Recent experience in pressure vessel materials irradiation

The reactor pressure vessel has been repeatedly cited as a primary concern in assessment of pressure boundary structural integrity and in planning for plant life extension programs. The life of the reactor pressure vessel will be limited by radiation-induced embrittlement; this is monitored in Westinghouse designed nuclear steam supply systems by testing samples of base metal, heat-effect-zone and weld metal in the form of Charpy V-notch, tensile, and fracture mechanics specimens which have been irradiated in surveillance capsules adjacent to the wall. The earliest reactor vessel material radiation surveillance program was the Yankee Rowe program which started in 1961. As data became available from power reactor surveillance and test reactor programs, estimates of radiation-induced changes in mechanical properties were predicted in the form of radiation damage trend curves which provide methods for calculating numerical estimates of changes in mechanical properties as a function of chemistry and fluence. For example, the proposed Revision 2 to Regulatory Guide 1.99 provides estimates of shifts in transition temperature as a function of copper and nickel content and fluence. Slight variations in chemical analyses for copper and/or nickel can result in limitation on heat-up and cool-down rates or compliance with regulatory rules, such as the PTS screening criteria. Automatic submerged arc welding was employed in the fabrication of reactor vessels in Westinghouse designed nuclear steam supply systems. The type of flux material utilized in the welding process is important because mechanical properties can differ depending upon what flux is used. This paper correlates the results from over 50 surveillance capsules with the welding practice and concludes that radiation damage trend curves can be developed for welding practice. By using trend curves based on welding practices, discrepancies in chemical analyses are eliminated and credibility is restored to structural integrity assessments.     

Fluence rate effects on irradiation embrittlement of model alloys

The Reactor Pressure Vessel (RPV) material of Nuclear Power Plants (NPP) is exposed to neutron irradiation during its operation. Such exposure generally induces degradation of the mechanical and physical properties of the materials: e.g. an increase of the ductile to brittle transition temperature (DBTT) and a decrease of the upper shelf impact energy.At a given irradiation temperature, dose and neutron spectrum, the sensitivity of materials to neutron irradiation depends on their chemical composition. In particular, elements like phosphorus, P, copper, Cu, and nickel, Ni play a key role in RPV steels.The effect of fluence rate on irradiation embrittlement of RPV materials is also a key issue for the correct interpretation of accelerated data and surveillance data in view of reactor pressure vessel life assessment of nuclear reactors. Much effort was done in the last decades to tackle such issues and quite contradictory results have been obtained.Model alloys can successfully be used to study embrittlement mechanisms and the effect of fluence rate. A parametric study of the response to neutron irradiation of 32 different model alloys with systematic 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%) was completed by some members of the European Network AMES. The irradiation of the 32 model alloys took place in the LYRA rig at the High Flux Reactor (HFR) of the Joint Research Centre, Petten, The Netherlands.Some model alloys were also irradiated in commercial reactors, namely in Rovno Nuclear Power Plant (NPP), Ukraine, and Kola NPP in Russia. Data available on these model alloys are presented and analysed in this paper, proving to be very important for the study of fluence rate effect.     

Master curve techniques to evaluate an irradiation embrittlement of nuclear reactor pressure vessels for a long-term operation

Irradiation embrittlement is a limiting condition for the long-term safety of a nuclear reactor pressure vessel (RPV). The first PWR in Korea is approaching its initial licensing life of 30 years. In order to operate the reactor for another 10 years and more, it should be demonstrated that the irradiation embrittlement of the reactor will be adequately managed by ensuring that the fracture toughness properties are above a certain level of the required safety margin. The RPV was designed by an old construction code and its beltline circumferential welds have suffered from an irradiation shift problem like other Linde 80 welds. The master curve method is considered as the most promising tool to characterize irradiated structural steels by using a fracture mechanics basis. In order to implement the master curve method for the assessment of an irradiation embrittlement of old power reactors for a continued long-term operation, three practical issues were emphasized in this investigation, which are the specimen geometry effects on the master curve results, the specimen reconstitution techniques in an old existing surveillance program, and the index temperatures of an irradiation embrittlement when compared with the conventional Charpy data.     

Pressure vessel steels: influence of chemical composition on irradiation sensitivity

Neutron irradiation of steels used in the construction of nuclear reactor pressure vessels can lead to the embrittlement of these materials, i.e. increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of nuclear power plants.

Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate) and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, i.e. the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field over the last 25 years.     

