GUIDE to the PERIODIC INSPECTION of NUCLEAR REACTOR STEEL PRESSURE VESSELS the Following States Are Members of the International Atomic Energy Agency

TECHNICAL REPORTS SERIES No. 99

Guide to the Periodic Inspection

Steel Pressure Vessels

J INTERNATIONAL ATOMIC ENERGY AGENCY.VIENNA, 1969

GUIDE TO THE PERIODIC INSPECTION OF NUCLEAR REACTOR STEEL PRESSURE VESSELS The following States are Members of the International Atomic Energy Agency:

AFGHANISTAN GHANA PAKISTAN ALBANIA GREECE PANAMA ALGERIA GUATEMALA PARAGUAY ARGENTINA HAITI PERU AUSTRALIA HOLY SEE PHILIPPINES AUSTRIA HUNGARY POLAND BELGIUM ICELAND PORTUGAL BOLIVIA INDIA ROMANIA BRAZIL INDONESIA SAUDI ARABIA BULGARIA IRAN SENEGAL BURMA IRAQ SIERRA LEONE BYELORUSSIAN SOVIET ISRAEL SINGAPORE SOCIALIST REPUBLIC ITALY SOUTH AFRICA CAMBODIA IVORY COAST SPAIN CAMEROON JAMAICA SUDAN CANADA JAPAN SWEDEN CEYLON JORDAN SWITZERLAND CHILE KENYA SYRIAN ARAB REPUBLIC CHINA KOREA, REPUBLIC OF THAILAND COLOMBIA KUWAIT TUNISIA CONGO, DEMOCRATIC . LEBANON TURKEY REPUBLIC OF LIBERIA UGANDA COSTA RICA LIBYA UKRAINIAN SOVIET SOCIALIST CUBA LIECHTENSTEIN REPUBLIC CYPRUS LUXEMBOURG UNION OF SOVIET SOCIALIST CZECHOSLOVAK SOCIALIST MADAGASCAR REPUBLICS REPUBLIC MALAYSIA UNITED ARAB REPUBLIC DENMARK MALI UNITED KINGDOM OF GREAT DOMINICAN REPUBLIC MEXICO BRITAIN AND NORTHERN ECUADOR MONACO IRELAND EL SALVADOR MOROCCO UNITED STATES OF AMERICA ETHIOPIA NETHERLANDS URUGUAY FINLAND NEW ZEALAND VENEZUELA FRANCE NICARAGUA VIET-NAM GABON NIGER YUGOSLAVIA GERMANY, FEDERAL REPUBLIC OF NIGERIA ZAMBIA NORWAY

The Agency's Statute was approved on 23 October 1956 by the Conference on the Statute of the IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957. The Headquarters of the Agency are situated in Vienna. Its principal objective is "to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world".

Permission to reproduce or translate the information contained in this publication may be obtained by writing to the International Atomic Energy Agency, Karntner Ring 11, P.O. Box 590, A-1011 Vienna, Austria.

Printed by the IAEA in Austria October 1969 TECHNICAL REPORTS SERIES No. 99

GUIDE TO THE PERIODIC INSPECTION OF NUCLEAR REACTOR STEEL PRESSURE VESSELS

REPORT OF A MEETING ON PERIODIC INSPECTION OF NUCLEAR REACTOR STEEL PRESSURE VESSELS HELD IN STOCKHOLM, SWEDEN 21 - 25 OCTOBER 1968

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1969 GUIDE TO THE PERIODIC INSPECTION OF NUCLEAR REACTOR STEEL PRESSURE VESSELS IAEA, VIENNA, 1969 ST.I/DOC/LO/99 FOREWORD

In accordance with the recommendation made by the Panel on Reactor Pressure Vessels, held by the International Atomic Energy Agency in 1966 at Pilsen, Czechoslovak Socialist Republic, and because of the growing importance of the subject, the IAEA organized a meeting of specialists on the Periodic Inspection of Nuclear Reactor Steel Pressure Vessels, from 21 to 25 October 1968, in Stockholm, Sweden. The enthusiastic response to this meeting, indicated by the presence of 45 specialists from 16 Member States of the Agency, demonstrated the importance of this topic. The purpose of the meeting was to afford an opportunity for the exchange of experiences in discussions on various factors affecting the periodic inspection and finally to reach an agreement on the draft of a document which could serve as a general guide for the inspection of power reactor steel pressure vessels. This guide is intended only to provide suggestions which may be followed or changed according to the particular case. It is hoped that it will be not only a useful reference work but will also serve as a starting point for more advanced work on the subject. The IAEA would like to thank the authorities in Sweden, the host country, who helped to arrange the meeting, the drafting committee, under the chairmanship of Mr. O. Kellermann, and Mr. E.C. Miller for guiding the discussions in a most productive way.

CONTENTS

INTRODUCTION 1

GUIDE TO THE PERIODIC INSPECTION OF NUCLEAR REACTOR STEEL PRESSURE VESSELS 3

1. Introduction 3 2. Factors contributing to vessel failure potential 5 3. Inspection requirements and planning for inspection 8 4. Methods of inspection 10 5. Inservice inspection programs 15 References 15

RECOMMENDATIONS 19

APPENDIX

PROPOSAL FOR RECURRING INSPECTIONS OF NUCLEAR STEEL PRESSURE VESSELS 23 by O. Kellermann and A. Tietze

Selected material from draft ASME CODE FOR INSERVICE INSPECTION OF NUCLEAR REACTOR COOLANT SYSTEMS 31

LIST OF PARTICIPANTS ' 59

INTRODUCTION"

This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards.. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be in- corporated into the early stages of design and inspection planning. How- ever, much of the contents of the Guide are also applicable, in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in an advanced stage of design or construction. The Guide comprises: A general discussion of factors, conditions, and potential failure mechanisms, for which recurring inspection should be provided; Brief descriptions of inspection techniques at present available and applicable to periodic inspection, with some evaluation of their relative effectiveness and reliability; A list of source references for material contained in the Guide and related material of general interest containing still more extensive bibliographic reference material; Recommendations for further activity; and An Appendix consisting of two examples of documents which are in the process of being considered for adoption as national codes or standards in their respective countries of origin. It is recognized that some phases of the subject matter of the Guide, in particular some inspection techniques and in-service monitoring systems, have received little attention in the text because they have not yet been developed to the point of being readily available for immediate application in reactor inspection programs. However, efforts have been made to provide suitable references to such potentially useful methods and systems. In at least one case - periodic overpressure testing - the material in the Guide may more nearly correspond to a compromise than a definite statement, but perhaps this reflects a general lack of agreement among specialists,, not limited to the participants in this meeting, who have strong but divergent opinions on the subject.

Provided by the Chairman of the meeting, Mr. E.C. Miller.

GUIDE TO THE PERIODIC INSPECTION OF NUCLEAR REACTOR STEEL PRESSURE VESSELS

1. INTRODUCTION

In October 1966, the International Atomic Energy Agency sponsored a meeting of a Panel of experts in Pilsen, Czechoslovak Socialist Republic, on Recurring Inspection of Nuclear Reactor Steel Pressure Vessels. The papers, discussions and recommendations of this Panel were published in IAEA Technical Report No. 81 [ l] . The Panel recommended, among other things, that a second meeting organized by the IAEA be held on the same subject, preferably in 1968, and that the IAEA, with assistance from the 1968 Panel and other sources, should prepare a Guide To The Periodic Inspection of Nuclear Reactor Steel Pressure Vessels. This Guide, and a recently published bibliography [ 2] , are the results of those recommendations.

1.1. Scope

This Guide serves as an "aide de memoire" to authorities concerned with the safe working of nuclear pressure vessels. It does not necessarily illustrate the minimum inspection requirements. The authority concerned should make a balanced choice from those suggestions in this guide which are applicable to the nuclear pressure vessel under consideration. Examples of such developed codes or standards are given in the Appendix. The Guide is intended to apply to the periodic, recurring, or continuing inspection of operating power reactor vessels of types now being built or planned for completion in the next few years. It attempts

(a) To identify the conditions for which inspection should be provided, based on consideration of the various potential modes of vessel failure or of materials properties deterioration; (b) To review and evaluate the applicability and effectiveness of available or proposed inspection methods and programs; (c) To recommend methods and controls which appear applicable, • practicable and necessary to monitor and detect possible failure mechanisms; (d) To recommend areas requiring further study, evaluation, research, or experimental development, including probabilities of the oc- currence of some types of failure modes and means for their prevention or amelioration; potential failure mechanisms for which available inspection techniques appear inadequate and for which the method should be improved or alternates developed; relatively new inspection techniques that hold promise for reactor vessel monitoring, but which are still in the developmental stages.

The Guide does not attempt to suggest the application of inspection controls to cover catastrophic acidents such as criticality incidents — ex-

3 4 GUIDE TO INSPECTION

plosions, core'failures, operating malfunction —-which can produce temperatures, pressure and dynamic forces far beyond those for which a reactor vessel, can be designed or built to withstand. The Guide recognizes that contributing factors can originate in auxiliary equipment pumps, valves, piping, pressurizers, turbines, heat exchangers, or reactor components control rod drives, fuel elements, core structure, resulting from corrosion, crud formation bearing wear, vibratory fatigue, or failure of electrical or mechanical control devices, which may contribute to vessel failures; however, the inspection recommendations in this Guide will be limited to the pressure-containing reactor vessel and those parts which are integral parts of the vessel.

1.2. Applicability of the Guide

It is expected that well over 200 central station electric power-gener- ating nuclear reactor steel pressure vessels of the types to which this Guide — subject to appropriate modification — may be applicable will be in operation throughout the world by 1975. These include steel pressure vessels for boiling-water, pressurized-water, and, to a limited extent, gas-cooled reactors. Also, many, but not all the recommendations will also be applicable to other types, generally with substantial modification in details.

1.3. Need for periodic inspection [ 1, 3]

Reactor vessels operate in intensely radioactive environments and are generally surrounded by massive concrete or other shielding, structural,, and insulating materials. These place severe limitations on accessibility for inspection from the external surfaces, particularly after the reactors have operated for a time. Similarly, the relatively fixed internal shielding and core support structure, the fuel element core and control rod complex, and the irregular, corrosion-tarnished, and radiation-activated weld overlay-clad surfaces severely restrict access to the vessel internal surfaces for direct or remote visual examination of even those surfaces which can be reached. The difficulty of access and other problems resulting from environment, design and operating characteristics can make periodic inspection of reactor vessels, using methods available and generally used for non-nuclear vessels, a difficult, only partially effective, and sometimes almost impossible operation. But these problems, plus the potentially serious consequences of vessel failure, make it even more imperative that programs should be established to provide for periodic or recurring inspection. No single set of rules can apply to all vessels, materials, and types of vessels; and this Guide is intended only to provide recommendations which can serve as a basis on which inspection programs or rules can be developed for individual reactors or reactor types.

1.4.' Information required [l]

1.4.1. Failure modes and design The establishment of an effective inspection program requires several types of basic information. These include understanding of the various potential vessel failure modes for which inspection should be provided, and of methods of design or operation which can reduce or delay 5 GUIDE TO INSPECTION

failure mechanisms and thereby lessen to some extent the frequency and extent of inspection coverage required.

