Prediction of Irradiation Embrittlement for Chinese Domestic A508-3 Steel

2010 ◽  
Author(s):  
Feng Lu ◽  
Rongshan Wang ◽  
Ping Huang ◽  
Haiyang Qian
Keyword(s):  
Nuclear Power ◽  
Surveillance Program ◽  
Primary Concern ◽  
Regulatory Requirements ◽  
Irradiation Embrittlement ◽  
Safety Standards ◽  
Rpv Steel ◽  
The Us ◽  
Calculation Results ◽  
Reactor Pressure

Reactor Pressure Vessel (RPV) is one of the most important components in nuclear power plant (NPP). The aging mechanism of primary concern for RPV is irradiation embrittlement, which can result in a decrease of fracture toughness of RPV steel. Prediction of irradiation embrittlement for a certain Chinese domestic manufactured A508-3 steel is performed. The calculation results given by the US safety standards, the French RCC-M standards and other international safety standards are compared with each other and compared to the data from commercial operation NPP surveillance program. The effect of neutron fluence is also investigated. Furthermore, the property of the steel against irradiation embrittlement is evaluated with the regulatory requirements in the relevant standards. It can be predicted that the Chinese domestic steel satisfies the requirements in these standards.

Surveillance Program for Irradiation Embrittlement of Reactor Pressure Vessels in Japan

2018 ◽  
pp. 202-218
Author(s):  
Takatoshi Hirota ◽  
Takashi Hirano ◽  
Masayuki Uchihashi ◽  
Tetsuya Toyoda ◽  
Shinichi Takamoto ◽  
...  
Keyword(s):  
Pressure Vessels ◽  
Surveillance Program ◽  
Irradiation Embrittlement ◽  
Reactor Pressure

Analysis of Dynamic Stress in the Reactor Core Barrel due to the Pressure Pulsations

2015 ◽  
Author(s):  
Igor Orynyak ◽  
Iaroslav Dubyk ◽  
Anatolii Batura
Keyword(s):  
Nuclear Power ◽  
Dynamic Stress ◽  
Reactor Core ◽  
Pressure Pulsations ◽  
Power Station ◽  
Coolant Pump ◽  
The Core ◽  
Calculation Results ◽  
Full Power ◽  
Reactor Pressure

This article presents vibrations analysis of the reactor core barrel caused by pressure pulsations induced by the main coolant pump. For this purpose, the calculations of the pressure distribution in the annulus between the core barrel and the reactor pressure vessel, bounded above by a separating ring were performed. Using transfer matrix method is obtained the solution of two-dimensional problem of pressure pulsations in the annulus between reactor core barrel and reactor vessel. The calculation results are compared with the pulsation pressure measurements made at commissioning unit 2 of the South Ukraine Nuclear Power Station. The distribution of pressure over the height of core barrel was obtained, which makes possible to estimate its strength for variant deformation of the core barrel as a beam, and in the case of deformation of the core barrel as a shell. The calculation results are used to assess the reliability of core barrel pre-load, which clamps the core barrel flange in place at the top, at full power operating.

Prediction of Radiation Embrittlement for Highly Irradiated Reactor Pressure Vessel Steels

2009 ◽  
Author(s):  
Hiroshi Matsuzawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Test Results ◽  
Prediction Formula ◽  
Irradiation Embrittlement ◽  
Nuclear Plants ◽  
Domestic Plant ◽  
Test Reactor ◽  
Reactor Pressure

There are 53 (fifty-three) nuclear power plants (both PWR and BWR type) are now under operating in Japan, and the oldest plant has been operating more than thirty years. These plants will be operated until sixty years for operation periods, and will be verified the integrity for assessment of nuclear plants for every ten years in Japan. Reactor Pressure Vessels (RPVs) are required to evaluate the reduction of fracture toughness and the increase of the reference temperature in the transition region. As the operating period will be longer, the prediction for these material properties will be more important. Recently the domestic prediction formula of embrittlement was revised based on the database of domestic plant surveillance test results for thirty years olds as the JEAC4201-2007 [7]. The adequacy for this prediction formula using for sixty year periods is verified by use of the results of the material test reactors (MTRs), but the effects of the accelerated irradiation on embrittlement has not been clear now. So, JNES started the national project, called as “PRE” project on 2005 in order to investigate how flux influences on the ΔRTNDT. In this project the RPV materials irradiated in the actual PWR plant have been re-irradiated in the OECD/Halden test reactor by several different fluxes up to the high fluence region, and the microstructual change for these materials will be investigated in order to make clear the cause of the irradiation embrittlement. In this paper the overall scheme of this project and the summary of the updated results will be presented.

