Sustained load crack growth in A533B-1 steel under neutron irradiation in a water environment

A study of the combined effects of radiation, water and temperature on sustained load crack growth behavior of reactor pressure vessel steel A533B-1 is reported. To complete this study wedge opening loading (WOL) T-type fracture toughness specimens were prepared from a sample of A533B-1 steel which had a copper content of 0.13%. The crack length change was measured after 939 hr of irradiation in a water environment. An electrical potential method was successfully used to measure the crack length of rusted radioactive specimens. Sustained load crack growth occurred at initial stress intensity factor K_Ii as low as . The value of stress corrosion cracking threshold factor K_Iscc after neutron irradiation in a water environment appears to be in the range of . The results of neutron irradiation in a water environment are to apparently increase the susceptibility of A533B-1 steel to stress corrosion cracking and hydrogen embrittlement.     

Reactor pressure vessel design of the high temperature engineering test reactor

The reactor pressure vessel (RPV) of the HTTR is 5.5 m (inside diameter), 13.2 m (inside height), and 122 mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc.2 1/4Cr–1Mo steel is chosen as the material for RPV. The temperature reaches about 400 °C at normal operation. The fluence of the RPV is estimated to be less than 1 × 10¹⁷ n/cm² (E > 1 MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr–1Mo steel for the RPV; embrittlement parameters, J-factor and are used.In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr–1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and , and nil-ductility transition temperatures, T_NDT, by drop weight tests, are shown. In-service inspection and results of R&Ds are also described.     

Reactor pressure vessel structural integrity research

Development continues on the technology used to assess the safety of irradiation embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack-tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil-ductility temperature (NDT) performs better than the reference temperature for nil-ductility transition (RT_NDT) as a normalizing parameter for shallow flaw fracture toughness data, (3) biaxial loading can reduce the shallow flaw fracture toughness, (4) stress based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow flaw fracture toughness because in-plane stresses at the crack-tip are not influenced by biaxial loading, and (5) an implicit strain based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation induced shift in Charpy V-notch vs. temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.     

Beremin模型预测尺寸效应对A508-3钢断裂韧性的影响

断裂韧性是用于表征反应堆压力容器（RPV）钢脆性状态的重要指标。在开展相关研究时,由于辐照空间小等原因,一般采用小尺寸紧凑拉伸（CT）试样。为掌握CT试样尺寸变化对国产RPV钢断裂韧性测试结果的影响,对国产A508-3钢的不同尺寸CT试样进行了测试分析,采用Beremin模型方法研究了尺寸效应对断裂韧性数据的影响,并建立了不同尺寸CT试样的断裂韧性数据归一化模型（TSM）。结果表明,同一温度下实验测得的断裂韧性值随试样尺寸的减小逐渐增大,不同样品通过标准方法得到的归一化数据存在偏差,本文建立的TSM可有效减小换算数据偏差。     

Assessment of toughness in long term service CrMo low alloy steel by fracture toughness and small punch test

Small punch test (SPT) is a miniature sample test technique which can evaluate in-service material properties with an almost non-destructive method. In this paper, the 2.25Cr1Mo steel samples serviced for 10 years in hydrogenation reactor (with temper embrittlement), 1.25Cr0.5Mo supper-pressure vapor pipe serviced for 14 years at 520 °C and several other low alloy steels have been studied by J_IC fracture toughness and SPT. The linear relationship between the small punch (SP) equivalent fracture strain and the fracture toughness of J_IC was created. The correlations applied to the experimental data indicated advantages of using SPT for the determining fracture toughness of in-serviced low alloy steels. Additionally, size affects the fracture pattern. Small punch samples of small size show dimple fractures whereas large fracture toughness samples show quasi-cleavage fractures.     

