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The brittle–ductile transition temperature (BDTT) and fracture toughness are useful indicators of the brittleness of reactor pressure vessel materials. However, current methods to analyze these parameters are expensive and time-consuming; moreover, the relevant mechanism is not very clear. Thus, in this study, a body-centered cubic crystal plasticity finite element model (CPFEM) based on the densities of dislocation and irradiation defects, which are full- and partial-absorption dislocation loops, was numerically implemented on A508-3 steel. Irradiation embrittlement was investigated in the CPFEM framework. The cleavage failure model for a representative volume element was proposed to calculate the BDTT, the results of which agreed well with experimental results in the literature for both unirradiated and irradiated samples. The risk for brittle failure is higher for irradiated samples at constant temperature. Furthermore, the proposed model combined with the Beremin cleavage fracture local approach, could estimate the fracture toughness of A508-3 steel at different temperatures before and after irradiation. The results obtained were consistent with reported experimental data, indicating that irradiation exacerbates the embrittlement. This work is beneficial to the calculation and evaluation of irradiation effect of RPV steel, providing reference for the research of reactor life extension. [ABSTRACT FROM AUTHOR]

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy's unique balance of high fracture toughness and high tungsten content; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. To alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni+ and He+ ion irradiations to mimic the expected displacement damage and He gas production expected after five years of service as materials for plasma facing components. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and a hexagonal W 2 C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation enhanced and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material's overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases. [Display omitted] [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Reduced-pressure electron beam (EB) plate butt welds were manufactured in two low-alloy pressure-vessel steels, SA508 Gr 3 Cl 1 and SA508 Gr 2, at two thicknesses in both steels, 30 mm and 130 mm. Transient temperatures during welding were recorded using thermocouple arrays. Residual stresses in the as-welded condition and after post-weld heat treatment were measured using diverse methods: neutron diffraction and the contour method at 30 mm thickness; and deep hole drilling and the contour method at 130 mm. Incremental centre hole drilling measurements were performed at 130 mm thickness to better understand near-surface stresses. Weld and heat-affected zone microstructures and microconstituents were evaluated using a combination of hardness mapping, optical microscopy and electron microscopy. The as-welded residual stresses exhibit the characteristic M-shaped distribution for hardenable steels, reaching 500–600 MPa in tension in both steels at both thicknesses. However, the modest changes to the chemical composition and the change in plate thickness both significantly influenced microstructures, mechanical properties and residual stress distributions. These sensitivities underline the need for physically faithful models. This extensive characterisation study enables the development and validation of models that predict the development of microstructures and residual stresses in EB welds in low alloy pressure vessel steels.

• The changes in fracture behavior of stainless steel with in neutron dose can be predicted from the presented method. • The multiaxial fracture strain model is determined by simulating tensile and fracture toughness test. • The neutron irradiation embrittlement constant is determined by simulating tensile test. • The neutron irradiation embrittlement constant is multiplied with the multiaxial fracture strain model to simulate fracture toughness tests, enabling the prediction of degradation in fracture toughness due to neutron irradiation. This paper proposes a numerical method for predicting the reduction in the fracture toughness with neutron dose using the concept of an embrittlement constant. Tensile and fracture toughness test data for type 304 stainless steel in the literature [ 41 , 42 ] are used to validate the proposed method. Fracture toughness tests are simulated using a stress-modified fracture strain damage model implemented in an ABAQUS user-subroutine. The true stress-strain curve and one point of the fracture strain locus are obtained from the tensile test simulation. The remaining parameters of the fracture strain locus are determined by simulating a fracture toughness test. The fracture strain locus is modified to reflect the irradiation effect on the fracture toughness using the proposed irradiation embrittlement constant C I. The function of the constant with the neutron dose is calculated from the tensile tests and simulations. Subsequently, the modified damage model is used to predict the reduction in the fracture toughness. The predicted simulation results are compared with experimental data, showing good agreement. [Display omitted] [ABSTRACT FROM AUTHOR]

The reactor pressure vessel is made of ferritic low alloy steel, which has fracture behaviors that include ductile to brittle transition temperature and decrease of fracture toughness in the beltline region of the reactor pressure vessel steel due to irradiation embrittlement. Decrease of fracture toughness due to irradiation during plant operation needs to be taken into account in the assessment of the reactor pressure vessel against pressurized thermal shock events. A large number of fracture toughness data were accumulated in domestic PWR plant surveillance program tests and knowledge on the fracture toughness evaluations, such as scattering, were obtained through research works. In this study, fracture toughness curve for the assessment of domestic reactor pressure vessels were developed taking into consideration the latest knowledge on fracture toughness evaluations, including investigations on fracture toughness data scattering and fracture toughness curve reliability.

