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The overarching goal of this research is to advance understanding of, model and predict embrittlement primarily due to hardening by precipitates in Reactor Pressure Vessel (RPV) steels as it pertains to 80-year life extension of fission reactors. The RPV is responsible for primary radioactive containment and pressurizing water in the nuclear reactor. The most immediate materials safety problem facing the RPV steels is neutron irradiation embrittlement, which is the reduction of fracture toughness. Embrittlement is caused by irradiation hardening (Δσy), mainly as a result of precipitation, and manifested as upward ductile-to-brittle transition temperature shifts (ΔT). The excess concentration of defects created under irradiation leads to radiation enhanced diffusion (RED), which greatly accelerates precipitation normally limited by extremely slow kinetics at ≈ 300ᴼC, or at typical reactor operating temperatures. Concomitantly, the radiation induced segregation (RIS) of solute elements leads to enrichment at microstructural features which likely plays a role in the heterogeneous nucleation of precipitates. At lower fluences (ϕt), or displacements per atom (dpa), the hardening features are nm-scale copper rich precipitates (CRPs) and solute defects vacancy complexes at trace impurity levels of > ≈ 0.06 wt% Cu. Effective Cu concentrations are less than ≈ 0.25 wt%. However, at higher ϕt much larger quantities of alloying elements precipitate to form Mn, Ni, Si nm-scale intermetallic precipitates known as MNSPs. Large volume fractions (fv) of MNSPs cause severe hardening and embrittlement. The variables controlling the formation and character of severely embrittling Mn-Ni-Si precipitates (MNSP) under neutron irradiation are neither fully understood nor explicitly treated in regulatory models. There are five primary questions addressed by this work. 1) What is the quantitative effect of temperature (Ti), fluence/dose (ϕt), flux/dose rate (ϕ), particle type, product form and alloy composition on precipitate size, number density, volume fraction, composition and magnetic character of CRPs and MNSPs? 2) Are very high Ni alloys, with high strength and toughness, irradiation tolerant at low Mn; and how does precipitation and hardening occur in steels with a much wider range of Mn, Ni and Si that conventional normalized and tempered bainitic steels?3) How can charged particle irradiations (CPI) compliment neutron data to gain insight into precipitation mechanisms?4) Are MNSPs enhanced or induced by irradiation (equilibrium versus non-equilibrium) and can this ongoing controversy by establishing their detailed character, long-time thermal stability, and formation mechanisms?5) How can these insights help build an improved high fluence, low flux predictive embrittlement model for steels used in in 80-year extended power reactor service?Atom probe tomography (APT), small angle neutron scattering (SANS) and transmission electron microscopy (TEM) show the presence of significant volume fractions of MNSPs in all RPV steels at the high 80-year fluence. Precipitation is reflected in Δσy, as characterized by tensile, shear and microhardness tests. Ni generally plays the strongest role in the formation of MNSPs. In the absence of sufficient Cu and CRPs, and at low to intermediate Ni, the MNSPs homogeneous nucleation rates are negligible in a defect free matrix. Heterogenous precipitation occurs at microstructural features, such as segregated network dislocation and irradiation induced interstitial loop sites. Units of % for compositions refers to atomic percent (%) unless otherwise noted. For typical RPV steel compositions (0.75% to 1.6% Ni with ≥ 0.8% Mn and 0.4 – 1.2% Si) MNSPs compositions are generally similar to G (Mn6Ni16Si7) and Γ2 (Mn2Ni3Si) phases. The formation of these phases is predicted by Calphad and has been confirmed by synchrotron X-Ray and TEM Diffraction measurements in a number of cases. However, in alloys with very low or very high Ni, coupled with variations in Mn and Si, other phase compositions are selected. For example, Ni-silicide type compositions are found in alloys with very low ≤ 0.24 Mn and high ≈ > 3% Ni. Notably at normal levels of > 0.8% Mn, very large MNSP fv form in 3.5% Ni steels at high fluence. The MNSP high Ni fv decreases approximately linearly with the decreasing alloy Mn content. Thus precipitation hardening is much lower in high ≈ 3.5% Ni and ≤ 0.24% Mn steels due to what is described as Mn starvation. For the so-called ATR-2 irradiation condition, which is the focus of this study, the high Ni fv ≈ 2.44 and 0.69% at 1.04 and 0.24% Mn, respectively. The resulting Δσy, which is well correlated with the √fv (as predicted by dispersed barrier hardening models), are 472 and 260 MPa.At high ϕ, a very high N and fv of MNSPs and CRP-MNSPs are observed in low Cu and Cu bearing alloys, respectively. In the latter case, Cu core Mn-Ni-Si shell CRP structures formed at lower dpa evolve into CRP-MNSP appendage co-precipitate features at high dpa. MNSP compositions formed in in rapid and convenient self-ion charged particle irradiation (CPI) are very similar to those found in neutron irradiations (NI). High dpa CPI produce fewer and larger precipitates than in NI. Further, higher dpa are needed to form the same precipitate fv for CPI versus NI conditions. The delayed precipitation is consistent with enhanced recombination of vacancies and SIA defects at the higher CPI dpa rates, which reduces the efficiency of RED. The MNSP grow slowly, but eventually reach large fv at very high dpa. Notably, fv correlates well with the G and Γ2 phase solute product, (Ni16Mn6Si7)(1/29) and (Ni3Mn2Si1)(1/6) and, at high dpa, is close to the equilibrium values, slightly modified by the Gibbs-Thomson effect. However, in steels with very low Mn and high Ni, Ni2-3Si silicide phase type precipitate compositions are observed; and when Ni is low, the precipitate compositions are close to the MnSi phase field. A comparison of dispersed barrier model predictions with measured hardening data suggests that the Ni-Si dominated precipitates are weaker dislocation obstacles than the G phase type MNSPs. Ultimately, fv from the CPI can be used to estimate y (and T) at lower service relevant dpa. While not quantitatively precise, this allows scoping studies of the embrittlement sensitivity of new RPV alloys. Post irradiation annealing (PIA) was used to clarify the irradiation induced versus enhanced controversy regarding dominant nanoscale Mn-Ni-Si precipitate (MNSP) formation mechanisms in pressure vessel steels. Radiation induced, non-equilibrium, MNSPs would dissolve under high temperature PIA, while radiation enhanced precipitates would be stable above a critical radius (rc). A Cu-free, high Ni steel was irradiated with 2.8MeV Fe2+ ions at two temperatures to generate MNSPs with average radii (r ̅) above and below an estimated rc for PIA at 425°Cup to 52 weeks. The complementary APT and Energy Dispersive X-ray Spectroscopy studies show MNSPs with r < rc dissolved, while those with r > rc slightly coarsened, consistent with thermodynamic predictions. Note, the MNSPs would be even more thermodynamically stable at much lower neutron irradiation RPV service temperatures around 290°C.Finally, the ultimate goal of this research is to create and analyze the high ϕt, intermediate ϕ ATR-2 database on both Δσy and microstructural changes in a large number of irradiated alloys. In the final section, ATR-2 results are integrated with a variety of other (mostly UCSB and surveillance) databases to develop a new high ϕt-low ϕ predictive embrittlement chemistry factor. Special emphasis is placed on the Δσy contributions of MNSPs, which are observed in a wide range of RPV steels at high ϕt. This directly informs the Odette Research Group’s development of an advanced embrittlement model which accurately predicts ΔT for low ϕ high ϕt conditions up to 80 full power years of operation.

