Small specimen predictions of fracture toughness for nuclear pressure vessel steels

The use of fracture mechanics in the fracture-safe design and continued safe operation of nuclear reactor pressure vessels has provided an incentive for the development of small specimens for obtaining pertinent fracture toughness data. Small specimens are required for economic reasons when a large number of heats are involved and for space limitation reasons such as in surveillance programs. Several approaches to obtaining fracture toughness from small specimens by either direct measurements or indirect correlations and calculations are reviewed, and their merits and limitations are discussed. Emphasis is placed on techniques which have been developed to determine static and dynamic fracture toughness from surveillance-type specimens. Recently developed techniques for obtaining J-initiation values from a single test specimen and methods for estimating lower and upper shelf fracture toughness from tensile properties are also presented.     

Characterization of tensile strength and fracture toughness of nuclear graphite NBG-18 using subsize specimens

The mechanical properties of NBG-18 nuclear grade graphite were characterized using small specimen test techniques and statistical treatment on the test results. New fracture strength and toughness test techniques were developed to use subsize cylindrical specimens with glued heads and to reuse their broken halves. Three sets of subsize cylindrical specimens of different sizes were tested to obtain tensile fracture strength and fracture toughness. The mean fracture strength decreased as the specimen size increased. The fracture strength data indicate that in the given diameter range the size effect is not significant and much smaller than that predicted by the Weibull moduli estimated for individual specimen groups of the Weibull distribution. Further, no noticeable size effect existed in the fracture toughness data. The mean values of the fracture toughness datasets were in a narrow range of 1.21-1.26 MPa√m.     

An indentation system for determination of viscoplastic stress-strain behavior of small metal volumes before and after irradiation

The development of fusion materials for the first wall in future fusion reactors requires methods for the investigation of irradiation effects on the mechanical properties of materials which are only available in small volumes. Depth and force reading hardness measurement (or indentation) is one of the candidates that have the potential to extract valuable information on the stress-strain behavior of a material. A modified commercial indentation device installed in a hot cell of a fusion materials laboratory (FML) in combination with a neural network based analysis method allows identifying the material parameters of a unified viscoplasticity model with nonlinear isotropic and kinematic hardening from small metal specimens. By investigation of the same material before and after irradiation the method provides the possibility to separate irradiation effects on modulus, hardening and viscous behavior.     

Mechanical properties and microstructural investigations of TIG welded 40 mm and 60 mm thick SS 316L samples for fusion reactor vacuum vessel applications

The development of advanced fusion reactors like DEMO will have various challenges in materials and fabrication. The vacuum vessel is important part of the fusion reactor. The double walled design for vacuum vessel with thicker stainless steel material (40–60 mm) has been proposed in the advanced fusion reactors like ITER. Different welding techniques will have to be used for such vacuum vessel development. The required mechanical, structural and other properties of stainless steels have to be maintained in these joining processes of components of various shapes and sizes in the form of plates, ribs, shells, etc. The present paper reports characterization of welding joints of SS316L plates with higher thicknesses like 40 mm and 60 mm, prepared using multi-pass Tungsten Inert Gas (TIG) welding process. The weld quality has been evaluated with non-destructive tests by X-ray radiography and ultrasonic methods. The mechanical properties like tensile, bend tests, Vickers hardness and impact fracture tests have been carried out for the weld samples. Tensile property test results indicate sound weld joints with efficiencies over 100%. Hardening was observed in the weld zone in non-uniform manner. Macro and microstructure studies have been carried out for Base Metal (BM), Heat Affected Zone (HAZ) and Weld Zone (WZ). Scanning Electron Microscopy (SEM) analysis carried out for the impact fractured specimens show ductile fracture. The microstructural study and ferrite number data indicate the presence of high content of delta ferrite in the weld zone as compared to the delta ferrite in base metal.     

Design plan and requirement of test module and testing items in IFMIF

The main purpose of present phase of IFMIF/EVEDA (International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities) is to produce a detailed report of IFMIF engineering design with the integrated design of all facilities in IFMIF. The main function of the IFMIF is to give the demanded design database for the licensing of DEMO reactors and further reactors, and it is achieved from the materials data set obtained from the high, medium, and low flux test modules (HFTM, MFTM and LFTM) of IFMIF. In the evaluation using small specimens, developments and guidelines of small specimen test technique or technology (SSTT) are also demanded for the achievements. This paper is summarized and also analyzed about the design plan status and requirements in these test modules from users, and testing items and test methodology in IFMIF.     

