A comparison of the use of the modified J-integral and the deformation J-integral in procedures for estimating the critical crack length for CANDU pressure tubes

The paper presents the results of a theoretical investigation whose objective has been to see whether there are advantages to be gained from using the modified J-integral in procedures for estimating the critical crack length for CANDU pressure tubes. For typical operation conditions, and with irradiated tubes having critical crack lengths over a wide range, it is shown that the slope of the modified J-integral J_M-Δa crack growth resistance curve for a pressure tube crack is only marginally greater than the slope of the corresponding deformation J-integral J_D-Δa curve; the results are expressed in terms of the parameter Z_*, which is dJ_M/da − dJ_D/da and the parameter Q, which is the fractional difference between dJ_M/da and dJ_D/da. In the light of these findings, there would appear to be little advantage to be gained in using J_M, rather than J_D, as a characterizing parameter for crack growth in a CANDU pressure tube.     

The HTR module pressure vessel unit, design criteria and safety philosophy

A modular-helium-cooled high temperature reactor system for the cogeneration of electricity and process heat has been developed by Siemens—Interatom.Design, manufacture and operation of the pressure vessel unit will conform to German nuclear codes and standards for LWR's, some deviations or peculiarities for their application to HTR's are unavoidable. These are for instance:

• - The main steam nozzle, through which the steam line at 530°C penetrates the steam generator pressure vessel with a nominal design temperature of 350°C.

• - The pressure test concept in which the preservice pressure test will be performed in complete accordance with the codes and standards at 1.3 times the design pressure of 70 bar using water. Afterwards, the presence of graphite structures, ceramic insulation and, of course, the pebble bed core has to be considered. Pneumatic pressure tests are performed at 1.1 times design pressure accompanied by more detailed ultrasonic examinations.

• - The position of operational material irradiation surveillance specimens has to be chosen carefully. Design postulates concerning the incrase of ΔRT_NDT will pe confirmed in a separate program.

In general, the requirements of the assured safety concept, aimed to rule out catastrophic failure of the pressure vessel unit during lifetime are fulfilled.     

Numerical simulation of post yield fracture mechanics experiments as a basis for the transferability to components

As a necessary step in the chain of transferability from small specimens to actual structures the numerical evaluations of two crack-growth resistance experiments on the basis of the J-integral and utilising sidegrooved compact specimens of different sizes, tested at room temperature and at 285°C are discussed. The necessary experimental and numerical techniques are presented:

• -The partial unloading technique as applied in the IWM is applicable with high accuracy and reproducability in the relevant temperature range up to operating temperature.

• -The J-evaluation combined with a node shifting and releasing technique as implemented in the IWM-version of ADINA proved to be a powerful and economic tool even for parameter studies.

The results of the experiments and of the numerical evaluations are presented as force-displacement diagrams and as J-integral vs. crack extension curves. The good qualitative and quantitative agreement supports the experimental evaluation of J from the force-displacement diagram and validitates the numerical procedures to be applied and extended to real structues.

Prediction of fatigue crack growth and experimental result on overlaod retardation

The phenomena of crack growth retardation are frequently observed under variable amplitude or irregular loading fatigue tests. This paper describes a prediction method on crack growth retardation caused by an overload during fatigue loads.The prediction reported in this paper is performed by the following procedure using the yield strength and vs. ΔK relationship of the material.

1. (1) Determination of the residual stress distribution caused by cyclic load and overload based on the Dugdale model.

2. (2) Determination of the effective residual stress intensity factor and effective stress intensity range (ΔK_eff).

3. (3) Prediction of the crack growth rate using ΔK_eff and vs. ΔK relationship of the material.

From the viewpoint to apply the prediction to a structural component, experiments have been carried out on steel pipes with an axial through thickness crack, which are subject to an overpressure during cyclic pressure. In the paper, the experimental results are compared with the prediction.     