1.4.2. Manufacturing history The complete history of the manufacture of the reactor pressure vessel should include reports regarding

(a) Testing and control, beginning with testing of material and ending with the final pressure test after erection. (b) Findings during the examination of material, welding, pressure testing, and total number, temperature and times of heat treatment cycles and other processes during manufacture. (c) Defects found and their location. Indication of whether or not repair is done and the way of evaluating repairs. (d) Observations during pressure test such as strain measurements, acoustic detection, etc. (e) Results from testing of samples of the material which have been exposed, as nearly as possible, to the same treatment as the material in the vessel during fabrication. These samples should be prepared from plates or other.product forms used in the vessel, preferably as prolongations of plates, forgings, welds and heat- affected zones. These samples as full thickness sections should be subjected concurrently to all heat treatments used on the plate and the fabricated pressure vessel. The samples should include base metal, welds and heat-affected zones. (f) "Fingerprinting" of the complete vessel should be done after the final acceptance pressure test in the field. These fingerprints should use the same methods, and the same inspection tools intended for later periodic inspection. 'This should provide for a possible comparison between the condition of the vessel in.the original state and in the future. In some cases it may be more convenient to conduct the pre-operational examinations for fingerprints in the shop, provided the method and techniques are the same as would be expected to be used later; this assumes no changes would be expected from shipping, erection and pre-operational tests, and that adequate records are provided.

2. FACTORS CONTRIBUTING TO VESSEL FAILURE POTENTIAL [1]

Identification of potential causes of vessel deterioration or failure, and, to the fullest possible extent, estimating the relative probability of failure resulting from such causes, are essential to determine the conditions for which inspection protection must be provided and to select appropriate inspection methods. Most of the factors listed below have caused or con- tributed to failures of comparable engineering structures, including pressure-containing equipment, not necessarily nuclear. In evaluating these factors it should be recognized that the probability of failure by some of the mechanisms listed is relatively remote, and that some mechanisms may have already been expected or provided for in design or fabrication; nevertheless, each such factor should be specifically reviewed for its possible contribution to failure. The relative improbability of reactor 6 GUIDE TO INSPECTION vessel failure by some of these mechanisms is such that the need for added inspection coverage becomes debatable, particularly when cost considerations tend to obscure the technical considerations of safety. Unfortunately, precise statistical failure probabilities cannot yet be assigned to operating power reactor vessels; in fact, most reactor incidents to date have resulted from causes or combinations of factors which were either unexpected or were thought highly unlikely to occur. The failure mechanisms of principal concern in periodic inspection control generally result from a combination of more than one of the factors listed below, and they may operate either concurrently or in sequence. The fundamentals and details of these mechanisms and factors are described in the Analysis and Discussion Sections of the 1966 IAEA Panel Report [ 1] and in the other References; they will not be repeated here.

2.1. Structural factors [ 4] Several factors listed below that are involved in the initial construction of a vessel may cause or contribute to failure. Most gross inadequacies will be detected in the course of fabrication, acceptance inspection or preoperational testing, but these controls are not necessarily infallible; further, some aspects of design, the demands placed on vessels in service, the extent of existing knowledge of these factors, and the state of vessel design and construction are such that unrecognized conditions can exist in the construction of any vessel which may contribute to deterioration or eventual failure. 2.1.1. Design deficiencies These deficiencies can include under-design or under-dimensioning, excessive localized stress concentrations, and the use of materials with insufficient mechanical properties or undetected inclusions or flaws located at points of critical stress concentration in the finished vessel. 2.1.2. Fabrication flaws In the course of vessel fabrication, several factors —mechanical working, heat treatment, welding'—can open up material flaws, generate cracks,' change properties of materials, develop points of localized stress concentration and introduce residual stresses. Such flaws can contribute to failure under the influence of operating environmental conditions, mal- operation in service and unexpected malfunctions.

2.2. Service deterioration of vessel materials [4] Deterioration may involve actual loss of structural material, changes in mechanical and metallurgical properties, stress-corrosion or hydrogen embrittlement. 2.2.1. Loss.of structural material High velocity coolants can corrode, or erode, or both, causing loss of metal by localized impingement or cavitation. Similarly, loose or .broken internals, .corrosion crud, or foreign objects may block coolant flow channels or, by vibration, abrade cladding, structural or core materials. These are factors which should largely be anticipated in design and fabrication control, but they also have sufficient probability to warrant continuing or recurring inspection control. 7 GUIDE TO INSPECTION

2.2.2. Changes in mechanical properties Relevant properties at room, operating temperatures and possible overheat conditions, such as tensile strength, yield strength, toughness and ductility, may be modified to varying extents by such 'mechanisms as overheating, strain aging, strain hardening, thermal or mechanical fatigue [5-14] , cyclic stressing, stress corrosion [15-17] , hydrogen embrittlement [17-19] and radiation effects [ 20-22] . The possibilities of each of these factors contributing to failure, alone or in combination with others, should be evaluated to determine the extent and type of inspection coverage required.

2.3. Mechanisms of vessel failure [ 4] Contributing causes of failure like those listed above (but not including major reactor system accidents or malfunction) are generally expected to operate by some relatively slow, progressive mechanism involving gradual loss of structural material (Sub-section 2.2.1), deterioration in metal properties (Sub-section 2.2.2), or actual growth of flaws by fatigue, local overstressing, residual stresses, .thermal stresses due to temperature differentials, particularly during startup and shutdown or scrams, and stress corrosion cracking or hydrogen cracking. These progressive mechanisms may continue to the point that some other mechanism takes over and triggers final failure by another mode, e.g.: (a) A simple overstress burst-type failure due to reduction in metal cross-section (Sub-section 2.2.1) or to reduction in metal strength properties (Sub-section 2.2.2). (b) Fast fracture [ 23-26] . When crack size, stress level and material toughness reach certain critical values, which are inter-related, a crack which previously has been stable will become unstable and propogate at' a speed and to an extent that is governed by the energy supply rate and the total amount of energy involved respectively. The critical size of the crack may be generated by various mechanisms, one of which is the subcritical growth of a fabrication flaw by the action of fatigue. The material toughness may be influenced by the reactor environment in such a way as to reduce the critical crack size for a particular stress level, and thus consideration may need to be given to the provision of suitable specimens in the material surveillance programs. The results of tests on such surveillance specimens can be interpreted through the use of a suitable theory of fracture analysis, the particular type of theory used depending on the type of steel. • • (c) Lessened vessel integrity due to failure or malfunction of associated equipment: corrosion crud formation resulting in blockage of fuel element cooling channels and deposition on heat transfer surfaces; stress corrosion cracking of austenitic stainless steel parts of vessels or auxiliaries such as control rod drive components, piping, nozzle safe-ends, core structures and core supports; . bolting relaxation and failure, fatigue, high loads, stress corrosion; 8 GUIDE TO INSPECTION

mechanical breakdown of auxiliary equipment —. control rod drives, pumps, valves, etc.;

fuel element ruptures.

3. INSPECTION REQUIREMENTS AND PLANNING FOR INSPECTION

The factors which can contribute to vessel deterioration and possible ultimate failure are many and varied. Fortunately, their net effects can be classed in a relatively small number of groups, such as those listed below, to establish the feasibility of inspection controls and to select the controls to be used. It should be recognized that adequate, practical and applicable inspection methods are not readily available for all these requirements. The individual methods of inspection and their limitations are discussed in greater detail in Section 4. ' The control, detection and compensation for failures of associated equipment and their consequences fall outside the scope of this Guide; nevertheless, there should be ample system performance monitoring provisions, redundancy in auxiliary components, and frequent testing of instrumentation circuits, • nuclear protective noise analysis, and boiling detection devices [ 30] . "••.••'

3.1. Loss of structural material

• This can usually be evaluated by a suitable combination of periodic visual examinations, dimensional inspection, and resonance or other ultrasonic techniques.

3.2. Changes in mechanical properties

3.2.1. Metallurgical effects of stress and temperature The subtleties of the metallurgically induced changes, in tensile properties and toughness, by such mechanisms as strain aging, temper embrittlement and radiation damage, may be evaluated by highly precise and sophisticated measurement of changes in physical properties rather than by generally established inspection methods; however, these are best accomplished on small specimens which are not readily adaptable to the mass of a reactor vessel or to parts with a complex geometry. This suggests the possible use of surveillance test, samples, but these create the further problem of duplicating the stress-temperature-environment parameters of the reactor; valid simulation of the stress conditions constitute the principal problem. It is possible that continued research and development on more precisely controlled ultrasonic methods and newer techniques such as acoustic spectrometry [31-33] may eventually establish the feasibility of direct measurement of property changes in operating reactor vessels, but they do not appear to be applicable with full confidence at present.

3.2.2. Neutron irradiation effects This subject has been actively and extensively studied in recent years, and partially standardized surveillance test programs have been established'for most reactors and reactor types going into service. These attempt to establish changes in tensile properties, notch impact toughness, 9 GUIDE TO INSPECTION and transition temperature behaviour of vessel materials (including welds and heat affected zones) as functions of accumulated fast neutron dosage at various stages of expected reactor operating life. These programs will contribute much valuable information to be taken into account in the near future. Radiation effects and surveillance problems and programs have been well covered in the literature, although the problems are by no means completely solved [ 20-22] . 3.2.3. Corrosion and hydrogen embrittlement [15-19] Although frequently considered as separate phenomena, the two appear to be closely related insofar as embrittlement is concerned. Hydrogen can be introduced into steel by general corrosion, electrolytic or galvanic action, contact with dissociated hydrogen ions, hydrogen produced by radio- lytic decomposition coupled with other environmental factors — pressure, temperature and chemistry of the corrodant. Once introduced into the iron lattice, hydrogen diffuses readily through it, and, if the hydrogen ions are sufficiently concentrated in the lattice, can embrittle the steel, assuming appropriate conditions of temperature and other environmental factors. Then the atomic hydrogen may be recombined and trapped as molecular hydrogen in laminations, cracks and other metal discontinuities. The hydrogen embrittlement may be a temporary phenomenon and, except for the molecular hydrogen accumulated in metal discontinuities, the hydrogen dissolved in the lattice dissipates readily when the source of hydrogen is removed. Another mechanism involves combination with carbon to form methane which accumulates in metal discontinuities, but this occurs at temperatures well above those of operating water-cooled reactors and will not be further considered here. Most stress-corrosion cracking — which is presumably hydrogen-related or hydrogen-assisted. occurs in stainless steels rather than in the lower strength alloys used in reactor vessels. Also, the structural steel in reactor vessels is generally clad with a layer of corrosion-resistant austenitic stainless steels, minimizing hydrogen pick-up.

3.3. Cracks and crack growth Assuming some progressively degrading mechanism operating under otherwise normal operating conditions of pressure and temperature, i.e. no major accident conditions such as explosions or violent pressure, temperature, or nuclear excursions, failures can be of two or three types: (a) The extension of a crack by some mechanism such as fatigue or stress-corrosion until it breaks through the pressure barrier and results in'a simple leak;' (b) Fast fracture, either brittle or ductile which requires that the material have a low fracture energy, and that it contain a crack of a critical size, this critical size being dependent on the applied stress and the fracture energy of the material at the metal temperature; however, unless the upper shelf energy is unusually low and the crack is located in a region of particularly unfavourable geometry, the critical crack size for a ductile tear may, in many cases, be large enough to enable leakage to occur well before fast fracture is likely to occur. The common factor in these types of failure is that.the vessel contains a growing crack. (The word 10 GUIDE TO INSPECTION

crack is used here in a broad sense to include any type of "crack- like", "two-dimensional" or "planar" type discontinuity.) It is conceivable that some mechanism such as radiation may progessive- ly reduce the fracture energy, and, with it, the required "critical crack size" to the size of the existing crack. Consequently, it is essential in establishing a continuing or recurring inspection program that provision be made to identify existing cracks or discontinuities, to estimate their size and orientation and to monitor their growth. A principal objective of a considerable number of non-destructive test methods is the detection and evaluation of cracks, although the methods differ in their abilities to identify cracks in different locations and of different geometries and orientations. The more commonly used non-destructive test methods for crack detection which appear to have application include visual examination, penetrant testing, magnetic particle testing, ultrasonic inspection and radiography. Electromagnetic methods are being developed which may have some application.

