Neutron induced damage in reactor pressure vessel steel: An X-ray absorption fine structure study

The radiation damage produced in reactor pressure vessel (RPV) steels during neutron irradiation is a long-standing problem of considerable practical interest. In this study, an extended X-ray absorption fine structure (EXAFS) spectroscopy has been applied at Cu, Ni and Mn K-edges to systematically investigate neutron induced radiation damage to the metal-site bcc structure of RPV steels, irradiated with neutrons in the fluence range from 0.85 to 5.0 × 10¹⁹ cm⁻². An overall similarity of Cu, Ni and Mn atomic environment in the iron matrix is observed. The radial distribution functions (RDFs), derived from EXAFS data have been found to evolve continuously as a function of neutron fluence describing the atomic-scale structural modifications in RPVs by neutron irradiations. From the pristine data, long range order beyond the first- and second-shell is apparent in the RDF spectra. In the irradiated specimens, all near-neighbour peaks are greatly reduced in magnitude, typical of damaged material. Prolonged annealing leads annihilation of point defects to give rise to an increase in the coordination numbers of near-neighbour atomic shells approaching values close to that of non-irradiated material, but does not suppress the formation of nano-sized Cu and/or Ni-rich-precipitates. Total amount of radiation damage under a given irradiation condition has been determined. The average structural parameters estimated from the EXAFS data are presented and discussed.     

Small-angle neutron scattering study on the effect of hydrogen in irradiated reactor pressure vessel steels

Hydrogen uptake can enhance the neutron embrittlement of reactor pressure vessel (RPV) steels. This suggests that irradiation defects act as hydrogen traps. The evidence of hydrogen trapping was investigated using the small-angle neutron scattering (SANS) method on four RPV steels. The samples were examined in the unirradiated and irradiated states and both in the as-received condition and after hydrogen charging. Despite the low bulk content of hydrogen achieved after charging with low current densities, an enrichment of hydrogen in small microstructural defects could be identified. Preferential traps were microstructural defects in the size range of ≈ > 10 nm in the unirradiated and irradiated samples. However, the results do not show any evidence for hydrogen trapping in irradiation defects.     

Monitoring the embrittlement of reactor pressure vessel steels by using the Seebeck coefficient

The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient () of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1 MeV, and fluencies ranging from 0 up to 4.5 × 10¹⁹ neutrons per cm², are presented. Within this range, it was observed that increased by ≈500 nV/°C and a linear dependency was noted between and the temperature shift ΔT_41 J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible.     

Grain boundary influence on displacement cascades in UO2: A molecular dynamics study

The influence of grain boundaries on the primary damage state created by a recoil nucleus in UO₂ matrix is studied here by molecular dynamics simulations. This study is divided in two steps: (1) the study of the structural properties of several symmetrical tilt boundaries for different misorientation angles ranging from 12.7° to 61.9°; and (2) the study of displacement cascades near these grain boundaries. For all the grain boundaries studied, the structure around the interface up to about 2 nm presents a perturbed but stable fluorite lattice. The type of defect at the interface depends directly on the value of the misorientation angles. For the small angles (12.7° and 16.3°) the interface defects correspond to edge dislocations. For higher misorientation angles, a gap of about 0.3 nm exists between the two halves of the bicrystal. This gap is composed of Schottky defects involving numerous vacancies along the interface. About 10 keV displacement cascades were initiated with an uranium projectile close to the interface. In all the cases, numerous point defects are created in the grain boundary core, and the mobility of these defects increases. However, cascade morphologies depend strongly on the grain boundary structure. For grain boundaries with edge dislocations, the evolution of the displacement cascades is similar to those carried out in monocrystals. On the other hand, cascades initiated in grain boundaries with vacancy layer defects present an asymmetry on the number of displaced atoms and the number of point defects created.     