低铜合金反应堆压力容器钢辐照脆化预测评估模型

反应堆压力容器（RPV）材料辐照脆化预测评估对保证核反应堆安全运行、预防重大灾难性事故的发生具有重要意义。通过深入了解RPV材料辐照损伤机理和分析国外较为成熟的RPV辐照脆化预测模型,揭示了国外有关压力容器辐照脆化预测模型对低铜RPV辐照脆化预测的不足及其原因。在此基础上,发展和建立了适用于低铜RPV辐照脆化趋势的预测模型CIAE-2009。利用辐照性能数据对CIAE-2009模型进行了验证。结果表明,CIAE-2009对低铜含量RPV材料辐照脆化趋势预测具有较高的准确性和可靠性。     

Irradiation embrittlement characterization of the EUROFER 97 material

The paper summarizes original results of irradiation embrittlement study of EUROFER 97 material that has been proposed as one candidate of structural materials for future fusion energy systems and GEN IV.Test specimens were manufactured from base metal as well as from weld metal and tested in initial unirradiated condition and also after neutron irradiation.Irradiation embrittlement was characterized by testing of toughness properties at transition temperature region - static fracture toughness and dynamic fracture toughness properties, all in sub-size three-point bend specimens (27 × 4 × 3 mm³). Testing and evaluation was performed in accordance with ASTM and ESIS standards, fracture toughness K_JC and K_Jd data were also evaluated with the “Master curve” approach. Moreover, J-R dependencies were determined and analyzed.The paper compares unirradiated and irradiated properties as well as changes in transition temperature shifts of these material parameters. Discussion about the correlation between static and dynamic properties is also given.Results from irradiation of EUROFER 97 show that this steel - base metal as well as weld metal - is suitable as a structural material for reactor pressure vessels of innovative nuclear systems - fusion energy systems and GEN IV. Transition temperature shifts after neutron irradiation by 2.5 dpa dose show a good agreement in the case of EUROFER 97 base material for both static and dynamic fracture toughness tests. From the results it can be concluded that there is a low sensitivity of weld metal to neutron irradiation embrittlement in comparison with EUROFER 97 base metal.     

Fracture assessment of a brittle RPV under cold loading

The oldest Swedish reactor is a boiling water reactor (BWR) with a vessel made of A302 Grade B with rather high Cu and Ni content. These elements have intensified the irradiation embrittlement in the beltline region so that RT_NDT of certain welds may exceed 100 °C at the end-of-life condition. A preliminary study of the fracture risk for the beltline region showed that the limiting loading case would be the cold over-pressurization of the reactor. The objective of this study was to develop a reliable methodology for fracture assessment of the aged reactor vessel under cold loading scenarios. The test program covered experiments on standard SEN(B) specimens and clad beams under uniaxial and biaxial loading. The test material was a reactor vessel steel prepared with a special heat treatment to simulate fracture toughness properties of the aged reactor. No significant effects of shallow crack and biaxial loading were observed on cleavage fracture toughness in different clad specimens. While the ASME K_Ic reference curve was shown to be overly conservative, the Master Curve methodology satisfactorily predicted the experimental outcomes of the test program. The Master Curve methodology indicated that a 20-mm deep surface crack was acceptable in the beltline region under a cold over-pressurization scenario. This value was three times greater than what a methodology based on the ASME K_Ic reference curve yielded.     

Fracture mechanics concepts for evaluation of irradiated reactor pressure vessel material behavior

Within the German research program Forschungsvorhaben Komponentensicherheit (FKS), irradiation experiments were performed with ferritic reactor pressure vessel (RPV) steels and welds. The materials cover a wide range of chemical composition and initial toughness to achieve different susceptibility to neutron irradiation. Different neutron flux was applied and the neutron exposure extended up to 8×10¹⁹ cm⁻². The change in material properties was determined by means of tensile, Charpy impact, drop-weight and fracture mechanics tests, including crack arrest. The results have provided more insight into the acting embrittlement mechanisms and shown that the fracture mechanics concept of the Code provides in general an upper bound for the material which can be applied in the safety analysis of the RPV.     

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.     

国产反应堆压力容器的辐照脆化行为及预测

反应堆压力容器（RPV）是保障核电站运行安全性、经济性的核心构件。对RPV的完整性评估而言辐照脆化是必须面对的问题。我国已开发了第三代设计寿命为60 a的核电站。当达到寿期末时,辐照脆化的行为是未知的,这给国产RPV的辐照脆化预测带来了困难。为研究高注量下的辐照脆化行为,对A508-3钢的材料力学性能试样进行辐照考验,辐照温度为(288±8) ℃,中子注量水平达到反应堆压力容器60 a寿期末的辐照水平1×10²⁰ cm^-2;开展拉伸、冲击和断裂韧性试验,分析辐照脆化行为,在EONY模型基础上,提出针对国产RPV钢的改进的辐照脆化模型。模型的有效性被试验数据证实,其可准确预测国内A508-3材料的辐照脆化趋势。     

