A Study on Pressure-Temperature Limit Curve for Beltline and Outlet Nozzle According to New Regulation Requirements

2021 ◽  
Author(s):  
Wan-Jun Ma ◽  
Kyung Seok Chung ◽  
Si-Hwa Jeong ◽  
Tae-Young Ryu ◽  
Jae-Boong Choi ◽  
...  
Keyword(s):  
Temperature Limit ◽  
Limit Curve ◽  
Outlet Nozzle

Parametric Study on Pressure-Temperature Limit Curve

2016 ◽  
Author(s):  
Shin-Beom Choi ◽  
Han-Bum Surh ◽  
Jong-Wook Kim
Keyword(s):  
Sensitivity Analysis ◽  
Fracture Mechanics ◽  
Crack Depth ◽  
Temperature Limit ◽  
Round Robin ◽  
Robin Problem ◽  
Limit Curve ◽  
Element Analysis ◽  
Probabilistic Fracture Mechanics ◽  
Key Parameter

The final goal of this study is to solve the round-robin problem for the safety of a reactor pressure vessel by adopting a finite element analysis and probabilistic fracture mechanics. To do so, a sensitivity analysis and a deterministic analysis should be conducted. This paper contains the results of the sensitivity analysis as intermediate results of a round-robin problem. Key parameters such as the initial Reference Temperature for Nil Ductility Transition, Ni contents, Cu contents, fluence, and input transient were chosen to conduct the sensitivity analysis. In addition, different values of crack depth to the thickness ratio are considered to develop FE models. Moreover, a series of FE analyses are carried out. As a result, each key parameter has an influence on RTNDT and KIc. This means that the P-T limit curve is shifted. If the value of each key parameter is increased, the P-T limit curve is moved to the right side. Therefore, the operating area of the P-T limit curve should be reduced. The results of this paper will be very helpful in enhancing our understanding of the P-T limit curve. In addition, it will be used to adjust the probabilistic fracture mechanics and solve the round-robin problem.

Determination of Pressure-Temperature Limit Curves Using Deterministic and Probabilistic Approaches

2006 ◽  
Vol 306-308 ◽  
pp. 333-338
Author(s):  
Sung Gyu Jung ◽  
In Gyu Park ◽  
Chang Soon Lee ◽  
Hyun Su Kim ◽  
Tae Eun Jin
Keyword(s):  
Nuclear Power ◽  
Probabilistic Approach ◽  
Operating Time ◽  
Safety Margin ◽  
Temperature Limit ◽  
Limit Curve ◽  
Level Increase ◽  
Asme Code ◽  
Potential Failure

Reactor pressure vessel (RPV) is the most critical component in nuclear power plant. RPV is subjected to radiation embrittlement, which is characterized as neutron fluence-dependent reduction in fracture toughness of the material. Therefore, risk for potential failure of RPV increases as operating time and fluence level increase. To prevent the potential failure, it is requested for RPV to operate in accordance with pressure-temperature (P-T) limit curve during operation. However, it has been reported that P-T limit curve which is typically developed in accordance with the procedure in ASME code is too conservative. Therefore, in order to investigate the conservatism of current P-T limit curve and develop more realistic one, probabilistic approach based on the risk was utilized in this paper. The resulting P-T limit curve is very higher than that from deterministic approach, and can be used as alternative operation limit of the RPV, because probabilistic P-T limit curve seems to have enough safety margin for potential failure of RPV.

Comparison of Pressure-Temperature Limit Curves Calculation

2011 ◽  
Author(s):  
F. Lu ◽  
H. Y. Qian ◽  
P. Huang ◽  
R. S. Wang
Keyword(s):  
Fracture Toughness ◽  
Nuclear Power ◽  
Temperature Limit ◽  
Nuclear Industry ◽  
Primary Concern ◽  
Limit Curve ◽  
Irradiation Embrittlement ◽  
Strong Basis ◽  
Asme Code ◽  
Made In