The effect of irradiation on the toughness of pressurized water reactor vessel steels under different service conditions

The different toughness tests performed on two pressure vessel steels with very different upper shelves served to make a number of observations concerning the shifts in the transition temperature due to the effect of irradiation, as well as changes in toughness with temperature in the ductile region.With respect to shifts in the transition temperature, the following was observed: the shift obtained with precracked charpy test specimens was narrower than with the others; the shift obtained with charpy V impact tests was substantially equal to that obtained with CT test specimens — wider in the case of steel A, but slightly narrower in that of steel H.With respect to toughness values in the ductile region: the toughness values obtained using precracked charpy test specimens are significantly higher than those obtained with CT test specimens for static tests; 25and 12.5 mm thick CT test specimens display comparable variations in J_1C and dJ/da, but with wide scattering; the effect of irradiation, if any, is of the same order of magnitude as the scattering of the results — however, a test temperature effect is observed; the variation in toughness with temperature is determined more easily by considering a J value corresponding to a stable crack propagation of 1 mm, so that ; this variation of J_Δal with temperature is substantially the same for both steels, or about −30% at 70 or 80°C, and −50% at 290°C.     

Analytical evaluation method of J-integral in creep-fatigue fracture for type 304 stainless steel

Creep-fatigue crack growth at the operating temperature of LMFBR can be characterized by ΔJ_F and J′ (same as C^*). Type 304 stainless steel, the main structural material of the Japanese LMFBR, shows notable cyclic hardening at elevated temperatures. Evaluation of these J-integrals with the finite-element method is strongly affected by the reference strain range when the cyclic hysteresis' is used as the stress-strain relation.In this paper, an evaluation method for ΔJ_F and J′ with a cyclic stress-strain curve (Δσ – Δ relation) is proposed and verified by experimental results. The evaluation method proposed here does not require cyclic calculations but is monotonic and the effect of the reference strain range is relatively small.     

PWR steam generator R&D in France: thermal-hydraulic and vibratory aspects

Concomitant with the launching of the French pressurized water reactor (PWR) nuclear power program, a large research and development (R&D) effort was initiated, devoted to the steam generators (SGs). This program, managed cooperatively by Framatome, the SG designer and manufacturer; Electricité de France (EDF), the French electrical utility; and the Commissariat à l'Energie Atomique (CEA), the French Atomic Energy Commission, primarily responsible for nuclear research; was focused on four main objectives:

1. (1) To obtain a better understanding of the physical phenomena existing in these steam generators and leading to SG performance alterations or operating life reductions.

2. (2) To test and validate improved design solutions for the model 51 Framatome steam generator, which was the first one designed under Westinghouse license.

3. (3) To test and validate new Framatome SG designs.

4. (4) To test and validate new, high-performance design tools.

This vast R&D program covers the following theses:

• - SG thermal-hydraulics,

• - SG tube vibration and wear,

• - SG materials (production, corrosion, etc.),

• - Primary and secondary fluid chemistry,

• - SG technology (manufacturing processes, NDT, etc.),

• - SG in-service inspection, and

• - SG maintenance.

These themes are too numerous to be dealt with in a single article. Consequently, the present article will focus on only the first two themes.     

Evaluation of hazards from industrial activities near nuclear power plants — deterministic and probabilistic studies

Among the potential hazards which could arise from industrial activity near nuclear power plants, fires and explosions of dangerous products are of particular concern. Indeed, thermal radiation from an adjacent fire could endanger the resistance of a plant's structures. Likewise, an accidental explosion would induce an overpressure wave which could affect buildings' integrity.

This paper presents the methodology developed by Electricité de France to evaluate the consequences of accidents affecting:

• - Industrial facilities: refineries, chemical and petrochemical plants, storage areas, pipelines of gaseous, liquid and liquefied materials.

• - Transportation routes (roads, railways, inland waterways) used to carry dangerous substances (solid explosives, liquid, gaseous or liquefied hydrocarbons).

