Fracture Assessment for Pressurized Water Reactor Vessel Nozzles Subjected to Pressure and Thermal Loading

2021 ◽  
Author(s):  
Qibao Chu ◽  
Qing Wang ◽  
Yonggang Fang ◽  
Wei Tan
Keyword(s):  
Crack Front ◽  
Thermal Loading ◽  
Structural Integrity ◽  
Safety Margin ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Main Challenge ◽  
Fracture Assessment ◽  
3D Finite Element Method

Abstract To ensure the structure integrity of the RPV, the main challenge is the embrittlement of beltline material. However, the stress of inlet or outlet nozzles of the RPV which are in general reinforced in comparison with the beltline, is more complex especially under the thermal loads. In recently studies, a lot of works have been done to show that the nozzle region may be more challenging under some conditions. In this paper, a fracture assessment for the RPV nozzles subjected to pressure and thermal loading is discussed using the software ABAQUS 6.12 and Zen Crack 7.9-3. It includes: SIF calculation based on 3D finite element method; structural integrity assessment under a typical LOCA transient; and the fatigue crack growth evaluation under cyclic loading situations. The results show that the SIF along the crack front is obviously asymmetric, and only to assess the safety of the deepest point along the crack front in the ASME and RCC-MR codes may be reconsider. If the KIa criteria is applied, under a typical LOCA transient, it is difficult to obtain an effective fracture safety margin for a 1/4 thickness crack, while based on the KIC criteria, the nozzle is shown to be safe in the case study. The shape of the surface elongated crack (which is often easily produced in the nozzle area) tends to be circle under the cyclic pressure loading situation which shows the crack shape assumed in the ASME and RCC-MR codes is reasonable.

Neutron Embrittlement Analysis of Qinshan 320MWe RPV on OLE

2017 ◽  
Author(s):  
Tao Hongxin ◽  
He Yinbiao ◽  
Cao Ming ◽  
Shen Rui
Keyword(s):  
Service Life ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Nuclear Safety ◽  
Radioactive Material ◽  
Pressurized Water Reactor ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Key Factor

One of the fundamental requirements on nuclear safety is to prevent the radioactive material from being released. Therefore, it is paramount to maintain the structural integrity of the pressure boundary of the reactor coolant system. The reactor pressure vessel (RPV), under high temperature, high pressure and high radiation in operation, is the most important as well as a Class I nuclear safety equipment. For a pressurized water reactor (PWR), the life of the RPV determines the service life of the entire nuclear power plant. The key factor controlling the life of a RPV is the accumulation of the neutron flux and which induces irradiation embrittlement degrading the anti-fracture capability of the RPV material. Several anti-fracture capability assessments carried out for the Qinshan 320MWe (QS1) RPV, such as (a) the structural integrity assessment against pressurized thermal shocks; (b) the fracture mechanics assessment under irradiation; (c) the P-T limit curves revised; (d) the evaluation of USE. They all demonstrated that the structural integrity of the QS1 RPV would be maintained for the extended service life.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

Environmental Crack Growth Predictions for Piping Components

2007 ◽  
Author(s):  
Sam Ranganath ◽  
Guy DeBoo
Keyword(s):  
Stainless Steel ◽  
Crack Growth ◽  
Test Data ◽  
Structural Integrity ◽  
Reference Curve ◽  
Pressurized Water ◽  
Water Reactor ◽  
Integrity Assessment ◽  
Asme Code ◽  
Reference Curves

Structural integrity assessment of reactor components requires consideration of crack growth. A key input to this is the development of reference stress corrosion crack (SCC) growth rate curves for use in the structural evaluation. The ASME Section XI Task Group on SCC Reference Curve is looking into available SCC data for stainless steel and nickel based alloys and associated weldment in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments. The test data show significant data scatter in crack growth rates (CGR). The conservative approach is to develop reference curves that bound all available data so that upper bound crack growth predictions. While this approach may be conservative, it may lead to excessive estimates of crack growth and result in unrealistic (and often meaningless) structural margin predictions. Selection of the appropriate SCC reference curves requires realistic interpretation of test data so that the predictions are consistent with field behavior and provide reasonable, but conservative assessment. This paper describes crack growth assessment for stainless steel piping and Alloy 600 safe end components with Alloy 182/82 welds in BWR environment. The results from the crack growth analysis for piping can be used to determine whether a proposed reference curve provides reasonable results. The objective is to use the piping and safe end crack growth predictions to develop optimal SCC Reference Curves for use in ASME Code evaluations.