The comparison of experimental values of the critical brittle temperature ΔTF and reference temperature ΔT0 of VVER-1000 reactor vessel weld metal with an elevated content of manganese and nickel is performed. ΔTF and ΔT0 values are defined proceeding from the standard impact bend Charpy and Charpy cracked fracture toughness specimen tests, respectively. Specimens were irradiated in industrial reactors in the frame of surveillance specimen program up to the fast (E ≥ 0.5 MeV) neutron fluences corresponding to the NPP long term operation period. The research results showed the shifts ΔTF and ΔT0 to agree with each other. Besides, it was discovered that in the range of over-design fluences the design embrittlement model has a tendency to underestimate the critical brittle temperature shift.

Post irradiation annealing (PIA) clarified the induced versus enhanced controversy regarding nanoscale Mn-Ni-Si precipitate (MNSP) formation in pressure vessel steels. Radiation induced MNSPs would dissolve under high temperature PIA, while radiation enhanced precipitates would be stable above a critical radius (rc). A Cu-free, high Ni steel was irradiated with 2.8MeV Fe2+ ions at two temperatures to generate MNSPs with average radii ( r ¯ ) above and below an estimated rc for PIA at 425 °C up to 52 weeks. Atom probe tomography and energy dispersive x-ray spectroscopy showed MNSPs with r rc slightly coarsened, consistent with thermodynamic predictions.