Probabilistic methods have been developed by analysis of actual accident statistics (e.g. risks induced by transportation routes) and realistic and representative accident scenarios have been set up. Five sequences have been identified:

• - Formation of a fluid jet at a breach.

• - Evaporation and possible formation of a liquid layer.

• - Atmospheric dispersion and drift of a gaseous cloud.

• - Heat radiation from fire.

• - Unconfined explosion of a gaseous cloud.

This paper gives an overview of the methods and the main assumptions used to deal with each sequence. Those methods, presently applied by Electricité de France, provide a coherent and realistic approach for the evaluation of the risks at nuclear power plants induced by industrial activity.     

Creep-fatigue crack propagation behavior in a surface cracked plate

A series of experiments were performed in order to clarify the surface crack growth behavior under creep-fatigue condition. Type 304 stainless steel was tested at 550°C and 650°C. Specimens were plates with a surface notch. Loading patterns were axial fatigue, bending fatigue, axial creep-fatigue and bending creep-fatigue. As results were obtained: (1) the beach mark method was available to measure the changes of the crack front shape after the test; (2) the electrical potential method was available to measure the changes of the crack front shape in real time; (3) the crack front shape was affected by the loading mode; and (4) ΔJ and ΔJ_c calculated from the proposed simplified method could characterize the surface crack growth rate.     

Recent design developments in the ASME boiler & pressure vessel code, Section III, Division 1, Nuclear Power Plant Components

This paper will review the significant new design changes and additions to the ASME Boiler & Pressure Vessel Code, Section III, Division 1, Nuclear Power Plant Components. Only those subsections related to the pressure boundary of the vessels will be reviewed for the following parts:

• Subsection NCA — General Requirements

• Subsection NB — Class 1 Components

• Subsection NC — Class 2 Components

• Subsection ND — Class 3 Components

• Subsection NE — Class MC Components

• Appendices

Starting with the current 1983 Edition of Section III, those items which have been added or revised in the five addenda since the edition was issued on July 1, 1983, will be reviewed. Some of the reasons for the additions and changes will be discussed     

European experiences of the proposed ASTM test method for crack arrest toughness of ferritic materials

The proposed ASTM test method for measuring the crack arrest toughness of ferritic materials using wedge-loaded, side-grooved, compact specimens was applied to three steels: A514 bridge steel tested at −30°C (CV30–50°C), A588 bridge steel tested at −30°C (CV30–65°C), and A533B pressure vessel steel tested at +10°C (CV30-12°C) and +24°C (CV30+2°C). Five sets of results from different laboratories are discussed here; in four cases FOX DUR 500 electrodes were used for notch preparation, in the remaining case HARDEX-N electrodes were used. In all cases, notches were prepared by spark erosion, although root radii varied from 0.1–1.5 mm. Although fast fractures were successfully initiated, arrest did not occur in a significant number of cases.The results showed no obvious dependence of crack arrest toughness, K_a, (determined by a static analysis) on crack initiation toughness, K₀. It was found that K_a decreases markedly with increasing crack jump distance, Δα/W. A limited amount of further work on smaller specimens of the A533B steel showed that lower K_a values tended to be recorded.It is concluded that a number of points relating to the proposed test method and notch preparation are worthy of further consideration. It is pointed out that the proposed validity criteria may screen out lower bound data. Nevertheless, for present practical purposes, K_a values may be regarded as useful in providing an estimate of arrest toughness — although not necessarily a conservative estimate.     

A study of fuel behavior in PWR design basis accident: an analysis of results from the PHEBUS and EDGAR experiments

As part of the French PWR safety study programme, fuel behavior during a design basis accident has been investigated in three parallel directions:

• - separate effect tests in the EDGAR apparatus for developing and validating cladding deformation models,

• - integral tests in PHEBUS for verifying codes,

• - development of fuel behaviour codes for plant calculation after assessment against experimental results. After describing the objectives and content of each of these programmes, the main findings are highlighted and discussed.

