Experience of irradiation embrittlement in the Loviisa reactor pressure vessels

The water gap between the wall and the core of the RPV (Reactor Pressure Vessel) in a VVER-440 plant is small compared with typical Western type LWR₅. The neutron fluence on the RPV wall is, consequently, much higher in a VVER-440 plant. In older VVER-440 plants the material of the RPV, especially the horizontal core weld, contains so much impurities (P- and Cu-content) that the irradiation embrittlement has become a problem. On bases of fracture mechanics analyses in Loviisa, IVO has been forced to make several measures to ensure safe operation of the plants. According to IVO's current understanding, both plants may be in operation for the design life without annealing of the RPVs.     

Neutron induced damage in reactor pressure vessel steel: An X-ray absorption fine structure study

The radiation damage produced in reactor pressure vessel (RPV) steels during neutron irradiation is a long-standing problem of considerable practical interest. In this study, an extended X-ray absorption fine structure (EXAFS) spectroscopy has been applied at Cu, Ni and Mn K-edges to systematically investigate neutron induced radiation damage to the metal-site bcc structure of RPV steels, irradiated with neutrons in the fluence range from 0.85 to 5.0 × 10¹⁹ cm⁻². An overall similarity of Cu, Ni and Mn atomic environment in the iron matrix is observed. The radial distribution functions (RDFs), derived from EXAFS data have been found to evolve continuously as a function of neutron fluence describing the atomic-scale structural modifications in RPVs by neutron irradiations. From the pristine data, long range order beyond the first- and second-shell is apparent in the RDF spectra. In the irradiated specimens, all near-neighbour peaks are greatly reduced in magnitude, typical of damaged material. Prolonged annealing leads annihilation of point defects to give rise to an increase in the coordination numbers of near-neighbour atomic shells approaching values close to that of non-irradiated material, but does not suppress the formation of nano-sized Cu and/or Ni-rich-precipitates. Total amount of radiation damage under a given irradiation condition has been determined. The average structural parameters estimated from the EXAFS data are presented and discussed.     

Positron annihilation study and computational modeling of defect production in neutron-irradiated reactor pressure vessel steels

Positron annihilation spectroscopy (PAS) and a computer simulation were used to investigate a defect production in reactor pressure vessel (RPV) steels irradiated by neutrons. The RPV steels were irradiated at 250 °C in a high-flux advanced neutron application reactor. The PAS results showed that mainly single vacancies were created to a great extent as a result of a neutron irradiation. Formation of vacancies in the irradiated materials was also confirmed by a coincidence Doppler broadening measurement. For estimating the concentration of the point defects in the RPV steels, we applied computer simulation methods, including molecular dynamics (MD) simulation and point defect kinetics model calculation. MD simulations of displacement cascades in pure Fe were performed with a 4.7 keV primary knock-on atom to obtain the parameters related to displacement cascades. Then, we employed the point defect kinetics model to calculate the concentration of the point defects. By combining the positron trapping rate from the PAS measurement and the calculated vacancy concentrations, the trapping coefficient for the vacancies in the RPV steels was determined, which was about 0.97 × 10¹⁵ s⁻¹. The application of two techniques, PAS and computer simulation, provided complementary information on radiation-induced defect production.     

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%.     

承压热冲击下材料特性对反应堆压力容器结构完整性的影响

介绍了承压热冲击(PTS)分析的背景和研究现状,阐述了基于确定性断裂力学的反应堆压力容器(RPV)结构完整性分析方法.分析了材料性能模式(线弹性和弹塑性)和辐照效应对PTS下RPV结构完整性的影响.     

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.     

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.     

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.     

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.     

