Reactor pressure vessel structural integrity research

Development continues on the technology used to assess the safety of irradiation embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack-tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil-ductility temperature (NDT) performs better than the reference temperature for nil-ductility transition (RT_NDT) as a normalizing parameter for shallow flaw fracture toughness data, (3) biaxial loading can reduce the shallow flaw fracture toughness, (4) stress based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow flaw fracture toughness because in-plane stresses at the crack-tip are not influenced by biaxial loading, and (5) an implicit strain based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation induced shift in Charpy V-notch vs. temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.     

UK developments in structural integrity aspects of NDT

Considerable advances have been made in the field of ultrasonic non-destructive testing over the past few years. In this paper some of these developments are reviewed in the context of the safe operation of a UK pressurised water reactor (PWR) by ensuring the structural integrity of the reactor pressure vessel (RPV).     

Acoustic monitoring techniques for structural integrity

A general review is given of the state of the art of acoustic monitoring and this is illustrated by specific applications including the detection of high pressure fluid leaks from pipelines or storage facilities, the detection of loose parts or rattling tubes in operating plant, and acoustic emission monitoring of crack growth by 3-D source location of secondary acoustic sources caused by fretting of the crack surfaces. An active acoustic method using tomographic techniques to detect anomalous conditions such as temperature ‘hot spots’ is also described.     

Overview of steam generator tube degradation and integrity issues

The degradation of steam generator tubes in pressurized water nuclear reactors, and, in particular, the stress corrosion cracking (SCC) of Alloy 600 tubes, continues to be a serious problem. Primary water SCC is commonly observed at the roll transition zone (RTZ), at U-bends and tube denting locations, and occasionally in plugs and sleeves. Outer-diameter SCC (ODSCC) and intergranular attack (IGA) commonly occur near tube support plate (TSP) crevices, near the tube sheet in crevices, or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of axial and circumferential cracking at the RTZ on both the primary and secondary sides. Outer-diameter stress corrosion cracking in TSP crevices, commonly consisting of segmented axial cracks with interspersed uncracked ligaments, is also becoming more common. Despite recent advances in inservice inspection (ISI) technology, a clear need still exists for quantifying and improving the reliability of ISI methods with respect to the probability of detection of the various types of flaws and their accurate sizing. These improvements are necessary to permit an accurate assessment of the consequences of leaving degraded tubes in service over the next reactor operating cycle. Degradation modes such as circumferential cracking, intergranular attack, and ODSCC at the TSP have affected a large number of tubes. New regulatory guidance is being developed that requires the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes will perform the required safety function over the next operating cycle. In formulating new guidance for the implementation of alternate repair criteria, the U.S. Nuclear Regulatory Commission is also evaluating the contribution to overall plant risk from severe accidents. Preliminary analyses are being performed for a postulated severe-accident scenario that involves station blackout and loss of primary feedwater, in order to determine the probability of failure for degraded tubes.     

Trends in the evaluation of the structural integrity of RPVs

The aim of this paper is to review recent trends, improvements and validations of methodologies for the assessment of reactor pressure vessel (RPV) integrity against the risk of leak or catastrophic failure, mainly deriving from the possible presence of crack-like defects at critical locations in the vessel wall.The first part of the work gives an overview of the input parameters, namely loading conditions, material properties and possible crack shape and dimensions, which are needed for a comprehensive fracture analysis of RPVs, discussing recent findings and still open questions about them.The next two sections are concerned with reviews of the presently available fracture approaches, related to both brittle and ductile fracture behaviour, and of probabilistic fracture mechanics methodologies.As conclusion, present limitations of methodologies for evaluation of RPV structural integrity and areas which need further improvements are outlined.     

Monitor of structural integrity for nuclear power plants

The database management, the load monitoring and the graphical user interface (GUI) are presented. The Combination of database management, load monitoring and GUI forms a useful tool for monitoring the structural integrity of nuclear power plants.     

Research on structural integrity of pressure boundary in light water reactor designs

This paper is a review of the recent researches performed and planned in Japan relevant to the structural integrity of the pressure boundary in light water reactor designs. Various aspects of relevant work on materials, pressure vessel and piping models are described.     