Observations on the effect of post-weld heat treatment on J-resistance curves of sa-106b seamless piping welds

Ontario Hydro has developed a leak-before-break (LBB) methodology for application to large diameter piping (21, 22 and 24 inch) Schedule 100 SA106B heat transport (HT) piping as a design alternative to pipe whip restraints and in recognition of the questionable benefits of providing such devices. Ontario Hydro's LBB approach uses elastic-plastic fracture mechanics (EPFM).In order to assess the stability of HT piping in the presence of hypothetical flaws, the value of the material J-integral associated with crack extension (J— R curve) must be known. In a material test program J-resistance curves were determined from various pipe heats and four different welding procedures that were developed by Ontario Hydro for nuclear Class 1 piping. The test program was designed to investigate and quantify the effect of various factors such as test temperature, crack plane orientation and welding effects which have an influence on fracture properties. An acceptable lower bound J-resistance curve for the piping steels and welds were obtained by machining maximum thickness specimens from the pipes and weldments and by testing side-grooved compact tension specimens. This paper addresses the effect of test temperature and post-weld heat treatment on the J-resistance curves from the welds.The fracture toughness of all the welds at 250°C was lower than that at 20°C. Welds that were post-weld heat treated showed high crack initiation toughness, J_lc, rising J-resistance curves and stable and ductible crack extension. Non post-weld heat treated welds, while remaining tough and ductile, showed comparatively lower J_Ic, and J-resistance curves at 250°C. This drop in toughness is possibly due to a dynamic strain aging mechanism evidenced by serrated load-displacement curves. The fracture toughness of non post-weld heat treated welds increased significantly after a comparable post-weld heat treatment.The test procedure was validated by comparing three test results against independent tests conducted by Materials Engineering Associates (MEA) of Lanham, Maryland. The J_Ic and J-resistance curves obtained by Ontario Hydro and MEA were comparable.     

Creep and creep-fatigue cracking behaviour of two structural steels

The present study deals with crack initiation and crack growth, not only under creep and creep-fatigue conditions but also under more complex thermomechanical cyclic loadings, in both 316L and 1Cr-1Mo-0.25V steel.In these creep ductile materials, most studies have focused on the creep crack growth rates, da/dt and load-geometry parameter C* correlations. In this paper, the creep crack initiation time is defined as the time T_i necessary for a defect to grow by a small critical distance X_c (X_c ≈ 50 μm for example). This initiation stage may represent a large part of the rupture life of a cracked component. The importance of such studies is discussed in the first part.In the second part, an attempt is made to present a simplified method based on the fracture mechanics of creeping solids to define the relevant load-geometry parameters for crack initiation and crack growth under creep-fatigue loadings. In particular, it is shown that da/dN-ΔK correlations apply only when the hold time t_h is smaller than the transition time t_tr between small-scale and large-scale viscoplasticity. Conversely, for long hold times, it is suggested that the T_i-C* correlation be used to predict the fatigue.     

PISC full scale reactor pressure vessel validation of non-destructive examination

The PISC III Programme involves validation of techniques and procedures and, within this programme, evaluation has now started on the ability to discriminate service induced defects from indications produced by fabrication defects in A 508 Class 2 material when sensitive techniques are used.Action No. 2 of PISC III: Full Scale Vessel Testing is designed for the performance demonstration of three groups of inspection procedures:

• - Mechanized ASME type procedures with variable recording level and complementary techniques

• - Industrial full ISI procedures (mechanized);

• - Several detailed evaluation procedures (generally mechanized) based on advanced techniques to be used on defective areas detected by usual inspection.

These procedures, typical for ISI in most of the cases, are applied in four situations which could be typical of old and new LWR pressure vessels:

• - vessel material and welds containing important service and fabrication defects but mixed with base material defects and small welding defects;

• - nozzle to shell welds with typical service defects, often well isolated and distant from other defective areas in rather clear material and/or welds;

• - nozzle inner radius defects;

• - artificially heat and unbranched fatigue defects in the test blocks assembled to simulate a PWR pressure vessel wall portion.