Reactor Pressure Vessel Integrity in Light of the Evolution of Materials Science and Engineering

2005 ◽  
Vol 473-474 ◽  
pp. 287-292
Author(s):  
Péter Trampus
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Materials Science ◽  
Reactor Pressure Vessel ◽  
Special Focus ◽  
Science And Engineering ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Materials Science And Engineering ◽  
Reactor Pressure

Structural integrity of the reactor pressure vessel of pressurized water reactors is one of the key safety issues in nuclear power operation. Integrity may be jeopardized during operational transients. The problem is compounded by radiation damage of the vessel structural materials. Structural integrity assessment as an interdisciplinary field is primarily based on materials science and fracture mechanics. The paper gives an overview on the service induced damage processes and associated changes of mechanical properties, the prediction of degradation and the assessment of the entire component against brittle fracture with a special focus on how the evolution of materials science and engineering has contributed to reactor vessel structural integrity assessment.

scholarly journals Fatigue Damage Management According to Performance Based Maintenance (PBM) Concept

2021 ◽  
Author(s):  
Masayuki Kamaya
Keyword(s):  
Fatigue Life ◽  
Crack Growth ◽  
Thermal Loading ◽  
Water Environment ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Term Operation ◽  
Primary Water ◽  
Crack Growth Curve

Abstract A maintenance concept of performance based maintenance (PBM) has been proposed by the current author. According to the PBM concept, inspection results are considered in determining the next inspection schedule. In this study, this concept was applied to fatigue degradation for stainless steel components in the pressurized water reactor (PWR) primary water environment. It is possible to estimate the fatigue life for the PWR water environment from that obtained in an air environment and the parameter Fen, which represents the ratio of the fatigue life in the air and PWR water environments. It was shown that the fatigue life prediction using Fen can be replaced by the crack growth analysis using the growth rate for the PWR water environment. Then, the crack growth was predicted for a thermal loading assuming the growth occurred in the PWR water environment. It was shown that the duration until the next inspection could be optimized based on the inspection results together with the crack growth curve. A long term operation before the inspection resulted in a longer duration until the next inspection.

Applicability of Fracture Toughness Curves Developed for Japanese Pressure Vessel Steels to Structural Integrity Evaluation

2015 ◽  
Author(s):  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Charpy Impact ◽  
Compact Tension ◽  
Transition Curve ◽  
Pressurized Water Reactor ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Reactor Pressure

Surveillance tests have been conducted on Japanese Pressurized Water Reactor (PWR) plants for more than 40 years to monitor irradiation embrittlement of reactor pressure vessel (RPV) beltline materials. Fracture toughness specimens are contained as well as tensile and Charpy impact specimens in a surveillance capsule and utilized for structural integrity evaluation. Therefore, a lot of fracture toughness data have been obtained by fracture toughness tests using such as Compact Tension (CT) and Wedge Opening Loading (WOL) specimens. More than one thousand data have been accumulated for both unirradiated and irradiated materials until 2013. Additionally, in terms of fracture toughness, Master Curve (MC) concept has been widely used for fracture toughness transition curve expression of ferritic steels. Considering such a situation, the new fracture toughness curves using Tr30, which denotes Charpy V-notch 30ft-lb transition temperature, as an indexing parameter were developed based on MC concept depending on product form for Japanese RPV steels in 2014. In this study, applicability of the newly developed curves of Japanese RPV steels to structural integrity evaluation is investigated. Especially, this paper focused on conservatism of the curves and the adequate margin to be added in evaluation of RPV integrity employing statistical methodology.

