Reactor Pressure Vessel Embrittlement Research Articles

Cu precipitation behaviour in long-term thermally aged, high-copper model reactor pressure vessel steels

Nanoscale Cu-rich precipitates that were formed in Cu-containing reactor pressure vessel (RPV) steels during service have a deleterious effect on mechanical properties, which can result in RPV embrittlement and limit reactor operation life. To understand the nanoscale precipitation mechanism, thermal aging at 370°C for up to 13200 h of high-copper RPV model steels was performed to produce nanosized Cu-rich precipitates. Then the Cu-rich precipitates were systematically investigated by atom probe tomography (APT). The changes in the mechanical properties of the steels were characterized by Vickers hardness test. The results show that the Cu-rich precipitates as determined by APT construction analysis have a spatial core-shell structure. The core is enriched with Cu atoms, and the region near the precipitate/matrix interface is enriched with Mn, Ni, Mo, and Si and C atoms. Cu-rich precipitates lead to precipitation strengthening and hardening/embrittlement effects. Using the Russell-Brown model, we estimated the hardening due to Cu-rich precipitates in the matrix and observed that the measured hardness and the estimated changes in hardness were in good agreement.

Non-hardening embrittlement mechanism of pressure vessel steel Ni-Cr-Mo-V welds during thermal aging

The mechanical performance of reactor pressure vessel materials is an important factor in the safety and economics of the operation of a nuclear power plant. The ductile-to-brittle transition temperature tested by Charpy impact test is the key parameter for evaluating the reactor pressure vessel embrittlement. In this article, the study of thermal aging embrittlement of temperature sets of reactor pressure vessel surveillance Ni-Cr-Mo-V steel weld metal was conducted by Charpy impact test. The thermal aging effect on the impact fracture behavior was analyzed. The impact test of the three batches of weld surveillance sample indicated that the weld metal embrittled during thermal aging. The study of impact fracture and Auger electron spectroscopy indicated that the element P segregated to the grain boundaries and lowered their cohesion strength during the long-term thermal aging. Therefore, the non-hardening embrittlement of Ni-Cr-Mo-V steel welds in a reactor pressure vessel caused by segregation of impurity elements P occurs during thermal aging.

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.

Accelerated cluster dissolution using electropulsing for ultrafast performance regeneration

Precipitation of copper-rich nanoclusters is the chief culprit, which is the cause of embrittlement of reactor pressure vessels. Here we provide a real sense of ``in situ'' pathway to investigate ultrafast performance regeneration under electropulsing. We find that the pulsed electric current can reduce the dissolution temperature of nanoclusters. The number of nanoclusters can be decreased drastically at relatively low temperature (430 °C), and the process takes only 3 min. Meanwhile, the toughness recovered more than 70%. The homogenization annealing treatment without pulsed electric current (430 °C/3 min) shows that the microstructure and mechanical properties do not change substantially.

Breaking Thermal Diffusion Limit for Ultrafast Performance Regeneration Using External Field

Precipitation of copper-rich nanoclusters is the chief culprit, which is the cause of embrittlement of reactor pressure vessels. Here we provide a real sense of in situ pathway to investigate ultrafast performance regeneration under electropulsing. We find that the pulsed electric current can reduce the dissolution temperature of nanoclusters. The number of nanoclusters can be decreased drastically at relatively low temperature (430 ℃), and the process takes only 3 min. Meanwhile, the toughness recovered more than 70%. The homogenization annealing treatment without pulsed electric current (430 ℃/3 min) shows that the microstructure and mechanical properties do not change substantially.

Development of pressure-temperature operation limits for a PWR vessel considering beltline shell and extended beltline nozzles

The pressure-temperature limits (P-T limits) of embrittled reactor pressure vessels (RPVs) should be determined for normal heat-up and cool-down operations based on fracture mechanics to ensure the structural integrity. In general, the P-T limits are mainly determined by the fracture toughness of beltline region material which suffers the highest level of neutron embrittlement. However, some other components such as nozzles which suffer insignificant neutron embrittlement may restrict the P-T limits as a result of higher stress concentration from their structural discontinuities. Therefore, beltline region material with the highest reference temperature as well as nozzles that have structural discontinuities need to be considered when developing the P-T limits of RPV. In this paper, the P-T limit calculations for a Taiwan domestic pressurized water reactor (PWR) vessel were performed considering both beltline shell and extended beltline nozzles according to the nonmandatory Appendix G to ASME Code Section XI. The three-dimensional finite element analyses were conducted to obtain the pressure and thermal stress distributions of extended beltline nozzles for P-T limits calculation. It is found that the beltline region still dominates the cool-down P-T limit, but the extended beltline region will partially control the heat-up limit. Further, the reference temperature ranges of extended beltline nozzles that may influence the P-T limits of RPV were also investigated. Present study is helpful for safe operation and also provides a reference for regulation of PWR plants in Taiwan.