Annealing for plant life management: Hardness, tensile and Charpy toughness properties of irradiated, annealed and re-irradiated mock-up low alloy nuclear pressure vessel steel

Hardness, tensile and Charpy properties of an irradiated (I) and irradiated-annealed-reirradiated (IAR) mock-up pressure vessel steel are presented. Spectrum tailored pressurized light water reactor (PWR) irradiation at 290°C by fast neutrons up to nominal fluences of 5 × 10¹⁹/cm² (E ≥ 1 MeV) in a swimming pool type reactor caused the hardness, tensile yield stress and tensile strength to increase. Embrittlement also occurred as indicated by Charpy toughness tests.

The optimum annealing heat treatment for the main program was determined using isochronal and isothermal runs on the material and measuring the Vickers microhardness. The response to an intermediate annealing treatment (460°C for 18 h), when 50% of the target fluence had been reached and then irradiating to the required end fluence (IAR condition) was then monitored further by Charpy and tensile mechanical properties. Annealing was beneficial in mitigating overall hardening or embrittlement effects. The rate of re-embrittlement after annealing and re-irradiating was no faster than when no annealing had been perfomed. Annealing temperatures below 440°C were indicated as requiring relatively long times, i.e. ≥168 h to achieve some reduction in radiation induced hardness for example.

The overall benefit of an appropriate annealing treatment has been demonstrated on a laboratory scale, and annealing is therefore indicated as being a useful tool in the overall strategy for achieving plant life management goals.     

A determination of the benefits of annealing irradiated pressure vessel weldments

The long-term benefit of annealing an irradiated reactor pressure vessel steel may be described in terms of a benefit factor, B. The benefit factor compares the mechanical properties of an annealed and reirradiated specimen with an equivalent specimen having no intermediate anneal. The benefit factor was determined using a series of microhardness specimens prepared from nuclear pressure vessel surveillance program materials. These specimens were annealed and then reirradiated in a test reactor. There was an obvious long-term benefit in the specimens annealed at 450°C. The long-term benefit was less obvious at 400°C and no significant benefit was noted at 350°C. The benefit factor may also be used as the basis of a surveillance program for an annealed pressure vessel. A strategy for such a surveillance program is described.     

A preliminary evaluation of irradiation damage in model alloys by electric properties based techniques

The most important effect of the degradation by neutron irradiation is a decrease in the ductility of reactor pressure vessel (RPV) ferritic steels. The main way to determine the mechanical behaviour of the RPV steels is tensile and impact tests, from which the ductile to brittle transition temperature and its increase due to neutron irradiation can be calculated. These tests are destructive and are regularly applied to surveillance specimens to assess the integrity of the RPV. The possibility of applying validated non-destructive aging monitoring techniques would however facilitate the surveillance of the materials that form the reactor vessel.The Institute for Energy of the Joint Research Centre has developed two devices, focussed on the measurement of the electrical properties which prove to give a good non-destructive assessment of the embrittlement state of ferritic steels. The first technique, called Seebeck and Thomson Effects on Aged Material (STEAM), is based on the measurement of the Seebeck coefficient, characteristic of the material and related to the microstructural changes induced by irradiation embrittlement. With the same aim, the second technique, named Resistivity Effects on Aged Material (REAM), measures instead the resistivity of the material.This paper explains (i) preliminary STEAM and REAM results and (ii) results compared with Charpy impact energy temperature shifts due to neutron irradiation. These results will make possible the improvement of such techniques based on the measurement of material electrical properties for their application to non-destructive irradiation embrittlement assessment.     

Isothermal aging test results (up to 100 000 h) of NiCrMoV steels for low-pressure steam turbine

Isothermal aging test of NiCrMoV rotor steel was carried out up to 100 000 h and change in the Charpy transition behavior was investigated. The test result revealed considerable embrittlement of the steels tested. Even at 343°C, the embrittlement of around 100°C was observed and higher temperature enhanced the embrittlement significantly. This embrittlement behavior strongly depends on the impurity contents of the materials and temper embrittlement parameter J factor or X well characterize the temper embrittlement susceptibility. Based on these test results, the amount of the temper embrittlement can be estimated from the information of chemistry. A correlation between the step cooling embrittlement and 100 000 h embrittlement was also found.     

Views of TAGSI on the effects of gamma irradiation on the mechanical properties of irradiated ferritic steel reactor pressure vessels

The paper reviews and analyses the effects of gamma irradiation dose on the properties of ferritic steels used in reactor pressure vessels (RPVs). It explains factors that affect the embrittlement of a RPV steel induced by combinations of fast neutrons, thermal neutrons, and gamma irradiation.

TAGSI were asked to consider the effects of gamma irradiation dose on the properties of steels used in reactor pressure vessels. TAGSI endorsed the use of the MCBEND code to calculate gamma fluxes and energetic gamma ray displacement cross-sections calculated using either Baumann or Alexander methods. TAGSI endorsed the calculation of the materials property changes due to an additional gamma dose using trend curves based on empirical correlation to neutron-induced damage (where k_γ1±0.25).     