Reactor Pressure Vessel (RPV) is one of the most important components in a nuclear power plant (NPP). The primary concern of aging mechanism for RPV is irradiation embrittlement. In order to prevent brittle fracture, during NPP heatup and cooldown processes, the pressure and temperature in RPV should be kept under the pressure-temperature (P-T) limit curve. The P-T limit curve method originated from a WRC bulletin in 1972 and was included in ASME Sec. XI App. G.. Since then, much effort for reducing the conservatism of the P-T limit curve calculation has been made in many countries. Technology developed over the last 30 years has provided a strong basis for revising the P-T limit curve methodology. Up to now, changes have been made in the latest version of the ASME and RCCM codes. In this paper, the P-T limit curve methodologies given by the ASME code, the RCCM code, and Chinese Nuclear Industry Standard EJ/T 918 are studied. The differences of the P-T curve methodologies in previous and current versions for the ASME and RCCM codes are discussed. Two P-T curve calculation methods based on the RCCM code Ver. 2007 are proposed, due to lack of specific description for the calculation method in the RCCM code. Comparison of the P-T curves obtained using methods from different codes is also performed. It shows that using static fracture toughness KIC instead of reference fracture toughness KIR to calculate P-T curves can increase acceptable operating region during NPP heatup and cooldown processes significantly. Comparing with the latest versions of the ASME and RCCM codes, the current Chinese Standard is more conservative.

The Pressure-Temperature Limit Curve of System-Integrated Modular Advanced Reactor Against Nonductile Failure

2007 ◽  
Author(s):  
Yoon-Suk Chang ◽  
Hyuk-Soo Chang ◽  
Sang-Min Lee ◽  
Jae-Boong Choi ◽  
Young-Jin Kim ◽  
...  
Keyword(s):  
Temperature Change ◽  
Sea Water ◽  
Three Dimensional ◽  
Design Procedure ◽  
Temperature Limit ◽  
Water Desalination ◽  
Electricity Production ◽  
Limit Curve ◽  
Crack Location ◽  
Three Dimensional Finite Element

A system-integrated modular advanced reactor is being developed for multi purposes such as electricity production, sea water desalination and so on in Korea. While ASME Codes provide simplified design and operation procedures to determine allowable loadings for pressure retaining materials in components, the procedures are applicable when a temperature change rate associated with startup and shutdown is less than about 56°C/hr. If the procedures are applied to a rapid temperature change, results would be overly conservative. The objective of this research is to assess an applicability of the simplified design procedures to reactor coolant system of the integrated modular reactor with the change rates of 56°C/hr and 100°C/hr. To investigate effects of cooldown rate, heatup rate and surface crack location, systematic three-dimensional finite element analyses are carried out. The resulting pressure-temperature limit curves are compared with those obtained from the ASME Sec. XI operating procedure as well as Sec. III design procedure. Thereby, it was proven that the specific design features significantly affect the safe design region in the pressure-temperature limit curve to prevent a nonductile failure.

Margins Assessment of Pressure-Temperature Limit Curves for a RPV

2018 ◽  
Author(s):  
Lv Feng ◽  
Zhou Gengyu ◽  
Qian Haiyang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Temperature Limit ◽  
Previous Method ◽  
Limit Curve ◽  
Irradiation Embrittlement ◽  
Temperature Range ◽  
Shell Temperature ◽  
Curve Method ◽  
Reactor Pressure

In order to prevent brittle fracture, the pressure and temperature in a reactor pressure vessel (RPV) is controlled by pressure-temperature (P-T) limit curves during the heat-up and cool-down processes. Nuclear power plants should update the P-T limit curves periodically, because of RPV material irradiation embrittlement. Too restricted P-T limit curves may cause difficulty of operating a reactor. The 2007 edition of the RCCM code Annex ZG provides a new method for defining the P-T limit curves. In this paper, two types of the P-T limit curves for a French type RPV are established by different methods, which are the current operation limits based on the 1993 edition of the RCCM code and the new proposed limits according to the 2007 edition of the code. The margins of the current P-T limit curves are evaluated by comparing with the new proposed limit curves. Furthermore, the reasonability of improvements of the new P-T limit curve method is discussed, and their individual effects are investigated, including the conventional defect size, the required material toughness and the stress intensity factor plastic correction. The present results indicate that the current P-T limit curves for the RPV studied are conservative and have about 25∼70 °C margin in the transition temperature range and about 10∼12MPa in the upper shell temperature range, depending on different conditions. The new P-T limit curve method, which not only removes some conservative assumptions in the previous method but also restricts some requirements, is more reasonable and can provide a relaxed operation window. Present results can be a reference for the nuclear power plant owner to release the operation limits and is helpful in enhancing our understanding of the P-T limit curve.