3.4. Leakage [ 30]

The timely detection of leakage is valuable in preventing larger failures. Continuous monitoring arrangements may be used for this purpose. In- spections during pressure tests should be made to locate leakage and its sources.

3.5. Vibrations

Vibrations of reactor internals may lead to their working loose or otherwise damaging the vessel. Pre-operational vibration measurements have been conducted on some reactors, but some of the difficulties have been the location of accelerometers and variation of test conditions from operating conditions.

4. METHODS OF INSPECTION

4.1. Design for accessibility

The need for accessibility for inspection depends on inspection tools and inspection methods which can be applied effectively. The conduct of an effective inspection requires accessibility to the parts or surfaces of interest, consistent with the characteristics and safety demands of the inspection methods used. Accessibility for repair is also essential, but its implementation lies outside the scope of this Guide. While some frequent periodic inspections will require only limited access to vessel surfaces, major inspections conducted at much longer intervals, or a possible later need for initially unexpected inspection, will probably require substantially complete and reasonably ready access to the vessel pressure boundary. The initial requirement, then, is that the vessel, its enclosure, and its internals be designed and built to provide sufficient access space to conduct inspections with existing planned inspection techniques, and with a minimum of inconvenience consistent with effective functional design. In the case of the external structure, shielding, and insulation, this may require perma- 992 GUIDE TO INSPECTION nent annular spaces round the vessel and nozzle extensions, or readily removable sections of these externals. The above suggestions regarding access are very general ones and subject to modification for each type of vessel; the important thing is that design should take into account the need for ready access to critical portions of the vessel for inspections by appropriate methods.

4.2. Nondestructive test methods [ 2-34]

The methods described below are the relatively common, conventional methods which are generally classified under this heading, although some of the pulications in the References are not thus limited. They will be discussed here primarily from the standpoint of their applicability and reliability within the specific context of this Guide. In considering such test methods, two essential points should be recognized and emphasized:

(a) Nondestructive test methods present flaws, cracks, defects or discontinuities as signals or indications, and not as direct repre- sentation of the exact size, type and orientation of the flaws, although competent operators can, by the exercise of manipulative skill, judgement and experience, make reasonably accurate descriptions and estimates of flaw size. (b) No one test is ideal for all types and orientations of flaws, and a combination of several tests will be far more effective in locating and defining a broader range of defects and evaluating their signifi- cance.

In this section of the Guide, visual examination will be treated as another nondestructive test method, which it is in fact. However, in suggested outline programs, such as in the Appendix, it is treated separately because of the much greater frequency with which it is used.

4.2.1. Visual examination [28] This is the most widely used tefet for evaluating soundness, surface condition, cleanness and cracks. While generally done with the unaided eye, sometimes special lighting, mirrors, magnifiers, horoscopes, remote television and fibre optics are used for remote examination. Visual examination is generaly fast, but results vary with the inspector1 s eyesight, judgement and experience. There are limitations to visual examination of the clad surfaces of reactor vessels because of tarnish or discoloration which can mask indications and radioactive contamination perniitting only remote visual examination, so that a meaningful examination may require that the surfaces be thoroughly cleaned and decontaminated. Penetrant, magnetic particle and some of the more sophisticated reflection and refraction methods sometimes serve the purposes of "visual inspection" or are a useful ad- junct. Dimensional measurement may also be considered to be a type of visual inspection.

4.2.2. Radiography [ 28] Except for visual examination, this is the oldest and most widely used method for nondestructive examination of welds. Within limits it records the internal soundness of materials, and is used to detect cracks and other discontinuities. It is widely used during fabrication, but because of the 12 GUIDE TO INSPECTION masking effect of radioactivity from a reactor which has been in operation, it is difficult to use in post-operational inspection. Its most serious limitation is perhaps the fact that it is only effective for cracks and planar discontinuities that lie within about 5° on either side of the radiation beam.

4.2.3. Ultrasonic examination [28] This is probably the most effective technique available at present for location of cracks and crack growth in recurring inspection of reactors. It requires skill, ingenuity, and judgement, but a really capable operator can locate cracks at almost any depth in the vessel wall, and can provide a reasonably accurate estimate of the size and the progressive relative growth from one examination to the next. Internal surface irregularities of the vessel, and metallurgical variables such as coarse-grained structure and clad interface, reduce its effectiveness. Special techniques may be required to conduct effective examination of complex areas and flaws near the surface. Considerable attention is being currently devoted to the development of special techniques, including remotely manipulated scanning devices [ 29] . This holds promise for periodic examination of important parts of reactor vessels, such as highly stressed nozzles, weld seams and weld overlay cladding. Current trends indicate that, assuming adequate provisions for access from, outside or inside, meaningful volumetric examination of many important parts should be feasible. Current ultrasonic techniques may be broadly classified into single probe and multiple probe methods. Single probe techniques and double probe techniques are well established in non-nuclear applications,, and are being considered for nuclear inspections. The single probe technique uses one transducer for transmitting and receiving. The multiple probe technique uses one or more transmitting transducers together with one or more receiving transducers.

4.2.4. Magnetic particle examination .[ 28] Magnetic particle examination is effective in detecting cracks and other linear-type surface or shallow sub-surface discontinuities in ferromagnetic materials. It can give false indications on rough surfaces; also, where prod contacts are used, arcing can damage the surface being examined.

4.2.5. Liquid penetrant examination [28] This, like magnetic particle examination, is sometimes considered as an aid to macroscopic or visual examination. It may be used on either magnetic or non-magnetic material, but is limited to cracks and other discontinuities that are open to the surface. The surface examined must be clean and dry. Surface roughness and variations in technique may make it difficult to differentiate spurious indications from actual defects.

4.2.6. Models or replications of surfaces In cases where surface examination, particularly by remote or limited access viewing, suggests or establishes that cracks or other surface flaws exist, it may be appropriate to make plastic model replications to examine in greater detail by direct visual observation. Once such a replication is made, it can be retained for comparison with replications at the same location at a later date. 13 GUIDE TO INSPECTION

4.2.7. Electric resistance examination Methods have been developed to measure metal thickness and detect cracks at or near the surface.

4.2.8. Other methods Methods and techniques employing acoustics, electromagnetics, infra- red, halography, optics and special electric-electronic phenomena are being developed. A constant watch should be kept for new methods and techniques to be used in inspection.

4.3. Surveillance programs [20-22]

As mentioned above, it is general practice at present to place a considerable number of tensile bars, impact, and possibly other specimens in appropriate locations in the reactor where they are subjected to the same or perhaps a higher level of fast neutron flux as the vessel material itself. These specimens are usually taken from cutouts or cut edges of plates, forgings, welds and other materials used in the actual vessel construction. In some instances reactor components may be tested. A representative number of samples are removed and tested at periodic intervals to determine radiation-induced changes in mechanical toughness, transition temperature and possibly other properties of the vessel materials. The use of surveillance programs to study the engineering aspects of irradiation embrittlement of reactor pressure vessel steels was the subject of a Consultants Meeting held under the auspices of the International Atomic Energy Agency in Vienna on 2-4 October 1967. Two of the resulting papers contain excellent reviews of the general subject of irradiation embrittlement and surveillance programs on the one hand [20], and the conclusions and recommendations of the working group on the other ( 21] . It appears desirable to continue surveillance monitoring of the effects of radiation on vessel materials as they are placed in service1, but where it can be established that two or more reactors are essentially duplicates in design, materials and operating conditions, consideration can perhaps be given to modification of this program, if there is reason for confidence that the results of the surveillance program from one vessel are truly representative of the others. As new information develops and new problems arise, it may be appropriate to modify the program by also monitoring changes in other mechanical and physical properties resulting from fast neutron irradiation or from other contributing factors in reactor environments, as mentioned in connection with hydrogen embrittlement, stress-corrosion cracking and strain-aging.

4.4. Overpressure testing [ 35-40]

An initial overpressure test is regularly performed on most pressure vessels upon completion of fabrication, as a leak test, verification of strength, a method of accomplishing stress redistribution, and a means of revealing improper design, gross flaws, faulty materials or suscepti- bility to fast fracture. Re-testing by overpressurizing is often required for non-nuclear vessels, gas cylinders and pipelines as a condition for

1 An existing surveillance test standard, E 185-66, has been issued by the American Society for Testing and Materials, and can be used as a possible guide. 14 GUIDE TO INSPECTION continued operation. This may be a regular scheduled inspection or a check after maintenance or repair operations. Periodic design-pressure or overpressure tests for nuclear vessels have been suggested and employed; it is very clear, however, that there is no unanimity on the desirability of such a test. Diametrically opposing views have been expressed. Credit has been claimed for improvements to the properties and integrity of the vessel by variously described mechanisms that have been given a variety of names. This test has been claimed to be a demonstration of the ability of the pressure vessel to operate for an extended period [ 36, 37, 40] . On the other hand, it has also been suggested that overpressure tests may lead to deformation of a vessel beyond dimensional limits and occasion mechanical or metallurgical damage to the structure and embrittlement due to strain aging or straining alone [ 38, 39] . Concern has been expressed for possibly premature failure of a vessel which otherwise might operate safely for its lifetime. There are opposing views regarding the effect of reactor operating temperature and the influence of the environment on the properties of the steel after an overpressure test. It has been claimed that the possibility of brittle fracture may be avoided by performance of the test at a temperature above the transition range. Judgement on these issues, in individual cases, is therefore a question of balancing benefits and risks.

4.5. Acoustic emission systems [41-46]

Major efforts are being made to develop passive systems which depend on the self-generation of sonic signals by release of acoustic stress waves when metal deformation or crack propagation takes place. Detectors may be placed at various distances from the stress wave source to detect energy releases and locate their points of origin by triangulation techniques. Most of these systems have been developed for applications other than the monitoring of flaw growth in reactor vessels, but some of them are developed to the point that they are routinely .and successfully used in aerospace applications and in monitoring the performance of rotating machinery. The process appears applicable to the continuous monitoring of a vessel for any crack growth during operation. It may be used in the initial overpressure test and in any recurring pressure re-tests during reactor shutdowns. Referring to Sub-section 4.4, if a periodic overpressure test is established as a requirement for continuing operation of a reactor vessel, a stress wave emission system used in conjunction with the overpressure test could be an appropriate measure. By acoustically monitoring the vessel as the test pressure increases, the pressure could be raised to the planned level providing no crack extension signal developed, or, if such warning signals did develop before reaching the intended pressure, the overpressurization could be discontinued upon the appearance of the warning signal.

