Study on mo steel for high temperature gas-cooled reactor material characterization of forged low Si mo steel

The Japan Atomic Energy Research Institute (JAERI) has carried out a series of research and development work related to the high temperature gas-cooled reactor (HTGR) and, accordingly the high temperature engineering test reactor (HTTR) will be constructed in the near future. As the reactor pressure vessel (RPV) material, Mo steel will be used. Material characterization tests have been carried out to evaluate the applicability of the Mo steel for the RPV and to prepare for the licensing. The present paper summarizes the fracture toughness behavior including K_Id and K_Ia, irradiation embrittlement susceptibility and degradation of steel due to the long term aging at high temperature of the forged low Mo steel. These tests reveal good fracture toughness which well meets the requirements of the ASME Code, low neutron irradiation embrittlement susceptibility, little embrittlement by long term aging and so on. The present test results demonstrate good applicability of forged low Mo steel to the RPV of HTGR.     

Behavior of mechanical properties of nickel-alloyed reactor pressure vessel steel under neutron irradiation and post-irradiation annealing

The effect of neutron irradiation and post-irradiation thermal annealing on tensile and impact properties of Cr–Ni–Mo steel used for WWER-1000 reactor pressure vessel (RPV) manufacturing was studied. A gap in yield stress and ultimate tensile stress fluence dependence at the fluence range of 0–3×10²³ neutrons m⁻² was observed while ductile-to-brittle transition temperature (DBTT) was continuously increasing with damage dose. The post-irradiation annealing recovery of tensile properties was found to be higher than the one of impact properties. Over-recovery of tensile properties due to 460 and 490°C post-irradiation annealings were observed. The annealing effectiveness of WWER-440 and WWER-1000 grades was compared. Nickel was supposed to affect both the radiation sensitivity and the post-irradiation residual DBTT shift of WWER-1000 type steel.     

Review of liquid metal corrosion issues for potential containment materials for liquid lead and lead–bismuth eutectic spallation targets as a neutron source

Lead (Pb) and lead–bismuth eutectic (44Pb–56Bi) have been the two primary candidate liquid metal target materials for the production of spallation neutrons. Selection of a container material for the liquid metal target will greatly affect the lifetime and safety of the target subsystem. For the liquid lead target, niobium–1 wt% zirconium (Nb–1Zr) is a candidate containment material for liquid lead, but its poor oxidation resistance has been a major concern. In this paper, the oxidation rate of Nb–1Zr was studied based on the calculations of thickness loss resulting from oxidation. According to these calculations, it appeared that uncoated Nb–1Zr may be used for a 1-year operation at 900°C at P_O2=1×10^–6 Torr, but the same material may not be used in argon with 5-ppm oxygen. Coating technologies to reduce the oxidation of Nb–1Zr are reviewed, as are other candidate refractory metals such as molybdenum, tantalum, and tungsten. For the liquid lead–bismuth eutectic target, three candidate containment materials are suggested, based on a literature survey of the materials’ compatibility and proton irradiation tests: Croloy 2-1/4, modified 9Cr–1Mo, and 12Cr–1Mo (HT-9) steel. These materials seem to be used only if the lead–bismuth is thoroughly deoxidized and treated with zirconium and magnesium.     

Monitoring the embrittlement of reactor pressure vessel steels by using the Seebeck coefficient

The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient () of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1 MeV, and fluencies ranging from 0 up to 4.5 × 10¹⁹ neutrons per cm², are presented. Within this range, it was observed that increased by ≈500 nV/°C and a linear dependency was noted between and the temperature shift ΔT_41 J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible.     

Microstructural stability of modified 9Cr–1Mo steel during long term exposures at elevated temperatures

This paper describes the stability of microstructure in normalized and tempered modified 9Cr–1Mo steel exposed to service temperatures in the range of 773–873 K for different time durations. A detailed microstructural and microchemical analysis of the secondary phases was carried out using optical and electron microscopy techniques. The microstructural observations, supported by hardness measurements showed that the lath morphology of the tempered martensite was retained even after 10 000 h of aging. The coarsening of M₂₃C₆ carbide was observed until 5000 h, when the Laves phase started appearing. The microstructural features observed are discussed in conjunction with the embrittlement observed in this steel on high temperature aging exceeding 5000 h.     

