Aging management for long-term interim storage casks

Stress and strain considerations in the design of mechanical components in nuclear power plants

The Korea Atomic Energy Research Institute (KAERI) has been collaborating with Brookhaven National Laboratory since 2007 to develop a realistic seismic risk evaluation system which includes the consideration of aging of structures and components in nuclear power plants (NPPs). This collaboration program aims at providing technical support to a five-year KAERI research project, which includes three specific areas that are essential to seismic probabilistic risk assessment: (1) probabilistic seismic hazard analysis, (2) seismic fragility analysis including the effects of aging, and (3) a plant seismic risk analysis. The understanding and assessment of age-related degradations of structures, systems, and components and their impact on plant safety is the major goal of this KAERI-BNL collaboration. Four annual reports have been published before this report as a result of the collaboration research.

Database for assessing the life expectancy of pressure-retaining components in nuclear power plants

Acoustic sensing technique provides an efficient measure to detect and diagnose incipient failures occurred in plant components. The study reported here investigates the feasibility of adaptive super-directive microphone array technique for detection and localization of anomaly in nuclear power plants. The technique extracts anomaly information of objective components from the acoustic signals obtained at geometrically arranged multiple microphones. A specific signal processing was used to obtain a super-directive sensitivity of the microphone array and to adapt the sensitivity following the acoustic environmental change. Two adaptive sensing algorithms have been developed and implemented in a personal computer, and their abilities to extract the objective acoustic signal have been tested through numerical simulation. For appropriate number of the microphones, the satisfactory performance of the sound extraction has been obtained within reasonable computational load. The practical applicability of this technique to component monitoring in nuclear power plants has been confirmed through the present study.

A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including regular in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the<br/> integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus,<br/> is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage the integrity issues on a nuclear power plant. In this paper, a web-based fatigue life evaluation system for primary components in nuclear power plant is proposed. This system provides engineering knowledge-information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procednres on primary components of a nuclear power plant.<br/>

Ageing of systems and components in nuclear power plants is a well known phenomenon that must be managed to ensure the continued safe operation of these plants. Much of the degradation due to ageing is controlled through periodic maintenance and/or components replacement. However, there are components such as electric cables that do not receive periodic maintenance or monitoring once they are installed. Electrical cable insulations are subjected in service to many operating environmental conditions simultaneously. These include oxidative atmospheres and elevated temperature which cause deterioration due to thermal ageing, degradation of mechanical and electrical properties due to irreversible changes in chemical structure, and thus shortening of insulation service life. Arrhenius model is used for addressing time-temperature ageing effects. The model used in chemical kinetics to predict the rate process has been extended to reliability on the assumption that failure rate is proportional to reaction rate to estimate the Time-To-Failure (TTF) at temperature. Arrhenius model is useful in predicting the life-span and in choosing parameters for accelerated ageing also. In order to allow extrapolation of short-term data to predict long-term performance, appropriate curves must be drawn through the short-term values. The short-term points are obtained by selecting some degree of deterioration (i.e. 50 % retention of Elongation of Break or 75% retention of dielectric strength, or so on), then ageing a given material at several elevated temperatures until the desired extent of degradation is achieved. This paper covers the case study on generating Arrhenius plots by obtaining short-term test data on tensile testing for extrapolation to long-term performance of various types of cables used in operating power plants.

Based on the German decision to minimise transport of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities several on-site storage facilities were licensed until the end of 2003. Because of the large amount of Type B(U) transport casks which are going to be used for long-term interim storage the question of timelimited Type B(U) licence maintenance during the storage period of up to 40 years has been discussed under different aspects. This paper describes present technical aspects of the discussion. A main aspect of qualification of transport casks for interim storage is the long-term behaviour of the metallic seal–lid system. Here results are presented from current long-term experimental tests with metallic 'Helicoflex' seals in which pool water is enclosed. This series of tests has been performed by the Federal Institute for Materials Research and Testing (BAM) on behalf of the Federal Office for Radiation Protection (BfS) since 2001. Finally, the paper presents a German concept for an exchange of experience, know-how and state-of-the-art between authorities and technical experts with regard to cask dispatch in nuclear facilities. BAM has taken over a central role in this so-called 'coordinating institution for cask dispatching information' ('KOBAF') which entails management of an online database of cask-specific documents and a technical working group meeting twice a year. The goal is to keep comparable technical standards for all nuclear sites and storage facilities which are going to load and dispatch casks of the same or similar types under the responsibility of different German state governments for the coming decades.

