On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ∼7 × 10¹⁹ n cm⁻². This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.     

The development of radiation embrittlement models for US power reactor pressure vessel steels

A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters-Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature - are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.     

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.     

Positron affinity for precipitates in reactor pressure vessel steels

The sensitivity of positron annihilation spectroscopy to irradiation-induced precipitates in reactor pressure vessel steels is discussed in the light of recent positron affinity and lifetime calculations. Carbide and nitride precipitates are found to trap positrons only if they contain metal vacancies. Copper precipitates are also attractive to positrons but they are probably detected through annihilation at the precipitate-matrix interface. These findings are related to available experimental data.     

Embrittlement recovery due to annealing of reactor pressure vessel steels

The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes.     

Suggested toughness/temperature relations for reactor pressure vessel steels

Relations are suggested for the means and standard deviations of three toughness measures for reactor pressure vessel steels: static initiation, dynamic initiation, and arrest. All of the relations are of the form: K_Ix = K_LS{1 + exp[(T − [RT_NDT + δT])/T_O]}, where K_Ix is the toughness measure of interest, K_LS is the lower-shelf toughness, T is the temperature, RT_NDT is the reference transition temperature, δT is a temperature shift, and T_O is a temperature which characterizes the breadth of the transition. The mean of K_LS differs for initiation and arrest and its standard deviation accounts for variation within a single heat. The mean of δT differs for all three toughness measures and its standard deviation accounts for heat-to-heat variability. However, it is shown that a value of T_o = 33.2°C can be used for all of the toughness measures. Finally, the lower bound curves of the ASME Boiler and Pressure Vessel Code are shown to represent toughness levels of low probability.     

Composition influences and interactions in radiation sensitivity of reactor vessel steels

A series of 4-way split laboratory melts of low alloy, pressure vessel steel, representing statistical combinations of specific impurity elements and/or alloying elements, are being evaluated after 288°C irradiation to 2 × 10¹⁹ n/cm², E > 1 MeV. The objective is to reveal interactions between elements (or interdependencies) in radiation sensitivity development. The primary base composition for melting was ASTM A 302-B or A 533-B steel; plates from the melt casts were heat treated to simulate 150 mm or thicker plate materials.This report summarizes preirradiation-postirradiation notch ductility determinations for six metls (24 composition variations) produced for the study. The melts were used to test the effects on radiation sensitivity of phosphorus impurities in combination with copper impurities and the significance of high/low levels of nickel, manganese, molybdenum and chromium alloying to steels containing a significant copper content ( 0.15 % Cu). Contributions of tin and arsenic impurities were also evaluated.One key finding is an inverse dependence of the phosphorus contribution to radiation sensitivity level, on copper content. Therefore, for the new (improved) steels containing a low copper content, phosphorus level should not be ignored in making projections of radiation sensitivity in service. Other important interactions between elements are reported.     

Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.     

Estimation of research reactor core parameters using cascade feed forward artificial neural networks

The pattern of the core reload program is very important for an optimize use of research reactors. Reactor safety issues and economic efficiency should be considered during pattern studies. In order to find the best core pattern for a research reactor, its reloading program should be solved as a multi-objective and constrained optimization problem. If considered objective functions of the optimization problem can be estimated in very short time, the optimal fuel reloading pattern can be used effectively. In this research a very fast estimation system for suggested core parameters has been developed using cascade feed-forward type of artificial neural networks (ANNs). Four main core parameters are suggested to optimize reactor core adequately. And also to get larger thermal fluxes in the desired irradiation box, a new flexible method was selected. A Software package has been developed to prepare and reform required data for ANNs training. The gradient descent method with momentum weight/bias learning rule has been used to train ANNs. To get the best conditions for considered ANNs training a vast study has been performed. It includes the effects of variation of hidden neurons, hidden layers, activation functions, learning and momentum coefficients, and also the number of training data sets on the training and simulation results. Some experimental convergence criteria are used to study them. A comparison selection rule has been used to adjust desirable conditions. Final training and simulation results show that developed ANNs can be trained and estimate suggested core parameters of research reactors very quickly. It improves effectively pattern optimization process of core reload program.     

Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the ‘BWRVIP-60 SCC disposition lines’ were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the ‘BWRVIP-60 SCC disposition lines’.     

Irradiation experiments related to reactor safety in particular pressure vessel steels

The dependence of neutron induced embrittlement of reactor pressure vessel steels on irradiation temperature and neutron exposure was investigated for steels with different copper content. A pronounced increase of the ductile to brittle transition temperature shift with decreasing irradiation temperature was found and quantitatively determined. The influence of the neutron energy spectrum and flux density on the embrittlement was not significant.Rigs for irradiating assemblies of fracture mechanics specimens (CT and WOL) up to 100 mm thickness and also for irradiation experiments under cyclic loading were developed. Irradiation experiments with these rigs are in progress.Creep experiments on canning tubes under different load conditions (uniaxial load and biaxial load under internal and external overpressure) as well as an irradiation device for investigating defective PWR fuel rods are briefly reported.     

