Nuclear Power Plants Age Research Articles

A Study on One-Time Inspection for Aging Management of Operating Nuclear Power Plants

One-time inspection is used to confirm the effectiveness of nuclear power plant aging management programs such as water chemistry, fuel oil chemistry, and selective leaching program that is designed to prevent or minimize aging degradation during the plant operation. This paper represents one-time inspection experience of selected components and susceptible locations within the scope of water chemistry and fuel oil chemistry aging management programs. Prior to performing one-time inspection, sample sizes of components to be inspected were determined based on combination of materials, environment and aging effect. Examination techniques and relevant operating experience were analysed to obtain technical basis for further implementation of one-time inspection aging management programs. Existing correlation curves between sample size and total population number were reviewed to prepare further inspections for entire operating plants and ensure safety level of components subject to one-time inspection program.

Flow induced vibration measurement of a full-length sodium fast reactor fuel bundle

Nuclear technology continues to be an integral part of the world’s integrated energy infrastructure to provide sustained carbon-free emissions form of energy. As existing nuclear power plants age, and newly conceived nuclear reactors are developed there are several phenomena that continue to limit the operations and lifetime of these systems including thermal fatigue, irradiation material degradation, and flow induced vibration. Of the new nuclear power plant companies, TerraPower is proactively addressing each of these phenomena through use of advanced design approaches and aggressive testing programs. This study details the outcomes resulting from the experimental testing program that supports design aspects of a sodium fast reactor fuel bundle and implications of flow induced vibration. Specifically, this study adds to the body of knowledge through three contributions: (1) The outcomes of this study result in the first-ever experimental data collected with sufficient temporal and spatial measurement resolution to validate multiphysics computational tools related to flow-induced vibration; (2) phenomenological comparisons of fluid–structure interactions are made between a straight ducted fuel bundle and a diluted duct fuel bundle; (3) pin-to-pin interactions are observed as they relate to different configurations within a single fuel bundle. The hydraulic characterization is compared against that of experimental data collected in previous studies to verify relevance and credibility of boundary conditions. The outcome of this work disseminates a new understanding of the vibrational response of a full-length SFR fuel bundle under representative hydraulic conditions to the community.

Hybrid Monte Carlo/deterministic activation calculation to support the decommissioning of PWRs: Validation against data from the thermal shield of the Enrico Fermi NPP

The progressive aging of nuclear power plants (NPPs) in the world is a relevant issue for the nuclear industry.The radiological characterization of the reactoris usually performed using a combination of sampling campaigns and computer simulations, and the latter are particularly useful if measurements are limited.This paper describes the development of a four-step, hybrid Monte Carlo/deterministic calculation scheme to evaluate the residual activation of a pressurized water reactor, focusing the analysis on the Enrico Fermi power plant in Italy. The scheme performance is evaluated using the ratio between the calculated specific activities and the measured ones (C/M ratio) for each point and isotope considered. The overall performance is satisfactory, as the scheme provides conservative results with a good accuracy, without excessive overestimations. The present computational scheme can be adapted to study the lifetime of a NPP, thus providing a useful support tool for the dismantling process in NPPs.