Effects of Neutron, Ion and Proton Irradiation on Nano-Indentation Hardness of RPV Steels

2020 ◽  
Vol 999 ◽  
pp. 39-46
Author(s):  
Cheng Liang Li ◽  
Guo Gang Shu ◽  
Jing Li Yan ◽  
Wei Liu ◽  
Yuan Gang Duan
Keyword(s):  
Proton Irradiation ◽  
Indentation Hardness ◽  
Irradiation Embrittlement ◽  
Irradiation Hardening ◽  
Nano Indentation ◽  
Rpv Steel ◽  
Hardness Tests ◽  
Increase Rate ◽  
Before And After ◽  
Reactor Pressure

The irradiation embrittlement damage of reactor pressure vessel (RPV) steel is one of its primary failure mechanisms. In this work, neutron, ion and proton irradiation experiments were carried on the same commercial RPV steels with the same irradiation fluence under the same temperature of 292°C. Then the nano-indentation hardness tests were performed on the RPV steel before and after irradiation. The results show that the irradiation hardening effects are observed by means of nano-indentation technique under the above three irradiations, and the hardening features are basically the same. While the max variation and increase rate are obviously different between those irradiations. It is found that the main reason of the above differences are caused by different energies of irradiation energetic particles, resulting in different types and quantities of defects. The conclusions in this paper are helpful to select and compare different irradiation experiments to the research of RPV steels irradiation embrittlement damage.

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.

Tool-Based Assessment of Reactor Trip Systems Availability and Safety Using Markov Modeling

2020 ◽  
pp. 175-203
Author(s):  
Oleg Odarushchenko ◽  
Valentyna Butenko ◽  
Elena Odarushchenko ◽  
Evgene Ruchkov
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Gap Analysis ◽  
Markov Modeling ◽  
Careful Selection ◽  
Regulatory Requirements ◽  
Safety Standards ◽  
Tools And Techniques ◽  
And Control

The accurate availability and safety assessment of a reactor trip system for nuclear power plants instrumentation and control systems (NPP I&C) application is an important task in the development and certification process. It can be conducted through probabilistic model-based evaluation with variety of tools and techniques (T&T). As each T&T is bounded by its application area, the careful selection of the appropriate one is highly important. This chapter presents the gap-analysis of well-known modeling approach—Markov modeling (MM), mainly for T&T selection and application procedures—and how one of the leading safety standards, IEC 61508, tracks those gaps. The authors discuss how main assessment risks can be eliminated or minimized using metric-based approach and present the safety assessment of typical NPP I&C system. The results analysis determines the feasibility of introducing new regulatory requirements for selection and application of T&T, which are used for MM-based assessment of availability and safety.

Industry Practice for the Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels in Japan

2009 ◽  
Author(s):  
Norimichi Yamashita ◽  
Masanobu Iwasaki ◽  
Koji Dozaki ◽  
Naoki Soneda
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Correlation Method ◽  
Surveillance Data ◽  
Surveillance Program ◽  
Irradiation Embrittlement ◽  
Term Operation ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material ageing issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation based on the appropriate recognition of the neutron irradiation embrittlement is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complementally use of surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old BWR RPV steels with high impurity content. Some discrepancies of PWR RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201–2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December, 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long-term surveillance program”, which is to be applied for the operation beyond 40 years. The use of the reloading, re-irradiation and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201–2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan.

Probabilistic Fracture Mechanics Analysis for Degraded Reactor Pressure Vessel in Pressurized Water Reactor Nuclear Power Plant

2014 ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsoung-Wei Chou
Keyword(s):  
Nuclear Power ◽  
Structural Reliability ◽  
The United States ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of PWR reactor pressure vessels over their long term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the United States 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. Further, two overcooling transients of steam generator tube rupture and pressurized thermal shock 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 reactor pressure vessels and can be also referred as its regulatory basis in Taiwan.

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.

Sign in / Sign up

Export Citation Format

Share Document