Effects of Alloy Elements on Irradiation Embrittlement of Reactor Pressure Vessel Steel: Insights from Neural Network Modelling

As one of the key components that can not be replaced in PWR, the safety and stability of reactor pressure vessel (RPV) steel determine the safety and economy of the reactor. The irradiation embrittlement of RPV steel is the limiting factors for the operation of PWR. The irradiation embrittlement of RPV steel is closely related to its alloy composition. Based on the machine learning method, the relationship between key alloy components (Cu/Mn/Ni/Si/P) and irradiation embrittlement of RPV steel was constructed. The results show that the relationship between the alloy composition and irradiation embrittlement is basically consistent with the traditional cognition. The irradiation embrittlement is sensitive to Cu content, and Cu-Ni has synergistic effect on irradiation embrittlement. In low Cu alloys, Mn-Ni and Ni-Si have synergistic effects on embrittlement.     

利用神经网络方法研究合金元素对反应堆压力容器钢辐照脆化的影响

反应堆压力容器（RPV）作为压水堆中不可更换的关键部件之一,其安全和稳定是决定反应堆安全经济运行的重要因素。RPV钢的辐照脆化问题是制约RPV在堆内安全服役的关键。RPV钢的辐照脆化与其合金成分关系密切。本文利用神经网络方法研究了RPV钢中关键合金成分（Cu、Mn、Ni、Si、P）与辐照脆化之间的关系。研究结果表明,基于神经网络方法得到合金成分与辐照脆化的关系与传统认知基本一致,辐照脆化对Cu含量最敏感,Cu-Ni对辐照脆化存在协同作用,低Cu合金中Mn-Ni、Ni-Si对脆化存在协同作用。     

The effects of the stainless steel cladding in pressurized thermal shock evaluation

Fracture mechanics analysis is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV), such as the pressurized thermal shock (PTS) analysis and P–T limit curve construction. However, the existence of stainless steel cladding, with different thermal, physical, and mechanical property at the inner surface of reactor pressure vessel complicates the fracture mechanics analysis. In this paper, the simple analytical treatment schemes to calculate the stress and resulting stress intensity factor at the tip of the flaws in the RPV with stainless steel cladding are introduced. For a reference thermal–hydraulic boundary condition, the effects of cladding thermal conductivity and thermal expansion coefficients on the stress intensity factor of surface flaws were examined. Also, the effects of cladding plasticity and thickness were quantitatively examined. The analysis results showed that the existence of the stainless cladding had significant impacts on the RPV failure probabilities.     

Failure assessment of RPV nozzle under loss of coolant accident

In case of a postulated loss of coolant accident (LOCA) of a reactor pressure vessel (RPV), the nozzle region experiences higher stresses and lower temperatures than the remaining part of the RPV. Thus, the nozzle is to be considered in the RPV safety assessment. For a LOCA event, three-dimensional elastic–plastic finite element calculations of stresses and strains in the intact RPV were performed. Using the substructure technique, fracture mechanics analyses were then carried out for several postulated cracks in the nozzle corner and in the circumferential weld below the nozzle. For different crack geometries and locations, the J-integral and the stress intensity factor were calculated as functions of the crack tip temperature. Based on the K_IC-reference curve and the J_R curve, both brittle and ductile instability of the postulated cracks were excluded. In order to reduce the expenses of three-dimensional finite element analyses for various crack geometries, an analytical procedure for calculating stress intensity factors of subclad cracks in cylindrical components was extended for cracks in the nozzle corner.     

Influence of the depth position on the neutron embrittlement of the VVER reactor pressure vessel steel 15Cr2MoV(A) consequences for the assessment of reactor safety

The dependence of the mechanical properties on the depth position in the unirradiated state and after irradiation up to neutron fluences of approximately 5 × 10¹⁸ and 70 × 10¹⁸ cm⁻² (E > 0.5 MeV) is tested on a forging made out of VVER 440 reactor pressure vessel (RPV) steel 15CrMoV. The near-surface position shows a higher strength and a lower transition temperature than the positions greater than 1/4 wall thickness. Irradiation does not change these differences in a significant manner. The testing of specimens from the 1/4 depth position within the surveillance programme, as normally laid down in the legal rules relating to nuclear power plants, results in a conservative safety assessment against brittle failure up to the EOL fluence. On taking into account fluence attenuation, the flux effect, etc., the toughness gradually increases from the inside to the outside of the wall after longer RPV operating times.     

Structural design of high temperature metallic components

In the high temperature engineering test reactor (HTTR), even at normal operation the service temperatures of class 1 metallic components reach temperatures above 900 °C when exposed to primary helium coolant of 950 °C. For these components, Hastelloy XR, which is the improved version of Hastelloy X, was developed and used for high temperature application.Some of the high temperature materials and their service temperatures, including Hastelloy XR, used for the class 1 and reactor internal metallic components of the HTTR are very well beyond the well-established Japanese elevated temperature structural design guideline. Moreover, at very high temperatures, where creep deformation is significant, the component design based on elastic analysis is impossible. Therefore, many research works on structural mechanics behavior were carried out to establish a high temperature structural design guideline and creep analysis methods. This paper reviews structural design of the high temperature components for the HTTR made of Hastelloy XR, 2 1/4Cr–1Mo steel, austenitic stainless steels SUS321TB and SUS316, and 1Cr–0.5Mo–V steel.     

