Evaluation of Irradiation Embrittlement of the Chinese RPV Steels

2019 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Binxi Wang ◽  
Yifeng Huang ◽  
Hui Li ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Master Curve ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Limiting Factor ◽  
High Fluence ◽  
Irradiation Embrittlement ◽  
Test Reactor ◽  
Reactor Pressure ◽  
Irradiation Resistance

Abstract Reactor pressure vessel (RPV) is considered to be irreplaceable, which is the most limiting factor for the lifetime of a nuclear power plant. This paper aims to introduce our project for the evaluation of the irradiation embrittlement for the Chinese RPV forging. The forging manufactured in China was irradiated in the high fluence engineering test reactor. Tensile tests, Charpy impact tests and fracture toughness tests in terms of master curve T0 were carried out for the material subjected to different irradiation fluences. Comparison of the mechanical properties of the irradiated materials and the materials without irradiation is made. The irradiation resistance of the materials in our project is also compared with the data for the irradiated RPV steels in the literatures.

Evaluation of Irradiation Embrittlement of the Weld Metals in the Chinese RPV

2020 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Yifeng Huang ◽  
Binxi Wang ◽  
Yan Yu ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Degradation Mechanism ◽  
Base Material ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Irradiation Damage ◽  
Limiting Factor ◽  
Irradiation Embrittlement ◽  
Weld Metals

Abstract Reactor pressure vessel (RPV) operates under high temperatures and pressures and is exposed to relatively high neutron radiation. RPV is considered to be irreplaceable, which is the most limiting factor for the lifetime of a nuclear power plant. As the most severe ageing degradation mechanism in RPV materials, irradiation embrittlement is a major issue affecting the integrity through the service life of a RPV. Our previous paper (ASME PVP2019-93615[1]) introduced our project for assessment of irradiation embrittlement of the materials for the Chinese RPVs to verify the 60-year design life, in which the specimens made of the RPV base material manufactured in China, the SA-508 Gr.3 Cl.1 forging, and the different types of weld metals were irradiated in the high fluence engineering test reactor (HFETR). The paper analyzed extent irradiation damage of the forging in terms of mechanical properties. As another part of the project, this paper concentrates on the evaluation of the weld metals in the same project. Tensile tests, Charpy impact tests and fracture toughness tests by master curve approach were carried out for the three types of weld metals subjected to different irradiation fluences (2.6E19n/cm2, 8.9E19n/cm2). Comparison of the mechanical properties of the irradiated and the unirradiated materials is made. The irradiation resistance of the weld metals in our project is also compared with the data in the literatures.

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.

Effect of Neutron Irradiation on the Mechanical Properties of an A508 CL2 and 15Kh2NMFA Irradiated in the NOMAD_3 Rig in the BR2 Cooling Water

2021 ◽  
Author(s):  
Inge Uytdenhouwen ◽  
Rachid Chaouadi
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Cooling Water ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Significant Loss ◽  
Irradiation Temperature ◽  
Irradiation Damage ◽  
Transition Curves ◽  
Reactor Pressure

Abstract The typical operating temperatures of a nuclear reactor pressure vessel in a PWR are between 290°C and 300°C. However, many BWRs and some PWRs operate at slightly lower temperatures down to 260°C. Most of the literature and neutron irradiation damage is therefore focused on those irradiation temperatures. It is well-known that the lower the irradiation temperature, the more neutron irradiation damage occurs, because no appreciable annealing happens below approximately 230°C. The NOMAD_3 irradiation consisted in total of 24 Charpy sized samples from an A508 Cl.2 forging and a 15Kh2NMFA material. They were irradiated to three various fluences between 1.55 and 7.90 × 1019 n/cm2 (E > 1MeV) at approximately 100°C. The hardening of the A508 Cl.2 was between 260 and 400 MPa which was much higher than the NOMAD_0 properties which were irradiated at approximately 280°C. The tensile tests of irradiated materials are all characterized by a significant loss of work hardening capacity leading to plastic flow localization promptly after the yield strength is reached. This affects also the shape of the Charpy impact transition curves. The radiation embrittlement derived from Charpy impact tests, ΔT41J, is up to 156°C for the highest fluence. For this irradiation, the embrittlement to hardening ratio was also around 0.43 +/−0.2°C/MPa as it was found in the previous campaign NOMAD_0. This paper discusses the tensile, hardness and impact properties of the NOMAD_3 irradiation campaign. It is compared to the NOMAD_0 with respect to effect of irradiation temperature and annealing recovery.

