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.

Probabilistic Assessment of Margins in Appendix G Methodology

2004 ◽  
Author(s):  
K. K. Yoon ◽  
H. P. Gunawardane ◽  
S. Rosinski
Keyword(s):  
Nuclear Power ◽  
New Technology ◽  
Probabilistic Approach ◽  
Safety Margin ◽  
Temperature Limit ◽  
Pressurized Thermal Shock ◽  
The Us ◽  
Asme Code ◽  
Type Code ◽  
Reactor Pressure

The pressurized thermal shock (PTS) re-evaluation program has been developed at the US NRC with the interaction of the EPRI Materials Reliability Program to revisit the PTS issue using new technology developed over the last decade. The results are encouraging as more realistic assessments of reactor pressure vessel (RPV) integrity are now possible using the NRC probabilistic fracture mechanics code FAVOR. This technology is now being considered for application to the ASME Code Section XI Appendix G methodology for pressure-temperature limit curves. This paper is a first attempt to use a FAVOR-type code to evaluate the true safety margin of nuclear power plant operational pressure temperature limits, especially low temperature over pressurization events. Preliminary results indicate that there is ample room for relaxing these limits based on this probabilistic approach.

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.

Evaluation of P-T Limit Curves According to Alternative Fracture Toughness Requirements

2014 ◽  
Vol 1051 ◽  
pp. 896-901
Author(s):  
Sin Ae Lee ◽  
Sung Jun Lee ◽  
Sang Hwan Lee ◽  
Yoon Suk Chang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Ductile Failure ◽  
Evaluation Procedure ◽  
Limit Curve ◽  
Asme Code ◽  
Temperature And Pressure ◽  
Reactor Pressure

During the heat-up and cool-down processes of nuclear power plants, temperature and pressure histories are to be maintained below the P-T limit curve to prevent the non-ductile failure of the RPV(Reactor Pressure Vessel). The ASME Code Sec. XI, App. G describe the detailed procedure for generating the P-T limit curve. The evaluation procedure is containing the evaluation methods of RTNDT using 10CFR50.61. However, recently, Alternative fracture toughness requirements were released 10CFR50.61a. Therefore, in this study, RTNDT of RPV according to the 10CFR50.61a was calculated and used for evaluation of P-T limit curve of a typical RPV under cool-down condition. As a result, it was proven that the P-T curve obtained from 10CFR50.61 is conservative because RTNDT value obtained from the alternative fracture toughness requirements are significantly low.

Effects of Major Parameters on the P-T Limit Curve Construction of Embrittled Reactor Vessel

2004 ◽  
Vol 261-263 ◽  
pp. 1647-1652
Author(s):  
Sung Gyu Jung ◽  
In Gyu Park ◽  
Chang Soon Lee ◽  
Myung Jo Jhung
Keyword(s):  
Fracture Toughness ◽  
Neutron Fluence ◽  
Crack Depth ◽  
Reactor Vessel ◽  
Sensitivity Analyses ◽  
Crack Orientation ◽  
Limit Curve ◽  
Asme Code ◽  
Potential Failure ◽  
Reactor Pressure

To prevent the potential failure of the reactor pressure vessel (RPV), it is requested to operate RPV according to the pressure-temperature (P-T) limit curve during the heat-up and cool-down process. The procedure to make the P-T limit curve was suggested in the ASME Code but it has been known to be too conservative for some cases. In this paper, the conservatism of the ASME Code Sec. XI, App. G was investigated by performing a series of sensitivity analyses. The effects of six parameters such as crack depth, crack orientation, clad thickness, fracture toughness, cooling rate, and neutron fluence were analyzed. The results of P-T limit curves are compared to one another.

Extending Temperature Limit of HMX: A Case Study for Perforation in the Gulf of Thailand

2021 ◽  
Author(s):  
Thum Sirirattanachatchawan ◽  
Phuvit Chaiwan ◽  
Pradondate Ut-ang ◽  
Wiwat Pattarachupong ◽  
Toon Puttisounthorn ◽  
...  
Keyword(s):  
High Temperature ◽  
Exposure Time ◽  
Oil And Gas ◽  
Operating Time ◽  
Safety Margin ◽  
Oil And Gas Industry ◽  
Temperature Limit ◽  
Explosion Hazard ◽  
Time Optimization ◽  
Two Parameters