Neutron embrittlement surveillance of the garigliano reactor vessel steel

The effects of neutron radiation on the pressure vessel of the Garigliano Nuclear Power Station have been analyzed on the basis of results of a reactor vessel material surveillance program of the plant. A high radiation embrittlement sensitivity was determined for the weld metal and for the A336 forging steel of the ring forging course just above the level of the fuel core. Both showed high copper and phosphorus contents, which accounted for the embrittlement sensitivity. The ring forging opposite the fuel core had a low copper and phosphorus content and revealed relatively low embrittlement. A neutron fluence of 6.3 × 10¹⁹n/cm² > 1 MeV was determined for the peak flux plane for 40yr of operation. However, the 40yr fluence for the ring forging at the top of the core level (3.5 × 10¹⁹n/cm² > 1 MeV) resulted in the highest final transition temperature because of the sensitivity of this steel. The measured Charpy-V shelf energy absorption values were plotted against yield stress values for comparable irradiations on the ratio analysis diagram (RAD). The analysis revealed that the pressure vessel steel properties would continue to degrade toward a condition of possible frangibility at the end of its life. This projection is based on an assumption of uniform embrittlement throughout the vessel wall thickness. Such uniformity does not exist; in fact, a sharp gradient exists in the steel such that ductility rises rapidly in the steel toward the outside wall as well as above and below the fuel core. Hence, because of this strong ductility gradient, the Garigliano reactor vessel should be able to operate safely over its intended design lifetime.     

Effect of the Cu-rich precipitates on thermal conductivity of RPV model steel

Cu-rich precipitates are the important influence factors for the irradiation embrittlement of the reactor pressure vessel model steels. The microstructure of the Cu-rich precipitates could be revealed by mechanical and magnetic properties. In this article, the effect of the Cu-rich precipitates on thermal conductivity was studied. The reactor pressure vessel (RPV) model steels were aged for different time at 500°C. The results show that the thermal conductivity of RPV model steel is first decreased and then increased during the experiment, with a minimum value at 48.33 ± 0.21 W·m^?1·K^?1 after being aged for 200 h. The changing thermal conductivity is decided by the synergistic effect of the following three factors: (1) the crystal structure transformation of Cu-rich precipitates, (2) the orientation relationship between the matrix and Cu-rich precipitates, (3) the content of Cu atoms in the matrix.     

Application of the Seebeck effect for monitoring of neutron embrittlement and low-cycle fatigue in nuclear reactor steel

The monitoring of neutron embrittlement and low-cycle fatigue in nuclear reactor steel is an important topic in lifetime extension of nuclear power plants. Among several material parameters that may change due to material degradation are the thermoelectric properties. Therefore, we investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel (RPV) steel. The Seebeck coefficients (SC) of several irradiated Charpy specimens made from Japanese reference steel JRQ were measured. The specimens suffered fluences from 0 up to 4.5 E¹⁹ neutrons/cm², with energies higher than 1 MeV. Measured changes of the SC within this range were about 500 nV/°C, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.5 E¹⁹ neutrons/cm². We obtained a linear dependency between the SC and the temperature shift ΔT₄₁ of the Charpy energy versus temperature curve, which is widely used to characterize material embrittlement.Similar measurements were performed on fatigue specimens made from the austenitic stainless steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying cyclic strain amplitudes of 0.28%. A clear correlation between the change of the SC and the accumulated plastic strain, i.e. number of cycles was obtained.Further investigations were made to quantify the size of the gauge volume in which the thermoelectric power (TEP), also called thermoelectric voltage, is generated. It appeared that the information gathered from a thermoelectric power measurement is very local. This fact can be used to develop a novel TEP-method acting as a thermoelectric scanning microscope (TSM).Finally, we conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. This disadvantage can partially be avoided by using the proposed new TSM.     