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.

Formation of a high density of Mn-Ni-Si nanoscale precipitates in irradiated reactor pressure vessel steels could lead to severe, unexpected embrittlement, which may limit the lifetimes of our nation’s light water reactors. While the existence of these precipitates was hypothesized over 20 years ago, they are currently not included in embrittlement prediction models used by the Nuclear Regulatory Commission. This work aims to investigate the mechanisms and variables that control Mn-Ni-Si precipitate (MNSP) formation as well as correlate their formation with hardening and embrittlement.A series of RPV model steels with systematic variations in Cu and Ni contents, two variables that have been shown to have a dominant effect on hardening, were irradiated in a series of test reactor and power reactor surveillance irradiations. Atom probe tomography (APT) measurements show that large volume fractions (fv) of MNSPs form in all the steels irradiated at high fluence, even those containing no added Cu, which were previously believed to have low sensitivity to embrittlement. It is demonstrated that while Cu enhances the rate of MNSP formation, it does not appear to significantly alter their saturation fv or composition. The high fluence MNSPs have compositions consistent with known intermetallic phases in the Mn-Ni-Si system and have fv very near those predicted by equilibrium thermodynamic models. In addition, X-ray diffraction experiments by collaborators shows that these precipitates also have the expected crystal structure of the predicted Mn-Ni-Si phases. Post irradiation annealing experiments are used to measure the hardness recovery at various temperatures as well as to determine if the large fv of MNSPs that form under high fluence neutron irradiation are thermodynamically stable phases or non-equilibrium solute clusters, enhanced or induced by irradiation, respectively. Notably, while post irradiation annealing of a Cu-free, high Ni steel at 425°C results in dissolution of most precipitates, a few larger MNSPs appear to remain stable and may begin to coarsen after long times. A cluster dynamics model rationalizes the dissolution and reduction in precipitate number density, since most are less than the critical radius at the annealing temperature and decomposed matrix composition. The stability of larger precipitates suggests that they are an equilibrium phase, consistent with thermodynamic models. Charged particle irradiations using Fe3+ ions are also used to investigate the precipitates which form under irradiation. Two steels irradiated to a dose of 0.2 dpa using both neutrons and ions show precipitates with very similar compositions. The ion irradiation shows a smaller fv, likely due to the much higher dose rate, which has been previously shown to delay precipitation to higher fluences. While the precipitates in the ion irradiated condition are slightly deficient in Mn and enriched in Ni and Si compared to neutron irradiated condition, the overall similarities between the two conditions suggest that ion irradiations can be a very useful tool to study the susceptibility of a given steel to irradiation embrittlement.Finally, the large fv of MNSPs that are shown to form in all steels, including those low in Cu, at high fluence, even those without added Cu, result in large amounts of hardening and embrittlement. A preliminary embrittlement prediction model, which incorporates MNSPs at high fluence, is presented, along with results from a recent test reactor irradiation to fluences representative of extended lifetimes. This model shows very good agreement with the data.

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.

The prediction of irradiation-induced transition temperature shift for RPV steels is an important method for long term operation of nuclear power plant. Based on the irradiation embrittlement data, an irradiation-induced transition temperature shift prediction model is developed with machine learning method XGBoost. Then the residual, standard deviation and predicted value vs. measured value analysis are conducted to analyze the accuracy of this model. At last, Cu content threshold and saturation values analysis, temperature dependence, Ni/Cu dependence and flux effect are given to verify the reliability. Those results show that the prediction model developed with XGBoost has high accuracy for predicting the irradiation embrittlement trend of RPV steel. The prediction results are consistent with the current understanding of RPV embrittlement mechanism.

International audience; The behavior of solute atoms in an undersaturated Fe-3.3at.%Ni model alloy upon heavy and light particle irradiations was investigated by Atom Probe Tomography. Whereas the first irradiations, performed using 27 MeV Fe ions at 400°C, highlight significant Ni enrichment on PD clusters, solute-rich features detected after 2.8 MeV protons H+ ions (400°C) and 1 MeV electron (320°C) irradiations are only associated with pre-existent microstructural defects. Nanoindentation tests shows no hardening in light particle irradiated samples, strongly suggesting that no dislocation loop formed. These observations put forward the importance of point defect clusters resulting from displacement cascades in Ni clusters’ formation.