Computer simulation of primary damage creation in displacement cascades in copper. I. Defect creation and cluster statistics

Atomic-scale computer simulation has been used to investigate the primary damage created by displacement cascades in copper over a wide range of temperature (100 K ? T ? 900 K) and primary knock-on atom energy (5 keV ? E_PKA ? 25 keV). A technique was introduced to improve computational efficiency and at least 20 cascades for each (E_PKA, T) pair were simulated in order to provide statistical reliability of the results. The total of almost 450 simulated cascades is the largest yet reported for this metal. The mean number of surviving point defects per cascade is only 15-20% of the NRT model value. It decreases with increasing T at fixed E_PKA and is proportional to (E_PKA)^1.1 at fixed T. A high proportion (60-80%) of self-interstitial atoms (SIAs) form clusters during the cascade process. The proportion of clustered vacancies is smaller and sensitive to T, falling from 30% to 60% for T ? 600 K to less than 20% when T = 900 K. The structure of clusters has been examined in detail. Vacancies cluster predominantly in stacking-fault-tetrahedron-type configurations. SIAs tend to form either glissile dislocation loops with Burgers vector b = 1/2<1 1 0> or sessile faulted Frank loops with b = 1/3<1 1 1>. Despite the fact that cascades at a given E_PKA and T exhibit a wide range of defect numbers and clustered fractions, there appears to be a correlation in the formation of vacancy clusters and SIA clusters in the same cascade. The size and spatial aspects of this are analysed in detail in part II [unpublished], where the stability of clusters when another cascade overlaps them is also investigated.     

Thermal effects in 10 keV Si PKA cascades in 3C-SiC

We present a molecular dynamics study of the influence of temperature on defect generation and evolution in irradiated cubic silicon carbide. We simulated 10 keV displacement cascades, with an emphasis on the quantification of the spatial distribution of defects, at six different temperatures from 0 K to 2000 K under identical primary knock-on atom conditions. By post-processing the simulation results we analyzed the temporal evolution of vacancies, interstitials, and antisite defects, the spatial distribution of vacancies, and the distribution of vacancy cluster sizes. The majority of vacancies were found to be isolated at all temperatures. We found evidence of temperature dependence in C and Si replacements and C_Si antisite formation, as well as reduced damage generation behavior due to enhanced defect relaxation at 2000 K.     

Neutron irradiation-induced dimensional changes in MEMS glass substrates

A study is made of radiation-induced expansion/compaction in Pyrex^® (Corning 7740) and Hoya SD-2^® glasses, which are used as substrates for MEMS devices. Glass samples were irradiated with a neutron fluence composed primarily of thermal neutrons, and a flotation technique was employed to measure the resulting density changes in the glass. Transport of Ions in Matter (TRIM) calculations were performed to relate fast (∼1 MeV) neutron atomic displacement damage to that of boron thermal neutron capture events, and measured density changes in the glass samples were thus proportionally attributed to thermal and fast neutron fluences. Pyrex was shown to compact at a rate of (in Δρ/ρ per n/cm²) 8.14 × 10⁻²⁰ (thermal) and 1.79 × 10⁻²⁰ (fast). The corresponding results for Hoya SD-2 were 2.21 × 10⁻²¹ and 1.71 × 10⁻²¹, respectively. On a displacement per atom (dpa) basis, the compaction of the Pyrex was an order of magnitude greater than that of the Hoya SD-2. Our results are the first reported measurement of irridiation-induced densification in Hoya SD-2. The compaction of Pyrex agreed with a previous study. Hoya SD-2 is of considerable importance to MEMS, owing to its close thermal expansivity match to silicon from 25 to 500°C.     

Analyzing the effect of displacement rate on radiation-induced segregation in 304 and 316 stainless steels by examining irradiated EBR-II components and samples irradiated with protons

Recent studies have indicated that, at temperatures relevant to fast reactors and light water reactors, void swelling in austenitic alloys progresses more rapidly when the radiation dose rate is lower. A similar dependency between radiation-induced segregation (RIS) and dose rate is theoretically predicted for pure materials and might also be true in complex engineering alloys. Radiation-induced segregation was measured on 304 and 316 stainless steel, irradiated in the EBR-II reactor at temperatures near 375 °C, to determine if the segregation is a strong function of damage rate. The data taken from samples irradiated in EBR-II is also compared to RIS data generated using proton radiation. Although the operational histories of the reactor irradiated samples are complex, making definitive conclusions difficult, the preponderance of the evidence indicates that radiation-induced segregation in 304 and 316 stainless steels is greater at lower displacement rate.     

