Neutron Irradiation Embrittlement Research Articles

Development of potassium doped tungsten plate for fusion reactor applications

Since fusion reactors including DEMO and beyond are expected to be operated for a longer period of time than ITER, plasma-facing materials (PFM) will suffer by longer term heat load and non-negligible neutron irradiation damage. Therefore, for tungsten (W) materials as PFM, improvement of low temperature brittleness and suppression of recrystallization embrittlement and neutron irradiation embrittlement are required. To address these issues, various modified W materials have been developed by applying alloying with solid solution elements and second-phase dispersion techniques such as solid particles and bubbles. As a result, potassium-doped W (K-doped W), in which K bubbles are dispersed, was found to be effective for all of the above issues. However, the materials used in these evaluations were relatively small and not large enough to fabricate, for example, ITER divertor monoblocks (12 mm thick). In the present study, K-doped W plate with a sufficient thickness (more than 12 mm) for application to fusion reactor components like divertor monoblocks were developed and the fundamental properties were evaluated.

Prediction of the binding energy of self interstitial atoms in alpha iron by a graph neural network

The estimation of neutron irradiation embrittlement is crucial for the long-term operation of nuclear power plants. The jump of self-interstitial atoms is an essential process for understanding embrittlement, and their binding energy is critical to estimating the activation energy for jumping. This study presents a fast estimation method for the binding energy of self-interstitial atoms in iron using a graph neural network. The network assigns a self-interstitial atom to a node and the relationship between self-interstitial atoms to an edge. The graph neural network was trained using a large dataset of binding energies between self-interstitial atoms (∼0.5 % density) calculated using an embedded atom method interatomic potential. The network was also trained using smaller datasets of binding energies between self-interstitial atoms (∼2.0 % density) calculated using first principles calculations. The graph neural network accurately predicts binding energy with a mean absolute error of 0.05–0.06 eV and coefficient of determination (R2) of 0.9 without overfitting, using the large interatomic potential dataset. When trained on smaller first principles datasets, the network predicted binding energies with a mean absolute error of 0.15–0.16 eV and coefficient of determination (R2) of 0.6 with some degree of overfitting. The developed graph neural network provides a fast and accurate estimation method for the binding energy of self-interstitial atoms in iron.

中性子照射したZr基金属ガラスの機械的特性

Neutron irradiation embrittlement behaviors of some Zr-Cu-Al and Zr-Cu-Ni-Al BMGs with various Zr-contents have been examined by Charpy impact tests under neutron fluences of 1~2 x 1024 n/m2 at relatively low and high temperatures of about 325 K and 550 K, respectively. In the case of the irradiation at 325 K, Charpy impact strength of Zr45Cu45Al10 BMG increased from 59 kJ/m2 up to 92 kJ/m2 accompanied with decreasing the hardness. In the Zr-rich BMGs of Zr50Cu40Al10, Zr55Cu30Ni5Al10, Zr59Cu31Al10, Zr65Cu20Ni5Al10 and Zr65Cu18Ni7Al10, impact strengths decreased from about 90 kJ/m2 down to about 30 kJ/m2 with decreasing the hardness. In the case of the irradiation at 550 K, impact strengths of the Zr-rich BMGs maintained relatively high values of about 60 ~ 90 kJ/m2. These results may be explained by amounts of excess free-volume in the neutron irradiated Zr-based BMGs.

Effects of pre-stress on the mechanical properties and microstructure of neutron-irradiated high-purity aluminum

Although structural materials in nuclear reactors are usually serviced under pre-stressed conditions, the irradiated mechanical properties and microstructures for prestressed metals have not been well understood due to the experimental challenges. In this study, the annealed high-purity aluminum was subjected to a low temperature (313 K) neutron irradiation under a stress of 1.5 MPa, which was ten percent of its unirradiated yield strength. Post-irradiation tensile tests were carried out at 300 K with a strain rate of 1.1 × 10−3 s−1. A strong radiation-induced hardening as well as a loss of uniform ductility was observed on all the specimens, and the pre-stress could defer irradiation hardening and embrittlement and decrease plastic instability stress, which was involved with a possible asymmetry distribution of voids induced by the pre-stress. Transmission electron microscopy (TEM) was conducted on the samples to quantify the void size and density as a function of pre-stress and dose. It was shown that the pre-stress could accelerate the void nucleation, increase the density but decrease the size of voids. Although previous studies have reported that the dislocation channel could be probably attributed to the micro-mechanism for irradiation embrittlement, none of dislocation channels were discovered in the deformed irradiated high-purity aluminum samples, which implied that the dislocation channel is unnecessary for irradiation embrittlement. In addition, a neutron irradiation embrittlement model for metals, which had been built on the basis of Johnson-Cook constitutive model, was found applicable to metals irradiated with pre-stress. The stimulated results also well agreed with the experimental data.