The role of reverse temper embrittlement on some low and high temperature crack extension processes in low carbon, low alloy steels: A review

The present paper attempts to review the influence or role that reverse temper embrittlement (RTE) plays on the low temperature (<400°C) environmentally assisted cracking (EAC) processes in aqueous environments and the high temperature ( > 400°C) creep dominated crack growth processes prevalent in low carbon, low alloy steels. RTE is a generic problem in iron based alloys and causes a marked reduction in the cohesion strength of grain boundaries due to the segregation of impurity solute atoms.

It has been shown that there is a strong possibility that RTE can promote or enhance EAC in low alloy steels (the heat affected zone in weldments being the most susceptible) subjected to aqueous environments at temperatures below about 350°C.

At higher operational temperatures, viz. above 400°C, the characteristics of creep embrittlement are discussed and three microstructural features which contributed to this phenomena are highlighted. Proposals aimed at mitigating creep embrittlement processes in low alloy steels are forwarded.

In certain steels it was recorded that two embrittlement processes were prevalent at operational temperatures in the range 450–550°C, viz. RTE and an irreversible creep embrittlement which was prevalent when large deformation occurred in service. Indeed both 2·25 Cr1Mo and Cr-Mo-V steels were severely temper embrittled at service temperatures of 430°C while the Cr-Mo-V steels did not exhibit evidence of RTE at higher service temperatures of around 530°C. However, the effects of RTE in promoting creep embrittlement are, as yet, unclear.     

A new approach to estimate irradiation embrittlement of pressure vessel steels

The idea of an engineering version of the local approach to brittle fracture as well as the possibility for it to be employmed to estimate irradiation embrittlement of pressure vessel (PV) steels is considered. Unlike the conventional temperature-shift-based methodology, the approach presented utilises the concept of stability of the ductile state of the metal. A new characteristic “parameter of mechanical stability” P_ms is proposed. This characteristic enables quantification of the level of stability of the ductile state of an irradiated PV steel in a specimen or in a reactor vessel with a crack at the specified level of loading. Within the framework of the proposed concept, a value for end-of-life fluence for a reactor PV is predicted by the condition of exhaustion of stability of a ductile state of a steel ahead of the crack (P_ms=1).     

Copper and phosphorus effect on residual embrittlement of irradiated model alloys and RPV steels after annealing

The dependence of the recovery of the transition temperature shift after annealing (475 °C, 100 h) on copper and phosphorus contents has been studied on irradiated reactor pressure vessel (RPV) materials. A set of model alloys with low nickel content, lower than 0.2 mass%, was used for the study. Copper and phosphorus contents were varied in a wide range: 0.005–0.99 and 0.002–0.039 mass%, respectively.     

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.     

The effect of nickel,manganese and copper on the irradiation sensitivity of low alloy steel welds

It was observed by LM Davies that ‘Eastern’ and ‘Western’ reactor pressure vessel steels have similar spans of irradiation sensitivity. In this paper the similarities and differences between these steels are explored by comparison between published data on VVER steels, and data and models developed on Western type welds with similar nickel contents to the VVER-440 and VVER-1000 steels. This comparison requires that allowance be made for irradiation temperature, dose rate and manganese effects. This can only be done approximately; even so, the results show encouraging similarities in behaviour between the Eastern and Western categories of steel.     

Influence of irradiation temperature on microstructure of EU batch of ODS Eurofer97 steel irradiated with neutrons

Abstract

A TEM investigation of an EU batch of oxide dispersed strengthened (ODS) Eurofer97 steel specimens, irradiated to 1 and 3 dpa at 300, 450 and 550°C in high flux reactor at Petten, has been performed to understand the influence of irradiation temperature on the characteristics of irradiation defects and, eventually, on the resulting mechanical properties of this material. Specimens irradiated at 300°C revealed the presence of a high density of black dot damage and small self-interstitial atom (SIA) dislocation loops causing substantial hardening and embrittlement. In contrast, negligible black dot damage, low density of large SIA loops and networks of dislocations are observed in specimens irradiated at 450 and 550°C. The lath martensitic structure and ODS particles remain unaffected after irradiation in all specimens. These results are discussed in view of possible activation of defect annihilation mechanisms to explain the observed recovery of mechanical properties at high irradiation temperatures.     