4.6. Leak detection [ 30]

Other types of pressure vessel monitoring besides those mentioned above have been proposed and used, particularly methods which can detect leaks in service, through the presence of excessive moisture or fission products, or by sonic identification, among others. Such systems are reasonably common and will not be further described or referred to here; however, it should be recognized that an aim in efforts to avoid catastrophic 15 GUIDE TO INSPECTION

fracture is to establish conditions such that if failure does occur, it will be a leak, not a break, and that this requires that leaks, when they appear, be readily recognized and identified.

5. INSERVICE INSPECTION PROGRAMS

Typical inservice inspection programs have been proposed by the USAS N-45 ad-hoc committee, AEC and ACRS (Draft ASME Standard - "Code for Inservice Inspection of Nuclear Reactor Coolant Systems") and the TUV of the Federal Republic of Germany ("Proposal for Recurring Inspections of Nuclear Steel Pressure Vessels"). These are published here as the Appendix. While these are representative of current views on inservice inspection requirements for water-cooled reactors, they should only be regarded as references. Their application to reactors currently in operation or construction will probably require modification to suit the type of reactor, its design, construction and operating features, the' administrative requirements of the regulatory and inspection agencies, and the developing state of iinspection techniques and practices.

REFERENCES

The available literature on the subjects involved is extensive, particularly on fatigue, stress corrosion, hydrogen embrittlement, nondestructive testing, fracture mechanics and radiation effects. An attempt has been made to limit the list below to material specifically and directly applicable to the text of the Guide (not necessarily in agreement with it), to survey articles relevant to the specific subjects, and to articles containing particularly good bibliographic lists, identified by "biblio".

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Recurring Inspection of Nuclear Reactor Steel Pressure Vessels, Technical Reports Series No. 81, IAEA, Vienna (1968). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, Recurring Inspection of Nuclear Reactor Steel Pressure Vessels (1960-1966), Bibliographic Series No.34, IAEA, Vienna (1968). [3] WHITMAN, G.D., ROBINSON, G.C., Ir., SAVOLAINEN, A. W. (Eds), Technology of Steel Pressure Vessels for Water-Cooled Nuclear Reactors, Rep. ORNL-NSIC-21 (1967) (biblio). [4] MILLER, E.C., The Integrity of Reactor Pressure Vessels, Rep. ORNL-NSIC-15 (1966) (biblio). [5] COFFIN, L.F., Jr., Low cycle fatigue: a review,, Appl. Mater. Res. 1 3 (1962) 129 (biblio). [6] ALMEN, J.O., BLACK, P.H., Basic theories relating to fatigue in metals, Metals Engng Q., (Feb. 1966) 9. [7] TUPPENY, W.H., Jr. (Ed.), High Temperature Low-Cycle Fatigue, Papers and Discussions in Appli- cations Session of SESA, Ottawa, 16-19 May 1967, Society for Experimental Stress Analysis, 21'Bridge Square, Westport, Connecticut 06880, USA. [8] MERKLE, J.G., MOORE, S. E., WITT, F.J., in Technology of Steel Pressure Vessels for Water- Cooled Nuclear Reactors (WHITMAN, G.D., ROBINSON, G.C., Jr., SAVOLAINEN, A.W., Eds), Rep. ORNL-NSIC-21 (1967) 321-3, 331-8 (biblio). [9] MILLER, E.C., The Integrity of Reactor Pressure Vessels, Rep. ORNL-NSIC-15 (1966) 17 (biblio). [10] AMERICAN SOCIETY OF MECHANICAL ENGINEERS, Criteria of Section III of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels (1963) (An up-dated revision will be available in 1969). [11] PICKETT, A.G., GREGORY, S.C., Cyclic Pressure Tests of Full-Size Pressure Vessels, W.R.C. Bulletin No. 135, Welding Research Council, 345 East 47th Street, New York 10017 (1968). [12] SALKIN, R. V., How Low Cycle Fatigue Embrittles Pressure Vessel Steels, CNRM Rep. No.8, Centre National de Recherches MStallurgiques, Li£ge, Belgium (Sep. 1966). [13] LANGER, B.F., Design of pressure vessels for low-cycle fatigue, J. bas. Engng 84 3 D (1962) 389. 16 GUIDE TO INSPECTION

[14] GROSS, J.H., PVRC Interpretive Report of Pressure Vessel Research, Section 2—Materials Considerations, W.R.C. Bulletin No. 101, Welding Research Council (1964) (biblio). [15] DEPAUL, D.J. (Ed.),"Stress corrosion", Ch. 10, Corrosion and Wear Handbook, USAEC (1959) 187 (biblio). [16] GRIESS, J.C., Stress-corrosion cracking of stainless steels, Nucl. Saf. 5 (Mar. 1963) 32. [17] WYLIE, R.D., "Properties and flaws which may affect the performance of nuclear reactor vessels", Recurring Inspection of Nuclear Reactor Steel Pressure Vessels, Technical Reports Series No. 81, IAEA, Vienna (1968) 43. [18] ROGERS, H.C., Hydrogen embrittlement of metals. Science N. Y. 159 (1968) 1057 (biblio). [19] SACHS, K., 'Delayed failure' and its prediction, New Scient. 39 (1968) 237.

Radiation effects and surveillance programs, while essential to recurring in-service inspection programs, are being given primary consideration by another group, also under the sponsorship of the International Atomic Energy Agency. The present meeting of specialists on periodic inspection accepts and supports the conclusions and recommendations of that group as listed below.

[20] HAVEL, S., "Irradiation embrittlement and surveillance programmes of reactor pressure vessel steels", prepared for the IAEA Consultants Meeting on Engineering Aspects of Irradiation Embrittlement of Reactor Pressure Vessel Steels, Vienna, 2-4 Oct. 1967 (PL-267/13). [21] INTERNATIONAL ATOMIC ENERGY AGENCY, Conclusions and Recommendations of the Working Group on Engineering Aspects of Irradiation Embrittlement of Reactor Pressure Vessel Steels, IAEA, Vienna, 2-4 Oct. 1967 (PL-267). [22] WHITMAN, G. D., ROBINSON, G.C., Jr., SAVOLAINEN, A. W. (Eds), Technology of Steel Pressure Vessels for Water-Cooled Nuclear Reactors, Rep. ORNL-NSIC-21 (1967) 221-281, 637-643 (biblio). [23] IRVINE, W.H., "Developments in the periodic inspection of primary circuit pressure components of nuclear reactors", Recurring Inspection of Nuclear Reactor Steel Pressure Vessels, Technical Reports Series No. 81, IAEA, Vienna (1968) 66-72. [24] WHITMAN, G.D., ROBINSON, G.C., Jr., SAVOLAINEN, A. W. (Eds), Technology of Steel Pressure Vessels for Water-Cooled Nuclear Reactors, Rep. ORNL-NSIC-21 (1967) 430-538 (biblio). [25] MILLER, E.C., The Integrity of Reactor Pressure Vessels, Rep. ORNL-NSIC-15 (1966) 23-34 (biblio). [26] PELLINI, W. S., Advances in Fracture Toughness Characterization Procedures and in Quantitative Interpretations to Fracture-Safe Design for Structural Steels, W.R.C. Bulletin No.130, Welding Research Council (1968) (biblio). [27] INTERNATIONAL ATOMIC ENERGY AGENCY, Non-destructive Testing in Nuclear Technology (Proc. Symp, Bucharest, 1965), 2 vols, IAEA, Vienna (1965). (Although this reviews nondestructive testing, the applications described relate largely to inspection of fuel materials and fuel elements rather than to reactor vessels.) [28] WHITMAN, G.D., ROBINSON, G.C., Jr., SAVOLAINEN, A. W. (Eds), Technology of Steel Pressure Vessels for Water-Cooled Nuclear Reactors, Rep. ORNL-NSIC-21 (1967) 609-628, 645-668 (biblio). [29] WYLIE, R.D. Surveillance of nuclear power plant components, Nucl. Engng Des. 8 (1968)'117. [30] OAK RIDGE NATIONAL LABORATORY and USAEC, Incipient Failure Diagnosis for Assuring Safety and Availability of Nuclear Power Plants (Proc. Conf. Gatlinburg, 1967), Rep. CONF-671011, USAEC (1967). [31] DICKINSON, W.W., An acoustic spectrometer system, Non-destructive Testing 1. (1968) 379. [32] GERICKE, O.R., Defect determination by ultrasonic spectroscopy, J. Metals, N.Y. 6 (1966) 932. [33] GERICKE, OR., Ultrasonic spectroscopy of steel, Mater. Res. Stand. 5 (1965) 23. [34] CROSS, B.T., HANNAH, K.J., TOOLEY, W.M., The Delta Technique —A Research Tool—A Quality Assurance Tool, Rep. TR68-11, Automation Industries, Boulder, Colo. (1968). [35] WHITMAN, G. D., ROBINSON, G.C., Jr., SAVOLAINEN, A. W. (Eds), Technology of Steel Pressure Vessels for Water-Cooled Nuclear Reactors, Rep. ORNL-NSIC-21 (1967) 629-37, 643-5, 664-6 (biblio). [36] NICHOLS,. R. W., The Use of Overstressing Techniques to reduce the Risk of Subsequent Brittle Fracture, Rep. IIS/IIW-287-67-Z (ex, X-409-67), U.K.A.E.A. Reactor Materials Laboratory, Culcheth (1967) (biblio). [37] NICHOLS, R. W., Overstressing as a means of reducing the risk of subsequent brittle fracture, Br. Weld. J. 15 (1968) 524. [38] HAWTHORNE, J.R., LOSS, F.J., The Effects of Coupling Nuclear Radiation with Static and Cyclic Service Stresses and of Periodic Proof Testing on Pressure Vessel Material Behavior, Rep. NRL 6620 (1967). 17 GUIDE TO INSPECTION

[39] BLAIR, J.S,, Here's why line pipe should not be tested to strains beyond the yield point. Oil Gas Int. 7 11 (1967) 62. [40] IRVINE, W.H., BEVITT, E., QUIRK, A., The Safety of Pressure Vessels against Fast Fracture, Rep. AHSB (S) R 107, U.K.A.E. A. Health and Safety Branch (1966). [41] PARRY, D.L., "Nondestructive flaw detection in nuclear power installations". Incipient Failure Diagnosis for assuring Safety and Availability of Nuclear Power Plants (Proc. Conf. Gatlinburg, 1967), Rep. CONF-671011, USAEC (1967) 107. [42] GREEN, A.T., HARTBOWER, C.E., "Stress-wave analysis technique for detection of incipient failure", Incipient Failure Diagnosis for assuring Safety and Availability of Nuclear Power Plants (Proc. Conf. Gatlinburg, 1967), Rep. CONF-671011, USAEC (1967) 127. [43] PEDERSEN, H.N., SPANNER, J.C., "Detection, location and characterization of flaw growth in metals using acoustic emission methods", Incipient Failure Diagnosis for assuring Safety and Availa- bility of Nuclear Power Plants (Proc. Conf. Gatlinburg, 1967), Rep. CONF-671011, USAEC (1967) 163. [44] GREEN, A, T., The detection of incipient failures in pressure vessel systems, Nucl. Saf. 1£ 1 (1969) (biblio). [45] SCHOFIELD, B.D., "Acoustic emission from metals — its detection, characteristics and sources". Physics and Nondestructive Testing (Proc. Symp. San Antonio, 1963), Southwest Research Inst., San Antonio, Tex. (1963) 63. [46] HATTON, P.H., Acoustic emission in metals as an NDT tool, Mater. Evaluation 26 7 (1968) 125.