Considerations about irradiation-induced precipitates in Soviet type reactor pressure vessel steels

Small angle neutron scattering (SANS) results on neutron irradiated Fe-Cu are presented and discussed and compared to positron annihilation results. An extended discussion is presented regarding a comparison of earlier positron annihilation and SANS measurements and their interpretation for different Soviet type reactor pressure vessel steels. It is suggested that the irradiation-induced precipitates contain vacancies and might be metal carbides.     

Observations of a steady state effect limiting radiation damage in reactor vessel steels

Reactor vessel material surveillance capsules which contain specimens of actual material used in the construction of a vessel are contained in nearly all operating reactors. These specimens monitor the changes in properties of the reactor vessel and assure that predicted changes based on trend curves which are used to set operating limits for the plant are conservative. Recently, data has been obtained from the Point Beach Unit No. 1 and Connecticut Yankee reactor vessel surveillance capsules exposed to neutron radiation for much longer periods of time, than those irradiated in test reactors and surveillance capsules which were removed at the first refueling and other early refueling outages. The data from these long time surveillance capsule exposures when compared to data from capsules removed from the same reactors earlier in life indicated that a limiting or steady state condition has resulted rather than a continuous embrittlement as predicted by trend curves. It is believed that the limited embrittlement or steady state condition which occurred from the surveillance capsule tests is due to a combination of relatively low neutron flux compared to that existing in test reactors which were the primary source of data used to establish trend curves and the longer exposure periods in the capsules that led to significant “annealing” during irradiation.     

Effect of carbon on irradiation-induced grain-boundary phosphorus segregation in reactor pressure vessel steels using first-principles-based rate theory model

In this paper, we incorporated the effect of carbon atoms on the irradiation-induced grain-boundary phosphorus segregation into the rate theory model by considering a carbon atom as a trap site of vacancies and self-interstitial atoms, and simulated the grain-boundary phosphorus coverage in the reactor pressure vessel steels, A533B steels which were neutron-irradiated using the Halden reactor. As a result, by selecting the sink strength of vacancies and self-interstitial atoms, the simulation reproduced the experimental grain-boundary phosphorus coverage that was measured using the scanning Auger electron microprobe analysis. It was observed that the grain-boundary phosphorus coverage does not depend on the dose rate regardless of the presence of carbon atoms. Furthermore, it was confirmed that vacancies scarcely transport phosphorus atoms to grain-boundaries as compared to the transport by self-interstitial atoms and it was found that carbon atoms influence the irradiation-induced phosphorus segregation by mainly suppressing the migration of vacancies.     

Positron annihilation lifetime study of JRQ and JPA irradiated and annealed reactor pressure vessel steels

The model reactor pressure vessel steels known as JRQ and JPA were manufactured in Japan for the IAEA neutron embrittlement research studies. These model alloys belong to the commercially used steel A533B-1 type and show relatively large changes in mechanical properties after relevant neutron irradiation. The neutron irradiation was performed by different neutron fluxes as well as different neutron fluences (up to about 150 × 10¹⁸ cm⁻² (E > 0.5 MeV)). For a better understanding of the neutron embrittlement, the Positron Annihilation Lifetime Spectroscopy (PALS) technique was applied in 2014. PALS measurement of irradiated specimens was performed using three detectors set-up due to induced ⁶⁰Co radioactivity of the studied specimens. We confirmed that the JPA steel, considered to be high-copper steel, is much more sensitive to defect creation due to neutron irradiation than the low-copper JRQ steel.     

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.     

Positron annihilation study and computational modeling of defect production in neutron-irradiated reactor pressure vessel steels

Positron annihilation spectroscopy (PAS) and a computer simulation were used to investigate a defect production in reactor pressure vessel (RPV) steels irradiated by neutrons. The RPV steels were irradiated at 250 °C in a high-flux advanced neutron application reactor. The PAS results showed that mainly single vacancies were created to a great extent as a result of a neutron irradiation. Formation of vacancies in the irradiated materials was also confirmed by a coincidence Doppler broadening measurement. For estimating the concentration of the point defects in the RPV steels, we applied computer simulation methods, including molecular dynamics (MD) simulation and point defect kinetics model calculation. MD simulations of displacement cascades in pure Fe were performed with a 4.7 keV primary knock-on atom to obtain the parameters related to displacement cascades. Then, we employed the point defect kinetics model to calculate the concentration of the point defects. By combining the positron trapping rate from the PAS measurement and the calculated vacancy concentrations, the trapping coefficient for the vacancies in the RPV steels was determined, which was about 0.97 × 10¹⁵ s⁻¹. The application of two techniques, PAS and computer simulation, provided complementary information on radiation-induced defect production.