A Detailed Analysis of the H.B. Robinson-2 Reactor Pressure Vessel Dosimetry Benchmark

The operation of many nuclear pressurized water reactors is being extended beyond their design lifetime limit. From the perspective of possible further lifetime extension, safety requirements are a priority. Therefore, the quantification of the neutron irradiation embrittlement of the reactor pressure vessel (RPV) is an important issue, as this is a guiding parameter that influences the reactor lifetime. In this context, the Institut de Radioprotection et de Sûreté Nucléaire developed a calculation scheme for the analysis of RPV aging under neutron irradiation, named VACS (vessel aging calculation scheme). VACS couples a deterministic approach (CASMO5 and SIMULATE5) to evaluate the full-core fission neutron source term and a Monte Carlo modeling (MCNP6) approach to model the neutron attenuation from the core to sites of interest (RPV, surveillance capsules, etc.). To ensure the reliability of aging predictions, this paper describes a detailed analysis of the neutron H.B. Robinson-2 reactor pressure vessel dosimetry benchmark. The results indicate that VACS shows satisfactory accuracy when the ENDF-B/VII.1 or JEFF-3.3 nuclear data libraries are used in the attenuation calculation. However, the use of ENDF-B/VIII.0 leads to significantly worse results.

Recent improvements of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessels

A probabilistic fracture mechanics (PFM) analysis code PASCAL was developed in Japan for probabilistic integrity assessment of reactor pressure vessels (RPVs) in pressurized water reactors (PWRs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. To strengthen the practical applications of PFM methodology in Japan, PASCAL has been upgraded to a new version, PASCAL5, which enables PFM analyses of RPVs in both PWRs and boiling water reactors (BWRs) subjected to a broad range of transients, including PTS and normal operational transients. In this paper, the recent improvements in PASCAL5 are described such as the incorporated stress intensity factor solutions and corresponding calculation methods for external surface cracks and embedded cracks near the RPV outer surface. In addition, the analysis conditions and evaluation models recommended for PFM analyses of Japanese RPVs in BWRs are investigated. Finally, PFM analysis examples for core region of a Japanese BWR-type model RPV subjected to two transients (i.e., low-temperature over pressure and heat-up transients) are presented using PASCAL5.

Analysis of Magnetic Nondestructive Measurement Methods for Determination of the Degradation of Reactor Pressure Vessel Steel.

Nondestructive magnetic measurement methods can be successfully applied to determine the embrittlement of nuclear pressure vessel steel caused by neutron irradiation. It was found in previous works that reasonable correlation could be obtained between the nondestructively measured magnetic parameters and destructively determined ductile-to-brittle transition temperature. However, a large scatter of the measurement points was detected even in the cases of the non-irradiated reference samples. The reason for their scattering was attributed to the local inhomogeneity of material. This conclusion is verified in the present work by applying three different magnetic methods on two sets of Charpy samples made of two different reactor steel materials. It was found that by an optimal magnetic pre-selection of samples, a good, linear correlation can be found between magnetic parameters as well as the ductile-to-brittle transition temperature with low scattering of points. This result shows that neutron irradiation embrittlement depends very much on the local material properties.

Uncertainty assessment for the displacement damage of a pressurized water reactor vessel

The neutron irradiation-induced mechanical property change is a major challenge for nuclear materials. For a Pressurized Water Reactor (PWR), neutron irradiation embrittlement of the reactor pressure vessel is one of the most important issues that determine the whole nuclear power plant lifetime. The present work focuses thus on the uncertainty assessment for the neutron irradiation displacement damage of a PWR vessel. The uncertainties of neutron flux, nuclear data, and atomic displacement models are propagated to Displacement per Atom (DPA) rates with a special emphasis on the importance of correlation matrices. In addition to the uncertainty propagated from the threshold displacement energy, the total uncertainty of damage energy rate is about 10% (at 1 σ ), whereas the uncertainty is largely underestimated if the correlations of damage cross section and neutron flux are not considered. Apart from the complete uncertainty propagation shown in this paper, combining the damage energy rate uncertainty and the displacement threshold energy uncertainty is recommended for estimating the calculated DPA rate uncertainty. Besides the uncertainty, the biases induced by the model defects are preliminarily estimated.

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

Abstract Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

Application of Probabilistic Fracture Mechanics to Reactor Pressure Vessel Using PASCAL4 Code

Abstract Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.