Mn-Ni-Si intermetallic precipitates (MNSPs) that are observed in some Fe-based alloys following thermal aging and irradiation are of considerable scientific and technical interest. For example, large volume fractions (f) of MNSPs form in reactor pressure vessel low alloy steels irradiated to high fluence, resulting in severe hardening induced embrittlement. Nine compositionally-tailored small heats of low Cu RPV-type steels, with an unusually wide range of dissolved Mn (0.06–1.34 at.%) and Ni (0.19–3.50 at.%) contents, were irradiated at ≈ 290 °C to ≈ 1.4 × 1020 n/cm2 at an accelerated test reactor flux of ≈3.6 × 1012 n/cm2-s (E > 1 MeV). Atom probe tomography shows Mn-Ni interactions play the dominant role in determining the MNSP f, which correlates well with irradiation hardening. The wide range of alloy compositions results in corresponding variations in precipitates chemistries that are reasonably similar to various phases in the Mn-Ni-Si projection of the Fe based quaternary. Notably, f scales with ≈ Ni1.6Mn0.8. Thus f is modest even in advanced high 3.5 at.% Ni steels at very low Mn (Mn starvation); in this case Ni-silicide phase type compositions are observed.

It has long been recognized that exposure to irradiation environments could dramatically degrade the mechanical properties of nuclear structural materials, i.e., irradiation-hardening and embrittlement. With the development of numerical simulation capability and advanced experimental equipment, the mysterious veil covering the fundamental mechanisms of irradiation-hardening and embrittlement has been gradually unveiled in recent years. This review intends to offer an overview of the fundamental mechanisms in this field at moderate irradiation conditions. After a general introduction of the phenomena of irradiation-hardening and embrittlement, the formation of irradiation-induced defects is discussed, covering the influence of both irradiation conditions and material properties. Then, the dislocation-defect interaction is addressed, which summarizes the interaction process and strength for various defect types and testing conditions. Moreover, the evolution mechanisms of defects and dislocations are focused on, involving the annihilation of irradiation defects, formation of defect-free channels, and generation of microvoids and cracks. Finally, this review closes with the current comprehension of irradiation-hardening and embrittlement, and aims to help design next-generation irradiation-resistant materials.

High resolution digital image correlation (HRDIC) has been used to quantify the effect of proton irradiation on strain localisation in Zircaloy-4. Confinement of slip to dislocation channels in irradiated material lead to intense, planar slip bands with high effective shear strain values within channels. More diffuse, homogenous slip was observed in non-irradiated material, with the highest strains measured near grain boundaries. By comparing experimental slip trace angles from HRDIC with theoretical slip trace angles determined using grain orientations from electron backscatter diffraction (EBSD), the active slip plane was determined. Understanding the localised deformation of irradiated materials relative to their microstructure is essential for valid predictions of the structural integrity, and therefore design life, of components in nuclear applications.

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

We present a microstructural study of a surveillance test specimen from a reactor pressure vessel steel of a Russian-type nuclear reactor after neutron irradiation, post-irradiation annealing and re-irradiation, using atom probe tomography (APT) and positron annihilation spectroscopy (PAS). The APT results showed the formation of Cu-rich solute nano-clusters (CRCs) during the initial irradiation and their subsequent coarsening during annealing. After re-irradiation, new CRCs have been observed. The irradiation-hardening almost recovered during annealing. However, by re-irradiation, hardening comparable to that by the initial irradiation was observed. The hardening due to the CRCs formed during the initial irradiation, estimated using the Russell–Brown model, was almost the same as that observed. However, the estimated hardening after the re-irradiation was about half of the measured one. The other hardening is attributed to the newly formed MDs by the re-irradiation, which was evidenced by PAS.

The purpose of this study is to investigate the internal friction (IF) behaviour of thermally aged Fe–Cu alloys that are model specimens simulating for the irradiation embrittlement of nuclear reactor pressure vessel (RPV) steels. Plate-shaped specimens of Fe–1 wt.% Cu alloy were quenched from 1123 K and thermally aged at 773 K for 10 3 min. The IF of the specimens was measured without magnetic field using the free resonant bending vibration method. Magnetic hysteresis loops and conductivity were also measured. The IF value, hysteresis loss, and conductivity increased under thermal aging, which suggests that the IF characteristics are closely related to the change of electromagnetic eddy-current loss due to the depletion of Cu solute atoms. This study demonstrates the possibility of the application of IF measurement as one of the tools for nondestructive characterization of degradation processes of irradiated RPV steels.