During design of Class I components in Nuclear Power Plants, cumulative usage factors (CUF) are conservatively calculated to estimate fatigue damage, and results must be below the limits of the applicable codes. Nevertheless, when these results are used to evaluate the possibility of using these components for an extended life, the results are frequently above code limits. Many Nuclear Power Plants have installed commercial fatigue monitoring systems at critical components in order to assess transient severity and cycle count for life extension fatigue calculations among other reasons. Since the commissioning of the system, unexpected operation modes and thermal stratification was discovered and evaluations needed to be done. Findings, interpretations and solving are presented in this paper.

The reactor coolant pump (RCP) is one of the most important components in nuclear power plants. It operated in high temperature, high-pressure, high speed and radiative environment, so a long-term security and reliable operations is required. Many internal flow analysis of RCPs was carried out, mainly foucs on steady and unsteady flow field at different operating points in RCP. The research about flow passage components, such as the blade of the RCP, is relatively few. When the RCPs operates in the nuclear power plant, the flow field lashed against the impeller of the RCP, results in a network of small cracks is found on the surface of impeller. For example, periodic vibration caused by a break in a guide vane leaded to cracking of two pieces blades of impeller in a large power plant in southwest of China, and this accident caused much economic loss. The computational method of stress due to the hydraulic reason is an important problem of the RCP. In this work, at first the CFD simulation including the case, guide vane, impeller, inlet and outlet at different operation points is studied, and the result of the pressure distribution on impeller blade is loaded on the impeller using fluid-structure interation (FSI) method. The result showed that the maximum von Mises stress appears on the trailing edge close to the impeller hub which has a large change in gradient of stress and which is prone to fatigue failure. The maximum stress on the impeller is mainly in proportion to the operating power. The maximum stress on the impeller have periodical characteristic, which is due to the number of blade of diffuser. All of these equip us with better understand of the fatigue and fracture of RCP, and it make sense to protect the fatigue damage and promote the stability of RCP.

Quantification of dislocations densities in zirconium hydride by X-ray line profile analysis

Reactor Pressure Vessel (RPV) is one of the most important components in nuclear power plant (NPP). The aging mechanism of primary concern for RPV is irradiation embrittlement, which can result in a decrease of fracture toughness of RPV steel. Prediction of irradiation embrittlement for a certain Chinese domestic manufactured A508-3 steel is performed. The calculation results given by the US safety standards, the French RCC-M standards and other international safety standards are compared with each other and compared to the data from commercial operation NPP surveillance program. The effect of neutron fluence is also investigated. Furthermore, the property of the steel against irradiation embrittlement is evaluated with the regulatory requirements in the relevant standards. It can be predicted that the Chinese domestic steel satisfies the requirements in these standards.

Extension of the list of load combinations that should be analyzed during the nuclear power plant (NPP) components seismic resistance justification is one of the important new regulatory requirements. The paper deals with the thermomechanical equipment for which seismic resistance assessment is performed using operating experience method, in particular the methodology of indirect assessment using the GIP (Generic Implementation Procedure) for the WWER reactors (GIP-WWER procedure). Review of GIP-WWER procedure has been performed. The paper presents the analysis of seismic screening evaluation work sheet based on the example of motor-operated valve. The disadvantages of the GIP-WWER procedure in terms of ensuring compliance with current regulatory requirements for combinations of operating and seismic loads have been identified and the ways of eliminating the revealed disadvantages have been offered. The supporting structures of the equipment in most cases are the metal structures. Impact of different NPP operating modes to the steel supporting structures of the equipment has been considered and the provisions of building codes governing the calculation of steel support structures have been analyzed. A minimum listing of operating and seismic loads combinations has been established, which should be considered in the justification of the seismic strength of steel support structures referred to the first and second seismic categories. The paper is a part of a series of publications devoted to consideration of regulatory requirements to combinations of operational and seismic loads in assessing the seismic resistance of NPP components.

