Microstructural Characterization of RPV Materials Irradiated to High Fluences at High Flux

Soneda, Naoki; Dohi, Kenji; Nishida, Kenji; Nomoto, Akihiro; Tomimatsu, Minoru; Matsuzawa, Hiroshi

doi:10.1520/stp49006s

Cited by 3 publications

(9 citation statements)

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“…The root of the volume fraction determined by 3DAP analysis is proportional to the transition temperature shift and the increase in hardness or yield strength. This relationship is widely found in surveillance data [59,69,71], MTR irradiation data [77,79] and also various changed particle irradiation data [37], and is irrespective of the Cu content and irradiation conditions. Figure 8 shows an example of the linear relationship between the cluster volume fraction and transition temperature shift [59].…”

Section: Characteristics Of Solute Clusterssupporting

confidence: 53%

“…However, recent microstructural analyses have provided new knowledge mainly from TEM observations and PA measurements. TEM observations with improved specimen preparation techniques using focused ion beam techniques confirmed that dislocation loops are commonly formed in irradiated commercial steels [58,59,77,[98][99][100][101]. Figure 10 shows example TEM images indicating interstitial dislocation loops in A533B steels irradiated with heavy ions [98].…”

Section: Matrix Damagementioning

confidence: 87%

“…This effect is more significant for higher Cu materials. Some SANS data suggest that P and Mn enhance 'precipitation' in high-Cu steels [62,82], while more recent 3DAP data show no such enhancement of cluster formation [77].…”

Section: Characteristics Of Solute Clustersmentioning

confidence: 98%

“…Interstitial dislocation loops are thought to be one of the microstructural features of SMD. It was pointed out in several papers that the contribution of visible dislocation loops to hardening estimated using an Orowan-type hardening model was smaller than that of solute clusters [77,100,104]. However, since the estimated contribution of dislocation loops to hardening or transition temperature shift exceeded 10% of the total value, dislocation loops might have non-negligible contribution to radiation embrittlement.…”

Section: Matrix Damagementioning

confidence: 99%

“…The cluster size and density increase with increasing dose in most cases. Figure 7 shows the 3DAP data of the density and size of solute clusters formed in A533B steels with various Cu and Ni contents irradiated in a MTR [77]. The Cu content has weak influence on the size but strong influence on the density.…”

Section: Characteristics Of Solute Clustersmentioning

confidence: 99%

See 4 more Smart Citations

Current understanding of radiation-induced degradation in light water reactor structural materials

Fukuya¹

2013

Journal of Nuclear Science and Technology

161

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Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.

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Section: Characteristics Of Solute Clusterssupporting

confidence: 53%

Section: Matrix Damagementioning

confidence: 87%

Section: Characteristics Of Solute Clustersmentioning

confidence: 98%

Section: Matrix Damagementioning

confidence: 99%

Section: Characteristics Of Solute Clustersmentioning

confidence: 99%

See 3 more Smart Citations

Current understanding of radiation-induced degradation in light water reactor structural materials

Fukuya¹

2013

Journal of Nuclear Science and Technology

161

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Statistical arguments towards the development of an advanced embrittlement correlation method for reactor pressure vessel materials

Nakasuji

Morishita

2019

Journal of Nuclear Science and Technology

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An embrittlement correlation method is one of the most important techniques used to ensure the integrity of pressure vessel steels in nuclear power plants. In Japan, the embrittlement correlation method is being addressed in accordance with the Japan Electric Association Code (JEAC 4201), which was developed using actual measured data on the irradiation embrittlement of pressure vessel steels. In the present study, to develop more reliable methodologies, statistical arguments were made concerning the embrittlement data. With regard to a set of residual data defined in JEAC 4201 as a collection of differences between the measured and calculated ΔDBTT (ductileto-brittle transition temperature shift) values, a statistical relationship between the population and samples was found, and then, the sampling errors in the mean values of the residuals were identified as key for establishing a more reliable correlation method. Using this relationship, it was noted that when predicting the amount of irradiation embrittlement of pressure vessel steels in a particular plant using the JEAC 4201 correlation method, the deviation associated with sampling errors needed to be corrected. Based on this finding, a more appropriate interpretation was found for the so-called MC correction in JEAC 4201, and moreover, a new correction method was developed within the framework of Bayesian estimation. This new correction method will be useful for establishing further advanced methodologies to manage irradiation embrittlement using the probabilistic risk assessment approach.

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Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

et al. 2021

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The cataloguing and revision of reactor pressure vessels (RPV) manufacturing and in-service inspection codes and their standardized material specifications—as a technical heritage—are essential for understanding the historical evolution of criteria and for enabling the comparison of the various national regulations, integrating the most relevant results from the scientific research. The analysis of the development of documents including standardized requirements and the comparison of regulations is crucial to be able to implement learned lessons and comprehend the progress of increasingly stringent safety criteria, contributing to sustainable nuclear power generation in the future. A novel methodology is presented in this work where a thorough review of the regulations and technical codes for the manufacture and in-service inspection of RPVs, considering the implementation of scientific advances, is performed. In addition, an analysis focused on the differences between irradiation embrittlement prediction models and acceptance criteria for detected defects (both during manufacturing and in-service inspection) described by the different technical codes as required by different national regulations such as American, German, French or Russian is performed. The most stringent materials requirements for RPV manufacturing are provided by the American and German codes. The French code is the most stringent with respect to the reference defect size using as a criterion in the in-service inspection.

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Microstructural Characterization of RPV Materials Irradiated to High Fluences at High Flux

Cited by 3 publications

References 0 publications

Current understanding of radiation-induced degradation in light water reactor structural materials

Current understanding of radiation-induced degradation in light water reactor structural materials

Statistical arguments towards the development of an advanced embrittlement correlation method for reactor pressure vessel materials

Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

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