RECOMMENDATIONS

I. In the opinion of the Meeting, it is no longer adequate for an assessment of the safety of a nuclear power reactor pressure vessel design to be made without examining the provisions made for effective periodic operational inspections. II. It is recommended that the owner of any such proposed plant submit, with the proposals, full details of the planned periodic inspections, the means whereby the inspections will be made, and the provisions to ensure that effective inspections are feasible, taking into account the specific designs. III. It is recommended that the owner be obliged to submit for approval these proposals for performing the recurring inspections and submission of records in accordance with the most up-to-date information. Such proposals should be examined by or on behalf of the National Regulatory body at a sufficiently early stage in the procedures. IV. For plants already in operation, or too far advanced in construction for the requirements to be embodied in the design stage, it is recommended that the owner investigate the best practical means available for effective periodic inspection as set out in this Guide and the attached Appendix. V. It is recommended that consideration be given to the preparation of guides or examples covering the recurring inspection of other reactor types, in addition to the examples in this Guide for water-cooled reactor vessels. VI. The recommendation of the 1966 Panel pertaining to repeated proof testing is reiterated. Additional work on such matters as the properties of steels under reactor operating conditions, the influence of overpressurizing on time-temperature-dependent properties of steels, and the application of a quantified model of elastic-plastic fracture mechanics is believed to be necessary to establish the relative advantages and disadvantages of proof testing. VII. Further emphasis is recommended on the development of new and improved nondestructive inspection techniques, due consideration being given to the possibility of remote operation of such devices. VIII. In view of the close connection between the integrity of vessel internals and the integrity of the pressure vessel itself, it is recommended that under the auspices of the Agency a meeting of experts should be convened to discuss pre-operational and recurring nondestructive testing and inspection of pressure vessel internals and, if possible, to prepare a document which could serve as a general guide on the basis of the discussions.

APPENDIX

PROPOSAL FOR RECURRING INSPECTIONS OF NUCLEAR STEEL PRESSURE VESSELS

O. KELLERMANN, A. TIETZE Institut fur Reaktorsicherheit der Technischen Uberwachungs-Vereine e.V., Cologne, Federal Republic of Germany

1. GENERAL RECOMMENDATIONS

1. 1. Necessity for recurring inspections

The necessity for periodic inspections is derived from the fact that in steel reactor pressure vessels under operation the storage of a high amount of energy and the accumulation of radioactive materials presents a risk for the people inside the plant and the environment. During its service life-time a steel reactor pressure vessel will possibly be exposed to influences whose single and combined effects cannot be calcu- lated with the accuracy desirable for nuclear safety. The most important influences are: stress, temperature, irradiation, hydrogen absorption and corrosive attack, all depending upon time. These influences may result in changes of material properties such as aging or formation of flaws and cracks. Recurring inspections need not be performed if stresses in the vessel wall are evidently lower at each spot than those minimum stresses necessary for the propagation of a brittle fracture.

1. 2. Purpose of recurring inspections

The purpose of recurring inspections usually consists in the early detection of alterations of the vessel material properties. This leads to the possibility of the prevention of such vessel failures during operation, especially of those failures that might have serious consequences. Alterations of the vessel material may be caused by wear, corrosion, erosion and different types of cracking. The vessel design should include calculations of critical crack lengths. The results of these calculations will help to define scope and intensity of recurring inspections and to interpret their results.

1.3. Pre-operational experiences

Results of pre-operational inspections and measurements, which will be important for the performance and the evaluation of recurring inspections, should be. suitably recorded (photographs of the vessel surface and documents of nondestructive tests, descriptions of special equipment, summaries of and extracts from stress fabrication and test reports). It is recommended that reference nondestructive tests should be performed on the vessel after installation of the core components and before starting up nuclear operation as numerous operational influences may handi-

23 24 APPENDIX cap the performance of inspections and therefore result in interpretation difficulties of inspection results. These statements are of special importance for all kinds of remote testing techniques .

1.4. Reactor and vessel design

The necessity of taking into account, at an early stage in the design of reactor and vessel, the needs of recurring inspection is very important. Therefore early agreement concerning the recurring inspection techniques to be applied should be arranged. The reactor construction should enable a 100% accessibility of the vessel wall from either outside or inside. Significant areas, such as nozzles, penetrations, weld seams, should be accessible from both sides. In this connection the application of remote techniques may be useful.

1. 5. Techniques applied

The choice of techniques to be applied in recurring inspections depends upon The reactor type and its operational conditions; Properties of the vessel and its internal and external structural parts; The type and size of flaws, and cracks, and the type of other material deteriorations to be detected; and The effectiveness of the test methods. There are two categories of techniques, integral techniques and local methods. Integral methods answer the question whether there is any (serious) change in the vessel property at all. Local methods are confined to areas and spots of the vessel. Integral methods are superior to local methods insofar as they should allow statements on the integrity of the whole vessel with one test arrangement. Hence the development of suitable integral techniques, e. g. acoustic emission and the possibilities of their applicability, that should be taken into account by reactor and pressure vessel designers. Integral techniques are: The overpressurizing test; Tests with surveillance specimens; Leak tests; Acoustic emission tests; and Combinations of the above-mentioned tests. Local methods are limited to certain areas or spots of the vessel wall. Their results, of course, refer only to these areas and spots. Local techniques are: Visual techniques; Optical techniques, using mirrors, endoscopes, telescopes, television; Replica methods; Dye-penetrant techniques; Magnetic particle techniques; Eddy current techniques; Infra-red techniques (thermal emission); Conductivity techniques; US techniques; and Transmission techniques. APPENDIX 25

1.6. Time intervals of inspection

It is not possible.to state the actual time intervals which may be applied to all reactor types.

It is recommended: To perform at least one first thorough operational inspection, e. g. as described in Sub-section 2.4, soon after the beginning of vessel operation to. detect incipient serious flaws originating from design or fabrication; and To use shutdown of reactor operation (refuelling, turbine inspections, other major repairs or modifications).

The following inspection plan is regarded as a minimum:

Yearly:

Inspection of all accessible (i. e. without dismounting of structural parts in addition to those necessary in connection with reactor operation) areas of the vessel interior and exterior (cladding, weld seams, areas of stress concentration, spots of corrosion, erosion, penetrations, nozzles). The results of the inspections shall be considered representative for the whole surface state. Inspections of other areas, should therefore follow if test results make it advisable.

Four-yearly:

Dismantling of internal and external structural parts to an extent that a 100% inspection of the inner surface and of significant areas of the outer surface (weld seams, penetrations, nozzles) can be performed.

Eight-yearly:

Overpressurizing test taking into account the toughness-behaviour of the vessel. Equivalent procedures and techniques may be substituted, in particular the overpressurizing test may be replaced by nondestructive methods. The results should be recorded, thus facilitating an objective comparison with earlier and later test results (see Sub-sections 2.2.6, 2.3.6 and 2.4. 7).

2. INSPECTION PLAN FOR STEEL PRESSURE VESSELS OF LIGHT- WATER-COOLED AND -MODERATED REACTORS

2.1. Introduction

The detailed recommendation of this inspection plan are the result of requirements by the recurring inspection. This means that the properties of the vessel and of its internal and external structural parts, and the requirements desired for the safety reasons, are equally taken into account Time intervals are, of course, modifiable to some extent, according to the 26 APPENDIX

operational conditions of the reactor. The application of remote NDT- techniques will be useful, and probably necessary to some extent for the performance of inspection to the recommended extent.

2.2. Yearly inspection

2.2.1. Formal examinations

2. 2. 1. 1. Specifications. Comparison of the plant with licensed specifications, especially with regard to number and arrangement of armatures, safety devices, pumps (circuit diagrams).

2. 2. 1. 2. Inspection of safety devices. Inspection, and if possible, testing of all armatures and safety devices.

2. 2. 1. 3. Operational report. Evaluation of all significant plant operation • records and documents, especially with regard to: Transgressions of pressure- and temperature limits; Specified velocities of pressure and temperature alterations; Transgressions of limited load and temperature cycle limits; Properties of the coolant contamination by gases, other solved sub- stances (electrical conductivity) and dispersed particles (Fe); and Operation time and retention of filters in the primary cooling cycle.

2.2.2. Head

100% inspection of the interior and exterior surface optically.

2.2.3. Flange

100% optical inspection of the bolts, nuts and supporting discs, random testing (20%) of bolts by dye-penetration.

2.2.4. Sealing surfaces

Visual inspection of the vessel flange and head flange sealing.

2. 2. 5. Vessel

Optical inspection of those internal surface (cladding) parts accessible during normal refuelling intervals. Optical inspection of the main coolant nozzle's interior. Optical inspection of the vessel's exterior, especially the welding seams and nozzles accessible during intervals (e. g. for " refuelling).

2.2.6. Documentation

The results of the tests mentioned under Sub-sections 2. 1. 1 to 2. 1. 5 are representative for the properties of the whole vessel. If the test results indicate any irregularities or flaws in the vessel material, further investigations and examinations should be added as described in Sub-sections 2. 3 and 2. 4. APPENDIX 27

The examiner should document the results (including photographs, sketches of equipment and arrangements) and compare them with earlier results before the vessel is used again for further operation. In some cases it might be necessary to repeat various inspections.

2.3. Four-yearly inspection

In addition to the examination described in Sub-section 2. 2 the following tests should be performed.

2.3.1 Head

Random tests (50%) of weld seams (penetrations) and cladding, using dye-penetrants. Each type of seam should be tested.

2.3.2. Flange

Optical inspection of all bolts, nuts and supporting discs, random testing (50%) of the surface by dye-penetration.

2. 3. 3. Vessel interior

The vessel internal structure parts are to be removed so that about 50% of the surface of the cylindrical part and of the bottom are able to be inspected. Areas with stress concentration (nozzles) or those influenced by corrosion or erosion should be tested additionally by dye-penetration. All welding seams (longitudinal, circumferential and nozzle welding seams) should be random (1 - 2m) tested by dye-penetration and 100% tested by US.

2.3.4. Sealing surfaces

100% optical inspection and dye-penetration testing of the flange and vessel sealing surfaces. -

2.3.5. Vessel exterior

100% optical inspection and dye-penetration testing of the following parts:

Nozzle welding seams of the core cooling and core spraying system; Flange welding seams; and The penetration holes in the bottom and control rod drive housings.

100% US testing of the longitudinal vessel welding seams and 50% of each circumferential welding seam. Random testing of the welding seams between vessel and vessel supporting constructions.

2.3.6. Documentation

The results of this inspection should be'documented in a suitable manner (see Sub-section 2. 2. 5) and compared with those of earlier inspections. If the interpretation of these results leads to serious doubts on questions of 28 APPENDIX safety the inspections should be followed by other tests, e. g. those mentioned under Sub-section 2.4 (eight-yearly inspection).

2.4. Eight-yearly inspection

In addition to the examinations outlined in Sub-sections 2. 2 and 2. 3 the following examinations should be performed:

2.4.1. Head

100% US testing of the welding seams.

2.4.2. Flange bolts

100% optical inspection and dye-penetrant examination of the bolts, nuts and supporting discs.

2.4.3. Vessel interior

100% optical inspection of the surface, especially in areas of high stress concentration (nozzles, penetrations, supporting rings, intra-shell transitions). Dye-penetrant testing of all welding seams and random testing by dye-penetration of areas of stress concentration.