Radiation damage in ZnO ion implanted at 15 K

Commercial O-face (0 0 0 1) ZnO single crystals were implanted with 200 keV Ar ions. The ion fluences applied cover a wide range from 5 × 10¹¹ to 7 × 10¹⁶ cm⁻². The implantation and the subsequent damage analysis by Rutherford backscattering spectrometry (RBS) in channelling geometry were performed in a special target chamber at 15 K without changing the target temperature of the sample. To analyse the measured channelling spectra the computer code DICADA was used to calculate the relative concentration of displaced lattice atoms.Four stages of the damage evolution can be identified. At low ion fluences up to about 2 × 10¹³ cm⁻² the defect concentration increases nearly linearly with rising fluence (stage I). There are strong indications that only point defects are produced, the absolute concentration of which is reasonably given by SRIM calculations using displacement energies of E_d(Zn) = 65 eV and E_d(O) = 50 eV. In a second stage the defect concentration remains almost constant at a value of about 0.02, which can be interpreted by a balance between production and recombination of point defects. For ion fluences around 5 × 10¹⁵ cm⁻² a second significant increase of the defect concentration is observed (stage III). Within stage IV at fluences above 10¹⁶ cm⁻² the defect concentration tends again to saturate at a level of about 0.5 which is well below amorphisation. Within stages III and IV the damage formation is strongly governed by the implanted ions and it is appropriate to conclude that the damage consists of a mixture of point defects and dislocation loops.     

Microstructural aspects of neutron embrittlement of reactor pressure vessel steels - A view from positron annihilation spectroscopy

A comprehensive review of positron annihilation studies of Cr---Mo---V reactor pressure vessel (RPV) steels (Soviet type 15Kh2MFA) in unirradiated and neutron irradiated states is presented. The influences of lattice defects, impurity atom distribution, irradiation temperature, flux and fluence of fast neutrons on positron annihilation parameters, especially during isochronal annealing, are discussed in terms of the positron trapping model. In contrast to the literature, where irradiation-enhanced Cu precipitates and solute coated microvoids are thought to be major defect types responsible for strengthening and hence embrittling of RPV steels, we suggest irradiation-induced precipitates, i.e. probably carbides, to play this role. Possibilities to probe this model are suggested.     

Study of PRIMAVERA steel samples by a positron annihilation spectroscopy technique

In the present article, a positron annihilation spectroscopy investigation of VVER-440/230 weld materials is discussed. Important characteristics of metals such as Fermi energy, concentration of electrons in the conduction band, size and concentration of defects were experimentally determined for three model materials with higher level of copper (0.16 wt.%) and phosphorus (0.027-0.038 wt.%). The impact of neutron irradiation and subsequent annealing on crystal lattice parameters was investigated. The experiments with the angular correlation of positron annihilation radiation (ACAR) complement the published positron annihilation spectroscopy (PAS) studies of the radiation treated VVER materials as well as previous experiments on PRIMAVERA materials. The availability of the experimental reactor to prepare strong ⁶⁴Cu positron sources provided for unique experimental conditions, such as good resolution of spectra (0.4 mrad) and reasonable short time of measurement (36 h). The present paper aims to contribute to further understanding of RPV (reactor pressure vessel) steels behaviour under irradiation conditions as well as annealing recovery procedures, which have already been applied at several VVER NPP units in Europe.     

Synthesis of atom probe experiments on irradiation-induced solute segregation in French ferritic pressure vessel steels

Microstructural changes due to neutron irradiation cause an evolution of the mechanical properties of reactor pressure vessels (RPV) steels. This paper aims at identifying and characterising the microstructural changes which have been found to be responsible in part for the observed embrittlement. This intensive work relies principally on an atom probe (AP) study of a low Cu-level French RPV steel (Chooz A). This material has been irradiated in in-service conditions for 0–16 years in the frame of the surveillance program. Under this aging condition, solute clustering occurs (Cu, Ni, Mn, Si, P, …). In order to identify the role of copper, experiments were also carried out on Fe–Cu model alloys submitted to different types of irradiations (neutron, electron, ion). Cu-cluster nucleation appears to be directly related to the presence of displacement cascades during neutron (ion) irradiation. The operating basic physical process is not clearly identified yet. A recovery of the mechanical properties of the irradiated material can be achieved by annealing treatments (20 h at 450°C in the case of the RPV steel under study, following microhardness measurements). It has been shown that the corresponding microstructural evolution was a rapid dissolution of the high number density of irradiation-induced solute clusters and the precipitation of a very low number density of Cu-rich particles.     