Plasticity Correction on Stress Intensity Factor Evaluation for Underclad Cracks in Reactor Pressure Vessels

Abstract Structural integrity assessment of reactor pressure vessels (RPVs) is essential for the safe operation of nuclear power plants. For RPVs in pressurized water reactors (PWRs), the assessment should be performed by considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. To assess the structural integrity of an RPV, a traditional method is usually employed by comparing fracture toughness of the RPV material with the stress intensity factor (KI) of a crack postulated near the RPV inner surface. When an underclad crack (i.e., a crack beneath the cladding of an RPV) is postulated, KI of this crack can be increased owing to the plasticity effect of cladding. This is because the yield stress of cladding is lower than that of base metal and the cladding may yield earlier than base metal. In this paper, detailed three-dimensional (3D) finite element analyses (FEAs) were performed in consideration of the plasticity effect of cladding for underclad cracks postulated in Japanese RPVs. Based on the 3D FEA results, a plasticity correction method was proposed on KI calculations of underclad cracks. In addition, the effects of RPV geometries and loading conditions were investigated using the proposed plasticity correction method. Moreover, the applicability of the proposed method to the case which considers the hardening effect of materials after neutron irradiation was also investigated. All of these results indicate that the proposed plasticity correction method can be used for KI calculations of underclad cracks and is applicable to structural integrity assessment of Japanese RPVs containing underclad cracks.

Nondestructive Investigation of Neutron Irradiation Generated Structural Changes of Reactor Steel Material by Magnetic Hysteresis Method

The neutron irradiation embrittlement of four different types of nuclear pressure vessel materials (three base metals and one weld material) were investigated by a magnetic nondestructive testing method, magnetic adaptive testing (MAT). The method is based on the measurement of minor magnetic hysteresis loops on Charpy specimens irradiated by neutrons in the BR2 reactor. Due to the neutron irradiation, the structure of the material was modified. The Charpy impact method is suitable for destructive characterization of material embrittlement. The results of Charpy impact test measurements at SCK CEN Belgian Nuclear Research Centre were compared with the nondestructively measured magnetic parameters. A definite correlation was found between magnetic descriptors and the ductile-to-brittle transition temperature (DBTT), regardless of the type of material or irradiation condition. The results suggest that this “calibration curve“ can be used to estimate the DBTT from non-destructive measurements.

Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels

Abstract In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the codes provided by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the probabilistic distributions of various influence factors related to aged degradation due to the long-term operation. In Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, we have developed a guideline for structural integrity assessment of RPVs based on PFM to improve the applicability of PFM in Japan and enable persons who have knowledge on fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body, explanation, and several supplements. The technical basis for PFM analysis is provided, and the latest knowledge is included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

The unified model for irradiation embrittlement prediction of reactor pressure vessel

The irradiation embrittlement of reactor pressure vessel (RPV) steel is the top priority for nuclear power operation safety. In this work, neutron irradiation experiments were performed on a commercial RPV steel at a high temperature using a neutron irradiation experimental reactor, then the mechanical and electromagnetic properties measurements were performed in a hot laboratory. The results show that there is a unified exponential function relationship between the variability of mechanical and electro-magnetic properties for RPV steels during neutron irradiation embrittlement, and the fitting accuracy results show that the prediction unified model is scientific, practical and effective. The theoretical basis of prediction unified model is that both mechanical and electromagnetic properties are external manifestations of the microstructure evolution for RPV under neutron irradiation. Finally, this paper points out that based on the prediction unified model, it can be developed into a non-destructive evaluation technique for RPV irradiation embrittlement with many advantages.

Improvements on Evaluation Functions of a Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessels

Abstract In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL was developed for structural integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. By reflecting the latest knowledge and findings, the evaluation functions are continuously improved and have been incorporated into PASCAL4 which is the most recent version of the PASCAL code. In this paper, the improvements made in PASCAL4 are explained in detail, such as the evaluation model of warm prestressing (WPS) effect, evaluation function of confidence levels for PFM analysis results by considering the epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions, and improved methods for KI calculations when considering complicated stress distributions. Moreover, using PASCAL4, PFM analysis examples considering these improvements are presented.

Performance of tungsten fibers for Wf/W composites under cyclic tensile load

Toughness improved tungsten-based composites are one of the currently considered material option for future fusion reactors capable to withstand both high heat flux and irradiation induced embrittlement. Today, fiber-reinforced composites (Wf/W) are being intensively studied as risk-mitigation materials to replace bulk tungsten which is susceptible to neutron irradiation embrittlement especially below 800 °C. Operation of a material as an element of a plasma facing component (i.e. divertor monoblock or first wall armour) implies not only high heat flux exposure but also thermal cyclic fatigue caused by repetitive oscillations of the heat loads due to the nature of the plasma and the limitations on the capacity of its confinement. In this work, we assessed the performance of potassium doped tungsten fibers under cyclic loading applied in tensile mode. Stress-controlled fatigue tests were performed at room temperature, 300 °C and 500 °C increasing the load from 50% of the yield strength up to the ultimate tensile strength of the studied fibers. It is revealed that significant cyclic hardening emerges as the fatigue stress limit exceeds the yield strength already within a few cycles. Despite the noticed cyclic hardening, the wire can sustain few hundreds of cycles without any detectable damage unless the cycle stress is increased to reach the value above the mean ultimate tensile strength. Given this observation, we have studied the impact of the cyclic stress (σC) on the rupture strength and total elongation of the wires exposed to twenty loading cycles varying test temperature in the range 23–500 °C. At room temperature, the rupture stress after cyclic deformation progressively increases with σC and saturates at 2.7 GPa with a moderate reduction of the total elongation, while the nominal ultimate tensile strength of the wire is 2.5 GPa. Thus, the strength of the wire is increased by 200 MPa, on average. At elevated temperature, the rupture stress after the cyclic deformation increases by more than 300 MPa.