To assess the combined effect of thermal aging and neutron irradiation on the cracking behavior of CF-8 cast austenitic stainless steel, crack growth rate (CGR) and fracture toughness J-R curve tests were carried out on compact-tension specimens in high-purity water with low dissolved oxygen. Both unaged and thermally aged specimens were irradiated at ∼320 °C to 0.08 dpa. Thermal aging at 400 °C for 10,000 h apparently had no effect on the corrosion fatigue and stress corrosion cracking behavior in the test environment. The cracking susceptibility of CF-8 was not elevated significantly by neutron irradiation at 0.08 dpa. Transgranular cleavage-like cracking was the main fracture mode during the CGR tests, and a brittle morphology of delta ferrite was often seen on the fracture surfaces at the end of CGR tests. The fracture toughness J-R curve tests showed that both thermal aging and neutron irradiation can induce significant embrittlement. The loss of fracture toughness due to neutron irradiation was more pronounced in the unaged than aged specimens. After neutron irradiation, the fracture toughness values of the unaged and aged specimens were reduced to a similar level. G-phase precipitates were observed in the aged and irradiated specimens with or without prior aging. The similar microstructural changes resulting from thermal aging and irradiation suggest a common microstructural mechanism of inducing embrittlement in CF-8. [ABSTRACT FROM AUTHOR]

[Display omitted] • Modest variations in SA508 steel chemistry have a marked impact on EB weld residual stresses. • Significant stresses remain after an ASME-compliant post-weld heat treatment. • Single-pass EB welds are likely to have thickness-dependent residual stress distributions. • Phase transformation kinetics must be incorporated into models for stress development. Reduced-pressure electron beam (EB) plate butt welds were manufactured in two low-alloy pressure-vessel steels, SA508 Gr 3 Cl 1 and SA508 Gr 2, at two thicknesses in both steels, 30 mm and 130 mm. Transient temperatures during welding were recorded using thermocouple arrays. Residual stresses in the as-welded condition and after post-weld heat treatment were measured using diverse methods: neutron diffraction and the contour method at 30 mm thickness; and deep hole drilling and the contour method at 130 mm. Incremental centre hole drilling measurements were performed at 130 mm thickness to better understand near-surface stresses. Weld and heat-affected zone microstructures and microconstituents were evaluated using a combination of hardness mapping, optical microscopy and electron microscopy. The as-welded residual stresses exhibit the characteristic M-shaped distribution for hardenable steels, reaching 500–600 MPa in tension in both steels at both thicknesses. However, the modest changes to the chemical composition and the change in plate thickness both significantly influenced microstructures, mechanical properties and residual stress distributions. These sensitivities underline the need for physically faithful models. This extensive characterisation study enables the development and validation of models that predict the development of microstructures and residual stresses in EB welds in low alloy pressure vessel steels. [ABSTRACT FROM AUTHOR]

F82H is the Japanese reference reduced-activation ferritic-martensitic (RAFM) steel for fusion blanket applications. The harsh environment of a fusion reactor, such as neutron irradiation and He/H damage, can result in significant degradation of F82H fracture toughness. Therefore, understanding the fracture toughness behavior of F82H in the fusion environment is critical to ensure the long-term safe operation of the fusion reactor. In this paper, we summarize seven irradiation campaigns of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) covering five variants of F82H steels, including F82H IEA, F82H Mod3, F82H doped with 1.4% natural Ni, F82H doped with 1.4% 58Ni, and F82H doped with 1.4% 60Ni. The irradiation temperatures covered the range from 220 °C to 530 °C and the neutron irradiation dose spanned 4 dpa to 70 dpa. The effects of neutron irradiation temperature, dose, materials composition, Ni doping, and He production on F82H fracture toughness are discussed. Our results showed that irradiation embrittlement monotonically decreased with increasing irradiation temperature until 400 °C for F82H IEA and F82H Mod3. F82H Mod3 showed better fracture toughness than F82H IEA both before and after neutron irradiation. We determined that 1.4% Ni alloying can be applied to F82H for simulating He effect in a fission reactor without jeopardizing the fracture toughness of the material. However, more studies are needed to understand the effect of high dose (>20 dpa) and He production on F82H fracture toughness. [ABSTRACT FROM AUTHOR]