Some issues by using the master curve concept

The master curve concept allows to quantify the variation of fracture toughness with the temperature throughout the ductile-to-brittle transition region. Limit curves of fracture toughness for defined failure probabilities and reference temperatures can be determined using this method. Thus, fracture mechanical values can be supplied for an integrity assessment of structural components. This paper presents the application of the master curve concept to the reference temperature determination over the thickness of the reactor pressure vessel (RPV) steel plate. It was shown that the master curve concept is applicable to the fracture mechanical characterisation of material with different microstructures using small test specimens. The influence of the materials homogeneity and the test temperature on the resulting reference temperature was investigated.     

The utilization of thorium for long-life small thermal reactors without on-site refueling

Thorium cycle has many advantages over uranium cycle in thermal and intermediate spectrum nuclear reactors. In addition to large amount of resources in the world which up to now still not utilized optimally, thorium based thermal reactors may have high internal conversion ratio so that they are very potential to be designed as long-life reactors without on-site refueling based on thermal spectrum cores. In this study preliminary study for application of thorium cycle in some of thermal reactors has been performed.

We applied thorium cycle for small long-life high temperature gas reactors without on-site refueling. Calculation results using SRAC code show that 10 years lifetime without on-site refueling can be achieved with excess reactivity of about 10% dk/k.

The next application of thorium cycle has been employed in long-life small and medium PWR cores without on-site refueling. Relatively high fuel volume fraction is also applied to get relatively hard spectrum, small size, and high internal conversion ratio. In the current study we have been able to reach more than 10 years lifetime without on-site refueling for 20–300 MWth PWR with maximum excess reactivity of a few %dk/k.

The last application of thorium cycle has been employed in long-life BWR cores without on-site refueling. Relatively high fuel volume fraction is applied to get relatively hard spectrum, small size, and high internal conversion ratio. In the current study we have been able to reach more than 10 years lifetime without on-site refueling for 100–600 MWth BWR with maximum excess reactivity of a few %dk/k.     

Correlation between irradiation-induced changes of microstructural parameters and mechanical properties of RPV steels

Radiation hardening, displayed by the yield stress increase, and irradiation embrittlement, described by the Charpy transition temperature shift, were experimentally determined for a broad variety of irradiation specimens machined from different reactor pressure vessel base and weld materials and irradiated in several VVER-type reactors. Additionally, the same specimens were investigated by small angle neutron scattering. The analysis of the neutron scattering data suggests the presence of nano-scaled irradiation defects. The volume fraction of these defects depends on the neutron fluence and the material. Both irradiation hardening and irradiation embrittlement correlate linearly with the square root of the defect volume fraction. However, a generally valid proportionality is only a rough approximation. In detail, chemical composition and technological pretreatment clearly affect the correlation.     

Comments on: “Studies on diversion cross-flow between two parallel channels communicating by a parallel slot — Part I” by A. Tapucu

Unique design techniques are needed for low activity ceramic materials in first wall/blanket regions of fusion reactors. A Weibull probabilistic design approach is used to characterize the scatter in the fracture strength and the size effect. Results indicate that ceramic first wall/blanket structures should be modular and each module should be proof tested. The ceramic materials should have high fracture strength, high Weibull modulus, and minimal strength degradation due to subcritical crack growth. The Weibull statistical analysis is coupled with finite element thermal and stress analysis and the probability of failure of ceramic first wall/blanket design concepts is predicted. The usefulness of the approach is demonstrated by optimizing the geometry of the structure to produce minimum probability of failure.     

