Irradiation temperature, flux and spectrum effects

Ballesteros, A.; Ahlstrand, R.; Christiane, Bruynooghe; Chernobaeva, A. A.; Kevorkyan, Yu. R.; Erak, D.; Zurko, D.

doi:10.1016/j.pnucene.2011.05.022

Cited by 32 publications

(9 citation statements)

References 4 publications

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“…When the Cu wt.% content is greater than 0.1%, the increase in the ΔRTDBT due to embrittlement by irradiation is greater the higher the Cu These studies on the influence of the chemical composition, neutron flux and irradiation temperature, under operating conditions of nuclear reactors are considered still valid today since they have been confirmed by analyzing the materials from reactors that have been in operation for decades [40][41][42][43]. Therefore, nowadays, it is a proven fact that the effect of neutron flux on the brittle-ductile behavior of the material is a complex phenomenon dependent on the composition of the alloy and on temperature [44].…”

Section: Main Technological Characteristics Influencing Irradiation Embrittlement Characteristicsmentioning

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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Section: Main Technological Characteristics Influencing Irradiation Embrittlement Characteristicsmentioning

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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“…It is well accepted that high flux has to be used carefully to predict the in-service behavior of RPV steels with high Cu content (e.g., [156,157,158,159]). Conversely, several studies showed that embrittlement of steels with low Cu is weakly sensitive to flux (e.g., [153,154,160,161,162]…”

Section: Flux Effectmentioning

confidence: 99%

60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

Duysen

Bellefon

2017

Journal of Nuclear Materials

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The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21 st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

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“…[1][2][3][4] Irradiation conditions such as dose, dose rate, temperatures, and material components influence irradiation microstructure. [5][6][7] RPV steels contain a certain amount of impurities, such as C, N, and O. Generally, carbon atoms are deemed the most critical impurities in iron to affect defect evolution.…”

Section: Introductionmentioning

confidence: 99%

“…[ 1–4 ] Irradiation conditions such as dose, dose rate, temperatures, and material components influence irradiation microstructure. [ 5–7 ]…”

Section: Introductionmentioning

confidence: 99%

An Object Kinetic Monte Carlo Simulation for Defect Evolution of Neutron‐Irradiated Reactor Pressure Vessel Steels: Carbon Sensitive Study

Zhang

Yang

et al. 2021

Physica Status Solidi (b)

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Herein, the defect evolution in the FeCMnNi model alloy is investigated by the gray‐alloy method in an object kinetic Monte Carlo (OKMC) model. Based on the parameters of the Fe–C system, the influence of Mn and Ni is implicitly considered, and the migration energy of mobile defects is modified. The number density of interstitial clusters and vacancy clusters in simulation has a good agreement with experimental results. The defect evolution features versus carbon and solutes concentration are also studied. The simulation results show that the number density of defect clusters increases with increasing carbon or solutes concentration. The average size of interstitial clusters decreases, while the average size of vacancy clusters is hardly affected by increasing impurity/solutes concentration. The defect features do not show a saturation tendency with increasing carbon/solutes concentrations. It also seems the influence of solutes concentration is more sensitive on defect features than carbon concentration in this FeCMnNi alloy.

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Irradiation temperature, flux and spectrum effects

Cited by 32 publications

References 4 publications

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

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

60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

An Object Kinetic Monte Carlo Simulation for Defect Evolution of Neutron‐Irradiated Reactor Pressure Vessel Steels: Carbon Sensitive Study

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