Evaluation method for structural integrity assessment in core disruptive accident of fast reactor

A procedure to evaluate the structural integrity of a reactor vessel under CDA condition has been developed based on simulative experiments under CDA loading using 1/20- and 1/10-scale steel models of the demonstration FBR in Japan (DFBR). For the experiments, a constant pressure source has been developed using a gas accumulator. Simulation analyses were also conducted using the dynamic analysis program AUTODYN-2D, which can simulate fluid-structure interaction and non-linear material behaviour in two-dimensional geometry. The actual dynamic material properties, whose strain rates were from 0.1 to 10 s⁻¹, were used in the analyses. The pressure history and the dynamic response of structures were compared between experimental results and the analysis, and the appropriateness of the numerical analytical method for the CDA problem was confirmed. In both experimental and numerical approaches, the effect of internal structures has been investigated, and a drastic strain reduction was observed owing to the energy absorption by internal structures. Furthermore, material tests and element tests including weld joint tests and burst tests were carried out to define a rationalised strain limit under multi-axial dynamic loading condition. Using the developed evaluation procedure, a preliminary evaluation for DFBR was carried out and the robustness of the reactor boundary in DFBR against CDA loading was also confirmed.     

Advances in PWR structural integrity assessment and materials technology in the United Kingdom

The UK PWR structural integrity and materials programme is considered under four headings: an overview of the programme, the focal points in some recent achievements and finally future work and facilities. This collaborative programme is being undertaken in the main by the United Kingdom Atomic Energy Authority and the Central Electricity Generating Board. The work is classified into five topic areas: fracture properties and testing procedures, metallurgical studies, environmentally assisted cracking, fabrication and repair, integrity methods and validation.The programme is related to the requirements of the project and also depends upon collaboration and exchanges with programmes elsewhere.     

Electric power research institute research in structural design and analysis

The structural research program of the Nuclear Safety and Analysis Department of the Electric Power Research Institute (EPRI) emphasizes an interdisciplinary approach to the study of structural interaction with fluid, soil, or other components. Current efforts involve establishing experimental benchmarks and multidimensional analysis options leading to the verifraction of simplified design methods.     

Prediction of structural integrity of steam generator tubes under severe accident conditions

Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating and design-basis accident conditions are reviewed. These rate-independent flow stress models are inadequate for predicting failure of steam generator tubes under severe accident conditions because the temperature of the tubes during such accidents can reach as high as 800°C where creep effects become important. Therefore, a creep rupture model for predicting failure was developed and validated by tests on unflawed and flawed specimens containing axial and circumferential flaws and loaded by constant as well as ramped temperature and pressure loadings. Finally, tests were conducted using pressure and temperature histories that are calculated to occur during postulated severe accidents. In all cases, the creep rupture model predicted the failure temperature and time more accurately than the flow stress models.     

Proactive approaches to assure the structural integrity of Boiling Water Reactor components

Pressure vessel components in operating Boiling Water Reactor (BWR) plants are subjected to a variety of loading and environmental conditions which could lead to degradation over time. The significant damage mechanisms such as fatigue, stress corrosion cracking (SCC) and irradiation embrittlement are considered in the design basis of the reactor components and thus provide adequate structural margins over the operating life of the plant. Nevertheless, when the design basis assumptions are exceeded, e.g., thermal cycles, vibratory loading or chemistry transients, cracking may occur in pressure boundary components. Several proactive measures are being implemented to address this concern and assure the structural margins in BWR plants. These measures include: (i) control of materials and design to mitigate SCC and improvement of the environmental conditions through the implementation of Hydrogen Water Chemistry, (ii) advances in automated ultrasonic inspection of the BWR pressure vessel and piping, (iii) improved monitoring techniques for tracking fatigue usage and SCC effects in the piping and in the core, and (iv) development and qualification of durable repairs and specialized techniques such as use of high purity materials and temper bead repair. This paper describes current progress in implementing these proactive approaches for Boiling Water Reactors.     

US NRC research on the integrity of piping in light water reactors

The paper describes the activities underway in NRC on the subject of LWR piping integrity as of the summer and fall of 1983. The paper is necessarily vague on certain topics of policy because they are either under review or are under development. Particularly in the area of BWR pipe cracking, events are very rapid so that positions and actions described in this paper may well be obsolete by the time it is published. Nevertheless, this paper is useful to show the intentions of NRC in the area of research for LWR piping, and it is also useful to document the status of the regulations on piping for which the research is being performed.     

