Commercial Reactor Pressure Vessel Research Articles

Effect of non-clustering element C and Mo on irradiation-induced clustering in reactor pressure vessel steels

Commercial reactor pressure vessel steels in nuclear power plant applications contain C and other minor non-clustering elements in addition to the clustering elements of Mn, Ni and Si. However, the effect of such non-clustering elements on irradiation-induced solute clustering in ferritic steels has seldom been addressed. Herein, we reveal that a commercial RPV steel and its Fe-Mn-Ni-Si model alloy with coarse/fine grains subjected to ion irradiation exhibits different irradiation-induced clustering, particularly in chemistry, mean size, and number density of irradiation-induced MnNiSi-clusters. The different chemical compositions of the clusters indicate different thermodynamics in driving solute clustering in RPV steel and its model alloy. The different number density and size evolutions of the clusters indicate different kinetics in governing irradiation-induced clustering. The presence of C and Mo, as non-clustering elements, is primarily responsible for the different clustering behaviors of Ni, Mn and Si in the two materials exposed to ion irradiation. The research finding is of significant implication for modelling and fundamental understanding irradiation-induced clustering in commercial RPV steels.

Effects of Neutron, Ion and Proton Irradiation on Nano-Indentation Hardness of RPV Steels

The irradiation embrittlement damage of reactor pressure vessel (RPV) steel is one of its primary failure mechanisms. In this work, neutron, ion and proton irradiation experiments were carried on the same commercial RPV steels with the same irradiation fluence under the same temperature of 292°C. Then the nano-indentation hardness tests were performed on the RPV steel before and after irradiation. The results show that the irradiation hardening effects are observed by means of nano-indentation technique under the above three irradiations, and the hardening features are basically the same. While the max variation and increase rate are obviously different between those irradiations. It is found that the main reason of the above differences are caused by different energies of irradiation energetic particles, resulting in different types and quantities of defects. The conclusions in this paper are helpful to select and compare different irradiation experiments to the research of RPV steels irradiation embrittlement damage.

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.

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.

Effects of Microstructure Variation on Tensile and Charpy Impact Properties in Heavy-Section SA508 Gr.3 Low Alloy Steels for Commercial Reactor Pressure Vessel

Effects of Microstructure Variation on Tensile and Charpy Impact Properties in Heavy-Section SA508 Gr.3 Low Alloy Steels for Commercial Reactor Pressure Vessel

Electrical resistivity measurement of Fe-0.6%Cu alloy irradiated by neutrons at 14–19 K

Electrical resistivity measurement is a useful experimental method for investigating the recovery of defects that are induced by irradiation in metals and alloys. In this study, an Fe-0.6%Cu alloy, used to model steel from old commercial reactor pressure vessels, was irradiated by neutrons at a low temperature range of 14–19 K with a dose of about 1.3 × 1020 neutrons/m2 (E > 0.1 MeV) in the Kyoto University Reactor (KUR); electrical resistivity measurement was performed during irradiation and after annealing of the irradiated sample from 20 K to 300 K to investigate the migration of point defects in the Fe-0.6%Cu alloy. The electrical resistivity was measured at 14–19 K. With the increase in the irradiation dose, the electrical resistivity increased linearly. Four peaks appeared at 70 K, 100 K, 150 K, and 260 K, in the change of electrical resistivity during annealing of the irradiated sample up to 300 K. The former two peaks were caused by the recombination of interstitials and vacancies, and the latter two peaks were caused by the formation of interstitial clusters and the migration of vacancies. Compared with previous electron irradiation results, the former two peaks represent new data, as does the ratio of recombination caused by close-pair and correlation to that caused by migrations of mixed-interstitials Fe–Cu and vacancies decreased in neutron irradiation.