Effects of neutron irradiation on elastic modulus of RPV steel

The embrittlement of reactor pressure vessel (RPV) steel owing to fast- neutron irradiation damage is one of its primary failure mechanisms. In this work, neutron irradiation experiments were performed on a commercial RPV steel at a high temperature (292 °C) using a neutron irradiation experimental reactor. Then the elastic modulus measurements and XRD analysis were performed on the RPV steel before and after irradiation in a hot laboratory. The results shows that the Young modulus and shear modulus of RPV steel do not change during neutron irradiation, the half-width of peak shape parameters and lattice constant of XRD diffraction peak remain unchanged, and no new phase is formed in RPV steel during irradiation embrittlement, the crystal structure of RPV steel does not change after neutron irradiation damage. It is found that the elastic modulus of the macro-mechanical properties and the micro-structure characterized by XRD analysis are consistent with each other during the neutron irradiation embrittlement process of RPV steel. Finally, this paper points out that the conclusion of neutron irradiation damage has no effect on the elastic modulus of RPV steel is of great significance to the structural integrity analysis and safety evaluation of RPV in the whole life of nuclear power plant.

Radiation-induced bcc-fcc phase transformation in a Fe[sbnd]3%Ni alloy

The issue of neutron irradiation embrittlement of Reactor Pressure Vessel steels must be considered for Nuclear Power Plant life extension. This phenomenon partly arises from the existing interactions between dislocations and nanometric clusters composed of Cu, P, Si, Mn and Ni. The latter alloying element, playing a key role in the evolution of solute enriched clusters under irradiation, is the focus of this publication. To assess the effect of Ni on microstructure evolution under irradiation, particle accelerator based experiments were conducted. An under-saturated Fe3at.%Ni alloy was irradiated with self-ions, at 673 K, up to ∼1.2 dpa. Then, the microstructural damage was characterized, at the atomic scale, by conventional Transmission Electron Microscopy, Scanning Transmission Electron Microscopy coupled to Energy Dispersive X-ray Spectroscopy and Electron Energy Loss Spectroscopy, while chemical features were investigated by Atom Probe Tomography. Informations obtained by combining these coupled techniques provide evidence for the formation of a FCC phase, containing 25 at.%Ni, which can be either the disordered γ phase or the ordered L12 type Fe3Ni phase. The metastable or stable state of this FCC phase is discussed in the light of what is known from the literature. It is the first time that this BCC-FCC phase transformation is observed in an under-saturated α-FeNi alloy and this likely occurred via a Radiation Induced Precipitation (RIP) mechanism. Ni atom segregation is observed on cavities, dislocation lines and dislocation loops. The latter constitute nuclei for precipitates, leading to the formation of an additional segregation site for Ni: the precipitate FCC - matrix BCC nearly coherent interface. Similar mechanisms are argued to be operating also in high Ni RPV steels under neutron irradiation.

Study on Dimensional Tolerances Required for Miniature C(T) Specimens

The Master Curve approach using miniature C(T) specimens with 4 mm thickness is promising for directly determining the reference temperature of reactor pressure vessel (RPV) steels because they can be taken from the broken halves of the Charpy specimens used in the surveillance program to monitor neutron irradiation embrittlement. The relative dimensional tolerances of standard C(T) specimens are provided in the present ASTM E1921 standard; consequently, the absolute dimensional tolerances are stricter for smaller specimens. In this study, the effect of the tolerances of key dimensions on the elastic–plastic equivalent stress intensity factor derived from the J-integral, KJ, was calculated using three-dimensional finite-element analysis. Even if the dimensional tolerances for the miniature C(T) specimens based on the present standard were mitigated in some degree (as examples, the tolerance of specimen thickness of ±0.08 mm was mitigated to ±0.1 mm; and the tolerance of specimen width of ±0.04 mm was mitigated to ±0.1 mm), the variations of KJ and the reference temperature were negligibly small. Furthermore, the use of the mitigated dimensional tolerances with adequate accuracy for evaluating the fracture toughness was ascertained.

Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.