Void swelling in a 9Cr ferritic/martensitic steel irradiated with energetic Ne-ions at elevated temperatures

In this work the void swelling behavior of a 9Cr ferritic/martensitic steel irradiated with energetic Ne-ions is studied. Specimens of Grade 92 steel (a 9%Cr ferritic/martensitic steel) were subjected to an irradiation of ²⁰Ne-ions (with 122 MeV) to successively increasing damage levels of 1, 5 and 10 dpa at a damage peak at 440 and 570 °C, respectively. And another specimen was irradiated at a temperature ramp condition (high flux condition) with the temperature increasing from 440 up to 630 °C during the irradiation. Cross-sectional microstructures were investigated with a transmission electron microscopy (TEM). A high concentration of cavities was observed in the peak damage region in the Grade 92 steel irradiated to 5 dpa, and higher doses. The concentration and mean size of the cavities showed a strong dependence on the dose and irradiation temperature. Enhanced growth of the cavities at the grain boundaries, especially at the grain boundary junctions, was observed. The void swelling behavior in similar 9Cr steels irradiated at different conditions are discussed by using a classic void formation theory.     

Photoconduction investigations in 75 MeV oxygen ion irradiated kapton-H polyimide

Photoconduction behaviour of 75 MeV oxygen ion irradiated (Fluences: 1.8 × 10¹¹, 1.8 × 10¹² and 1.8 × 10¹³ ions/cm²) kapton-H polyimide film in the visible region has been investigated at different temperatures ranging 400-2500 °C and at various electric fields ranging 40-600 kV/cm. A photoinduced exciton formation is the major source for providing charge carriers through thermolization and field-assisted dissociation processes. An attempt has been made to fit the field dependence of the steady state photocurrent to one of the several possible conduction mechanisms. In the high and low fluence (1.8 × 10¹³ and 1.8 × 10¹¹ ions/cm²) irradiated samples there exists a possibility of Poole-Frankel type of photoconduction mechanism, whereas at intermediate fluence (1.8 × 10¹² ions/cm²) a Schottky type photoconduction mechanism may be operative. The log I_ps versus 1/T plots consist of two straight lines with a knee point around 800-1000 °C. The activation energy estimated from the slope of these lines is field dependent varying from 0.40 to 0.73 eV and 0.18 to 0.23 eV above and below the knee point, respectively. This indicates the presence of more than one type of trapping levels in irradiated kapton-H polyimide.     

Evolution of the nanostructure of VVER-1000 RPV materials under neutron irradiation and post irradiation annealing

A high nickel VVER-1000 (15Kh2NMFAA) base metal (1.34 wt% Ni, 0.47% Mn, 0.29% Si and 0.05% Cu), and a high nickel (12Kh2N2MAA) weld metal (1.77 wt% Ni, 0.74% Mn, 0.26% Si and 0.07% Cu) have been characterized by atom probe tomography to determine the changes in the microstructure during neutron irradiation to high fluences. The base metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 14.9 × 10²³ m⁻² (E > 0.5 MeV), and the weld metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 11.5 × 10²³ m⁻² (E > 0.5 MeV). High number densities of ∼2-nm-diameter Ni-, Si- and Mn-enriched nanoclusters were found in the neutron irradiated base and weld metals. No significant copper enrichment was associated with these nanoclusters and no copper-enriched precipitates were observed. The number densities of these nanoclusters correlate with the shifts in the ΔT_{41 J} ductile-to-brittle transition temperature. These nanoclusters were present after a post irradiation anneal of 2 h at 450 °C, but had dissolved into the matrix after 24 h at 450 °C. Phosphorus, nickel, silicon and to a lesser extent manganese were found to be segregated to the dislocations.     