2.4.4. Sealing surface

100% optical inspection and dye-penetrant examination of the surfaces.

2.4.5. Vessel exterior

100% US testing of the vessel welding seams in the vessel body. 100% optical inspection and dye-penetrant testing of the nozzle and penetration welding seams. US testing of these parts as far as possible. Optical inspection and dye penetration testing of control rod drive housings. Oscillation measurements upon control rod drive housings under operation conditions.

2. 4. 6. Overpressurizing test

A hydrostatic pressure test should be performed with a pressure

PT = PD X 1. 3 where

' PD design pressure, hydrostatic test pressure.

A pneumatic pressure test should be performed with a.pressure

P^ = PD X 1. 1 where Pj = . pneumatic test pressure. APPENDIX 29

If the tests described up to Sub-section 2.4, 5 have given no indication of propagating cracks or serious corrosion and erosion phenomena the overpressurizing test may be omitted. Where there is any doubt, this test should be performed, as it is a valuable proof for the integrity of the vessel during future operation.

2.4,7. Documentation

The results should be recorded as described under Sub-sections 2. 2. 6 and 2. 3. 6 to make it possible, for objective comparisons to be made with earlier and later results of vessel inspections.

Selected Material from

Draft ASME

CODE FOR INSERVICE INSPECTION

NUCLEAR REACTOR COOLANT SYSTEMS *

(Dated October 1968)

[Extracted from Code for Inservice Inspection of Nuclear Reactor Coolant Systems with the permission of THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS, United Engineering Center, 345 East 47th Street, New York, N. Y. 10017]

Any part of this Code may be quoted. Credit lines should read: "Extracted from Code for Inservice Inspection of Nuclear Reactor Coolant Systems with the permission of The American Society of Mechanical Engineers, United Engineering Center, 345 East 47th Street, New York, N.Y. 10017".

PRELIMINARIES

Inservice inspection philosophy In the development of the Code an attempt was made to itemize those systems where inservice inspection was essential for the protection of the public health and safety. The extent of these systems considered essential include those portions of the pressure-containing components up to and including the outermost containment isolation valve which could isolate the system in the event of a rupture of the pressure-containing boundary. The extent of the system so considered is generally as shown in the attached Figures, for typical BWR and PWR plants, and is shown in red. The diagrams of these plants have been extracted from Preliminary Safety Analysis Reports and rearranged to place all the system components covered by the Code in a definable outline. One exception to the area shown in red is the secondary side of the shell of steam generators.

# Material has been deleted from the original document where such material relates exclusively to administrative provisions in the country of origin. Inserted substitute material is identified by enclosure in square brackets.

31 32 APPENDIX

Areas selected for examination

The areas most predominantly selected for examination are those associated with welds in the pressure-containing components. Although the impression might be created that undue attention is placed on the examination of weld joints, it must be pointed out that such an examination of the weld joint calls for a volumetric examination which, when conducted, examines a reasonable amount of the base, material, including the heat affected zone. Therefore, the most convenient locations to satisfy the requirements of examining both base metal and weld joints are locations at the weld joints, and these are singled out. Especially singled out are the dissimilar metal weld joints, which are subject to additional stresses imposed by the differences in the thermal expansion of the two, or more, alloys used. Additional areas requiring attention are the weld repairs in the base material which are not associated with the weld seams. It is the intention of the Code that records be maintained of their location to provide definite knowledge of their presence, location and extent, so that if signs of distress are encountered during plant operation they can be related to the original fabricated condition and standards.

Representative sampling

Two basic philosophies were involved in the selection of areas for the examinations:

(1) The highest service factor related to operating conditions, and (2) Random areas, selected to provide an assessment of the general • overall condition.

The first category, covering the high service factor areas, include such areas as those having significantly higher stresses due to operating conditions, such as the knuckle joint between the reactor vessel head and the flange, the inside radius section of primary nozzles, or other areas with high service factors, such as those related to abrupt changes in geo- metrical configuration, or vessel material exposed to significant amounts of neutron fluence which may create some changes in the material and its behavior. The inspection of these high service factor areas generally involve periodic examinations. In the case of the second category, that of random area selection, an attempt was made to provide sufficient examinations to assess the general overall condition. Examples in this category are representative longitudinal and circumferential weld seams on the reactor vessel, internal vessel supports, weld seams in piping, pumps and valves, and bolting in the pressure-containing boundary. The one item in the nuclear facility which could not be unequivocally adapted to the foregoing examination philosophies was the reactor pressure vessel. Due to the problems involving access, combinations of materials, limitations on the examination techniques which could be applied; etc., several departures or exceptions had to be provided. For example, representative examinations of the seam welds, opposite the reactor core, requires the use of remote mechanisms or techniques which have not yet been fully developed, and a reasonable period of time was allotted for the development of the techniques, but access to conduct the examination (based APPENDIX 33 on the results of current programs) is required. Another departure, this time from the standpoint of the type of examination, is exemplified by the austenitic welds of the internal brackets attached to the reactor vessel wall. These brackets are not required to receive a volumetric examination (since no known technique presently exists) as are the other welds in the vessel wall, but a more frequent and comprehensive visual examination is specified.

Preoperational examinations

The most appropriate approach to establish a base line is to conduct preoperational examinations prior to the initial operation of the facility. The results of subsequent examinations can then be compared to the original condition to determine if changes have occurred. If-the examination shows no changes on specific components, then no action would be required; however, if changes have occurred then an evaluation of the indications would be required to determine necessary action. The intent of the Code is that preoperational examinations be as closely representative of the examinations to be performed later as is practicable. In some cases it may be more convenient to conduct the preoperational examinations in the shop, provided the method and techniques are the same as would be expected to be used later; this assumes no changes would be expected due to shipping, erection, and preoperational tests and that adequate records are provided. For example, shop radiographs of the weld seams in a valve body would serve since the valve bodies can be re- radiographed and compared later; but radiographs of the weld seams of the reactor vessel prior to the shop hydrotest would not be considered adequate, since changes may occur as a result of the hydro, and radiography is not considered as a practical method for the inservice inspection of the vessel.

Evaluation of results and repairs

Evaluation of indications from the examinations is related to the acceptance standards involved in the original construction. In most cases the acceptance standards are specified in Section III of the ASME Boiler and Pressure Vessel Code, the Nuclear Power Piping Code, and the Code for Pumps and Valves for Nuclear Power. The evaluation is for the purpose of assessing whether or not the condition indicated meets the original acceptance standards, so the proper disposition can be made. The intent of the Code is to provide for, on a case by case basis, a review of the indications that show potential signs of distress and relate the condition to the service requirements to which the component is subjected. For the case where the review shows that no changes have occurred, disposition would require nothing further. For the case where the evaluation shows potential signs of distress, but the condition meets (is within) the original acceptance standards, disposition may require nothing more than the examination of an additional number of components or like areas. Where the evaluation indicates that the condition does not meet the acceptance standards, disposition may require additional examinations of this and like areas. It may be determined that the conditions can be tolerated, or that repair or replacement is necessary. In the event that repair or replace- 34 APPENDIX APPENDIX 35 36 APPENDIX

t- Z X < 111 t^ >- "" <-i H> uiZ « 0¥ <£— h|_ UI-ul it/ l< — u k OK-. ,5 J u 0>-< ?>- X >0. u VI <3 I- 1 4 I < APPENDIX 37 38 APPENDIX ment is required, the intent of the Code is that restoration should be governed by the original acceptance standards.

Provisions for access

It was recognized that provisions must be made in the design and construction of the nuclear facility to include allowance for access to permit the conduct of'the examinations, and the provisions for access are included in the very first section of the Code. The committee attempted to provide the maximum flexibility to the j designer in the design and arrangements of the components to provide the space to conduct the examinations which, for example, may be from the interior or the exterior or any combination of interior and exterior. The provisions for access are directly related to the methods and equipment available to perform the examinations, some of which have not yet been fully developed, but are included in current programs sponsored by the Edison Electric Institute and TVA, the Pressure Vessel Research Committee and the AEC.

Implementation

Reviews of early drafts prompted consideration of the implementation of the proposed Code to existing nuclear power facilities, or to plants in the process of construction. These considerations resulted in the following statement issued by the ACRS on July 15, 1968: "With regard to the implementation of the proposed inservice inspection standard the ACRS believes that routine ISI requirements of reactor pressure vessel in use or under construction should continue to reflect recognition of built-in limitations on the inspectability of such vessels, except where evidence of pressure vessel deterioration dictates otherwise". It is the intent of the Committee that full implementation of this Code would be effective approximately one year after the first publication date in the ASME Journal Mechanical Engineering. Furthermore, it is the strong conviction of the Committee that implementation of the Code is applicable only to those components and plant facilities contracted for after the effective date, or voluntary adoption, of the Code. In the case of components and plant facilities in use, under construction, or which have been contracted for prior to the adoption of the Code, it is not the intent of this Committee that the requirements are applicable. For these cases, each should be studied to find a practical alternate means of assuring the health and safety of the public.

Balance of the plant No attempt was made in this Code to establish the inservice inspection criteria for the pressure vessels and piping in the conventional sections of the nuclear facility; i.e., outside the steam generator or containment isolation valves. It should be pointed out however, that jurisdiction of the pressure vessels and piping in this conventional section of the facility will come under the jurisdiction of the State, Province, City or Municipality where the Boiler and Pressure Vessel Code has been made effective. The pressure vessels and piping and safety devices will be inspected in ac- APPENDIX 39 cordance with the requirements of the authorities having jurisdiction at the plant site. The insurance companies providing inspection service will meet the requirements of the authorities having jurisdiction over the pressure vessels and piping, but may require more frequent inspections than required by any State or local regulations where experience demonstrates the necessity. Where conditions observed warrant additional investigations, it is anticipated they will be conducted following the techniques of the Inservice Inspection Code. Similarly, components should be inspected, insofar as possible, as is now done in conventional fossil fueled plant. Corrective recommendations will be submitted for any deficiencies or substandard conditions which are observed.

INTRODUCTION

The Code for Inservice Inspection of Nuclear Reactor Coolant Systems consists of a number of sections that collectively constitute the Code. Hereinafter in this introduction and in the text of this Code, when the word "Code" is used without being identified with another specific code section, the word "Code" denotes Inservice Inspection Code. The Code contains basic requirements for the access requirements to conduct the examinations, and basic requirements for the examinations which, when conducted under the supervision of a qualified inspector, constitutes an inspection of the pressure-containing components of the reactor coolant system. The designer is cautioned that the Code is not a design handbook. The Code does not do away with the need for the engineer or competent engineering judgement, for in some cases the access to be designed is based upon requirements of remotely operable examination equipment, some of which has not been fully developed. Attention of users of the Code is directed to the fact that the numbering of the sections and the material thereunder may not be consecutive. Such, discontinuity is recognized. It is not the result of editorial or printing errors. An attempt has been made, insofar as possible, to follow a uniform outline in the various sections. Due to the fact that the complete outline may cover phases not applicable to a particular section, the Code has been . prepared with gaps in the numbering. It is believed that in this way, cross referencing between sections is made easier and use of the Code is facilitated since the same subject in general appears under the same number and subnumber in all sections. The Code is under the direction of [the USA Standards] Committee N-45 and the administrative sponsorship of The American Society of Mechanical Engineers. [The American Society of Mechanical Engineers] is organized to keep the Code up to date in context and in step with the developments in examination techniques and usage. Revisions are issued periodically, and new editions are published at three- to four-year intervals, depending on conditions. All requests for interpretations or suggestions for revisions should be addressed to the Secretary, [ASME Ad-Hoc Committee on Inservice Inspection], care of the American Society of Mechanical Engineers, United Engineering Center, 345 East 47th Street, New York, New York 10017. 40 APPENDIX

ISI-100 GENERAL REQUIREMENTS

110 Scope This Code defines for the purposes of ensuring the protection of the public health and safety, the rules and requirements for inservice inspection of pressure-containing components of the reactor coolant system, and portions of associated auxiliary systems and emergency core cooling systems in nuclear power facilities with light water cooled and moderated reactors. The Code covers classification of areas subject to inspection, within defined system boundaries, fixes responsibilities, provisions for accessibility, definitions of examination methods, examination techniques and procedures, personnel qualifications, frequency of inspection, records mapping, evaluation of inspection results, disposition, and repair requirements. The inspections required by this Code are intended to be performed generally during a reactor refueling outage or other plant shutdown periods. The inspections consist of visual and nondestructive examinations of those pressure-containing components which are within the system boundaries defined in Section ISI-120, and which under foreseeable malfunctions or interactions could potentially impair the safe continued operation of the nuclear power facility.