Catalytic conversion of fast pyrolysis oil to hydrocarbon fuels over HZSM-5 in a dual reactor system

The catalytic conversion of fast pyrolysis bio-oil to hydrocarbon fuels was studied over HZSM-5 at atmospheric pressure. Experiments were conducted in a dual reactor system having two reactors in series. The temperatures in these reactors were in the range 340–400°C (first reactor) and 350–450°C (second reactor). The bio-oil was co-processed with tetralin in all the runs. The objective was to maximize the organic distillate product with a high concentration of aromatic hydrocarbons. The maximum amount of organic distillate in the effluent from the second reactor was 21 wt% of the bio-oil feed and the highest concentration of aromatic hydrocarbons was 76 wt% of the distillate. The dual reactor system was particularly beneficial when the temperature in the first reactor was low. Thus, with the first reactor at 340°C, the yields of organic distillate and aromatic hydrocarbons were 15–16 wt% and 8–11 wt% of wood, respectively, which are nearly two-fold compared to those from a single reactor system operated at 340°C (7.8 wt% and 4.8 wt%). Under the above conditions, the coke plus char yields were 25–26 wt% of wood which are up to 10 wt% lower than from the single reactor system at 340°C (29 wt%).     

Radiation embrittlement modelling for reactor pressure vessel steels: I. Brittle fracture toughness prediction

Modelling for the irradiation effect on brittle fracture toughness of reactor pressure vessel (RPV) steel is performed on the basis of the probabilistic model for fracture toughness prediction proposed by the authors earlier. The irradiation effect on parameters controlling plastic deformation and brittle fracture of RPV steels is analyzed. The physical mechanisms are considered which control the cleavage microcrack nucleation for RPV steels in the unirradiated and irradiated states and also in state after post-irradiation annealing. Prediction of the temperature dependence of brittle fracture toughness is performed as applied to irradiated 2.5Cr–Mo–V reactor pressure vessel steel. Modelling of the fluence effect and the phosphorus and copper content effect on brittle fracture toughness is carried out. It is shown that the probabilistic model based on a new formulation for brittle fracture criterion allows the adequate modelling for the irradiation effect on fracture toughness for RPV steel. Application of alternative models is discussed for fracture toughness prediction for irradiated RPV steels.     

Corrosion study of nickel for alkaline water electrolysis

A study of the behaviour of nickel in the conditions of advanced electrolysis (130–180°C—40% KOH) shows the detrimental effect of oxygen in general corrosion. Corrosion can be limited by restricting sulfur content in nickel. No stress corrosion cracking is observed with nickel in KOH. Nickel appears to be more sensitive to cathodic hydrogen embrittlement than to gaseous hydrogen embrittlement.     

Investigation of embrittlement of step cooling heat treatment on pressure vessel steel

Abstract

The effects of thermal aging and step cooling embrittlement on the impact toughness of a reactor pressure vessel steel SA533B quenched and tempered (QT) with and without post-weld heat treatment (PWHT) have been studied. Charpy impact testings were conducted on the aged plates at 350°C for 5000 h to evaluate whether the embrittlement was induced by step cooling heat treatment. The results show that thermal aging increases the ductile–brittle transition temperature in both QT and PWHT states but dramatically decreases the upper shelf energy in QT state and has less effects on the PWHT state. By comparing the correlation between thermal aging embrittlement and step cooling embrittlement for both QT and PWHT states in steel, it is found that the step cooling heat treatment can obviously promote further embrittlement of the base metal in QT state but has little influence on the impact toughness in PWHT and thermal aged state. Further analysis indicates that the step cooling heat treatment cannot promote steel embrittlement at some heat treatment states. Finally, a new method is proposed to evaluate the degree of step cooling embrittlement of the pressure vessel steel.     

On the ΔT41 criterion for irradiation shift

Current methodologies for assessing the effects of neutron irradiation on the ductile to brittle transition in reactor pressure vessel steels include the use of the Charpy-based ΔT₄₁ irradiation shift criterion. This criterion is commonly used in the assessment of irradiation shift for setting limits to the pressure/temperature operating diagram and in defect assessment methodologies. In both of these cases the Charpy-based temperature shift is assumed to produce an equivalent shift in the fracture toughness/temperature curve. Based on the known behaviour of steels, this study shows that this assumption may not be justified when it is applied to static fracture toughness measurements. The results of this analysis indicate that the Charpy-based shift will underestimate the shift in fracture toughness by an amount proportional to the hardening produced by irradiation. Application of the analysis to the USA database on plate and forging materials shows that the experimentally measured differences between Charpy-based and fracture toughness-bassed temperature shifts are consistent with the anticipated effects of irradiation hardening on fracture behaviour. A yield stress augmented Charpy-based shift criterion improves the equivalence in shift values as determined by the Charpy-based and fracture toughness-based approaches. The implications for the use of the current ΔT₄₁ criterion in safety assessments are discussed.