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。     

Prediction of lower bound fracture toughness for lifetime evaluation of reactor pressure vessels

The potential damage of embrittlement in service is a very important problem of MnMoNi steels used for the nuclear reactor pressure vessel. A decrease of critical flaw size may occur when embrittlement proceeds. The remaining lifetime of the reactors should be assessed taking into account the embrittlement of the steel paying special attention to the degradation of dynamic fracture toughness. The present study introduces the basic concept of the remaining lifetime assessment. Examined was a small specimen fracture toughness test for measuring the dynamic fracture toughness of nuclear reactor pressure vessel (RPV) steels. The result was applied in the measurement of the dynamic fracture toughness of 12 heats of RPV steels. The test results were analyzed to find more practical applications and a method is presented to predict the lower bound dynamic fracture toughness using the Charpy impact test and tensile test results.     

Pressure vessel steel fracture toughness in the regime from room temperature to 400°C

The fracture toughness of steels that are susceptible to dynamic strain aging shows a minimum at temperatures higher than the upper shelf starting temperature. This phenomenon is caused simultaneously by strain aging and plastic deformation. The first aim of the present work is to analyze the effect of dynamic strain aging on the fracture toughness values of three pressure vessel steels in the temperature range between room temperature and 400°C. Fracture mechanics tests were carried out on A533 GB, A516 G70 and 304L steels to obtain the following parameters: J_IC, CTOD_m and the J-R curves. These values were compared against those available in the present references, and good agreement was found. Charpy V notch tests were also carried out on A516 G70 steel at the same test temperatures as for the fracture mechanics tests to analyze the effect of the strain rate. The critical wide stretch zones of the 304L steel specimens were also measured to verify another author's hypothesis about a toughness drop at the upper shelf temperature.     

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.     

Fracture toughness prediction for highly irradiated RPV materials: From test results to RPV integrity assessment

New fracture toughness data are represented for highly irradiated RPV materials that were obtained by testing standard compact specimens with thickness of 12.5 mm and 25 mm and pre-cracked Charpy specimens machined from the RPV decommissioned. Two advanced engineering methods, the Master Curve and the Unified Curve, are applied for treatment of the test results. Application of the dependence of fracture toughness K_JC on test temperature T predicted with the Master Curve and the Unified Curve methods on the basis of surveillance specimens testing is discussed for RPV integrity assessment when the reference K_JC(T) curve is recalculated to the crack front length of the postulated flaw that is considerable larger than thickness of surveillance specimens. The prediction of the K_JC(T) curve transformation caused by neutron irradiation is considered.     

Effect of thermal aging on fracture toughness of RPV steel

The effect of thermal aging on mechanical properties and fracture toughness was investigated on pressure vessel steel of light water reactors. Submerged are welded plates of ASME SA508 C1.3 steel were isothermally aged at 350°C, 400°C and 450°C for up to 10,000 hrs. Tensile, Charpy impact and fracture toughness testings were conducted on the base metal and the weld heat affected zone (HAZ) material to evaluate whether thermal aging induced by the plant operation is critical for the integrity of the pressure vessel or not. Tensile properties of the base metal was not changed by thermal aging as far as the thermal aging conditions were concerned. Relatively distinct degradation was observed in fracture toughness J_IC and J-resistance properties of both the base metal and the weld HAZ material, while only slight changes were observed in Charpy impact properties for both of them. However, it was concluded that the effect of thermal aging estimated by 40–80 years of plant operation on fracture toughness of both materials is small.     

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。      

Experimental and numerical investigations of the warm-prestressing (WPS) effect considering different load paths

The effect of warm prestressing has been investigated representative for the core weld metal of the RPV Stade. Model experiments on CT specimens show a significant rise of effective fracture toughness K_eff after warm prestressing and the conservative WPS hypothesis, ‘no failure, if ∂K_I/∂t≤0’, is verified. Partial unloading and reheating show no influence on the effective fracture toughness K_eff. The magnitude of the WPS effect as a function of warm prestress level and temperature, path of unloading and cooling can be predicted using a modified Beremin model with temperature dependent parameters. It is shown that the Weibull stress is an appropriate crack tip loading parameter for decreasing load paths.