Design techniques for ceramics in fusion reactors

Unique design techniques are needed for low activity ceramic materials in first wall/blanket regions of fusion reactors. A Weibull probabilistic design approach is used to characterize the scatter in the fracture strength and the size effect. Results indicate that ceramic first wall/blanket structures should be modular and each module should be proof tested. The ceramic materials should have high fracture strength, high Weibull modulus, and minimal strength degradation due to subcritical crack growth. The Weibull statistical analysis is coupled with finite element thermal and stress analysis and the probability of failure of ceramic first wall/blanket design concepts is predicted. The usefulness of the approach is demonstrated by optimizing the geometry of the structure to produce minimum probability of failure.     

Influence of the specimen configuration and the insert material on fracture toughness characterisation with reconstituted specimens

Specimen reconstitution techniques offer the possibility to obtain fracture toughness measurements when only small amounts of material are available. In order to obtain extra information from charpy specimens, an electron-beam weld reconstitution method is established to obtain compact tension specimens (CT) from the broken halves of the charpy ones. Three types of reconstituted CT specimens with different weld configurations are tested in order to analyse the influence of specimen configuration on fracture toughness evaluation. The validity of the fracture toughness characterisation is analysed by comparing J-integral resistance curves (J–R curves) of specimens with insert and those of reference specimens without insert.     

小堆结构材料全生命周期数据管理系统设计研究

小型模块化反应堆（简称小堆）结构材料具有种类繁多、来源广泛和格式多样等特点。基于现代信息技术及大数据背景,结合小堆结构材料数据的特殊性,从材料数据管理角度出发,设计构建了一个覆盖小堆结构材料全生命周期的专用数据管理系统,实现从碎片化数据获取到海量数据集成、处理并融合的转变。整个系统不仅实现了自定义数据库设计,还实现了小堆结构材料全生命周期数据的管理和应用,满足用户数据查询、数据检索、可视化分析等多种需求,有利于推进小堆结构材料数据管理向规范化、智能化发展。同时该数据管理系统突破了多尺度材料数据管理技术瓶颈,增强了材料数据的安全性和可靠性,为数字化小堆研发设计提供了重要支持。      

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

Environmentally assisted cracking (EAC) or, in other words, stress corrosion cracking (SCC) of in-core materials has become an increasingly important reason for the downtime and maintenance costs of nuclear power plants (NPPs). Use of small size specimens for stress corrosion testing of irradiated materials is necessary because handling of high dose rate materials is difficult and the availability of these materials is limited. A drawback of using small size specimens is that they do not in some cases fulfil the requirements of the relevant testing standards and sometimes their limited load-bearing capacity prevents corrosion fatigue tests and tests with static loading at reasonable K_I values. The test results show that the ductile fracture resistance curves of a Cu–Zr–Cr alloy are, to some extent, independent of the specimen geometry and size. However, the curves of small specimens deviate from the curves of larger specimens at high J values (large plastic zone relative to the remaining ligament) or when the crack growth exceeds about 30% of the remaining ligament. The size dependency of the tested Cu–Zr–Cr alloy seems to be a consequence of decreasing stress triaxiality as the size of the specimen is decreased. The results of the SCC tests of sensitized SIS 2333 stainless steel (equal to AISI 304) specimens in simulated boiling water reactor (BWR) water show that the plastic deformation of the remaining ligament of the specimen has no significant effect on the environmentally assisted crack growth rate. This indicates that stress corrosion testing is not limited by the specimen size. The size dependency in SCC tests should be further studied by conducting tests using various specimen sizes.     

Trends in computational intelligence in nuclear engineering

Several changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations.     

Current understanding of radiation-induced degradation in light water reactor structural materials

Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.     

Economic benefits of advanced materials in nuclear power systems

A key obstacle to the commercial deployment of advanced fast reactors is the capital cost. There is a perception of higher capital cost for fast reactor systems than advanced light water reactors. However, cost estimates come with a large uncertainty since far fewer fast reactors have been built than light water reactor facilities. Furthermore, the large variability of industrial cost estimates complicates accurate comparisons. Reductions in capital cost can result from design simplifications, new technologies that allow reduced capital costs, and simulation techniques that help optimize system design. It is plausible that improved materials will provide opportunities for both simplified design and reduced capital cost. Advanced materials may also allow improved safety and longer component lifetimes. This work examines the potential impact of advanced materials on the capital investment cost of fast nuclear reactors.     

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.     

