Apparent Embrittlement Saturation and Radiation Mechanisms of Reactor Pressure Vessel Steels

Pachur, D.

doi:10.1520/stp28202s

Cited by 13 publications

(8 citation statements)

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“…Therefore, it would be recommended that the brittle fracture should be specially monitored under 150 °C. This agrees with the conclusions of Pachur [67] that showed as an irradiation temperature of 150 °C produces the greatest fragility of the material, being this less for a higher temperature. This is due to the fact that an increase in the irradiation temperature favors the repair of defects produced by neutron bombardment through a process of annealing the material [67].…”

Section: T (°C) Kic Meansupporting

confidence: 92%

“…According to the methodology outputs, it would be recommended that the brittle fracture should be specially monitored under 150 • C. This agrees with the conclusions of Pachur [67] that showed an irradiation temperature of 150 • C produces the greatest brittleness of the material, this being less for a higher temperature since irradiation temperature favors the repair of defects by annealing.…”

supporting

confidence: 75%

“…ASME FFS-1 [52] provides, therefore, a threshold of K IC equal to 220 MPa • √ m, valid for ferritic steels, aligned with the requirement for the ASME B&PV code [38] reference curve. On the other hand, from 150 • C, the irradiation embrittlement is less severe due to annealing effects [67], as demonstrated by experimental works [55,68,69]. Figure 7 shows the predicted K IC for every studied material.…”

Section: Step 4-validation Of Limit Conditions and Results Based On T...mentioning

confidence: 79%

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Multicriteria Analytical Model for Mechanical Integrity Prognostics of Reactor Pressure Vessels Manufactured from Forged and Rolled Steels

Rodríguez-Prieto

Callejas²,

Primera

et al. 2022

Mathematics

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The aim of this work is to present a new analytical model to evaluate jointly the mechanical integrity and the fitness-for-service of nuclear reactor pressure-vessels steels. This new methodology integrates a robust and regulated irradiation embrittlement prediction model such as the ASTM E-900 with the ASME Fitness-for-Service code used widely in other demanding industries, such as oil and gas, to evaluate, among others, the risk of experiencing degradation mechanisms such as the brittle fracture (generated, in this case, due to the irradiation embrittlement). This multicriteria analytical model, which is based on a new formulation of the brittle fracture criterion, allows an adequate prediction of the irradiation effect on the fracture toughness of reactor pressure-vessel steels, letting us jointly evaluate the mechanical integrity and the fitness-for-service of the vessel by using standardized limit conditions. This allows making decisions during the design, manufacturing and in-service of reactor pressure vessels. The results obtained by the application of the methodology are coherent with several historical experimental works.

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Section: T (°C) Kic Meansupporting

confidence: 92%

supporting

confidence: 75%

Section: Step 4-validation Of Limit Conditions and Results Based On T...mentioning

confidence: 79%

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Multicriteria Analytical Model for Mechanical Integrity Prognostics of Reactor Pressure Vessels Manufactured from Forged and Rolled Steels

Rodríguez-Prieto

Callejas²,

Primera

et al. 2022

Mathematics

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“…Since in a study conducted by Mager 20 no dependence was observed between the neutron flux value and the embrittlement of the material for neutron flux rates 21 ranging from ¼ 2.5 Â 10 18 n/cm 2 to ¼ 8.8 Â 10 19 n/cm 2 . Pachur 22 demonstrated that embrittlement is maximized at 150 C for the typical neutron flux range of nuclear reactors (from ¼ 10 18 n/cm 2 to ¼ 10 20 n/cm 2 ). 14 The irradiation embrittlement of RPVs materials is quantified in terms of the change in ductile-to-brittle transition temperature (ÁRT NDT ), usually represented as ÁT 41J , which is the temperature at which the energy absorbed in a Charpy impact is 41 J.…”

Section: Stage a Prediction Models Analysis And Sl Methodology Definmentioning

confidence: 99%

Selection of candidate materials for reactor pressure vessels: Application of irradiation embrittlement prediction models and a stringency level methodology

Rodríguez-Prieto¹,

Camacho²,

Sebastián³

2017

Proceedings of the Institution of Mechanical Engineers, Part L:

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The selection of materials for the reactor pressure vessel manufacturing is a complex process that involves great responsibility because small differences in chemical composition can adversely affect the manufacturing process and the in-service behavior of the material. Thus, it is recommendable to perform previous materials pre-selection stages based on the state-of-the-art knowledge, integrating research results with standardized requirements and using simplifier materials selection methodologies like the stringency level method. To address this issue, an evaluation of the influence of chemical composition on the shift of the ductile-to-brittle transition temperature has been performed using the most used and consolidated prediction models that are R.G. 1.99 Rev.2, NUREG/CR-6551, and ASTM E 900-02. A proposal of maximum limits for copper, nickel, and phosphorous to avoid irradiation embrittlement has been presented to carry out the process. The results have been analyzed by using the stringency level methodology to support the decision process. To this end, a materials data collection has been carried out to analyze the requirements described by 20 different specifications of materials from first to fourth generation of light water reactors, covering the main designs of pressurized reactors from Western Europe, North America, Japan, and Russia. It can be concluded that more recently developed materials exhibit more stringent requirements than earlier developed materials.

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“…Mager [38] did not find apparent dependence between the neutron flux value and the embrittlement of the material, for neutron flux rates from φ = 2.5 × 10 18 n/cm 2 to φ = 8.8 × 10 19 n/cm 2 . Porter [36] and Kangilaski [37] determined that the irradiation temperature for which the embrittlement of the material is maximum takes place at temperatures below 230 • C. Pachur [39] showed that an irradiation temperature of 150 • C produces the greatest fragility of the material. This is because an increase in the irradiation temperature favors the repair of defects produced by neutron bombardment, through a process of annealing the material [14].…”

Section: Review Of the State Of The Art Of The Rpvs Steel Embrittlement For Suitable Materials Selection History Of Irradiation Embrittlementioning

confidence: 99%

Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

Rodríguez-Prieto

Frigione

Kickhofel³

et al. 2021

Sustainability

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The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

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Apparent Embrittlement Saturation and Radiation Mechanisms of Reactor Pressure Vessel Steels

Cited by 13 publications

References 0 publications

Multicriteria Analytical Model for Mechanical Integrity Prognostics of Reactor Pressure Vessels Manufactured from Forged and Rolled Steels

Multicriteria Analytical Model for Mechanical Integrity Prognostics of Reactor Pressure Vessels Manufactured from Forged and Rolled Steels

Selection of candidate materials for reactor pressure vessels: Application of irradiation embrittlement prediction models and a stringency level methodology

Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

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