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

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.

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.

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.

Effects of Delta - Ferrite Content on the Integrity of Reactor Vessel Cladding

2010 ◽  
Author(s):  
Ho-Sang Shin ◽  
Jin-Ki Hong ◽  
Koo-Kab Chung ◽  
Hae-Dong Chung ◽  
Gwang-Yil Kim ◽  
...  
Keyword(s):  
Electron Microscopy ◽  
Scanning Electron Microscopy ◽  
Reactor Vessel ◽  
Bending Test ◽  
Ferrite Content ◽  
Delta Ferrite ◽  
Bending Tests ◽  
Asme Code ◽  
Two Phases ◽  
Scanning Electron

As the design life of new nuclear power plant increases, the austenitic stainless cladding integrity of reactor vessel becomes one of the new concerns. Since 1970’s, there have been some specific recommendations on delta ferrite content of austenitic cladding of reactor vessels and welds. It has been known that the delta ferrite is beneficial for reducing micro-fissure in welds, though the high delta ferrite content increases the probability of embrittlment of welds. In this study, the mechanical and microstructural properties of austenitic weld metals with the limit values of the recommended range (5 ∼ 18 FN) of the delta ferrite control on low alloy steels were characterized by using bending test and scanning electron microscopy. The base metal was ASME Code Sec. II specification SA 508 Gr. 3 Cl. 1 plate and weld materials were EQ308L and EQ309L strips. Four kinds of cladding were deposited with submerged arc welding process on SA508 cl.3 plates. The bending tests were performed through ASME code Sec. IX and the microstructure of fractured surfaces was analyzed by scanning electron microscopy (SEM). In bending tests, there were no fractures except the highest delta ferrite content specimens (28FN). From the SEM observation of fractured surfaces, cracks initiated from the interface between austenite and ferrites phases in the cladding layer and propagated through the continuous interfaces between two phases. For specimens without continuous interfaces of two phases, though the cracks were observed in the interface of phases, the propagation of cracks was not observed. From the test results, continuous interfaces between austenite matrix and ferrite phase provide the path for crack propagation. And the delta ferrite content affects the integrity of cladding of reactor vessel.

Structural Integrity Evaluation of a Reactor Vessel Outlet Nozzle Weld Inlay

2012 ◽  
Author(s):  
N. L. Glunt ◽  
A. Udyawar ◽  
C. K. Ng ◽  
S. E. Marlette
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Repair Process ◽  
Reactor Vessel ◽  
Nuclear Industry ◽  
Outlet Nozzle ◽  
Weld Material ◽  
Asme Code ◽  
Primary Water

Nickel-base weldments such as Alloy 82/182 dissimilar metal (DM) butt welds used in Pressurized Water Reactor (PWR) nuclear power plant components have experienced Primary Water Stress Corrosion Cracking (PWSCC), resulting in the need to repair/replace these weldments. The nuclear industry has been actively engaged in inspecting and mitigating these susceptible DM butt welds for the past several years. Full and Optimized Structural Weld Overlay as well as Mechanical Stress Improvement Process (MSIP®) are some of the mitigation/repair processes that have been implemented successfully by the nuclear industry to mitigate PWSCC. Three conditions must exist simultaneously for PWSCC to occur: high tensile stresses, susceptible material and an environment that is conducive to stress corrosion cracking. These mitigation/repair processes are effective in minimizing the potential for future initiation and crack propagation resulting from PWSCC by generating compressive residual stress at the inner surface of the susceptible DM weld. Weld inlay is an alternative mitigation/repair process especially for large bore nozzles such as reactor vessel nozzles. The weld inlay process consists of excavating a small portion of the susceptible weld material at the inside surface of the component and then applying a PWSCC resistant Alloy 52/52M repair weld layer on the inside surface of the component to isolate the susceptible DM weld material from the primary water environment. The design and analysis requirements of the weld inlay are provided in ASME Code Case N-766. This paper provides the structural integrity evaluation results for a typical reactor vessel outlet nozzle weld inlay performed in accordance with the ASME Code Case N-766 design and analysis requirements. The evaluation results demonstrate that weld inlay is also a viable PWSCC mitigation and repair process especially for large bore reactor vessel nozzles.

Design of a Weld Overlay for a Large Bore Pipe Nozzle to Optimize Residual Stress

2010 ◽  
Author(s):  
Douglas E. Killian
Keyword(s):  
Residual Stress ◽  
Compressive Residual Stress ◽  
Reactor Vessel ◽  
Inside Surface ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Surface Flaws ◽  
Asme Code ◽  
Dissimilar Metal Weld ◽  
Metal Weld

Full Structural Weld Overlays (FSWOL) have been used successfully in the nuclear power industry for a number of years to mitigate and repair small (4″) to medium (10″) bore welded piping components susceptible to primary water stress corrosion cracking (PWSCC). Mitigation is provided by the creation of compressive residual stress on the inside surface of the pipe as layers of weld overlay are deposited over the outside surface of the pipe. ASME Code Case N-740-2 requires that these overlay designs provide adequate structural integrity considering the growth of postulated 75% through-wall inside surface flaws by PWSCC and cyclic fatigue. Application of this repair procedure to larger diameters components such as 30 inch reactor vessel nozzles is not practical due to the large amount of weld metal (overlay thickness) which would be required to satisfy the design requirements of a FSWOL and the associated demands on implementation schedule and exposure to radiation. An alternate procedure is currently being considered for these larger components which utilizes an Optimized Weld Overlay (OWOL) design based on a reduced thickness and smaller postulated flaw. In particular, ASME Code Case 754 specifies, in part, that 50% through-wall inside surface flaws be shown to be acceptable. Furthermore, an OWOL would continue to provide mitigation of materials susceptible to PWSCC by requiring that the thickness of the overlay be sufficient to induce compressive residual stress on the inside surface. This paper presents results of finite element analysis for an optimized weld overlay on a large bore (30″) reactor vessel coolant nozzle dissimilar metal weld, with particular attention to the incremental development of residual stress with each layer of weld metal. Through numerical simulation of the complete fabrication history, including repair of the original dissimilar metal weld, hydrostatic testing, and completion of the nozzle safe end-to-pipe joint prior to implementation of the overlay, the pre-overlay state of stress is defined for use as the basis for evaluating the stress improvement provisions of the weld overlay process. Results are obtained for both kinematic and isotropic hardening rules to study the effect of these two extreme measures of material characterization on the development of residual stress. Additional results are presented to study the sensitivity of the welding simulations to material yield strength and mesh refinement. Predicted stresses are also compared to measured data from a full scale mockup of a large bore reactor vessel nozzle with an optimized weld overlay.

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.

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.

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