원자력발전 고방사선구역 소구경 노즐에 대한 자동화검사 기술

원자력발전소의 노후화에 따른 이종금속용접부에서 결함 발생 사례가 지속적으로 보고되고 있으며, 국내에서도 원자력발전소 가동연수 증가에 따라 이종금속부에 결함 발생도 점점 증가할 것으로 판단된다. 국내에서는 증기발생기 배수 노즐 이종금속용접부, 원자로냉각재계통(RCS) 고온관 시료 채취 노즐에서 결함 사례가 보고되었고, 원자력발전소의 인력 접근이 제한적인 고방사선구역 내 소구경 노즐의 이종금속용접부에 대한 검사 시 인력 투입을 최소화 하는 신뢰성 있는 자동화 비파괴검사 가능 기술 및 시스템 개발 필요성이 증대되고 있다. 본 연구에서는 원자력발전소의 소구경 노즐 결함 검출을 위해 1) beam simulation을 통한 최적 검사 탐촉자 설계, 2) 소구경 이종용접부 검사용 multi-directions UT 최적 검사기술, 3) 원격제어 automatic inspection system을 개발하였으며, 표준결함시편을 이용하여 개발된 기술 및 시스템의 결함 검출능을 검증하였다. 개발된 최적기술과 시스템은 실제 발전소에서 발생된 결함 (RCS 고온관 시료 채취 노즐)에 대한 검사에 적용하여 결함을 검출함으로써 개발 검사 기술 및 시스템의 적합성이 검증되었다. 개발된 기술은 원전의 다양한 소구경 건전성 평가에 활용함으로써 원자력발전 안전성 확보에 기여할 수 있을 것으로 기대된다. Defects in dissimilar metal welds are reported to be on the increase during the operating lifespan and aging of nuclear power plants. In Korea, reported cases of defects due to dissimilar metal welds include the drain nozzle of a steam generator and RCS hot tube sampling nozzles. Therefore, there is an urgent need to develop a reliable automated nondestructive inspection technique and a system for the inspection of dissimilar metal welds of small diameter nozzles in a high radiation area of a nuclear power plant. In this study, to ensure effective defect inspection of small diameter nozzles (RCS high-temperature tube sampling nozzle) of a nuclear power plant, three different methods were developed. These include: (1) optimum inspection probe design by beam simulation, (2) multi-directions UT optimum inspection technique for the inspection of small diameters of different welded parts, and (3) remote control automatic inspection system. The developed technique and systems have been verified to be suitable for use in the inspection of defects in smaller diameter nozzles in nuclear power plants.

Frequency domain reflectometry for remaining useful life estimation of instrumentation and control cables

As the existing fleets of nuclear power plants age, a greater emphasis has been placed on technologies to quantify the remaining useful life of safety-critical components and to ensure that these systems will operate as intended when called upon, especially in emergency situations. Some of the most important, yet often overlooked components, are the cables that provide the signal paths for instrumentation and control systems used to ensure safe and efficient operation of nuclear power plants. Instrumentation and control cable aging, in particular, is primarily concerned with degradation of the polymer material by thermal oxidation while exposed to heat, humidity, radiation, and other environmental stressors. Consequently, as the polymer becomes embrittled, it cracks and becomes susceptible to moisture intrusion that can cause shorts and shunts in the cable circuits. This article describes the development of two prognostic models to estimate the remaining useful life of nuclear power plant cables. The first model uses data compiled in an existing polymer aging database as background for introducing the remaining useful life model development methodology. The second model uses data generated from accelerated cable aging experiments performed by the authors and explores the use of in situ electrical measurements such as frequency domain reflectometry for predicting the remaining useful life of instrumentation and control cables typically found in nuclear power plants.

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.

Slow strain rate tensile tests on irradiated austenitic stainless steels in simulated light water reactor environments

Irradiation stress corrosion cracking (IASCC) is a critical degradation mechanism for reactor internal components that contributes to the safe and economic operation of reactors. As nuclear power plants age and irradiation dose increases, IASCC becomes an increasingly important issue because of its potential impact on the integrity of reactor internal components. In this study, slow strain rate tensile tests were conducted on irradiated tensile specimens at strain rates between 3 and 7×10−7s−1 to evaluate cracking susceptibility of austenitic stainless steels in simulated light water reactor (LWR) environments. Significant increases in yield strength were observed for all irradiated specimens and a dose dependence of irradiation hardening was obtained at temperatures relevant to LWRs. Ductility and strain hardening capability were also found decrease rapidly with the increase of dose. After the tests, the specimens were examined using a scanning electron microscopy to characterize fracture morphology. The area fractions of non-ductility fracture were used to evaluate the IASCC susceptibility along with the tensile properties. IASCC susceptibility was also compared for several stainless steels irradiated in the Halden and BOR-60 reactors. A possible neutron spectrum effect was discussed.

