Reactor Pressure Vessel Surveillance

IAEA-202

REACTOR PRESSURE VESSEL SURVEILLANCE

PROCEEDINGS OF A TECHNICAL COMMITTEE MEETING ORGANIZED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY WITHIN THE FRAMEWORK OF THE INTERNATIONAL WORKING GROUP ON RELIABILITY OF REACTOR PRESSURE COMPONENTS (IWG-RRPC) HELD IN PLZEN, CZECHOSLOVAKIA, 17-18 MAY 1976

(a# AA TECHNICAL DOCUMENT ISSUED BY THE (is INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1977 IAEA INTERNATIONAL WORKING GROUP ON

RELIABILITY OF REACTOR PRESSURE COMWPONETS (IWG-RRPC)

Technical Committee Meeting on "Reactor vessel surveillance: results of programmes conducted and proposals for revision"

Chairmen: Dr. Karel Mazanec, Corresponding Member of the Academy of Science, CSSR.

Dr. Len Steele, Naval Research Laboratory, U.S.A.

Ing. S. Havel, Nuclear Research Institute, Res, CSSR.

Scientific Secretary: I.K. Terentiev, IAEA

Hosted by the Czechoslovak Atomic Energy Commission and the

Skoda National Corporation

Printed by the IAEA in Austria November 1977 PLEASE BE AWARE THAT ALL OF THE MISSING PAGES IN THIS DOCUMENT WERE ORIGINALLY BLANK The IAEA does not maintain stocks of reports in this series. However, microfiche copies of these reports can be obtained from INIS Microfiche Clearinghouse International Atomic Energy Agency Kmntner Ring 11 P.O. Box 590 A-1011 Vienna, Austria on prepayment of US $0.65 or against one IAEAmicrofiche service coupon. Introduction

The Technical Committee meeting on "Reactor Vessel ourveilance: Results of Programmes Conducted and Proposals for Revision" was convened by the IAEA within the programme of activities of the International Working Group on Reliability of Reactor Pressure Components. On the invitation of the Czechoslovak Atomic Energy Commission the meeting was held in Plzen on 17-18 i:ay 1976. It was hosted by the AEC of the CSSR and Skoda National Corporation.

The meeting was attended by 51 participants from 16 countries and 3 international agencies. 22 reports were presented and discussed.

Professor I.S. Zheludev, IAEA Deputy Director General, opened the meeting and addressed the participants. The meeting was also addressed by Mr. Neumann, Chairman of the Czechoslovak Atomic Energy Commission and Mr. Erbal, Production Director of Skoda National Corporation, Plzen. Professor Karel Mazanec of Ostrava University, chaired the meeting. Dr. L.E. Steele of the Naval Research Laboratory, USA, and Mr. S. Ravel, Director of the Nuclear Research Institute in Rez, CSSR, served as co- chairmen.

On the basis of reports presented and discussions during the sessions, recommendations on "Surveillance of Reactor Pressure Vessels for Irradiation Damage" were prepared. Some comments on these recommendations were received later only from Mr. Prantl, Switzerland, concerning the wording of the certain recommendations. These comments were taken into consideration by the Secretariat while preparing the final version. Contents

Introduction

Session I General reports reviewing national programmes on reactor vessel surveillance

1.1 "Surveillance as a complement to irradiation embrittlement studies: status and needs", L.E. Steele, Naval Research Laboratory, USA. 1

1.2 "Surveillance progrmnmes prepared and carried out during production and exploitation of the A-1 Nuclear Reactor Pressure Vessel", M. Brumovsky, R. Filip, Skoda Works and J. Cervasek, M. Vacek, Nuclear Research Institute, Rez, CSSR. 11

1.3 "Present Status of Surveillance Tests for Nuclear Reactor Vessels in Japan", S. Miyazono, JAERI, Japan. 37

1.4 "Material Surveillance Programme of Pressure Vessel Steels in India", K.S. Sivaramakrishman, Bhabha Atomic Research Centre, Trombay, Bombay. 39