Consolidation of scientific and technological expertise to assess the reliability of reactor pressure vessel embrittlement prediction in particular for the arctic area plant (COBRA)

The evaluation and prognosis of reactor pressure vessel (RPV) material embrittlement in WWERs and the allowable period of their safe operation are performed on the basis of impact test results of irradiated surveillance specimens. The main problem concerns the irradiation conditions (irradiation temperature, neutron flux and neutron spectrum) of the surveillance specimens that have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV wall condition. In particular, the key issue is the possible difference between the irradiation temperature of the surveillance specimens and the actual RPV wall temperature. It is recognized that the direct measurement of the irradiation temperature by thermocouples during reactor operation is the only way to obtain reliable information. In addition, the neutron field's parameters in the surveillance specimens location have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level.The COBRA project, which started in August 2000 and had a duration of 3 years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contained state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The selected reactor for the experiment was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings, the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5 ± 4 °C. Uncertainties were evaluated also with experimental work carried out in the WWRSZ research reactor and by finite element modelling of surveillance capsules. The results obtained show that there is not need to perform temperature correction when surveillance data of irradiated specimens are used for embrittlement assessment of WWER-440(213) reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals 2.7 × 10¹² cm⁻² s⁻¹ with E > 0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.     

Assessment of irradiation response of WWER-440 welds using samples taken from Novovoronezh unit 3 and 4 reactor pressure vessels

The results of the study on Novovoronezh unit 3 and 4 (NV NPP-3 and 4) reactor pressure vessel (RPV) radiation embrittlement measured using subsize impact specimens (5×5×27.5 mm³) fabricated from samples taken from the corresponding RPV walls are presented. The post-irradiation annealing effectiveness and the embrittlement kinetics of Novovoronezh unit 3 and 4 RPV welds under re-irradiation are discussed. Ductile-to-brittle transition temperatures (DBTT) obtained using standard Charpy (TT^10×10) and subsize impact (TT^5×5) specimens of trepans cut out from Novovoronezh unit 2 RPV are compared. A new relation between TT^10×10 and TT^5×5 has been developed.     

A parametric study on pressure–temperature limit curve using 3-D finite element analyses

In order to operate a reactor pressure vessel (RPV) safely, it is necessary to keep the pressure–temperature (P–T) limit during the heatup and cooldown process. While the ASME Code provides the P–T limit curve for safe operation, this limit curve has been prepared under conservative assumptions. In this paper, the effects of conservative assumptions involved in the P–T limit curve specified in the ASME Code Sec. XI were investigated. Three different parameters, the crack depth, the cladding thickness and the cooling rate, were reviewed based on 3-D finite element analyses. Also, the constraint effect on P–T limit curve generation was investigated based on J–T approach. It was shown that the crack depth and constraint effect change the safety region in P–T limit curve dramatically, and thus it is recommended to prepare a more precise P–T limit curve based on finite element analysis to obtain P–T limit for safe operation of a RPV.     

反应堆压力容器辐照监督无损评估技术研究

反应堆压力容器（RPV）的辐照脆化问题是核安全的重中之重,影响到核电厂的安全性、经济性与公众信心。介绍了传统RPV辐照监督方案,讨论了现行技术的局限性,梳理了RPV辐照监督无损评估技术国外研究进展与存在问题,在实验与理论研究的基础上创新性地提出了中子辐照条件下RPV钢力学性能预测统一模型,并形成了基于电磁性能的RPV辐照监督无损评估技术,进一步完善后具有较好的工程应用前景。同时指出了开展RPV钢电磁性能实验研究,既有助于从一个全新的角度理解与再认识国产RPV钢长寿期服役时的辐照脆化行为,又有利于揭示RPV钢辐照脆化机理,丰富辐照脆化的基础理论。      

Brittle fracture analysis of the Kozloduy Unit 1 RPV

The safety of the RPV of the Bulgarian NPP Kozloduy Unit 1 was analysed within EC-financed contracts according to a pressurized-thermal-shock- (PTS-) procedure applied in Germany (Erve, M., Hertlein, R., 1991. Post SMiRT Seminar No 11, August 1991), considering the most relevant transients and taking into account the actual embrittlement in the core weldment. The paper reports on the main aspects of the PTS-procedure, determining the acceptable transition temperature (T_K^a-evaluation) to exclude brittle fracture, and compares the main results with the fluence related transition temperature (T_K^F) of the material got from sampling from the weldment concerned. Testing of the toughness properties by small size Charpy-V-notch specimens revealed only a small irradiation effect in comparison to the properties after the recovery annealing performed in 1989. This could be explained by the fact that only small values of Cu-content in the weld metal were confirmed, thus balancing the expected influence of the relatively high P-content. The main conclusion is: assuming a defect size of 10×60 mm, the evaluation shows, for KNPP 1 after the 18th cycle for the screening transient, a sufficient margin in the T_K^a-value to the actual material properties and—from the technical point of view—thus, recovery annealing is not necessary for the time being. Further embrittlement of the RPV will be covered by an additional surveillance program with samples accelerated re-irradiated in a Russian NPP. Proper operator actions during PTS events can further improve the situation with respect to loading of the RPV during transients, thus increasing the safety margins.     

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

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