Abstract The current explosive limit chart using in the oil and gas industry published with unclear condition of safety factor and consequence of overexposing temperature and time, resulting in many published papers disclose the possibility of expanding the safe-operating envelope of HMX. HMX is preferable because it typically provides deeper penetration than HNS but less stability at high temperature. Therefore, this study aims to maximize use of HMX for hollow gun perforation in typical environment in the GOT. The explosive temperature limit depends on two parameters, exposure time and temperature. The maximize use of HMX could achieve by, either ways, reducing the exposure time or extending the temperature limit line. Firstly, the operating time optimization is doable by using statistic record of the depth perforated by HNS and practical running speed together with a 20% safety margin. Secondly, expanding the temperature limit of HMX is a precise task because the HMX once exceeding the stability temperature, the perforating performance losses and explosion hazard arises due to thermal decomposition. However, this could be creditable by integrating the published explosive testing results over the current operating-envelope and applying a safety margin. The represented operating time, counting from running in hole to tool on surface, for perforation with E-line unit in the high-temperature environment could reduce by an hour. This operating time allows the temperature limit of HMX increasing by only 7 F, which considering as insignificant. The integrating result of three published paper indicated no explosive deflagration happens if the temperature is below the "Fiasco line" – introduced by a company, however, the operating time longer than 200 hours is not incorporated. By applying safety margin, the new operating envelope of HMX in the hollow carrier proposes between the typical and the Fiasco line. Combing time optimization and the new line, the HMX temperature limit extends from 375 F to 394 F for 2.2 operating hours. This new criterion has been applied successfully since 2018; 325 m of HMX achieved perforation condition without an indication of misfire or catastrophic self-detonation, resulting in reduce 50% of HNS consumption. In conclusion, the new temperature cutoff is valid for maximizing the use of HMX with a reasonable safety margin.

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.

Small Bore Piping System Fatigue — Determination of Favored Crack Location and Related Admissible Vibration Criterion

2012 ◽  
Author(s):  
Julien Berger ◽  
Jacques Guillou ◽  
Emilie Poirier
Keyword(s):  
Nuclear Power ◽  
Piping System ◽  
Mean Square ◽  
Weld Quality ◽  
Root Mean Square Velocity ◽  
Crack Location ◽  
Piping Systems ◽  
Potential Failure ◽  
Analytical Expressions

Each year, a few cracking of small bore piping systems occurs in French Nuclear Power Plant (NPP) due to vibratory fatigue. The particular design of such small bore piping system (basically composed of a socket, a small pipe and a valve with or without tubing) allows the crack initiation at two different locations of the socket weld toes: on the main pipe or on the small bore piping. The favored location depends on specific features, from geometrical and material characteristics to the weld quality. The knowledge of the favored location with its related vibratory limit is an important issue in terms of inspection plans, plant availability and integrity of safeguard systems. The main purpose of the present work is to provide analytical expressions for estimating the location of a potential failure by cracking fatigue and its initiation. This threshold is given by a vibratory limit expressed in terms of root mean square velocity. Inherent uncertainties (e.g. preload, weld quality…) impose to consider the proposed expressions as guidelines for the prediction of the crack location and initiation.

Comparison Between Different Calculation Methods for Determining Bolting Up Moments

2016 ◽  
Author(s):  
Alexander Mutz ◽  
Manfred Schaaf
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
European Standard ◽  
Calculation Algorithm ◽  
Operational Conditions ◽  
Class 1 ◽  
Asme Code ◽  
Floating Type

There are several different standards for flange calculation used in the European and so in the Suisse context. The European Standard EN 1591-1 that is used for the calculation of bolted flanged joints and EN 13555 in which the determination of the required gasket characteristics are defined were reissued in 2013 and in 2014, respectively. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components in Suisse nuclear power plants it is also used for class 1 flange connections. A standard for the determination of the required gasket characteristics is not well established which leads to a lack of clarity. As a hint, different m and y values for different kind of gaskets are invented in ASME BPVC Section III. As cited in the Note of table XI-3221.1-1 the values m and y are not mandatory. In Switzerland, mainly the ASME BPVC should be used for the calculation of flange connections. The aim of the ASME Code is more or less not the tightness of the flanges but the integrity. Therefore, stresses are derived for dimensioning the flanges. Following loads are not considered neither for calculation of stresses nor for calculation of tightness. Considering the experience with flanges in general it could be asked, if it is more useful to look at the tightness than at the stresses. The codes KTA 3201.2 and KTA 3211.2 regulate the calculation of flange connections in German nuclear power plants. Stresses in floating type and in metal-to-metal contact type of flange connections and the tightness are calculated for the different load cases. In this paper, the differences in the calculations are shown between KTA 3211.2, ASME BPVC, Section III, Appendix 11, EN 1591-1 and Finite element calculations. In all load cases leakage shouldn’t occur. Therefore, internal pressure and temperature in test and operational conditions after bolting-up are also considered for the stress calculation if it is possible in the calculation algorithm.

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