Microstructural Evolution of Nuclear Power Steel A508-III in the Creep Process at 800°C

The A508-III steel is widely used to manufacture the lower heads of commercial reactor pressure vessels (RPV). In severe accident, the reactor core in the RPV begins to melt and meanwhile the technology of in-vessel retention (IVR) exerts its role. In this case the inner surface of RPV will expose to temperatures over a phase transition temperature. However, the significant nonlinear feature of creep curve of A508-III steel suffered heterogeneous damage was not studied. In this work, the creep tests were performed for the steel at the phase transition temperature of 800°C. The microstructural evolution at different creep stages was characterized by scanning electron microscopy and transmission electron microscopy. The results show that, at the second creep stage, more coarsening second phase particles occur in the steel. With the creep processing, the grain size and diameter of second phase particles increase. At the tertiary creep stage, the grain size increases significantly, and the second phase particles coarsen during the process of atom migration. In addition, Micro-cracks and voids also come into being in the situation and they can become larger by combing each other during the creep process. At this stage, the growth of cavities and second phase particles coarsening become the main mechanism of creep damage. The trend of microstructural evolution is consistent with the creep constitutive equation obtained for the A508-III steel at the phase transition temperature of 800°C. The results obtained provide indispensable foundation to establish the relationship between the macroscopic creep and microscopic damage.

Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

Characterization of transition behavior in SA508 Gr.4N Ni–Cr–Mo low alloy steels with microstructural alteration by Ni and Cr contents

Abstract SA508 Gr.4N Ni–Cr–Mo low alloy steel, in which Ni and Cr contents are higher than in commercial reactor pressure vessel (RPV) steels, may be a candidate RPV material with the improved strength and toughness due to its tempered martensitic microstructure. This study aims at assessing the effects of microstructural factors with alloying element contents on the transition properties of Ni–Cr–Mo low alloy steels. Model alloys with different Ni and Cr contents were fabricated and their Charpy impact toughness and fracture toughness were examined in the transition region according to ASTM E23 and E1921 standard procedures, respectively. The test results showed extensive improvements of both impact toughness and fracture toughness with increasing Ni and Cr contents. However, Ni content was more effective on the impact toughness than on the fracture toughness, while Cr content was more effective on the fracture toughness. In order to identify a difference in effects of alloying elements contents on the fracture toughness and impact toughness, the relations between the transition properties and the scale of the microstructural features such as packets and carbides are discussed in detail.

The increase of irradiation-induced yield strength in model alloys and its correlation with transition temperature shifts and similitude to reactor pressure vessel materials

A series of model alloys are studied to understand irradiation embrittlement phenomena and similitude in behaviours with commercial Reactor Pressure Vessel (RPV) materials. Important results and prediction models for a Ductile-to-Brittle Transition Temperature (DBTT) shift are drawn from the data. Some of the issues have always been the lack of tensile properties and the impossibility to establish a correlation between the transition temperature and the hardening of the alloys. Recently, within the FRAME project, tensile properties have been measured on both fresh and irradiated specimens. It is these new data that are analysed in this paper to investigate the irradiation-induced yield strength increases of the model alloys and compare them with existing data from RPV steels.

Mechanical Property Changes in Reactor Vessel Materials Thermally Aged for 209 000 H at 260°C

Abstract Surveillance materials consisting of a SA-508 Class 2 forging, a Mn–Mo–Ni Linde 80 submerged-arc weld, and an SA-533, Grade B, Class 1, correlation monitor material were thermally aged on a commercial reactor pressure vessel. The materials were exposed to a thermal environment of 260°C for 209000h. This temperature is below the range (minimum of 370°C) where the effects of long-term thermal aging are typically considered relevant. Charpy impact, master curve transition temperature, upper-shelf fracture toughness, and tensile testing were conducted to evaluate the long-term thermal aging changes in material properties. Small changes in the impact properties were observed for all the materials, but were generally within the 95 % confidence bounds for typical Charpy data. Upper-shelf energy also showed small variations, but a general decrease for all materials was not seen. Fracture toughness testing at the upper shelf indicated that the upper-shelf toughness had increased, however the data is scattered. Master curve T0 testing in the transition region showed little change in the forging and plate results; however an improvement in the transition temperature of the weld metal was measured.

