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.     

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.     

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.     

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).     

A comparison of Western and Eastern nuclear reactor pressure vessel steels

The task was essentially to compare the irradiation response of `East' and `West' steels. Since the plates and forgings of pressure vessels must be welded together, it is obvious that the strength requirements of the welds and heat affected zones (HAZ) can be no less demanding than those of the plates and the forgings themselves, particularly as experience has shown that the most likely location for flaws is in the welds or their HAZs. These and the highly stressed regions of the reactor pressure vessel (RPV) are important because neutron irradiation degrades the mechanical properties of steels.After comparing the various designs, manufacture and materials of the various RPVs, a comparison was made of the irradiation response of these different steels. The role of mitigating the change in mechanical properties on irradiation by thermal annealing was also considered.Particular codes/guides could only be used for the predicting results underpinning their own database because a major difference between these national codes/guides is that the elements conferring irradiation sensitivity are different for the two cases considered, i.e. Russian codes [1] (PNAE G-7-002-86) and the USNRC guide [2] (RG 1.99 Rev. 2). In the former, copper and phosphorus are significant, while copper and nickel are identified as significant in the latter case.Predictions were compared for `real' materials used in NPPPVs whose compositions were known. The irradiation response of these steels is coincidentally similar. The essential difference in behaviour is in the lifetime fluence. Eastern steels are irradiated to a much higher fluence than Western steels. Differences in the predictions of the Eastern–Western codes/guides are a reflection of differences in the concentration of deleterious elements and pessimisms of the various codes/guides, particularly at low concentrations of deleterious elements where they are most conservative. Thirdly, and on a `fitness for purpose' basis, the shift in transition temperature produces a limitation to the lifetime of the earlier Eastern RPVs. However, by thermally annealing the RPV to mitigate the effect of neutron irradiation, where the conditions to recover the mechanical properties of both Eastern and Western steels are nearly the same, the operational life of these older Eastern plants has been extended. Life assurance of these plants has, therefore, become practicable.This aspect of RPV technology, which is currently being considered in the US, could extend the operational life of nuclear power plants and thereby reduce the cost of the electricity generated.     

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.     

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.     

Characterization of high strength and high toughness Ni–Mo–Cr low alloy steels for nuclear application

The reactor pressure vessels of PWRs have mostly been made of SA508 Grade 3 (Class 1) low alloy steels which have revealed moderate mechanical properties and a moderate radiation resistance for a 40 or 60 year operation. The specified minimum yield strength of the material is 345 MPa with a ductile–brittle transition temperature of about 0 °C. While other materials, most of which are non-ferrous alloys or high alloyed steels for a higher temperature application, are being developed for the Generation-4 reactors, low alloy steels with a higher strength and toughness can help to increase the safety and economy of the advanced PWR systems which will be launched in the near future. The ASME specification for SA508 Grade 4N provides a way to increase both the strength and toughness by a chemistry modification, especially by increasing the Ni and Cr contents. However, a higher strength steel has a deficiency due to a lack of operating data for nuclear power plants. In this study, experimental heats of SA508 Grade 4N steels with different chemical compositions were characterized mechanically. The preliminary results for an irradiation embrittlement and the HAZ properties are discussed in addition to their superior baseline properties.     

Characterisation of as cast model Mn–Si–Cr–Ni steels for reactor pressure vessel applications

Abstract

Initial results are reported from a study aimed to investigate the role and influence of the elements Cr, Ni, Mn and Si on the radiation stability of reactor pressure vessel steels. Twelve as cast model ferritic steels with basic composition typical of those used in Russian WWER-1000 and Western PWR reactor pressure vessel materials were subjected to Charpy impact, magnetic Barkhausen noise (MBN), Vickers hardness tests and SEM examination. Higher Cr content in model steels was found generally to give increased RMS values independent of Mn and Si contents. The ductile–brittle transition temperatures (DBTT) and hardness values of the model steels were found to be independent of composition. Two steels, with low concentration of Ni and high concentration of Cr or vice versa , showed high transition temperatures (?16 and ?42°C respectively). An additional heat treatment to improve the properties is being considered for these compositions. The correlation between DBTT and MBN results has potential for rapid determination of the effect of composition and irradiation on the steel properties. The next stage of the assessment will investigate the effect of irradiation of the model steels to accumulated neutron fluences of ～10¹⁹ cm^?2.     