1.5 "Westinghouse Nuclear Europe Reactor Vessel Surveillance Programme", T.R. Mager. !47

1.6 "Reactor Vessel Surveillance: Present Practice and Future- Trends in Switzerland", G. Prantl, T. Varga, D.H. Njo. $1

1.7 "PWR Pressure Vessel Surveillance Programme in Belgium", Ph. Van Asbroeck. 71

1.8 "A Utility Review of Irradiation Surveillance Programs and Industry Responsibilities", T.D. Keenan, Yankee Atomic Electric Company, Westboro, ,iassachusetts, USA. 79 1.9 "Contribution to the question of surveillance programs for nuclear reactor pressure vessels", M1.Brumovsky, Skoda Works, Plzen, CSSR. 95

1.10 "Reactor Vessel Material Surveillance Program", Draft version, presented by the delegation of Italy. 99

1.11 "Comments on Reactor Vessel Surveillance Programmes in the Federal Republic of Gerfmarny", E. Bazant, BBR, FRG. 1ll

Session II Results of surveillance programmes

2.1 "Evaluation of surveillance specimens and in-service inspection of tubes of A-1 reactor heavy water calandria", P. Mrkous, M. Brumovsky, J. Prepechal, Skoda Works, CSSR. 113

2.2 "Scope and results of the Reactor Vessel Radiation Surveillance Program of the Nuclear Power Plant Beznau I", E. Sandona, P. Pliss, Switzerland. 125

Session III Surveillance Requirements and Criteria for Analysis

3.1 "Brittleness, Presupposition (criteria) for reactor vessel brittle fracture", E. Bazant, BBR Mannheim, FRG. 139

3.2 "Analysis of mechanical property data obtained from nuclear pressure vessel surveillance capsules", J.S. Perrin, Battelle Memorial Institute, Columbus, Ohio, USA. 163

3.3 '"ew methods for determining radiation embrittlement in reactor vessel surveillance", R.A. Wullaert, Practure Control Corporation USA. 173 3.4 "Evaluation of the Maine Yankee reactor belt line materials" R.A. Wullaert, J.W. Sheckherd, R.W. Smith, USA. 193

3.5 "Materials surveillance program for C - E NSSS reactor vessels" J.J Koziol, Combustion Engin'ering Inc., Windsor, Connecticut, USA. 217

3.6 "Report about Acoustic emission Analysis on the Reactor Pressure Vessel of the First Austrian Nuclear Power Plant", K.K. Wischin, Austria. 231

3.7 "In-service inspection of the VVR-S reactor", F. Jonak, L. Kaisler, Nuclear Research Institute, Rez, CSSR. 233

3.8 "US NRC Research Programs on Fracture Toughness for Surveillance Applications and Requirements for Neutron Dosimetry and Analysis", C.2. Serpan, NRC, USA. (Text is not available).

3.9 "Materials surveillance programme for Babcock & Wilcox produced NSSS reactor vessels", A.L. Lowe, USA, (Text is not available).

Conclusions and Recommendations of the Meeting 255

List of the participants of the Meeting 259 SURVEILLANCE AS A COMPLEMENT TO IRRADIATION EMBRITTLEMENT STUDIES: STATUS AND NEEDS L. E. Steele Naval Research Laboratory

ABSTRACT The history of the study of radiation embrittlement of reactor pressure vessel steels has gone through three stages in the USA -

1. A scientific curiosity, 2. Empirical or laboratory evaluation of typical steels, and 3. Integration of the scientific and empirical to advance status and evolve standard techniques. The current stage is one in which surveillance data compliments the laboratory studies which characterized Stage 3. The early USA surveillance programs were generally analyzed by the same people who were the primary laboratory investigators. An effort must be made to continue this type of collaboration as a useful two-way learning procedure though it will become more and more difficult as nuclear power is broadly commercialized. The current status of both types of USA programs will be presented to encourage the most advantageous use of data from both sources.