Pressure-temperature limit curve for reactor vessel evaluated by ASME code

2002 ◽  
Vol 14 (2) ◽  
pp. 191-208 ◽  
Author(s):  
Myung Jo Jhung ◽  
Seok Hun Kim ◽  
Sung Gyu Jung
Keyword(s):  
Temperature Limit ◽  
Reactor Vessel ◽  
Limit Curve ◽  
Asme Code

Evaluation of Pressure-Temperature Limit Curve Considering Revised Stress Intensity Factor Method

2020 ◽  
Author(s):  
Yunjoo Lee ◽  
Hyosub Yoon ◽  
Kyuwan Kim ◽  
Jongmin Kim ◽  
Hyunmin Kim
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Pressure Vessel ◽  
Temperature Limit ◽  
Reactor Pressure Vessel ◽  
Flaw Size ◽  
Limit Curve ◽  
Asme Code ◽  
Reactor Pressure

Abstract Pressure-Temperature limit methodology is based on the rules of Appendix G in Section XI of the ASME Code in accordance with the requirements of 10 CFR 50, Appendix G, and the Appendix G in Section XI method refers to Welding Research Council (WRC) Bulletin 175 (WRC175). Flaw size is an important factor to protect the reactor pressure vessel from brittle failure but is not explicitly documented in WRC175. However, according to the recent change of Appendix G, the ¼ thickness (¼T) flaw size is postulated in the surface of the nozzle inner corner for the evaluation of Pressure-Temperature limit. In this paper, stress intensity factor is computed by using 3D finite element analysis (FEA) considering ¼T corner cracks of inlet nozzle and outlet nozzle in reactor pressure vessel. The result is compared with the stress intensity factor using influence function in the ASME Code. The results of stress intensity factor in accordance with the ASME Code are more conservative than those of the 3-D FEA with a crack. The allowable pressure and operation region in Pressure-Temperature limit curve are affected by the calculation methods of stress intensity factor.

Demonstrate ASME Section XI Appendix G Margins for Pressurized Water Reactor Inlet and Outlet Nozzle Corner Regions

2018 ◽  
Author(s):  
Anees Udyawar ◽  
J. Brian Hall ◽  
Justin Webb ◽  
Alexandria Carolan
Keyword(s):  
Stress Intensity ◽  
Stress Intensity Factors ◽  
Temperature Limit ◽  
Flaw Size ◽  
Pressurized Water Reactor ◽  
Intensity Factors ◽  
Outlet Nozzle ◽  
Element Analysis ◽  
Pressurized Water ◽  
Water Reactors

Since the implementation of pressure-temperature (P-T) limit curves in the 1960s for light water reactors, the P-T limit curves have been based on the limiting locations in the reactor coolant system, which are typically the irradiated reactor pressure vessel (RPV) region adjacent to the core (beltline) and the closure head flange. Recently, it has been questioned as to whether the reactor vessel inlet or outlet nozzle corners could be more limiting due to the stress concentration at these locations. The discussion presented in this paper provides technical justification that the RPV nozzle corner P-T limit curves are bounded by the traditional P-T limit curves for the pressurized water reactors (PWRs). The current approach in evaluating the Pressurized Water Reactor Inlet and Outlet nozzle corner regions with respect to plant heatup and cooldown Pressure Temperature Limit Curves contains a number of conservatisms. These conservatisms include postulation of a large 1/4T flaw at the nozzle corner region, use of RTNDT (reference nil-ductility temperature), and fracture toughness prediction based on plane strain fracture toughness. The paper herein discusses several factors that can be considered to improve the pressure temperature limit curves for nozzle corners and increase the operating window for nuclear power plant operations. Prior to the 2013 edition, the ASME Section XI Appendix G did not require the use of a 1/4T flaw for the nozzle corners; furthermore, a smaller postulated flaw size is permissible. Based on inspection capability and experience, a smaller flaw size can easily be justified. The use of a smaller flaw size reduces the stress intensity factors and allows for the benefit of being able to take advantage of increased material toughness due to the loss of constraint at the nozzle corner geometry. The analysis herein considers the calculation of stress intensity factors for small postulated nozzle corner flaws based on a 3D finite element analysis for Westinghouse PWR inlet and outlet nozzle corner regions. The Finite Element Analysis (FEA) stress intensity factors along the crack front are used in the determination of allowable pressures for the cooldown transient Pressure-Temperature limit curves.

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