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.     

Aging and life prediction of cast duplex stainless steel components

Case duplex stainless steels, used extensively in nuclear, chemical, and petroleum industries because of higher strength, better weldability, higher resistance to stress corrosion cracking, and soundness of casting, are susceptible to thermal aging embrittlement during service at temperatures as low as 250°C. Recent advances in understanding the aging mechanisms, kinetics, and mechanical properties are presented, with emphasis on application of the material in safety-significant components in a nuclear reactor. Aging embrittlement is primarily due to spinodal decomposition of ferrite involving segregation of Fe, Cr, and Ni, and precipitation of M₂₃C₆ on ferrite-austenite boundaries or in ferrite. Aging kinetics are strongly influenced by synergistic effects of other metallurgical reactions that occur in parallel with the spinodal decomposition, i.e. clustering of Ni, Mo, and Si and G-phase precipitation in ferrite. A number of methods are outlined for estimating end-of-life aging, depending on several factors such as degree of permissible conservatism, availability of component archive material, and methods of estimating and verifying the activation energy of aging.     

Enhancement of the quality of the reactor pressure vessel used in light water power plants by advanced material,fabrication and testing technologies

Fracture safe assessment of nuclear reactor pressure vessels (RPV) is based upon an adequate stress analysis, reliable material characteristics and acceptable defect sizes.Problems may arise concerning inhomogeneities, low toughness and crack phenomena as observed in the base material and heat affected zone (HAZ). Therefore, efforts have been made to develop a steel which would be both non-susceptible to embrittlement and/or cracking in the HAZ, and have a higher upper-shelf toughness of base and HAZ material. Tests have been made on inhomogeneities and defects and also on improvement of chemical composition, the steel-making process, welding procedures and the optimum temperature cycle and level for stress-relief heat treatment.To solve these problems, common testing methods were supplemented by tangential-cut techniques, small HAZ-tensile test procedures and HAZ-simulation techniques.Results indicate that 50 per cent of 100 investigated component-strength welds are affected by micro stress-relief cracking (SRC) on a micro- and millimetre scale. The 22 NiMoCr 37 steel with optimised chemical composition, and the 20 MnMoNi 55 steel are both resistant to stress-relief embrittlement and SRC. Specific welding techniques are found to limit SRC and proposals for optimum stress-relief temperatures are given.For the generation of new components, the fracture-safe analysis can now be based completely upon homogeneous and high upper-shelf base materials including the HAZ.     

Overview of hydrogen embrittlement in high-Mn steels

Hydrogen and fuels derived from it will serve as the energy carriers of the future. The associated rapidly growing demand for hydrogen energy-related infrastructure materials has stimulated multiple engineering and scientific studies on the hydrogen embrittlement resistance of various groups of high performance alloys. Among these, high-Mn steels have received special attention owing to their excellent strength – ductility – cost relationship. However, hydrogen-induced delayed fracture has been reported to occur in deep-drawn cup specimens of some of these alloys. Driven by this challenge we present here an overview of the hydrogen embrittlement research carried out on high-Mn steels. The hydrogen embrittlement susceptibility of high-Mn steels is particularly sensitive to their chemical composition since the various alloying elements simultaneously affect the material's stacking fault energy, phase stability, hydrogen uptake behavior, surface oxide scales and interstitial diffusivity, all of which affect the hydrogen embrittlement susceptibility. Here, we discuss the contribution of each of these factors to the hydrogen embrittlement susceptibility of these steels and discuss pathways how certain embrittlement mechanisms can be hampered or even inhibited. Examples of positive effects of hydrogen on the tensile ductility are also introduced.     

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.     

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.     

Aging monitoring via magnetic property change of pressure vessel materials

The coercive force and magnetic remanence in A533B steel were measured in terms of the magnetic hysteresis energy loss. It was found that magnetic hysteresis energy loss in the tested materials are increased linearly with irradiation fluences of 1017 n/cm² range and decreasing with thermal annealing temperature below 1200K. The results can be explained in terms of the influence of pinning mechanisms on magnetic domain structures and these are related to changes in pressure vessel steel mechanical properities. It seems that this techniques can be successfully applied for development of a nondestructive testing method for monitoring of embrittlement in nuclear power pressure vessel materials. Applicability of this method for higher exposure levels is discussed.     

Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor

In this paper, three‐dimensional (3D) power distribution of newly designed small nuclear reactor core has been achieved by using neutron kinetic/thermal hydraulic (NK/TH) coupling. This is pressurized water reactor‐based small nuclear reactor in which plate type fuel element has been used and the core of the reactor has hexagonal type geometry. This paper depicts the design of the reactor core by using coupling approach of neutronics(Neutron Kinetic) and thermal hydraulic studies. For this purpose, neutronic analysis has been obtained by using lattice physics code, i.e. HELIOS and neutron kinetic code, i.e. REMARK. HELIOS code gives the cross‐section data which is being used as input to the REMARK code. At the same time, THEATRe code was used for the thermal hydraulic analysis of the reactor core. In the coupling process, some data (fuel temperature, moderator temperature, void fraction, etc.) from THEATRe code has been used in conjunction with HELIOS and REMARK codes. After finalizing the NK/TH coupling, 3D evaluation of the power distribution of the reactor core has been achieved and is included in the paper. The purpose of this paper is to evaluate the design and get the normal operational behavior of the reactor core by NK/TH coupling approach. Copyright © 2012 John Wiley & Sons, Ltd.     

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0·24 w/o copper, but different nickel contents at 0·63 w/o and 0·19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982°C (1800°F)) and to simulate Regulatory Guide 1·99 database materials (austenitized at 871°C (1600°F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260°C and 288°C) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an average temperature effect of 38°C for a 28°C difference in irradiation temperature. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel has little effect on irradiation response. No apparent microstructure effect on irradiation response was noted and the HSST-02 material's response to irradiation was similar to results from power reactor and other test reactor tests, thus qualifying the Ford Test Reactor for irradiation experiments such as those conducted for the Yankee Atomic program.     

Combination of radiation and hydrogen damage of reactor pressure vessel materials

The hydrogen embrittlement of low-alloyed base steel, austenitic cladding and heat affected zone (HAZ) of a reactor pressure vessel was measured for both unirradiated and irradiated materials. The fracture toughness decreased with both hydrogen charging and neutron irradiation; the shift of the fracture toughness-temperature transient curve is influenced by both damage processes. The plastic zone in hydrogen-charged material becomes smaller. The total elongation of both CrMoV and CrNiMoV HAZ decreases with increasing hydrogen content. This influence is pronounced in the HAZ after a weld process without subsequent annealing, a total loss of plasticity being observed in this case. The properties of the austenitic layer are not influenced at comparable hydrogen contents.     

Experience of surveillance schemes adopted for magnox steel reactor pressure vessels

Abstract

Surveillance or monitoring schemes are recognised to be an important part of any strategy to demonstrate that reactor pressure vessels used in civil nuclear power stations are operated within a safe regime. In the paper the authors describe the experience obtained from the surveillance schemes adopted for the UK's magnox nuclear power stations that were constructed with C–Mn steel reactor pressure vessels. These power stations were constructed in the late 1950s and 1960s and the last ceased generating in 2006. During the lifetime of the fleet with steel pressure vessels, there were developments in testing, observed changes in properties and understanding of radiation damage process that challenged the safety cases to support the operation of the stations. At the time the reactors were designed the concept of fracture toughness was only beginning to be investigated yet, during the lifetime of the stations, fracture toughness testing was successfully adopted as an input to fracture mechanics based assessment of the steel vessels. Over the operating life, a series of challenges emerged that were successfully addressed, including both hardening and non-hardening embrittlement, the latter due to impurity phosphorous segregation in weld metal and contributions from thermal nuclear embrittlement. These challenges led to the adoption of sophisticated statistical techniques to assess changes in embrittlement properties of the most critical construction material – submerged arc weld metal. A large scale sampling and testing programme of submerged arc weld metal removed from a decommissioned reactor pressure vessel validated the assessment process. As a result of successfully addressing these, and other challenges when the last two steel pressure vessel stations closed in December 2006, they had achieved lifetimes of nearly 40 years.