RPV material investigations of the former VVER-440 Greifswald NPP

The real toughness response of RPV material can only be determined after the final shut down of the NPP. Such a chance is given now by investigating material from the former Greifswald NPP (VVER-440/230).In the first part the paper deals with fast neutron fluence calculations and retrospective dosimetry based on Niobium. Unfortunately, a second neutron reaction besides ⁹³Nb(n,n’) leading to ^93mNb-activity is the reaction ⁹²Mo(n,γ)⁹³Mo. Based on the found Nb and Mo contents in the RPV material, it turned out that the ^93mNb generation on the Mo path mostly dominates over the fast neutron induced generation from Nb.The comparison between the calculated and the measured ^93mNb activities typically resulted in deviations of 50%. Possible reasons for the observed differences are discussed.In the second part first results of fracture mechanic investigations are reported. SE(B) specimens from three thickness positions were tested and evaluated according to the test standard ASTM E1921-05. Cleavage fracture toughness values, K_Jc, were determined and Master Curve based reference temperatures (T₀) were evaluated. The T₀ measured at the inner surface of the RPV did not represent the conservative condition. The T₀ of disc 1-1.3 located between the surface and 1/4 thickness is about 40K higher compared with those of the surface.The measured K_Jc values are not enveloped by the 5% fractile indexed with T₀ according to the Master Curve concept. However, the 5% fractile indexed with the VERLIFE reference temperature RTT_o that includes an additional margin envelops the measured K_Jc values. Therefore the VERLIFE lower bound curve conservatively describes the fracture toughness of the investigated weld metal.     

Microstructural characteristics and embrittlement phenomena in neutron irradiated 309L stainless steel RPV clad

The effects of neutron irradiation on the microstructural features and mechanical properties of 309L stainless steel RPV clad were investigated using TEM, SEM, small tensile, microhardness and small punch (SP) tests. The neutron irradiations were performed at 290 °C up to the fluences of 5.1 × 10¹⁸ and 1.02 × 10¹⁹ n/cm² (>1 MeV) in Japan Materials Testing Reactor (JMTR). The microstructure of the clad before and after irradiation was composed of main part of fcc austenite, a few percent of bcc δ-ferrite and small amount of brittle σ phase. After irradiation, not only the yield stress and microhardness, but SP ductile to brittle transition temperature (SP-DBTT) were increased. However, the increase in SP-DBTT is almost saturated, independent of the neutron fluence. Based on the TEM observation, the origin of irradiation hardening was accounted for by the irradiation-produced defect clusters of invisible fine size (<1-2 nm), and the shift of SP-DBTT was primary due to the higher hardening and the preferential failure of δ-ferrite. The embrittlement of the clad was strongly affected by the initial microstructural factors, such as the amount of brittle σ phase, which caused a cracking even in an early stage of deformation.     

堆内捕集器对压力容器中子辐照影响分析

严重事故下为实现堆内熔融物滞留,可采用堆内捕集器(IVCC)的策略。捕集器属压力容器的一部分,属不可更换设备,需长期在堆内受中子辐照。本文通过对典型压水堆压力容器模型和带IVCC的压力容器模型的比较,发现IVCC不会改变压力容器内快中子通量,不会对压力容器的辐照造成影响。且IVCC使得压力容器的热中子通量明显减小,降低了压力容器的整体辐照水平。这说明IVCC对压力容器的辐照性能不会产生不利影响,反而有助于防止压力容器的老化。     