Recently, annealed specimens of pure copper have been tensile tested in a fission reactor at a damage rate of 6 × 10−8 dpa/s with a constant strain rate of 1.3 × 10−7 s−1. The specimen temperature during the test was about 90 °C. The stress response was continuously recorded as a function of irradiation time (i.e. displacement dose and strain). The experiment lasted for 308 h. During the dynamic in-reactor test, the specimen deformed and hardened homogeneously without showing any sign of yield drop and plastic instability. However, the specimen yielded a uniform elongation of only about 12%. The preliminary results are briefly described and discussed in the present note.

Surveillance specimens of 15Kh2MFA reactor pressure vessel (RPV) steel and Sv-10KhMFT weld metal were investigated using internal friction, electrical resistivity and Mossbauer spectroscopy measurements in order to obtain information on their irradiation embrittlement and recovery annealing mechanisms. Internal friction measurements showed that carbon becomes trapped by irradiation-induced defects (fluence=4.1×10 19 n/cm 2 , E >1 MeV, T =265 °C). During the recovery annealing (450 and 475 °C/1 h) carbon does not redissolve into the solid solution. Although irradiation usually increases the electrical resistivity of the RPV steels, in this study irradiation decreased the electrical resistivity of both the base and the weld metals. This can be due to a decrease in the dislocation density or precipitation under irradiation. The annealing temperature range for recovery was 300–600 °C for the base and the weld metals according to the electrical resistivity measurements. Some evidence of irradiation-induced carbide formation was achieved by the Mossbauer spectroscopy measurements.

In the unirradiated condition the Ti6Al4V (α+β) alloy has slightly higher tensile strength and noticeably lower ductility compared to that of the Ti5Al2.5Sn (α) alloy both at 50 and 350 °C. The fracture toughness behaviour of both alloys is similar at ambient temperature. At 350 °C, on the other hand, the fracture toughness of the (α) alloy is lower compared to that of the (α+β) alloy. Neutron irradiation at 50 °C to a dose level of 0.3 dpa caused hardening, plastic instability and a substantial reduction in fracture toughness of both alloys. Irradiation at 350 °C resulted in a substantial hardening and a significant decrease in the fracture toughness in the (α+β) alloy due to irradiation induced precipitation whereas only minor changes in the tensile and fracture toughness behaviour were observed in the (α) alloy. The tensile and fracture toughness properties of the (α+β) alloy are more strongly affected by neutron irradiation compared to that of the (α) alloy.

The tensile and fracture behaviour of CuCrZr and CuAl25 IG0 alloys joint to 316L(N) stainless steel by hot isostatic pressing (HIP) have been determined in unirradiated and neutron-irradiated conditions. The tensile and fracture behaviour of copper alloy HIP joint specimens are dominated by the properties of the copper alloys, and particularly, by the strength mismatch and mismatch in strain hardening capacities between copper alloys and stainless steel. The test temperature, neutron irradiation and thermal cycles primarily affect the copper alloy HIP joint properties through changing the strength mismatch between the base alloys. Changes in the loading conditions i.e. tensile, bend ( J I ) and mixed-mode bend ( J I / J II ) lead to different fracture modes in the copper alloy HIP joint specimens.

For the Ringhals 3 and 4 pressurized water reactors (PWR) reactor pressure vessels (RPV), results from the irradiation surveillance program are available also for neutron fluences, which cover long-term operation (LTO). These standard surveillance results are based on the RTNDT concept. The belt-line welds of both RPVs have an elevated nickel-content of 1.6 wt. % and, as a consequence, irradiation response is higher than predicted by model equations. Therefore, the mechanical testing program has been expanded, exceeding the requirements of the standard testing program and covering both base and weld materials. To improve the understanding of the material behavior, extended Master Curve testing was performed on PCCV and subsize SE(B) specimens from irradiation surveillance capsules with the help of specimen reconstitution technique. Special care has been taken on the limited amount of weld material within the available broken Charpy halves before specimen reconstitution. Results have been compared to existing data on similar base and weld materials from the German research programs CARISMA and CARINA. Late-blooming effects or sudden saturation effects are not observed for base or weld materials under LTO conditions. The data for the four different weld materials of similar chemical composition indicate that not only the chemical composition, but also other influencing factors like, e.g., the welding heat treatment, may be important for the reference temperature of the unirradiated state as well as for the irradiation behavior. To investigate this effect in more detail, a future investigation program will be discussed including manufacturing of a surrogate weld material with the same chemical composition as in Ringhals 3 and 4 RPV. The influence of heat treatment condition can be investigated by applying different heat treatments and subsequently performing test reactor irradiation and mechanical testing. Specimen reconstitution will be required due to limited space inside the test reactor irradiation capsules.