Simulation of Acoustic-Structure Interaction by Using Finite Element Analysis in Reactor Pressure Vessel of PWR

2018 ◽  
Author(s):  
Koji Maeta ◽  
Keisuke Matsuyama ◽  
Hirokazu Sugiura ◽  
Shigeyuki Watanabe ◽  
Hideyuki Morita ◽  
...  
Keyword(s):  
Structural Integrity ◽  
Acoustic Analysis ◽  
Structural Response ◽  
Scale Model ◽  
Pressurized Water Reactor ◽  
Pressure Pulsations ◽  
Element Analysis ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Pressure Distributions

The reactor coolant pump (RCP) in a pressurized water reactor (PWR) plant generates pressure pulsations at multiple frequencies. These pressure pulsations excite the acoustic modes inside the reactor vessel (RV), and cause significant acoustic loads on the reactor internals (RIs). For verifying the structural integrity of the RIs, it is important to predict the acoustic loads, which is used for vibration analysis of the RIs. Traditionally, an analytical method, assuming that structures are rigid, has been used in order to predict the acoustic pressure distributions inside the RV [1]. However, water in actual PWR plant is heavy enough to influence structural response, so that it is required to use methods for a coupled structural acoustic system. In this article, the coupled structural-acoustic analysis using the commercial software ANSYS is proposed in order to predict the acoustic loads, and the applicability of this method is discussed. The structural-acoustic interactions inside the RV are investigated by element tests and the scale model test. The acoustic pressures measured by these tests are compared with the calculated results.

Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel

2016 ◽  
Vol 138 (3) ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsiung-Wei Chou
Keyword(s):  
Structural Reliability ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of pressurized water reactor (PWR) reactor pressure vessels (RPVs) over their long-term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the U.S. is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Furthermore, two overcooling transients of steam generator tube rupture (SGTR) and pressurized thermal shock (PTS) addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR RPVs and can also be referred as its regulatory basis in Taiwan.

Effect of Uneven Crack Front on Crack Tip Mechanics and the Implication to Stress Corrosion Crack Growth

2009 ◽  
Author(s):  
He Xue ◽  
Zhanpeng Lu ◽  
Hiroyoshi Murakami ◽  
Tetsuo Shoji
Keyword(s):  
Growth Rate ◽  
Crack Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Crack Front ◽  
Pressurized Water Reactor ◽  
Element Analysis ◽  
Pressurized Water ◽  
Water Reactor ◽  
Nickel Base

Uneven crack fronts have been observed in laboratory stress corrosion cracking tests. For example, cracking fronts of nickel-base alloys tested in simulated boiling water reactor (BWR) and pressurized water reactor (PWR) environments could exhibit uneven crack front. Analyzing the effect of an uneven crack front on further crack growth is important for quantification of crack growth. Finite-Element analysis shows that the local KI distribution can be significantly affected by the shape and size of the uneven crack front. Stress intensity factor at the locally extended crack front can be significantly reduced. Since generally there is a nonlinear CGR versus KI relationship, it is expected that crack growth rate at the locally extended crack front can be significantly different from those in the neighboring areas. There could be several patterns for the growth of an uneven crack front. For example, once initiated, the crack growth rate in areas other than the locally protruded front would become higher and then the whole crack front would tend to become uniform. On the other hand, if the crack growth in other areas is still low, there is a possibility that the crack growth rate at the front tip would slow down.

Mechanistic Formulation of PWSCC Growth Rates of Ni-Base Alloys and Weld Metals

2011 ◽  
Author(s):  
Zhanpeng Lu ◽  
Tetsuo Shoji ◽  
He Xue ◽  
Chaoyang Fu
Keyword(s):  
Thermal Activation ◽  
Structural Integrity ◽  
Thermal Activation Energy ◽  
Evaluation Procedure ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Weld Metals ◽  
Primary Water ◽  
Integrity Analysis

Several Ni-base alloys and their weld metals such as Alloy 600 and Alloy 82/182 suffered from stress corrosion cracking in pressurized water reactor primary water environments. Materials Reliability Program (MRP) proposed a CGR disposition curve in a report MRP 55 for PWSCC of thick-section Alloy 600 materials. This deterministic CGR equation has been adopted by Section XI Nonmandatory Appendix O of the ASME Boiler and Pressure Code for flaw evaluation. MRP also proposed a CGR disposition curve in MRP report 115 for PWSCC of Alloy 82/182/132 weld metals. In the same fashion, JSME and JNES also provided CGR disposition curves in the flaw evaluation procedure in structural integrity analysis. Stress intensity factor (K), temperature and thermal activation energy are included in both MRP 55 and MRP 115 reports. Both MRP 55 and MRP 115 are engineering-based rather than mechanism-based. The fundamental correlations such as crack growth rate vs. K are quantified based on the theoretical model and screened experimental data, which are compared to the reported disposition curves and used for improving the prediction.

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