Design issues and implications for structural integrity of fusion power plant components

The criteria and implications for successful design, licensing and power plant operation are assessed, and imposed constraints and limitations are examined. The design of a reliable fusion power plant is dependent on the availability of licensed nuclear materials and the structural-thermal loading conditions during normal and abnormal events. Various conditions in a tokamak lead to structural damage and possible failure. Taking into consideration all the possible structural failure mechanisms, the most likely are combinations of fatigue and creep. Issues encountered in the fusion environment are the significant amount of irradiation creep, the large ratio of helium production to displacement damage, and the degradation of fatigue strength and ductility, effects which are even encountered at low temperatures. Design codes distinguish between failure criteria under steady and transient loads, and lay down rules for failure prediction under combined creep-fatigue conditions. Currently, there are no established fusion specific licensing processes or component design codes. Any limits imposed on designs or performance are taken from existing design codes developed by the fission industry. There is a need to initiate the process of defining and developing tools for the design and licensing of fusion components and facilities to ensure nuclear safety.     

Recent Japanese activities for non-destructive examination and structural integrity of reactor components

Nuclear power generation has been spotlighted as a substitute for petroleum. Efforts are being exerted for the development of more reliable nuclear power plants. This paper describes the present status of such development activities in Japan, especially to enhance reliability of non-destructive flaw detection and to improve component integrity through the cooperative efforts of experts in materials, structural design, fracture mechanics, fabrication and non-destructive examination.     

Phase 2 of the international piping integrity research group programme

The results of phase 1 of the International Piping Integrity Research Group (IPIRG-1) programme have been widely reported. The significance of the results is reviewed briefly, in order to put the phase 2 programme into perspective. The success of phase 1 led the participants to consider further development and validation of pipe and pipe component fracture analysis technology as part of another international group programme (IPIRG-2). The benefits of combined funding and of the technical exchanges and interactions are considered to be of significant advantage and value. The phase 2 programme has been designed with the overall objective of developing and experimentally validating methods of predicting the fracture behaviour of nuclear reactor safety-related piping, to both normal operating and accident loads. The programme will add to the engineering estimation analysis methods that have been developed for straight pipes. The pipe system tests will expand the database to include seismic loadings and flaws in fittings, such as bends, elbows and tees, as well as “short” cracks. The results will be used to validate further the analytical methods, expand the capability to make fittings and extend the quasi-static results for the USNRC's new programme on short cracks in piping and piping welds. The IPIRG-2 programme is described to provide a clear understanding of the content, strategy, potential benefits and likely significance of the work.     

Some aspects of work in the U.K. Relating to PWR structural integrity

Until recently much of the effort on structural integrity has been focussed on the pressure vessel. More recently it has been recognised that the pressure vessel does not, by exemption remove the need for attention to be paid to the other components all of which have their interesting features.This overview of U.K. work relating to PWR, gives consideration to the need, requirement and scope of the programs. Finally some specific topics are discussed in detail.Emphasis is placed on the vital issue of the need for validation of assessment procedures and techniques. In this respect there is an element of dependence on overseas programs which have to be continually reviewed in the light of U.K. needs.The programs have been formulated against the background of specific U.K. requirements which are to be met and the vast amount of work being undertaken in other countries. Much of the work is of a generic nature and where specific aspects are being considered they usually have a safety as well as an economic implication.     

Evaluation method of check-valve integrity during sudden closure using thermal-hydraulic and structural analyses

The Advanced Thermal Reactor (ATR), a boiling-water-cooled, heavy-water-moderated, pressure-tube type reactor, has a series of check-valves just upstream of a water drum that serves as a distributing header for fuel channels. In the case of a hypothetical guillotine break upstream of the water drum, the check-valve integrity is a key issue for reactor safety during rapid closure. Pipe break experiments adjacent to and far from the water drum were conducted to measure the deformation of the valve and the water hammer behavior. Fastest disk impact was observed in the case of an adjacent break and water hammer was observed in the case of a break far from the valve. An evaluation method of valve integrity after the rapid closure to be taken over by the simple method is proposed. In the present study, the thermal-hydraulic analysis is conducted using the HITSL code with a dynamic check-valve model, and structural analysis is conducted by the DYNA3D code. Both behaviors relating thermal-hydraulics and structure are traced by these codes.