Numerical Simulation of Thermal Stratification in Cold Legs by Using OpenFOAM

*,† During a small-break loss-of-coolant accident in pressurized water reactors (PWRs), emergency core cooling system (ECCS) is actuated and cold water is injected into cold legs. Insufficient mixing of injected cold water and hot primary coolant results in thermal stratification, which is a matter of concern for evaluation of pressurized thermal shock (PTS) in view of aging and life extension of nuclear power plants. In this study, an open source CFD software, OpenFOAM, is used to simulate mixing and thermal stratification in the cold leg of ROSA/LSTF, which is the largest thermal-hydraulic integral test facility simulating PWR. One of the cold-leg is numerically simulated from the outlet of primary coolant pump to the inlet of downcomer. ECCS water is injected from injection nozzle connected at the top of the cold leg into the steady-state natural circulation flow under high-pressure and high-temperature conditions. The temperature distribution in the cold leg is compared with experimental and FLUENT’s results. Effects of turbulent flow models and secondary flow due to the elbow section of the cold leg are discussed for the case with the single-phase natural circulation. Injection into a two-phase stratified flow is also simulated and predictive and numerical capabilities of OpenFOAM are discussed.

Strategic aspects of aging nuclear infrastructure: game theory and the Chernobyl event

Game theory can provide an analytical framework for investigation of the strategic conflicts surrounding critical infrastructure. Immediately after the 1986 Chernobyl event, for example, plant management, political appointees and world media struggled over the timing and accuracy of information release. The aging of nuclear power plants and the increasing difficulties faced by management and regulatory agencies in assuring high plant reliability virtually guarantee the recurrence of such contests. Standard operating procedures and conservative plant design were once considered sufficient to guarantee plant safety, but after the 1979 Three Mile Island event management practice was also recognised as crucial. We contend that crisis behaviour can be best understood in the context of the strategic interaction among regulatory agencies, plant management and the media. Strategic issues lead naturally to game-theoretic models, which focus on the interaction of action choices and information. The Chernobyl event provides an illustrative case study to show that game analysis can provide insights into decision-makers' best responses to nuclear accidents.

Research Activities for Nuclear Power Plant Aging Promoted by PLEC, JAPEIC, Japan

In order to perform research activity for aging countermeasure of nuclear power plant effectively, Plant Life Engineering Center (PLEC) was established in Japan Power Engineering and Inspection Corporation (JAPEIC) in April 2000 sponsored by Ministry of International Trade and Industry (MITI, presently METI). Outlined activities of PLEC are as follows. Results of technical survey for research and development for aging phenomena have been summarized in a table (Research Map) categorizing them into “inspection and monitoring”, “evaluation method for aging” and “preventive maintenances and refurbishment”. Necessary research themes have been extracted from the Research Map consulting to experts of the specified research area and they are summarized into Medium and Long-term Research Perspective (Research Perspective), which contains prioritized research themes and outlined specification of each theme. Several new research themes proposed by various organizations and selected by PLEC as effective for the regulation activities are identified and proposed to be funded by METI every year. This paper also provides outlines and obtained results of aging related research projects currently conducted by JAPEIC sponsored by METI.

ICONE11-36580 Research activities for nuclear power plant aging promoted by PLEC, JAPEIC, Japan

In order to perform research activity for aging countermeasure of nuclear power plant effectively, Plant Life Engineering Center (PLEC) was established in Japan Power Engineering and Inspection Corporation (JAPEIC) in April 2000 sponsored by Ministry of Economy, Trade and Industry (METI, called as MITI at that time). Results of technical survey for research and development for aging phenomena have been summarized in a table (Research Map) categorizing them into "inspection and monitoring", "evaluation method for aging" and "preventive maintenances and refurbishment". Necessary research themes have been extracted from the Research Map consulting to experts of the specified research area. Medium and long-term research perspective (Research Perspective) has been established which contains prioritized research themes and outlined specification of each theme. Several new research themes proposed by various organizations and selected by PLEC as effective for the regulation activities of METI are identified and proposed to be funded by METI every year. There are about ten on-going research programs funded by METI. Their progress and performance are evaluated annually to improve their efficiency including their alteration, abolition and integration. This cycle of research is going to be attained successfully. Technology Advisory committee composed of members from various field of nuclear power including prefectural and municipal governments supervises the PLEC activity to concentrate national wide potentials and to secure transparency, openness and neutrality. This paper also provides an outline of the aging related research projects currently conducted by JAPEIC under the auspices of METI.

