (PWHT) of the girth welds and its boundary conditions could also play an important role in determining welding residual stress fields at the plant’s normal operating conditions. Sensitivity analyses were conducted and the technical observations and comments are provided.

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