Many components in conventional and nuclear power plant, aero-engines, chemical plant etc., operate at temperatures which are high enough for creep to occur. These include plain pipes, pipe bends, branched pipes etc., the manufacture of such components may also require welds to be inserted in them. In most cases, only nominal operating conditions (i.e., pressure, temperatures, system load, etc.) are known and hence precise life predictions are not possible. Also, the proportion of life consumed will vary from position to position within a component and the plant. Hence, nondestructive techniques are adopted to assist in making decisions on whether to repair, continue operating or scrap certain components. One such approach is to use scoop samples removed from the components to make small creep test specimens, i.e., sub-size uniaxial creep test specimens, impression creep test specimens, small punch creep test specimens, and small ring (circular or elliptical) creep test specimens. Each specimen type has its own unique advantages and disadvantages and it may not be obvious which one is the most appropriate test method to use. This paper gives a brief description of each specimen and associated test type and describes their practical limitations. The suitability of each of the methods for determining “bulk” material properties is described and it is shown that an appropriate test type can be chosen.

Corrosion-fatigue (CF) is widely recognized as one of the most complex and deleterious degradation mechanisms threatening the structural integrity and long-term operation (LTO) of primary circuit components in Nuclear Power Plants (NPPs). As global nuclear fleets age and transition toward Subsequent License Renewal (SLR) periods of 80 years, the limitations of traditional, empirical fatigue models—which often rely on conservative, air-based design curves—have become increasingly apparent. This research presents a comprehensive quantitative assessment framework that bridges the gap between theoretical materials science and practical structural health monitoring by integrating high-fidelity numerical simulations with advanced experimental datasets. The study focuses on quantifying the synergistic effects of cyclic mechanical loading and the aggressive electrochemical environment of high-temperature, high-pressure reactor coolant, specifically targeting the vulnerabilities of 316L stainless steel and Alloy 690.The methodology utilizes a dual-pillar approach: first, conducting rigorous fatigue tests in a high-pressure recirculating autoclave loop to establish empirical crack initiation and growth rates ( ) under simulated reactor water chemistry. These experiments employ Direct Current Potential Drop (DCPD) sensors for sub-millimeter precision in real-time crack monitoring. Second, these physical results are synthesized into a Multiphysics Finite Element Analysis (FEA) framework that incorporates the Slip-Dissolution/Oxidation Model to simulate the localized electrochemical conditions at the crack tip. A key finding of this research is the identification of a critical "threshold" in the loading waveform, where tensile rise times exceeding 45 seconds were found to accelerate environmental damage by a factor of 15 compared to fast-transient cycles.Furthermore, the study quantifies the impact of Manganese Sulfide (MnS) inclusions, demonstrating that 85% of cracks initiate at micro-pits with a critical depth-to-width ratio of 0.8. By applying Monte Carlo simulations, the research establishes a probabilistic distribution for the Remaining Useful Life (RUL) of critical components, identifying that fluctuations in local strain rates account for 45% of total predictive variance. The ultimate output of this study is a calibrated "Digital Twin" framework that achieved a 98% accuracy in real-time crack growth prediction. These results provide a technically defensible foundation for risk-informed maintenance, suggesting that proactive, data-driven monitoring can safely extend the operational life of nuclear assets while maintaining the highest margins of structural safety and regulatory compliance.

Since the mid-90s, GRS has been dealing systematically with the assessment of the influences of ageing on the safety of nuclear power plants. This also involves the generic evaluation of operating experience. In the following, the procedure, methodological approaches and relevant results are presented in an exemplary manner for passive mechanical components. The results show that the number of events influenced by ageing phenomena that have so far occurred in German plants is low and that measures were taken where necessary to detect, monitor and control any changes due to ageing. As new insights may emerge over time concerning the assessment of the ageing behaviour of the components, it is considered necessary to continue with the ongoing generic evaluation of operating experience in order to obtain timely indications of any safety-relevant changes.