The Effect of Neutron Flux on Radiation-Induced Embrittlement in Reactor Pressure Vessel Steels

Abstract The effect of neutron flux level is a longstanding concern for determining the applicability of test reactor data or high lead-factor surveillance data to the prediction of embrittlement in commercial reactor pressure vessels (RPV). However, as operating reactors reach higher fluences, the question of flux effects is becoming increasingly relevant for situations such as embrittlement attenuation through the (RPV) and the comparison of boiling and pressurized reactor RPVs. In spite of its technological importance and extensive experimental and theoretical investigation, the issue of neutron flux effects on radiation-induced embrittlement in RPV steels remains unresolved. For neutron fluxes much greater than RPV operating conditions (>100 times higher), a regime exists in which the effects of flux are well defined experimentally and are well predicted by current models. These same models predict that the effect of flux should be relatively weak for values near those obtained in commercial RPVs, but may increase at lower fluxes. However, it is difficult to obtain sufficient data at low to intermediate fluxes to unambiguously determine the effects of flux since the time required to reach the desired fluences is necessarily long. In order to demonstrate the potential effects of flux on RPV embrittlement, and to help interpret the available experiments, a kinetic embrittlement model has been used in an analysis of the primary variables involved. The results of the analysis are consistent with a modest effect of flux on embrittlement for fluxes near those typical of commercial RPVs.

Irradiation embrittlement of model alloys and commercial steels: analysis of similitude behaviors

The deep understanding of the irradiation embrittlement of the pressure vessel of nuclear reactors is a key issue for the plant lifetime assessment and life extension through mitigation methods like annealing and much effort have been done in the last decades to tackle such complex issue. The reactor pressure vessel (RPV) material of nuclear power plants is exposed to neutron irradiation during its operation. Such exposure is generally inducing a degradation of the mechanical and physical properties of the materials, e.g. an increase of the ductile to brittle, DBT, transition temperature and a decrease of the upper shelf energy. The different response of materials to neutron irradiation, even many factors are also playing significant role, is mainly due, for a given exposure, to the chemical composition of the materials. In particular, for the RPV steel, elements like phosphorus, P, copper, Cu, and nickel, Ni, are playing a key role. A parametric study of the response to neutron irradiation of 32 different model alloys with parametric variation of elements (Ni from 0.004 to ∼2 wt%, P from 0.001 to 0.039 wt%, Cu from 0.005 to ∼1 wt%) has been recently completed within the frame of the European Network AMES and EC-JRC AMES Institutional project [1]. Such study on model alloys reveals to be a fundamental tool to understand the individual role of each element and synergisms. To demonstrate the usefulness of the study to commercial RPV steels, an analysis of the results and the similitude of behavior between model alloys and available RPV commercial steels has been carried out and the results are presented in this paper.

Effect of neutron flux on low temperature irradiation embrittlement of reactor pressure vessel steel

Miniature Charpy V-notch impact test specimens of commercial reactor pressure vessel (RPV) steels having high and low copper contents were irradiated at the different irradiation positions with neutron flux levels of ∼6 × 10 14, ∼7 × 10 15, and ∼8 × 10 16 n m −2 s −1 ( E > 1 MeV) to fluence levels ranging from ∼6 × 10 21 to ∼7 × 10 22 n m −2 ( E > 1 MeV) at temperatures of about 50°C to 150°C in the Japan Materials Testing Reactor (JMTR). The results showed that the radiation-induced increases in ductile-to-brittle transition temperature (ΔDBTT) at a neutron flux level of ∼6 × 10 14 n m −2s −1 were greater than those for neutron flux level of ∼7 × 10 15 n m −2 s −1. The neutron flux effect on embrittlement tended to be more pronounced in the lower neutron fluence range of ∼6 × 10 21–∼1 × 10 22 n m −2 than in the higher fluence level of ∼7 × 10 22 n m −2, and also to be larger for the low copper steel than for the high copper steel, although the ΔDBTT for the high copper steel was larger than that for the low copper steel regardless of neutron fluence or flux. The displacement dose rate effect identified by the data converted to the ΔDBTT for the full size Charpy specimens from those for the miniature Charpy specimens was consistent with that based on the comparison of the results in the literature.

Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: an atom probe study

Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally-aged (~100 000 h at 280°C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate and forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentration in the matrix after the stress relief heat treatment. The composition of cementite carbides aged for 100 000 h were compared to the Thermocalc™ prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.

High-temperature creep and tensile data for pressure vessel steels SA533B1 and SA508-CL2

Steel materials, SA533B1 and SA508CL-2, are widely used in the construction of commercial reactor pressure vessel lower heads. Previously, knowledge of the high temperature structural properties of SA533B1 was limited to temperatures below 922 K. In support of severe accident condition analyses, tensile and creep tests in the temperature range 900–1473 K were performed at the Idaho National Engineering Laboratory (INEL) under the US Nuclear Regulatory Commission Lower Head Failure Analysis Program. These creep and tensile test data are discussed and compared with previously developed lower temperature data for SA533B1 material. There is also additional discussion of SA508CL-2 creep data previously developed at the INEL.

Heavy-section steel irradiation program summary

In light-water-cooled nuclear power plants, a steel reactor pressure vessel (RPV) forms the primary containment boundary for the nuclear fuel assembly. Since a failure of the RPV carries the potential of major contamination release and severe accident, it is imperative to safe reactor operation to understand and be able to accurately predict realistic failure models of the vessel material. A major known risk arises from irradiation-induced degradation of the RPV fracture resistance which occurs as a result of normally occurring neutron flux during the service life of the reactor. Better understanding of the irradiation degradation of the vessel material will provide better input to improve the engineering models vital for prediction of RPV integrity and safe service life. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water RPVs. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness ( K 1c and J 1c), crack-arrest toughness ( K 1a), ductile tearing resistance (d J/da), Charpy V-notch impact energy, dropweight nil-ductility temperature (NDT), and tensile properties. Models based on observations of radiation-induced microstructural changes using field ion and high-resolution transmission electron microscopy provide a firmer basis for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI Program are highcopper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. In addition, a limited effort Will focus on stainless steel weld overlay cladding, typical of that used on the inner surface of RPVs, since its postirradiation fracture properties have the potential for strongly affecting the extension of small surface flaws during overcooling transients. Results from the HSSI studies will be integrated to aid in resolving major regulatory issues facing the U.S. Nuclear Regulatory Commission which involve RPV irradiation embrittlement such as pressurized-thermal shock, operating pressure-temperature limits, low-temperature overpressurization, and the specialized problems associated with low upper-shelf welds. Taken together the results of these studies also provide guidance and bases for evaluating both the aging behavior and the potential for plant life extension of light-water RPVs.

Effect of mechanical constraint on the fracture characteristics of thick-section steel

Operating limits for commercial reactor pressure vessels are often selected to take advantage of the brittle to ductile fracture toughness transition which has been observed to evolve above the NDT temperature for thin section, low alloy steels. Linear elastic fracture mechanics (LEFM) data obtained with thin sections and extrapolated to project thick section fracture behavior have raised concern that this transition to high ductility may be severely depressed or eliminated near reactor operating temperatures. Research programs at NRL emphasizing thick section dynamic tear (DT) studies and those at Westinghouse relating to thick section LEFM investigations have resolved this issue. These studies have shown that thick section steels of shelf toughness and strength class similar to A533-B continue to exhibit a significant toughness transition within a relatively narrow temperature inverval above NDT. Size-effect studies with the DT test indicate that the mechanical constraint associated with a 12-in thick section elevates the corresponding FTE temperature approximately 70°F (39°C) above that for thin sections. For the steels investigated the DT energy curve may be divided into three toughness regimes, ranging from frangible to ductile, which are shown to indicate similar transition features for specimens of 5 8 in. to 12 in thickness.