Bayesian Statistical Analysis on Chemical Composition Contributing to Irradiation Embrittlement at High Fluence Region

2016 ◽  
Author(s):  
Hisashi Takamizawa ◽  
Yutaka Nishiyama
Keyword(s):  
Statistical Analysis ◽  
Chemical Composition ◽  
Probability Distribution ◽  
Input Data ◽  
High Fluence ◽  
Irradiation Embrittlement ◽  
Ni Content ◽  
Individual Probability ◽  
Si Content ◽  
Reactor Pressure

It has been accepted that neutron irradiation embrittlement of reactor pressure vessel is caused by irradiation-induced formation of solute clusters (SCs) and matrix damages (MDs). In the present study, to analyze the contribution of chemical composition contained in SCs to irradiation embrittlement at high fluence region, statistical analysis using the Bayesian nonparametric (BNP) method was performed for Japanese PWR surveillance data. The significance of P, Si and Mn contents, which are not necessarily included in embrittlement correlations unlike the Cu and Ni content, was evaluated. The BNP method can learn the complexity of the statistical model itself from the input data and infer the predicted data with individual probability distribution of predict condition. The result suggested that irradiation embrittlement was most affected by the Si content in three examined elements at high fluence region.

Design and Fabrication of Reactor Pressure Vessel for High Temperature Engineering Test Reactor (HTTR)

2004 ◽  
Author(s):  
Yukio Tachibana ◽  
Shigeaki Nakagawa ◽  
Tatsuo Iyoku
Keyword(s):  
Pressure Vessel ◽  
Temper Embrittlement ◽  
Reactor Pressure Vessel ◽  
The Body ◽  
Inlet Temperature ◽  
Irradiation Embrittlement ◽  
Fabrication Procedure ◽  
Test Reactor ◽  
Service Inspection ◽  
Reactor Pressure

The reactor pressure vessel (RPV) of the HTTR is 5.5 m in inside diameter, 13.2 m in inside height, and 122 mm and 160 mm in wall thickness of the body and the top head dome, respectively. Because the reactor inlet temperature of the HTTR is higher than that of LWRs, 2 1/4Cr-1Mo steel is chosen for the RPV material. Fluence of the RPV is estimated to be less than 1×1017 n/cm2 (E>1 MeV), and so irradiation embrittlement is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the RPV using embrittlement parameters, J-factor and X. In this paper design, fabrication procedure, and in-service inspection technique of the RPV for the HTTR are described.

Master Curve Approach for Some Japanese Reactor Pressure Vessel Steels

2008 ◽  
Author(s):  
Minoru Tomimatsu ◽  
Takashi Hirano ◽  
Seiji Asada ◽  
Ryoichi Saeki ◽  
Naoki Miura ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steels ◽  
The World ◽  
Reactor Pressure ◽  
Plant Components

The Master Curve Approach for assessing fracture toughness of reactor pressure vessel (RPV) steels has been accepted throughout the world. The Master Curve Approach using fracture toughness data obtained from RPV steels in Japan has been investigated in order to incorporate this approach into the Japanese Electric Association (JEA) Code 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components”. This paper presents the applicability of the Master Curve Approach for Japanese RPV steels.

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.

Crack Initiation and Arrest of Irradiated RPV-Steels: A National Project on German RPV Weld and Base Materials

2004 ◽  
Author(s):  
Elisabeth Keim ◽  
Reinhard Langer ◽  
Hilmar Schnabel ◽  
Hieronymus Hein
Keyword(s):  
Crack Initiation ◽  
Safety Assessment ◽  
Master Curve ◽  
Charpy Impact ◽  
Pressurized Water Reactors ◽  
Charpy Impact Energy ◽  
Pressurized Water ◽  
Base Materials ◽  
Water Reactors ◽  
Reactor Pressure

In Germany the procedure which has to be applied for the safety assessment of the reactor pressure vessel is based on the RTNDT concept. The Master Curve concept (based on T0) has the advantage compared to the RTNDT concept that the basic tests are fracture toughness tests instead of Charpy impact energy or Pellini tests. By means of the recently initiated German project CARISMA (Crack Initiation and Arrest of Irradiated Steel Materials), a data base will be created on pre-irradiated original materials of the four generations of German nuclear pressurized water reactors, which allows to examine the consequences if the Master Curve instead of the RTNDT concept will be applied.

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.

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