Fracture mechanics issues for irradiated blanket structures

This paper reviews some of the factors that will affect fracture behavior of fusion reactor structures and summarizes some component life predictions based on linear elastic fracture mechanics analysis. The review includes discussion of the environments to which the components will be subjected, the response of materials to these environments, the time dependent nature of the structural response, and the fracture related failure mechanisms.Radiation environments and complex loading conditions in a fusion reactor cause a variety of material phenomena. These phenomena include irradiation swelling and creep, strength changes due to matrix hardening, helium embrittlement, and surface effects such as sputtering and blistering.The interaction of thermal creep, irradiation creep, and swelling results in complex time, temperature, and neutron fluence dependent stress histories in first wall and blanket structures. These effects reduce compressive thermal stresses during the burn portion of a reactor operating cycle and result in residual tensile stress during the non-burn portion of the cycle. The cyclic nature of these stresses, particularly in a tokamak reactor, and the presence of undetected flaws provide a basis for the application of fracture mechanics. Linear elastic fracture mechanics analysis techniques have been applied to predict component life for several conceptual tokamak fusion reactor designs. These analyses show that the structural life may be limited by growth of initial flaws to a coolant leakage. Results indicate that for neutron wall loadings below 2 to 3 Mw/m², life is likely to be controlled by stresses during the burn period and, at higher wall loadings, by residual stresses during the non-burn period.Fracture toughness properties tend to be reduced by irradiation. Therefore, brittle fracture will be a potentially critical failure mode. Fatigue crack growth and fracture characteristics of the design will affect the operating mode of a reactor and influence the performance of different types of reactors. Tests are currently planned to develop material crack growth and fracture toughness data [1] for candidate alloys because these properties have been shown to be important.     

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.     

Fracture mechanical properties and neutron irradiation effects of fuel compacts for the HTTR

To simulate the nuclear fuel for High Temperature Engineering Testing Reactor (HTTR), fuel compact models using SiC-kernel coated particles instead of UO₂-kernel coated particles were prepared under the same conditions as those for the real fuel compact. The mechanical and fracture mechanics properties were studied at room temperature. The thermal shock resistance and fracture toughness for thermal stresses of the fuel compact were experimentally assessed by means of arc discharge heating applied at a central area of the disk specimens. These model specimens were then neutron irradiated in the Japan Material Testing Reactor (JMTR) for fluences up to 1.7 × 10²¹n/cm² (E ·> 29 fJ) at 900°C ± 50°C. The effects of irradiation on a series of fracture mechanical properties were evaluated and compared with the cases of graphite IG-110 used as the core materials in the HTTR.     

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。     

The effect of neutron irradiation on mechanical properties of YAG laser weldments using previously irradiated material

The objective of this study is to make clear the effect of neutron irradiation on mechanical properties of laser weldments using irradiated material. This estimation is necessary for the application to joining coolant piping of the ITER blanket. Irradiation testing was performed at Japan Material Testing Reactor (JMTR). On the irradiation condition for weldments using irradiated material, fast neutron fluence was 1.4 × 10²⁴ n/m², which corresponds to a displacement damage rate of 0.26 displacement per atom (dpa) and irradiation temperature 200 °C. The results of this study show that tensile properties of all weldments changed into that of base material by the effect of neutron irradiation. The results of hardness tests show that irradiation hardening at an irradiation damage dose of 0.3 dpa is almost same as that at irradiation damage 0.6 dpa. It is concluded that irradiated weldments using irradiated material were moved toward irradiated base material on tensile and hardness properties up to 0.6 dpa. On the other hand, tensile properties of base material were changed by the effect of neutron irradiation up to about 0.3 dpa, and with much less change from 0.3 dpa to 0.6 dpa. It is inferred that the effect of neutron irradiation of SS316LN-IG almost saturated up to 0.3 dpa.     

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.     

Positron annihilation lifetime study of JRQ and JPA irradiated and annealed reactor pressure vessel steels

The model reactor pressure vessel steels known as JRQ and JPA were manufactured in Japan for the IAEA neutron embrittlement research studies. These model alloys belong to the commercially used steel A533B-1 type and show relatively large changes in mechanical properties after relevant neutron irradiation. The neutron irradiation was performed by different neutron fluxes as well as different neutron fluences (up to about 150 × 10¹⁸ cm⁻² (E > 0.5 MeV)). For a better understanding of the neutron embrittlement, the Positron Annihilation Lifetime Spectroscopy (PALS) technique was applied in 2014. PALS measurement of irradiated specimens was performed using three detectors set-up due to induced ⁶⁰Co radioactivity of the studied specimens. We confirmed that the JPA steel, considered to be high-copper steel, is much more sensitive to defect creation due to neutron irradiation than the low-copper JRQ steel.