Department of energy’s annealing prototype demonstration projects

One of the challenges utilities face in addressing technical issues associated with the aging of nuclear power plants is the long-term effect of plant operation on reactor pressure vessels. These vessels are exposed to neutrons during the operation of a reactor. For certain plants, this exposure can cause embrittlement of some of the vessel welds, which can shorten the useful life of the vessel. This reactor pressure vessel embrittlement issue has the potential to affect the continued operation of a number of US pressurized water reactor plants. However, the properties that are degraded by long-term irradiation can be recovered through a thermal annealing treatment of the vessel steel. Although a dozen Russian-designed and several US military vessels have been successfully annealed, US utilities concluded that an annealing demonstration using a US reactor pressure vessel was a prerequisite before annealing a licensed US nuclear power plant. In May 1995, the Department of Energy and Sandia National Laboratories initiated a program to evaluate the feasibility of annealing US licensed plants using two different heating technologies. One team completed its annealing prototype demonstration in July 1996, using an indirect gas-fired furnace at the uncompleted Public Service of Indiana’s Marble Hill nuclear power plant in southern Indiana. The second team’s annealing prototype demonstration using a direct heat electrical furnace at the uncompleted Consumers Power Company’s nuclear power plant at Midland, Michigan, was scheduled to be completed in early 1997, but has now been delayed indefinitely. This paper describes the Department of Energy’s annealing prototype demonstration program and the results to date for each project.

Design concepts for the reactor protection and control process instrumentation digital upgrade project at the Donald C. Cook Nuclear Plant Units 1 and 2

As the nation's nuclear power plants age, the need to consider upgrading of their electronic protection and control systems becomes more urgent. Hardware obsolescence and mechanical wear out resulting from frequent calibration and surveillance play a major role in defining their useful life. At Cook Nuclear plant, a decision was made to replace a major portion of the plant's protection and control systems with newer technology. This paper describes the engineering processes involved in this successful upgrade and explains the basis for many decisions made while performing the digital upgrade.

Slowing the aging of nuclear power plants

As nuclear power plants need more maintenance as they age. However, the reactor core also gets hotter. The operator of an aging nuclear plant is caught between conflicting obligations: to increase the maintenance of an ever hotter facility, yet keep the exposure of each worker to an unprecedentedly low level. The solution chosen by the US utility industry is chemical decontamination-more specifically, decontamination of the system that cools the nuclear reactor, namely, the full primary, or reactor coolant, system. The process is under way at the Indian Point 2 pressurized-water reactor plant of the Consolidated Edison Company. The process consists of removing radioactive oxide deposits from the cooling system using a two step process using alkaline permanganate and Can-Derem. The authors describe the process and its use at Indian Point 2.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Good practices for effective maintenance to manage aging of nuclear power plants

Maintenance is the primary means of detecting and correcting the effects of aging degradation at nuclear power plants. The effectiveness of the maintenance program has a direct impact on the operational safety of nuclear power plants. As part of the Nuclear Regulatory Commission NPAR program, the effectiveness of maintenance programs to manage aging were examined for four different governmental and industrial organizations. These were: the U.S. commercial aviation industry, the U.S. Navy ballistic submarine program, the U.S. Air Force B-52 bomber program, and the Japanese nuclear power industry. The purpose was to identify effective maintenance practices that could be adapted by commercial nuclear power plants to assist in managing equipment aging. The key maintenance areas discussed in this paper include the development of systematic maintenance programs, the application of appropriate and effective preventive maintenance tasks and inspections, the collection and evaluation of performance and operating records to assess the effectiveness of the maintenance programs, and the use of feedback and corrective actions to improve the maintenance programs. Also examined is the issue of diminishing manufacturing sources.

Irradiation-assisted stress corrosion cracking as a factor in nuclear power plant aging

Irradiation-assisted stress corrosion cracking (IASCC) is a form of intergranular stress corrosion cracking (IGSCC) that has occurred infrequently during the last two decades in in-core components of nonsensitized austenitic stainless steel. IASCC is apparently a time-dependent phenomenon, which does not manifest itself until a minimum residence time or threshold fast neutron fluence has been attained in an appropriate material/environment combination. Until now, IASCC has been observed in components that are readily removable, such as fuel rod and absorber tube cladding. The problem is being mitigated in the BWR by alloy impurity control. Hydrogen water chemistry is another highly promising remedy currently under qualification. An extensive effort is underway to improve our understanding of IASCC.

International conference on nuclear power plant aging, availability factor, and reliability analysis

International conference on nuclear power plant aging, availability factor, and reliability analysis