The paper summarizes the PISC II programme results which stress the characteristics of capable NDT techniques, in opposition to material characteristics like acceptable base material defects. It describes the full scale pressure vessel components available to conduct the PISC III exercise with improved ultrasonic techniques.     

Short-term degradation mechanisms of piping

The operation of PWR-type EDF plants has shown that components have been subjected to loadings higher than the design basis loads. For example, localized degradations on nuclear system pipes were found after relatively short times (10-10⁴ hours). The main damage mechanisms involved are:

• - erosion—cavitation arising from the type of hydraulic flow,

• - vibrational fatigue arising from the flow or operation of mechanical

• - corrosion fatigue occurring in some confined spaces (dead ends).

This paper addresses these damage modes and the mitigating steps taken to cope with them, together with the initiatives taken for future reactors.     

The use of energy absorbers to protect structures against impact loading

When a flying missible impacts a fixed structure, the interface loading is dependent on the deformation characteristics of both impacting and impacted bodies. If both are too rigid to accommodate the amount of gross deformation required to neutralize the incoming kinetic energy, or if such energy absorption has a chance to proceed in uncontrolled and unreliable ways, then there is a need to interpose a specifically designed “energy absorber” between missile and structure, from which a well-defined load time history can be derived during the course of impact.

The required characteristics of such an energy absorption material are:

• the capability to accommodate large permanent deformation without structural failure; and

• the reliable and controlled “load-deformation” (or “stress-strain”) behaviour under dynamic conditions, with an aim at an optimal square shape curve.

Consideration must also be given to environmental or other disturbing effects, like temperature, humidity, and “out of plane” loading. A short survey is presented of the wide range of energy absorbers already described in technical papers or used in a number of practical safety applications within varied engineering fields (from vehicle crash barriers to high energy pipe whipping restraints). However, with such open a literature, information is usually lacking in the specific data required for design analysis.

The following “energy absorption” materials and processes have thus been further experimentally investigated, with an a aim at pipe whipping restraint application for nuclear power plants:

1. (1) plastic extension of austenitic stainless steel rods;

2. (2) plastic compression of copper bumpers; and

3. (3) punching of lightweight concrete structures.

Dynamic “stress-strain” characteristics have been established for stainless steel bars at several temperatures under representative loading conditions. For this purpose, a test rig has been specifically designed to incorporate a number of adjustable parameters and to behave as a representative “slice” of an actual pipe whipping restraint; typical strain rates are in the 10 sec⁻¹ range. The behaviour of copper bumpers has been compared under static and dynamic conditions (using a conventional drop weight test (DWT) machine); as no significant strain rate effects were emphasized, only static tests have been further developed. The DWT rig was used again to investigate crushing or punching of cellular concrete under varying geometries and loading conditions. To remedy certain deficiencies of the regular commercial grades of cellular concrete, special lightweight mixtures have been studied to optimize material toughness and provide a wider range of specific resistance.Results of this experimental program are presented and discussed. The use of energy absorbers is then illustrated for a few typical pipe whipping restraints. The design of restraints is based on real dynamic characteristics of “energy absorption” material as produced by the test program. To derive design loads of restraints, a number of methods can be used ranging from a simplified “energy balance” graph to sophisticated plastodynamic computer analysis. Typical results are presented and discussed to compare the efficiency of these alternative methods.     

Standardized French PWR programs and the application of mechanical analysis

The implementation of the French PWR construction programs is marked by the following factors:

• - the construction of similar plants by series, together with the standardization of plant layout and component design,

• - the various aspects of the plant operation policy adopted by Electricité de France (EdF), the main client,

• - the special provisions of French licensing regulations,

• - the continuous development of technical experience required to support an active export policy.

The effect of these factors on the organization of the programs and on technical achievements will be examined, with special attention being given to the mechanical aspects, which constitute the main subject of the SMiRT conference.