A model to evaluate the hardening effect of solute clusters in Fe-based alloys

Solute cluster induced hardening plays a dominant role in the irradiation embrittlement of reactor pressure vessel. For Cu-rich clusters, the Russell and Brown model (1972) is often applied for their correlation with hardening, but many parameters of the model are taken as adjustable to fit specific experimental data. In this work, a correlation is attempted and discussed between Cu-rich clusters and hardening in a series of reactor pressure vessel model alloys. Compared with the Cu clusters of the same size, Cu-rich clusters containing Ni and Mn may have lower obstacle strength and cause lower hardening.

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.

Irradiation effect on impact fracture behavior of A508-3 steel in ductile-to-brittle transition range

Abstract The mechanical performance of reactor pressure vessel (RPV) materials is an important factor in determining the safety and economics of the operation of a nuclear power plant. The ductile-to-brittle transition temperature (DBTT) tested by Charpy impact test is an important parameter for evaluating the RPV embrittlement. In this paper, the Charpy impact test of RPV steel prior and after irradiation was carried out at different temperatures. The fracture morphology was observed by scanning electron microscopy. The irradiation effect on the impact fracture behavior was analyzed in the DBTT temperature zone. Results show that in the DBTT range, the crack initiation and propagation energy are obviously reduced for the irradiated material. After irradiation, the dimple area on the fracture surface is significantly reduced, and the distance from the crack initiation source to the gap is decreased. For the RPV steel with low Cu steel content, matrix damage becomes the main factor on material irradiation embrittlement.

Modular system for probabilistic fracture mechanics analysis of embrittled reactor pressure vessels in the Grizzly code

Abstract In light water reactor (LWR) nuclear power plants, the reactor pressure vessel (RPV) plays an essential safety role, and its integrity must be ensured during a variety of transient loading conditions. These can include off-normal conditions such as a pressurized thermal shock (PTS), as well as transients encountered during normal startup, shutdown, and testing of the reactor. Exposure to irradiation and elevated temperatures embrittles the RPV’s steel over time, making it increasingly susceptible to failure due to propagation of fractures that could initiate at the locations of flaws introduced during the manufacturing process. As long-term operation scenarios are being considered for LWRs in the United States, it is important to have a flexible simulation tool that can be used to perform probabilistic evaluations of RPV integrity under a wide variety of conditions and incorporate improved predictive models of RPV steel embrittlement. The Grizzly code is being developed to meet these needs. This paper describes Grizzly’s modular architecture, provides results of benchmarking studies of various components of Grizzly, and demonstrates the application of Grizzly on a model that includes plume effects that are difficult to represent in other codes being used in current practice.

Reactor pressure vessel embrittlement: Insights from neural network modelling

Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

Effects of neutron irradiation on resistivity of reactor pressure vessel steel

The embrittlement of reactor pressure vessel (RPV) steel owing to fast-neutron irradiation is one of its primary failure mechanisms. In this work, neutron irradiation tests were performed on an RPV steel at a high temperature (565 K) using a neutron irradiation test reactor. In addition, resistivity measurements were performed on the RPV steel both before and after irradiation in a hot laboratory using the four-probe method. The results showed that the resistivity of the RPV steel exhibits nonlinear behaviour with respect to the radiation fluence and that the nonlinearity becomes more pronounced with an increase in the radiation fluence. For instance, when the radiation fluence is 0.1540 dpa and the excitation current is increased from 0.2 mA to 200 mA, the resistivity of the RPV steel decreases by as much as 67.12%. During irradiation embrittlement, the resistivity increases with the fluence. When the radiation fluence is greater than 0.116 dpa, the increase in the resistivity accelerates. When the radiation fluence is less than 0.116 dpa and when an excitation current of 2 mA or 20 mA is used, the relationship between the resistivity and the radiation fluence for the RPV steel is a quadratic one, whereas that between the rate of change in the resistivity and the radiation fluence is a linear one. Thus, the resistivity of RPV steel can be used to characterise its degree of irradiation embrittlement, and resistivity measurements can be employed as a nondestructive evaluation technique for monitoring the degree of irradiation damage experienced by in-service RPV steel.

Thermodynamic and kinetic modeling of Mn-Ni-Si precipitates in low-Cu reactor pressure vessel steels

Formation of large volume fractions of Mn-Ni-Si precipitates (MNSPs) causes excess irradiation embrittlement of reactor pressure vessel (RPV) steels at high, extended-life fluences. Thus, a new and unique, semi-empirical cluster dynamics model was developed to study the evolution of MNSPs in low-Cu RPV steels. The model is based on CALPHAD thermodynamics and radiation enhanced diffusion kinetics. The thermodynamics dictates the compositional and temperature dependence of the free energy reductions that drive precipitation. The model treats both homogeneous and heterogeneous nucleation, where the latter occurs on cascade damage, like dislocation loops. The model has only four adjustable parameters that were fit to an atom probe tomography (APT) database. The model predictions are in semi-quantitative agreement with systematic Mn, Ni and Si composition variations in alloys characterized by APT, including a sensitivity to local tip-to-tip variations even in the same steel. The model predicts that heterogeneous nucleation plays a critical role in MNSP formation in lower alloy Ni contents. Single variable assessments of compositional effects show that Ni plays a dominant role, while even small variations in irradiation temperature can have a large effect on the MNSP evolution. Within typical RPV steel ranges, Mn and Si have smaller effects. The delayed but then rapid growth of MNSPs to large volume fractions at high fluence is well predicted by the model. For purposes of illustration, the effect of MNSPs on transition temperature shifts are presented based on well-established microstructure-property and property-property models.

Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

Abstract The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

Positron Annihilation and TEM Characterization of Cu-Enriched Clusters in the Ferritic Steels Containing Copper

Nanosized Cu-enriched clusters formed in Cu-containing reactor pressure vessel (RPV) steels during service have a deleterious effect on mechanical properties, which can result in RPV embrittlement and limit reactor operation life. To understand the effect of Cu-enriched clusters (CECs) behavior on mechanical properties, thermal aging at 370°C for times up to 13200 h was performed on the high-copper ferritic steels. The microstructure evolution of CECs was investigated by positron annihilation spectroscopy (PAS) and transmission electron microscope (TEM). The results show that the CECs formed after aging times up to 3000 h, which composition is composed of Fe, Cu, Ni, Mn, and Si, are 9R structures. The CECs lead to precipitation hardening/embrittlement effects. The changes of hardness have a linear relationship with transition temperature shifts.

Mechanical properties and microstructure of long term thermal aged WWER 440 RPV steel

The integrity assessment of the Reactor Pressure Vessel (RPV) is essential for the safe and Long Term Operation (LTO) of a Nuclear Power Plant (NPP). Hardening and embrittlement of RPV caused by neutron irradiation and thermal ageing are main reasons for mechanical properties degradation during the operation of an NPP. The thermal ageing-induced degradation of RPV steels becomes more significant with extended operational lives of NPPs. Consequently, the evaluation of thermal ageing effects is important for the structural integrity assessments required for the lifetime extension of NPPs.As a part of NRG's research programme on Structural Materials for safe-LTO of Light Water Reactor (LWR) RPVs, WWER-440 surveillance specimens, which have been thermal aged for 27 years (∼200,000 h) at 290 °C in a surveillance channel of Armenian-NPP, are investigated. Results from the mechanical and microstructural examination of these thermal aged specimens are presented in this article.The results indicate the absence of significant long term thermal ageing effect of 15Cr2MoV-A steel. No age hardening was detected in aged tensile specimens compared with the as-received condition. There is no difference between the impact properties of as-received and thermal aged weld metals. The upper shelf energy of the aged steel remains the same as for the as-received material at a rather high level of about 120 J. The T41 value did not change and was found to be about 10 °C. The microstructure of thermal aged weld, consisting carbides, carbonitrides and manganese-silicon inclusions, did not change significantly compared to as-received state. Grain-boundary segregation of phosphorus in long term aged weld is not significant either which has been confirmed by the absence of intergranular fracture increase in the weld. Negligible hardening and embrittlement observed after such long term thermal ageing is attributed to the optimum chemical composition of 15Cr2MoV-A for high thermal stability.

Calculation of neutron and gamma fluences on VVER reactor pressure vessel

Embrittlement is one of the most important effects affecting reactor pressure vessel (RPV) aging. RPV is irradiated with neutrons and gammas, especially fast neutrons, which mainly lead to embrittlement of RPV during operation lifetime of nuclear reactors. Therefore, the radiation-induced embrittlement of the RPV should be carefully evaluated. In this paper, a preliminary calculation was performed using the MCNP5 code to identify the areas in the RPV of the VVER-1000/V320 reactor where the neutron and gamma fluxes are maximum. Also, the neutron and gamma fluence distributions on the RPV were investigated and evaluated along with their energy spectra. These calculations are the starting point for the evaluation of radiation damage to RPV of VVER reactors.

Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (TK) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in TK shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the TK shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime.

Crystal plasticity modeling of irradiation effects on flow stress in pure-iron and iron-copper alloys

The mechanistic modeling of irradiation induced embrittlement of reactor pressure vessel steels strongly depends on the precise evaluation of flow stress behavior. This requires accurate characterization of change in both the yield strength as well as the strain-hardening capacity. A dislocation-density based crystal plasticity model is thus developed in this work to quantify these variations with irradiation. The model considers the interaction between dislocations and irradiation induced defects such as self-interstitial atomic loops, vacancy clusters and precipitates to obtain flow stress variations in irradiated ferritic alloys. The model is calibrated and validated for polycrystalline pure-iron and iron-copper alloys, neutron-irradiated to different dose levels under typical pressure vessel operating conditions. A comparison with experimental results show that the model is able to quantify the changes in flow stress behavior accurately. At 0.2 dpa a loss of strain-hardening capacity beyond 2% strain is also obtained from the model. The yield strength increase with irradiation obtained from the model is also compared with analytical strengthening models based on Orowan’s equation.

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.