Saurindranath Majumdar

Materials integration issues for high performance fusion power systems

One of the primary requirements for the development of fusion as an energy source is the qualification of materials for the frost wall/blanket system that will provide high performance and exhibit favorable safety and environmental features. Both economic competitiveness and the environmental attractiveness of fusion will be strongly influenced by the materials constraints. A key aspect is the development of a compatible combination of materials for the various functions of structure, tritium breeding, coolant, neutron multiplication and other special requirements for a specific system. This paper presents an overview of key materials integration issues for high performance fusion power systems. Issues such as: chemical compatibility of structure and coolant, hydrogen/tritium interactions with the plasma facing/structure/breeder materials, thermomechanical constraints associated with coolant/structure, thermal-hydraulic requirements, and safety/environmental considerations from a systems viewpoint are presented. The major materials interactions for leading blanket concepts are discussed.

ITER structural design criteria and their extension to advanced reactor blankets

Applications of the recent ITER structural design criteria (ISDC) are illustrated by two components. First, the low-temperature-design rules are applied to copper alloys that are particularly prone to irradiation embrittlement at relatively low fluences at certain temperatures. Allowable stresses are derived and the impact of the embrittlement on allowable surface heat flux of a simple first-wall/limiter design is demonstrated. Next, the high-temperature-design rules of ISDC are applied to evaporation of lithium and vapor extraction (EVOLVE), a blanket design concept currently being investigated under the US Advanced Power Extraction (APEX) program. A single tungsten first-wall tube is considered for thermal and stress analyses by finite-element method.

Performance limits for fusion first-wall structural materials

Key features of fusion energy relate primarily to potential advantages associated with safety and environmental considerations and the near endless supply of fuel. However, high-performance fusion power systems will be required in order to be an economically competitive energy option. As in most energy systems, the operating limits of structural materials pose a primary constraint to the performance of fusion power systems. In the case of fusion power, the first-wall/blanket system will have a dominant impact on both economic and safety/environmental attractiveness. This paper presents an assessment of the influence of key candidate structural material properties on performance limits for fusion first-wall blanket applications. Key issues associated with interactions of the structural materials with the candidate coolant/breeder materials are discussed.

Designing against low-cycle fatigue at elevated temperature

Abstract Factors that are important in determining low-cycle fatigue damage at elevated temperature are discussed. The linear damage rule for computing creep-fatigue damage is shown to be unsatisfactory in many situations. The damage-rate equations developed at Argonne National Laboratory have been generalized to include multiaxial creep-fatigue under complicated loading histories. Available creep-fatigue data under combined axial-torsion loading can be explained in a consistent manner by the damage-rate approach.

Effects of phase geometry and volume fraction on the plane stress limit analysis of a unidirectional fiber-reinforced composite☆

Abstract Limit theorems of plasticity are applied to unidirectional fiber-reinforced composite materials to determine bounds to plastic limit conditions for the composite in average stress space. For this purpose a representative volume element (RVE), sufficiently large compared to the scale of inhomogenity of the composite, is chosen and analyzed by limit analysis methods. Upper and lower bounds to the plane average stress limit condition for the composite are derived for the following geometries which are presented in descending order of generality. 1. (a) Only volume fraction of phases are known. 2. (b) Volume fraction and fiber cross-section shape are known, but the size and distribution are arbitrary. 3. (c) Deterministic periodic array. The bounds are shown to be improved at the cost of generality of fiber geometry and distribution. An evaluation is made of the effects of fiber volume fraction and distribution geometry on the composite limit condition.

Importance of strain rate in elevated-temperature low-cycle fatigue of austenitic stainless steels☆☆☆

Abstract Extrapolation of elevated-temperature, tensile-hold fatigue life of types 304 and 316 stainless steel is obtained by the use of four existing life predictive methods. The results show that, although the calculated lives for the different methods are similar for short hold-time tests, they can vary greatly from one method to another when extrapolated to long hold-time situations. Methods that do not take into account the effects of strain rate provide optimistic values as opposed to the more pessimistic values projected by the methods that account for strain-rate effects.

Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach

At present, the fatigue life of nuclear reactor components is estimated based on empirical approaches, such as stress/strain versus life (S∼N) curves and Coffin-Manson type empirical relations. In most cases, the S∼N curves are generated from uniaxial fatigue test data, which may not truly represent the multi-axial stress state at the component level. Also, the S∼N curves are based on the final life of the specimen, which may not accurately represent the mechanistic time-dependent evolution of material behavior. These discrepancies lead to large uncertainties in fatigue life estimations. We propose a modeling approach based on evolutionary cyclic plasticity that can be used for developing finite element models of nuclear reactor components subjected to multi-axial stress states. These models can be used for more accurately predicting the stress-strain evolution over time in reactor components and, in turn, fatigue life. The model parameters were estimated for 316 stainless steel material, which are widely used in U.S. nuclear reactors. The model parameters were estimated for different test conditions to understand their evolution over time and their sensitivity to particular test conditions, such as the pressurized water reactor coolant condition.Copyright

Status report on assessment of environmentally assisted fatigue for LWR extended service conditions

.......................................................................................................................................i List of Figures ..............................................................................................................................v List of Tables ...............................................................................................................................x Abbreviations ...............................................................................................................................xi Acknowledgements ......................................................................................................................xii

Pressurized Water Reactor Environment Effect on 316 Stainless Steel Stress Hardening/Softening: An Experimental Study

In USA there are approximately 100 operating light water reactors (LWR) consisting fleet of both pressurized water reactors (PWR) and boiling water reactors (BWR). Most of these reactors were built before 1970 and the design lives of most of these reactors are 40 years. It is expected that by 2030, even those reactors that have received 20 year life extension license from the US nuclear regulatory commission (NRC) will begin to reach the end of their licensed periods of operation. For economical reason it is be beneficial to extend the license beyond 60 to perhaps 80 years that would enable existing plants to continue providing safe, clean and economic electricity without significant green house gas emissions. However, environmental fatigue is one of the major aging related issues for these reactors, and may create hurdles in long term sustainability of these reactors. To address some of the environmental fatigue related issues, Argonne National Laboratory (ANL) with the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program trying to develop mechanistic approach for more accurate life estimation of LWR components. In this context ANL conducted many fatigue experiments under different test and environment conditions on 316 stainless steel (316SS) material that is or similar grade steels are widely used in US reactors. Contrary to the conventional S∼N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to understand material ageing more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help to develop computer based advanced modeling tools to better extrapolate stress-strain evolution of reactor component under multi-axial stress states and hence to help predicting their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In another paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.Copyright

Online Stress Corrosion Crack Monitoring in Nuclear Reactor Components Using Active Ultrasonic Sensor Networks and Nonlinear System Identification: Data Fusion Based Big Data Analytics Approach

The current state of the art nondestructive evaluation (NDE) techniques used in nuclear reactor structural inspection are manual labor intensive, time consuming, and only used when the reactor has been shut down. Also, despite periodic inspection of plant components, a failure mode such as stress corrosion crack can initiate in between two scheduled inspections and can become critical before the next scheduled inspection. In this context, real time monitoring of nuclear reactor components is necessary for continuous and autonomous monitoring of component structural health. In this research, an active ultrasonic based on-line monitoring (OLM) framework is developed which can be used for real-time monitoring of degradation of nuclear power plant systems, structures, and components. Nonlinear system identification technique such as Bayesian Gaussian Process technique method is investigated to estimate the structural degradation in real-time. Active broadband ultrasound input is used for damage interrogation and a multi-sensor configuration is implemented to improve spatial resolution of state estimation. The damage index at any particular time is computed using nonlinear techniques such as Gaussian Process probabilistic modeling and the necessity of sensor data fusion is evaluated. The framework was demonstrated through the monitoring of an anomaly trend in a nuclear reactor steam generator tube undergoing stress corrosion cracking (SCC) testing.Copyright