Understanding of copper precipitation under electron or ion irradiations in FeCu0.1 wt% ferritic alloy by combination of experiments and modelling

This work is dedicated to the understanding of the basic processes involved in the formation of copper enriched clusters in low alloyed FeCu binary system (FeCu0.1 wt%) under irradiation at temperature close to 300 °C. Such an alloy was irradiated with electrons or with ions (Fe⁺ or He⁺) in order to deconvolute the effect of displacement cascades and the associated generation of point defect clusters (ion irradiations), and the super-saturation of mono-vacancies and self-interstitial atoms (electron irradiation). The microstructure of this alloy was characterised by tomographic atom probe. Experimental results were compared with results obtained with cluster dynamic model giving an estimation of the evolution of point defects (free or agglomerated) under irradiation on the one hand and describing homogeneous enhanced precipitation of copper on the other hand. The comparison between the results obtained on the different irradiation conditions and the model suggests that the point defect clusters (dislocation loops and/or nano-voids) created in displacement cascades play a major role in copper clustering in low copper alloy irradiated at 573 K.     

Method based on isotope ratio mass spectrometry for evaluation of carbon activation in the reactor graphite

The general idea of this work is to introduce an evaluation method to restore the irradiation parameters of graphite or other carbonaceous materials using experimental and modelling results of ¹³C generation in the irradiated material. The method is based on coupling of stable isotope ratio mass spectrometry and computer modelling of the reactor core to evaluate the realistic characteristics of the reactor core such as the neutron fluence in any position of the reactor graphite stack or other graphite constructions.The generation of carbon isotopes ¹³C and ¹⁴C in the irradiated graphite of the RBMK-1500 reactor has been estimated by modelling of the reactor core with computer codes MCNPX and CINDER90. Good agreement of simulated and measured Δ¹³C/¹²C values in graphite of the central part of the reactor core indicates that the neutron flux (1.40 × 10¹⁴ n/cm² s) is modelled accurately in the graphite sleeve of the fuel channel. The simulated activity of ¹⁴C is compared with the one measured by the β spectrometry technique. Results indicate that production of ¹⁴C from ¹⁴N in the RBMK-1500 reactor is considerable and has to be taken into account in order to make proper evaluation of ¹⁴C activity. Measured ¹⁴C specific activity values correspond to 15 ± 4 ppm impurity of ¹⁴N in graphite samples from the RBMK-1500 reactor core.     

Characterization of neutron-irradiated ferritic model alloys and a RPV steel from combined APT, SANS, TEM and PAS analyses

Understanding the behavior of reactor pressure vessel (RPV) steels under irradiation is a mandatory task that has to be elucidated in order to be able to operate safely a nuclear power plant or to extend its lifetime. To build up predictive tools, a substantial experimental data base is needed at the nanometre scale to extract quantitative information on neutron-irradiated materials and to validate the theoretical models. To reach this experimental goal, ferritic model alloys and French RPV steel were neutron irradiated in a test reactor at an irradiation flux of 9 × 10¹⁷ nm⁻² s, doses from 0.18 to 1.3 × 10²⁴ nm⁻² and 300 °C. The main goal of this paper is to report the characterization of the radiation-induced microstructural change in the materials by using the state-of-the-art of characterization techniques available in Europe at the nanometre scale. Possibilities, limitations and complementarities of the techniques to each other are highlighted.     

Effect of ion beam irradiation on magnetic and structural properties of Pt/Cr/Co multilayers

This paper discusses the effect of ion beam irradiation on the magnetic and structural properties of Pt/Cr/Co multilayers. We observe Co-Cr-Pt ternary alloy phase formation in 1 MeV N⁺ ion irradiated [Pt (2.5 nm)/Cr (0.8 nm)/Co (3.0 nm)]_×6/Si multilayers for a fluence of 1 × 10¹⁶ ions cm⁻² and beyond. The observed phase formation is accompanied by an enhancement in the average grain size, surface roughness and coercivity. Monte Carlo simulation has been performed to study ion-induced defect evolution and atomic displacements to correlate the above observed effects.