120 System boundary subject to inspection For pressurized water reactor (PWR) and boiling water reactor (BWR) facilities, the inspection requirements of this Code shall apply to all pressure-containing components within the following system boundaries; (a) The reactor coolant system1, including the reactor vessel, its associated control rod housings, and all other pressure-containing components and piping which are subject to reactor coolant system conditions during normal reactor operation (including startup, hot standby, and shutdown). For BWR facilities, the reactor coolant system boundary extends to and includes the outermost containment isolation valve2 in the main steam and feedwater lines, and the safety valves within the containment. 2 (b) Portions of the associated auxiliary systems which connect to the reactor cooling system. When such systems penetrate the containment the boundary extends to and includes the first positive acting containment isolation valve outside the containment. However, when this system contains two block valves4 both of which are normally closed during normal

1 Reactor coolant system means that system through which reactor coolant flows continuously during normal reactor operation for the purpose of removing heat from the reactor core.

2 Containment isolation valves are those valves in system piping which penetrates the primary reactor containment, and which can serve to isolate the system inside of containment from those portions of the same system located outside of containment,

' Associated auxiliary systems are those systems in which reactor coolant is diverted from the reactor coolant system either continuously or intermittently in support of normal reactor operation.

4 System block valves are stop and check valves (other than containment isolation valves) which can serve to block flow from the reactor coolant system. APPENDIX 41 reactor operation, the boundary extends to and includes the second of the two valves (the second valve must be positive acting) regardless of whether or not the lines penetrate the containment. (c) Portions of emergency core cooling systems5 connected to the reactor coolant system, which penetrate containment, extending to and including the first positive acting containment isolation valve outside containment. For systems not penetrating containment, the system boundaries extend to and include the second of two block valves normally closed during normal reactor operation.

(d) Excluded are components or piping within the system boundaries of (a), (b), and (c), which in the postulated event of a rupture, result in the loss of reactor coolant not in excess of the capability of normal make-up systems6 to supply water to the reactor coolant system for the interval of time required to permit a reactor shutdown and orderly cool-down without requiring the operation of emergency core cooling systems.

130 Responsibility Performance of the inservice inspections required by this Code shall be the responsibility of the owner of the nuclear reactor facility. This responsibility shall include: (a) the development and preparation of written examination instructions and procedures, including diagrams and/or system drawings delineating the identification and extent of areas of components subject to examination; (b) verification of the qualifications of personnel who perform the examination required by this Code for the specific levels or responsibility in accord with Section ISI-220; (c) the maintenance of adequate records of inspections performed, such as radiographs, diagrams, drawings, inspection data, and personnel qualifications; (d) the recording of all information and inspection results which provide a basis for evaluation, and facilitate comparison with the results from subsequent inspections; and (e) the retention of all inspection records for the service lifetime of the component examined.

140 Accessibility 141 General provisions for access The design and arrangement of the system components and piping shall be the responsibility of the owner, or his designated agent, and shall include allowance for adequate clearances to permit the conduct of the examinations

5 Emergency core cooling systems are those systems which enable, as one of their functions, the injection of water into the reactor vessel in the event of breaks in the system boundaries of (a), (b) and (c).

6 Normal make-up systems serve to replenish reactor coolant inventory as required to make up for leakages or withdrawal from the reactor coolant system during normal reactor operation, (e.g., chemical and volume control systems in PWR plants, and control rod drive systems in BWR plants). 42 APPENDIX specified in Table I of Section ISI-260 at a frequency specified in Section ISI 240. The owner shall determine that considerations are given in the designs7 and arrangements for the following: (a) access for inspection personnel and/or examination personnel and equipment necessary to conduct the examinations; (b) removal and storage of structural members, shielding components, insulating materials, and other equipment and components required to perform the specified visual observations, examinations, and tests; (c) installation and support of handling machinery (e.g., hoists or other handling equipment) where required to facilitate removal, disassembly, and storage of equipment, components and other materials; (d) conduct of alternate examinations to those specified herein in the event structural defects or indications are revealed which may require such alternate examinations; (e) performance of necessary operations associated with repair or replacement of system components and piping, in the event structural defects or indications are revealed which may require such repairs or replacements. 142 Specific provisions for access The owner shall be responsible to determine that provisions in the design and arrangement of the reactor coolant system, associated auxiliary system, and emergency core cooling systems will allow access, at least, to the following areas, to permit the examination by the methods listed under Section ISI-260, Table I; (a) Reactor vessel (1) Essentially 100 percent of the surface areas of the reactor vessel pressure containing membrane, either from the inside or outside surfaces of the vessel, or a combination thereof. (2) The interior of the reactor vessel, including the space at the bottom head to permit the detection of mechanical or structural failure of reactor internals. (3) At the following specific areas: (A) longitudinal and circumferential shell and head weld joints, (B) closure flange to shell and head transitions, bolt holes and ligaments, (C) flange bolting and nuts, (D) weld joints at vessel nozzles and penetrations, (E) weld zones of internal load carrying structural attachments, (F) weld zones of vessel external supports, (G) selected areas of vessel interior clad surfaces, (H) control rod housings assembly weld joints, mechanical joints, and weld joints to vessel head.

7 Design considerations other than access provisions may be applicable to specific system components and piping to render inservice inspections practicable (e.g., surface finish of components subject to crud or corrosion product buildup, material selection to minimize activation in service, shielding from irradiation effects, etc.). APPENDIX 43

(b) Other ASME Section III Class A Code vessels8 The following areas either from the inside or outside surfaces, or a combination thereof: (1) longitudinal and circumferential shell and head joints including secondary side of steam generators and large diameter ferritic steel pipe, (2) flange to shell and head transitions, bolt holes, and ligaments, (3) flange bolting and nuts, (4) weld joints of vessel nozzles and penetrations, (5) weld zones of vessel external supports, and (6) selected areas of vessel interior clad surfaces.

(c) System piping, Class I Nuclear Power Piping Code9 The inspections of the exterior surfaces of the system piping at the following locations: (1) pipe-to-pipe (and to fitting) weld joints, (2) pipe-to-vessel nozzle (and to other pressure containing components) weld joints, (3) selected lengths of longitudinal weld joints in pipe and fittings, normally exposed at the circumferential welds, (4) pipe-to-branch piping weld joints, (5) piping mechanical joints, (6) piping support and hangar attachments welds.

(d) System pumps and valves, 'Class I Nuclear Pump and Valve Code10 The following areas either from the inside or outside surfaces, or a combination thereof: ; i (1) pressure-containing seam welds in pump casings (or other pressure containing components), (2) pressure-containing seam welds in valve bodies (including system safety valves, relief valves, and containment isolation valves), (3) pressure-retaining bolting, and flange bolting ligaments in the components of (1) and'(2).

I ISI-200 INSERVICE INSPECTION

201 Definitions "Examination" - denotes all visual examination and nondestructive testing such as radiography, ultrasonic, liquid penetrant, and magnetic particle methods. "Inspection" - denotes a witness or performance of examinations by an Inspector having jurisdiction over the nuclear power facility at the plant site. 8 ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Vessels.

9 ASME Code for Pressure Piping, Nuclear Power Piping, LISAS B31.7.

10 ASME Code for Pumps and Valves for Nuclear Power (to be issued). 44 APPENDIX

"Inspector" - [the term "Inspector" has a unique connotation in the USAS Committee document. For the purposes of other countries, it should be the duly authorized representative of the responsible Inspecting or Certifying Authority] .

202 Duties of the Inspector It is the duty of the Inspector to witness, or otherwise verify all the examinations required by this Code, and also to have made any additional investigations necessary to verify that all applicable requirements have been met. It is the duty of the Inspector to assure himself that the nondestructive examination methods used follow the techniques specified in ASME Boiler and Pressure Vessel Code Section III - Nuclear Vessels in IX-300 of Appendix IX, and that the examinations are performed in accord with written qualified procedures and by personnel employed by the owner or his agent and qualified in accordance with ASNT-TC-1A11, and Supplements and Section ISI-220 of this Code. The Inspector has the right at any time to require requalification of any procedure or operator when the Inspector has reason to believe the requirements are not being met. The examination records shall be certified by the Inspector only after he has satisfied himself that all the requirements have been met and that the records are correct.

203 Qualification of Inspectors, Engineering Specialists and Inspecting Agencies The inspection required by this Code shall be performed, (a) where the reactor is in the United States, by an Inspector employed by a State or Municipality of the United States or an Inspector regularly employed by an insurance company authorized to write boiler and pressure vessel insurance in the United States; or (b) where the reactor is in Canada, by an Inspector employed by a Canadian Province or where authorized by the Province in which the reactor is installed, by an Inspector regularly employed by an insurance company licensed to write boiler and pressure vessel insurance in that Province; or (c) by an Inspector employed by other enforcement authorities in the United States or Canada having jurisdiction over the designated reactor facility. Any Inspector who performs inspections required by the Code, shall have been first qualified by written examination pursuant to the legislation __ or rules of a State or Municipality of the United States, the legislation of a Canadian Province or the rules of another authority having jurisdiction over a nuclear reactor at the installation location and which has adopted this Code. The Inspector shall not be an employee of the owner.

" ASNT-TC-1A and Supplements is a "Recommended Practice for Nondestructive Testing Personnel Qualification and Certification" published by the American Society for Nondestructive Testing, 914 Chicago Avenue, Evanston, Illinois 60202. APPENDIX 45

204 Access for Inspector The owner shall arrange for a duly authorized Inspector to have access to all parts of the plant, when requested. The owner shall keep the Inspector informed of the progress of the preparatory work necessary to permit inspections and shall notify him reasonably in advance when the components will be ready for any required inspections.

210 Examination techniques and procedures Methods, techniques and procedures for the inservice inspections are entitled "visual", "surface", and "volumetric" in the appropriate sections of this Code. Each term describes a category, permitting a selection of different methods or procedures to accommodate varying degrees of accessibility and radiation levels, and the automation of equipment to perform the examinations.

211 Visual examination A visual examination is employed to provide a report of the general condition of the part, component or surface to be examined, including such conditions as scratches, wear, cracks, corrosion or erosion on the surfaces, or alignment or movement of the part or component.