At this time about 25 USA nuclear power reactors have operated long enough to have provided initial surveillance or dosimetry results. An effort will be made to summarize the general status of these in order to: 1. Provide complimentary data to laboratory studies.

2. Assess directions in handling the problems of radiation embrittlement.

3. Note lessons learned for improving surveillance efforts in the future.

4. Identify possible research tasks for the future to support in-service surveillance and other measures. 5. Justify facts advancing surveillance requirements to status of national codes and standards. 6. Justify facts requiring changes in current national codes and standards.

A plan will be presented along with an introduction of each member of the USA delegation for systematic presentation of the status of reactor vessel surveillance in the USA.

INTRODUCTION

The problem of neutron radiation embrittlement was recognized in the USA at about the time commercial nuclear power began, the late 1950s, but data which would significantly affect pressure vessel

1 and reactor design came only in the mid-to-late 1960s. For this reason, though there were doubts enough to begin surveillance program planning, it was well after 1970 before the research data significantly affected design of new nuclear plants through limits on composition and on properties of pressure vessel steels.

Surveillance data became available in the mid-to-late 1960s but were too sparse and represented mainly one-of-a-kind plants and therefore were of limited value to complement research data and influence changes to minimize radiation embrittlement effects. Nevertheless, data were adequate to validate research data and therefore put force into the prior voluntary standards for reactor vessel surveillance.

The growing volume of surveillance data available in the 1970s adds a new dimension complementing greatly the available research data. This is particularly important as the rules or guides of the U. S. Nuclear Regulatory Commission have become more definitive in recent years.

CRITICAL ELEMENTS OF SURVEILLANCE The critical elements of surveillance include: 1. The fracture behavior of steels used to construct pressure vessels,

2. The influence of neutron radiation on fracture performance of steels, 3. Measurement of the radiation incident on the vessel (peak point if not uniform on inner vessel circumference), 4. Measurement of the specific fracture response of the steels (all components) of a vessel to radiation, and

5. Operating implication of foregoing factors (operation to minimize influence of these factors). All of these factors have been considered in the application of vessel surveillance though there is room for improvement in each.

SUMMARY OF STATUS FROM RESEARCH VIEWPOINT

The fracture of steels has been the subject of much research and technology in recent years and these advancements impact directly on the question of reactor vessel integrity. On the positive side, empirical studies clarified the point of transition called the nil ductility transition temperature and the influence of temperature above this point and steel thickness as well. Advancements in the understanding of linear elastic fracture mechanics advanced to the benefit of quantitative analysis of flawed steel structures and, when full thickness tests are included, provides the basis for full vessel criteria for fracture prevention. This is not to say all the essential background has been completed, but rather that the right way has been pointed out. The greatest remaining need is to understand fracture in the ductile portion of the transition curve, more systematic full section tests for all vessel components, and better small specimen representation of full section.

2 The influence of neutron radiation on steel fracture has been advanced through years of research studies to a point of relative maturity but has been limited largely by complexity of conducting full section irradiated tests and by lack of the opportunity for a statistically based irradiation study on real reactor vessel steels. The greatest accomplishment has been the verification of residual element effects (copper and phosphorus especially) and the advancement of this factor to the point of standardization limitations as well as to inclusion in regulatory guides. The latter factors have major influences on surveillance implications; putting a critical view on older reactors and possibly providing the impetus for improved future steels and hence reduced needs for surveillance for reactors constructed in the future.

Advances in the techniques for measurement of vessel radiation exposure have been great in support of research experiments. Major studies were necessary to validate irradiation embrittlement studies conducted in research or testing reactors for projection to the power reactor condition. Critical aspects were to measure neutron flux, fluence and spectrum and to relate these to vessel steel response thereto. In addition, reflecting the relatively mature status of our research knowledge, in these areas, a series of standards have been published in support of their application.

The relative response of individual steels or steel components is especially critical and is related to the important factor of steel composition and its influence on radiation embrittlement which was noted above. Experience in research programs indicates high sensitivity in some weld metals containing high copper levels. Not all of this can be assigned to the copper level. A more systematic study of various steels typical of those in service is needed to aid in the projection of embrittlement and to guide establishment of regulatory criteria.