Most of the existing local approaches for cleavage fracture derive from the assumptions that global failure is a weakest-link event and that only the tail of the size distribution of micro-crack initiating features is significant. This appears to be sufficient in predicting lower shelf toughness under high constraint conditions, but may fail when attempting to predict toughness in the transition region or for low constraint conditions when using the same parameters. While coupled ductile damage models with Beremin-like failure probability could be useful in the transition region, uncoupled models with “a posteriori” probability calculations are advantageous to the engineering community. Cleavage toughness predictions in the transition regime, which can be extended to low constraint conditions, are here made with improved criterion for particle failure and experimentally based size distribution of initiators for specific RPV steel. The model is shown to predict experimentally measured locations of cleavage initiators. Further, the model predicts the fracture toughness in a large part of the transition region and accurately predicts the measured shift with irradiation, although further works is required to improve the model’s predictions at higher levels of fracture toughness and for low constraint conditions, which are expected to be achieved by a better accounting for plasticity. All results are obtained without changes in model parameters. This suggests that the model can be used as a tool for analysing toughness changes due to constraint and temperature driven plasticity changes.

The evolution of interstitial dislocation loops in V-4Cr-4Ti during heavy ion and neutron irradiation was investigated in comparison with that during electron irradiation at temperature ranging from 407 K to 673 K. Temperature and damage rate dependence of the loop number density by heavy ion and neutron irradiations agreed well with those by electron irradiation obeying Sink and Recombination Dominant kinetics at low and high temperature, respectively. Heavy ion and neutron irradiations induced higher loop number density than that of electron irradiation at 563 - 673 K. The difference may be attributed to cascade-enhanced vacancy loop formation.

Irradiation embrittlement rate as well as the re-embrittlement rate after annealing of pressure vessel steels can in general be described by the copper, phosphorus and nickel contents of the materials. In the current study embrittlement and re-embrittlement rate of 32 different model alloys were measured using sub-size CH-V specimens of dimensions 3 × 4 × 27 mm. The impurity and alloy elements of the model alloys were varied within the limits Cu: 0−1.0 wt-%, P = 0−0.040 wt-% and Ni: 0−2.0 wt-% relatively evenly. The specimens were originally irradiated in the high flux reactor in Petten into a neutron fluence of 0.3×1019n/cm2, E >1 MeV and re-irradiated after the annealing treatment in the Loviisa power reactor into a fluence of 1.0 × 1019n/cm2, E > 1 MeV. In the paper the irradiation and re-irradiation embrittlement rate has been modeled based on the Cu, P and Ni contents of the alloys. Re-irradiation rate was found to be clearly lower than the original irradiation rate except for small corner in the element space, i.e., no nickel, low copper and high phosphorus alloys. Phosphorus coefficient in the re-irradiation shift function is relatively two times higher than in the original irradiation shift function. Many of the alloys in the IAI-condition show intergranular type fracture (IGF). However, the effect of the change of the fracture mode from cleavage to IGF on transition temperature shift remains open. The dependence of irradiation embrittlement, annealing and re-embrittlement behavior on the copper, phosphorus and nickel contents is still not solved, one reason being the relatively limited variation of the key elements in the materials. Model alloys offer a good variation matrix of the elements even, if the measured data cannot be directly applied to real steels.

A model based on key embrittlement mechanisms is proposed for the analysis. The advantages of such semi-mechanistic model, when compared to non-mechanistic models, is that it allows improved fitting of data and permits the visualization of the relative contribution to embrittlement of the various damage components. Data from a set of model alloys with parametric variation of the Cu, P and Ni contents, irradiated in the High Flux Reactor Petten (The Netherlands) and Kola Nuclear Power Plant (Russia), have been thoroughly analysed. The low Ni model alloys results are studied in detail showing the semi-mechanistic model capabilities and their potential for application to commercial steels and welds; in particular for those used in Russian pressure water reactor (VVER). The additional effect of Ni and its influence in the model are also presented in this article.