(a) Direct visual examination Direct visual examination may be performed when access is sufficient to place the eye within 24 inches of the surface to be examined and at an angle no less than 30 degrees with the surface to be examined. Mirrors may be used to improve the angle of vision. Lighting, in addition to the general room lighting, shall be provided to illuminate the area to be examined at right angles and at oblique angles to expose cracks or evidence of corrosion or erosion. Resolution shall be considered adequate when the combination of access, lighting, angles of vision, either unaided or corrected, can resolve a black line, 1/32 of an inch wide, on an eighteen percent neutral grey card placed on the surface to be examined, or in a situation similar to the area to be visually examined.

(b) Remote visual examination Remote visual examination may be substituted for direct visual examination when, for any reason, it is desirable. Remote visual examination may include visual aids such as telescopes, periscopes, borescopes, fiber optics, or TV cameras and monitoring systems, with or without attachments for permanent recording. Such systems shall demonstrate the ability to provide a resolution at least equivalent to that obtainable by direct visual'observation. Mirrors, movable lights, or rotating optics, or any combination thereof, may be employed to display cracks, surface scratches or evidence of corrosion, erosion, alignment or movement.

(c) Replication Surface replication methods shall be considered acceptable provided the surface resolution is at least equivalent to that obtainable by the visual observation. 46 APPENDIX

(d) Cleaning Visual examinations which require clean surfaces or decontamination for valid interpretation of results are to be preceded by appropriate cleaning processes. 19 212 Surface examination A surface examination is specified to delineate or verify the presence of surface or near surface cracks or discontinuities and may be conducted by applying either a magnetic particle examination'or a liquid penetrant examination where the surface condition, material, and accessibility permit such an examination. Neither method is permitted when the surface is in direct contact with water during the time the examination must be conducted.. (a) Magnetic particle examination Magnetic particle examination provides a method for the detection of rounded discontinuities, cracks, and other linear discontinuities in welds, plates, forgings, tubular products and castings. Sensitivity is greatest for surface defects and diminishes rapidly with depth below the surface. Magnetic particle examination is applicable only to ferromagnetic materials. The procedure for the application shall conform tp Appendix IX-350 of Section III of the ASME Boiler and Pressure Vessel Code [or the applicable national Code or Standard] . (b) Liquid penetrant examination Liquid penetrant examination provides a method of nondestructive examination for the detection of discontinuities open to the surface in ferrous and nonferrous materials which are nonporous. Typical discontinuities detectable by this method are cracks, seams, laps, cold shuts, and laminations. A liquid penetrant is applied to the surface to be examined ' and allowed to enter such openings, the excess penetrant is removed, the part is dried and a developer applied. The developer is wetted or stained by the penetrant entrapped in the discontinuities and increases the evidence of the discontinuities so that they may be seen either directly or by the use of "black light". The procedure for the application of liquid penetrant examination shall conform to Appendix IX-360 of Section III of the ASME Boiler and Pressure Vessel Code [or the applicable national Code or Standard].

213 Volumetric examination A volumetric examination is for the purpose of indicating the presence of subsurface discontinuities with a method or technique capable of examining the entire volume of metal contained beneath the surface to be examined. Methods such as radiography, ultrasonic examination, or other newly developed techniques, may be employed provided the method is demonstrated to be capable of detecting subsurface discontinuities. (a) Radiographic examination Radiographic techniques, em'ploying energy sources such as X-rays, gamma rays or thermalized neutrons, may be utilized with appropriate

12 Since cleaning specifications and standards are not yet fully developed, it is recommended that special investigations and precautions be taken before any cleaning process is used. APPENDIX 47 image recording techniques such as photographic film or papers, "electro- static systems, direct image orthocons, or image converters. For radiographic examinations employing either X-ray equipment or radioactive iso- topes and photographic films, the procedure shall be as specified in Appendix IX-330 "Radiographic Examination" of Section III of the ASME Boiler and Pressure Vessel Code [or the applicable national Code or Standard] . Other radiographic techniques shall be demonstrated to be at least equivalent to the sensitivities described in Appendix IX-330. (b) Ultrasonic examination Ultrasonic examination, using the pulse-echo method, shall be as described in Appendix IX-340 "Ultrasonic Examination of Welds" of the ASME Boiler and Pressure Vessel Code [or the applicable national Code or Standard] . Other ultrasonic procedures, or newly developed techniques, may be substituted for the method described above, provided the methods are demonstrated to be at least equivalent to those obtainable by the pulse- echo method. For the ultrasonic examination of materials other than welds, ASTM-E-114 "Ultrasonic Testing by the Reflection Method Using Pulsed Longitudinal Waves Induced by Direct Contact" and ASTM-A-435 "Longi- tudinal Wave Ultrasonic Inspection of Steel Plates for Pressure Vessels" are referenced as procedural documents for these examinations [ and may be used] .

214 Alternate examinations Alternate examination methods, combination of methods, or newly developed techniques may be substituted for the methods of ISI-Sections 211, 212 and 213 provided the results yield demonstrated equivalence or superiority. The acceptance standards for these alternate methods shall be subject to the requirements specified in Section ISI-311.

220 Qualifications of nondestructive examination personnel Personnel performing nondestructive examination operations shall be qualified in accordance with ASNT-TC-1A, Supplements and Appendices, as applicable for the examination technique and methods used. All nondestructive examinations shall be performed and the results evaluated by qualified nondestructive examination personnel. The assignment of responsibilities to individual personnel shall be at the discretion of the owner, or the owner's agent. The emphasis on the qualification test shall be on the individual's ability to perform the nondestructive examination in accordance with the applicable procedure for the intended application. For nondestructive examination methods that involve more than one operation or type, it is permissible to use personnel qualified to perform one or more operations. As an example, one person may be used who is qualified to conduct the examinations and another may be used who is qualified to interpret and evaluate the examination results. For nondestructive examination methods not covered by ASNT-TC-1A documents, personnel shall be qualified by the owner, or his agent to , comparable levels of competency by subjection to comparable examinations on the.particular method involved, for example, leak testing. The practical 48 APPENDIX

portion of the ASNT Qualification shall be performed using the owner's procedure(s) on part(s) representative of the owner's plant. The Inspector has the duty to verify the owner's or his agent's certification of an operator in accordance with ASNT-TC-1A and has the preroga- tive to audit the program and require requalification of any operator when the Inspector has reason to question the performance of that operator.

230 Examination requirements The owner shall be responsible for ensuring that the detailed examinations of all pressure-containing components within the system boundaries defined in Section ISI-120 as required in Table I of Section'ISI-260 are performed in accordance with qualified written procedures in accordance with Section ISI-210, and that the requirements contained herein are met. All detailed examinations, listed in Table I, shall be performed completely, once, as a preoperational examination requirement prior to initial plant startup, except that the examinations shall be extended, in all cases, to include essentially 100% of the pressure-containing welds. Shop examination records of individual system components or piping may serve in some instances, in lieu of the in-plant preoperational examinations, provided: (1) such examinations are conducted under conditions and with equipment and techniques equivalent to those which are expected to be employed for subsequent inservice examinations, and (2) the shop examination records are or can be documented and identified in a form consistent with those required in Section ISI-600. The owner shall be responsible for the compilation of appropriate examination records of this preoperational examination for the purpose of comparison with the examination results of subsequent inservice inspections.

240 Inspection intervals

241 Required interval The required examinations and inspections indicated in Table I of Section ISI-260 shall be completed during each 10 year interval. These intervals, which represent calendar years after the reactor facility has been placed into commercial service, shall not be exceeded by more than one year for inservice inspections periods to be concurrent with plant outages. Inspections on a more frequent schedule are at the discretion of the owner. For plants that are out of service continually for one year or more, the inspection period may be extended for one year or more, in full year increments.

242 Inspection program It is intended that the inservice examinations be performed during normal plant outages such as refueling shutdowns or maintenance shutdowns during the inspection interval. Except as specified in Section ISI-251 for examinations categories A, B, and L, at least 25 percent of the required examinations shall have been completed after one-third of the inspection APPENDIX 49 interval has expired (with credit for no more than 33-1/3 percent if additional examinations are completed) and at least 50 percent after two- thirds of the inspection interval has expired (with credit for no more than 66-2/3 percent). The remainder shall be completed at the end of the inspection interval.

243 Successive inspections The examination of Items (Table I, Section ISI 260) in each Examination Category, as selected during the initial inspection interval, shall be repeated in the same order, on a rotational basis, during the successive inspection intervals.

244 Additional examinations Examinations which reveal structural defects shall be extended to include an additional number (or areas) of system components or piping (in the same category or of the same type) equal to that initially examined. In the event further unacceptable structural defects are revealed, all remaining system components or piping (in the same category or of the same type) shall be examined.

250 Examination categories

251 Areas and extent of examinations The areas of the pressure-containing components requiring examination are those considered for inspection or examination because of their materials, geometry, stress levels, applied loads, environment, and type of fabrication. Every area falls within one or more of the following examination categories (as indicated in Table I) and shall.be examined at least to the extent specified herein: Examination category Areas and extent of examinations

A Pressure boundary welds in the reactor vessel wall opposite the length of the reactor thermal shield where used, or the effective length of the reactor core where the thermal shields are not used. Included in this category are the shell longitudinal and circumferential seams and base material defect weld repairs (except shallow repairs whose depth does not exceed the lesser of 3/8 inch depth or 10% of nominal thickness) in the region of the reactor core. The areas to be examined include the weld metal and base metal for one plate thickness beyond the weld. The individual examinations may be performed at or near the end of the inspection interval, and shall cover at least 10 percent of the length of each longitudinal seam and 5 percent of the length of each circumferential seam. In any case, for any individual area examined, the length of weld examined shall not be less than one wall thickness. 50 APPENDIX

When the longitudinal seams in the region of the reactor core have received an exposure to neutron fluence in excess of 1019nvt (En of 1 Mev or above), the length of weld to be examined shall be increased to, at least, 50 percent.

Examination category Areas and extent of examination

B Pressure boundary welds in vessels including base material defect weld repairs (except shallow repairs whose depth does not exceed the lesser of 3/8 inch depth of 10 percent of nominal thickness). The areas subject to examination shall include the weld metal and base metal for one plate thickness beyond the weld. The examinations shall cover at least 10 percent of the length of each longitudinal and 5 percent of the length of each circumferential seam. For weld seams on the reactor vessel the individual examination may be performed at or near the end of the inspection interval.

Examination category Areas and extent of examination

C Pressure boundary welds, vessel to flange and head to flange. For vessel to flange and head to flange welds the extent of the individual examinations performed over the period of the inspection interval shall cumulatively cover 100 percent of each circumferential seam. The number and extent of areas examined during each inspection shall provide a representative sampling of the entire weld.

Examination category Areas and extent of examination

D Pressure boundary nozzles in vessels. Areas subject to examination include the nozzle attachment weld, integral extension of nozzle inside vessel (or pipe), but does not include nozzle-to-pipe weld. Excluded are seal welds in pressure boundaries. The extent of examination of each nozzle shall cover 100 percent of the nozzle-to-shell weld, and 100 percent of the inner radius section of the nozzle-to-shell juncture. All the nozzles shall be examined over the period of the inspection interval. The number of nozzles examined during each . inspection shall be the total number divided approximately equally among the expected number of examinations. APPENDIX 51 during the inspection interval and selectively distributed • to yield a sampling of each category of nozzles.