WWER-1000 base metal reference steel and its characterisation

The surveillance programmes of western power reactors include, in many cases, standard reference materials in addition to actual pressure vessel steels. These are specimens cut from standard steel plates (Heavy Section Steel Technology, JRQ, etc.) that are similar in composition and heat treatment to the base material in the respective reactor pressure vessels, and are supposed to serve as a monitor by comparing the radiation embrittlement of the plant-specific material to the reference material, and to detect anomalies in the radiation environment of the surveillance capsules.A correlation monitor material for the eastern WWER-1000 (similar as the JRQ for western reactors) is needed in order to determine the reliability of accelerated data for the validation of reactor pressure vessel surveillance data. Reference materials should be well characterised in terms of irradiation behaviour (transition temperature shift, non-destructive signal, etc.). The magnitude of the sensitivity to irradiation for this material should be measurable for the selected exposures. In this subject the IAEA is launching a new co-ordinated research programme. Material is already manufactured, and the JRC-IE has become its custodian. A detailed plan for characterisation of the reference steel is set up, including irradiation conditions, post-irradiation testing techniques and implementation plan. It is expected the participation of several research institutes worldwide in a round robin, which will allow a better comprehension of WWER-1000 steel's behaviour and will be considered as a benchmarking between different laboratories.The JRC-Institute for Energy in collaboration with the Russian Research Centre – Kurchatov Institute is performing the “as received” material characterisation by both destructive methods and non-destructive techniques.The non-destructive techniques used at the JRC-IE premises are novel methods specially developed for non-destructive assessment of the embrittlement state of materials, as the STEAM method and the measurement of magnetic properties. The STEAM technique (Seebeck and Thomson effects on aged material), is based on the measurement of the Seebeck coefficient. The magnetic properties evaluation is done through Barkhausen noise and permeability measurements.This paper presents a preliminary analysis of the results obtained by all involving laboratories.     

Use of the Master Curve methodology for real three dimensional cracks

At VTT, development work has been in progress for 15 years to develop and validate testing and analysis methods applicable for fracture resistance determination from small material samples. The VTT approach is a holistic approach by which to determine static, dynamic and crack arrest fracture toughness properties either directly or by correlations from small material samples. The development work has evolved a testing standard for fracture toughness testing in the transition region. The standard, known as the Master Curve standard is in a way “first of a kind”, since it includes guidelines on how to properly treat the test data for use in structural integrity assessment. No standard, so far, has done this. The standard is based on the VTT approach, but presently, the VTT approach goes beyond the standard. Key components in the standard are statistical expressions for describing the data scatter, and for predicting a specimens size (crack front length) effect and an expression (Master Curve) for the fracture toughness temperature dependence. The standard and the approach, it is based upon, can be considered to represent the state of the art of small specimen fracture toughness characterization. Normally, the Master Curve parameters are determined using test specimens with “straight” crack fronts and comparatively uniform stress state along the crack front. This enables the use of a single K_I value and single constraint value to describe the whole specimen. For a real crack in a structure, this is usually not the case. Normally, both K_I and constraint vary along the crack front and in the case of a thermal shock, even the temperature will vary along the crack front. A proper means of applying the Master Curve methodology for such cases is presented here.     

A design study on MOX-fueled small fast reactors for standardization of a small fast nuclear reactor system

A way of development to standardize a small fast nuclear reactor system, which is considered one of the suitable concepts at next generation for satisfying such needs as generality, small dependence on natural resources, safety and non-proliferation, is proposed. This process consists of three steps : the first is to demonstrate the basic system within a short period based on current techniques, the second is to achieve greatly higher economy, and the final is to standardize the commercial system that can economically compete with or overcome current light water reactors. A technical investigation is conducted on the performance of a mixed-oxide (MOX)-fueled small fast reactor with a reflector-driven reactivity control system to satisfy the needs at the first step, considering plenty of accomplishments on the MOX fuel and its advantage for limiting the duration of development to the level required at the stage. The results obtained from a series of neutronic and thermal-hydraulic calculations show the feasibility of a small fast reactor that produces the electric power of about 50MW, achieves about two-year consecutive operation with high safety performance and is greatly flexible for updating the system. A mixed-nitride-fueled core is found to be promising past the first stage.