The implications of radiation embrittlement to reactor operation has been limited to projections based upon research and design data. Nevertheless, conservative application of these data provides a technique for minimizing the potential effects by controlling vessel temperature and applied stress during normal startups and shutdowns. This technique is especially applicable to the pressurized water reactor where pump heat may be used to gradually increase vessel temperature before initiating nuclear power. This technique has little value for minimizing the effects of low energy ductile upper shelf however, a problem which is probably the most critical if we are to assure vessel reliability. The fullest implication of radiation embrittlement to reactor operation must await pertinent surveillance results. SUMMARY OF STATUS FROM SURVEILLANCE VIEWPOINT While surveillance is not designed to provide research data, such results can contribute significantly to the fullest understanding of radiation embrittlement of vessel steels. Combining research and surveillance results permits an assessment of the status of knowledge as well as future needs. The fracture behavior of vessel steels as evaluated for surveillance has been based largely on small Charpy V-notch specimens. Thus, the projection of effects in terms of vessel fracture potential has been based on extrapolation from large unirradiated tests, on linear elastic fracture mechanics evaluations, and on radiation produced

3 changes based on small surveillance specimens. This three-way analysis is useful but is not adequate for the direct quantitative evaluation desired. The ideal would be an evaluation of the fracture behavior of the poorest component steel in a vessel - an irradiated K curve for that material. The best hope for reaching this goal involves experiments on large test specimens coupled with an acceptable quantitative analysis based upon a smaller specimen which has been corre- lated with the larger specimens over the transition and upper shelf regions.

For the ultimate test of neutron radiation on steel fracture, large irradiated fracture mechanics specimens must be tested for AT (irradiation induced change) and for low shelf (radiation induced shelf drop) conditions. Acceptable small correlation specimens which describe the heavy section results must be found for surveillance if direct fracture performance of the vessel is to be defined from surveillance.

Standard techniques for describing peak radiation exposure which have been defined for research experiments are generally adequate for surveillance but the flux at peak locations should be determined by dosimetry surveillance runs. Further, projection by computer analysis from surveillance location to vessel wall offers ,a chance for misinterpretation, as does the use of a 1 MeV cutoff for defining damaging neutron fluence which should be modified to encom- pass all neutrons >0.1 MeV.

One of the major contributions of surveillance has been in defining the specific response of various vessel steels or components of steels to service radiation. Generally, such results have vali- dated the research results relative to composition (copper and phosphorus) effects but on the actual steels used in vessel construction. The most startling observation has been on welds of vessels constructed before about 1970 wherein copper was used as a coating on weld rods and therefore were high in copper and sensitive to radiation embrittlement. Such results are clear from initial surveillance data from plants such as Maine Yankee, which contained high copper welds and a related high level of embrittlement, a major vote for standards to control copper in vessel steels. By contrast, where steel composition has been controlled (especially copper, phosphorus, and sulfur) as in the plate and forgings of later USA reactors, the sensitivity to radiation and the upper shelf toughness are superior to all earlier steels and hence offer the utlimate answer for vessel surveillance - improve steels to the point where surveillance is no longer needed or is minimal at most. The meshing of projections from both research and surveillance in this case represents one of the most conclusive and positive results of the whole area of steel embrittlement study.

The application of surveillance results have been used in many cases to establish operating guidelines for reactor startup and shutdown. It is believed that such guides, which often are applied routinely while major changes in vessel toughness may be years away, would not have been so applied if it were not for a well developed body of data on radiation embrittlement from research experiments obtained in years of study.

4 APPLICABLE NATIONAL CODES, REGULATORY GUIDES, AND STANDARDS Because of the relative maturity of this whole area of study as noted above and its importance to reactor safety, a series of major national codes, guides, and standards have been issued. For use of reference and comparison, Table 1 provides a listing of key documents, their title and basis. In addition to these there are a series of ASTM standards which relate to the question of radiation embrittlement and surveillance in a secondary way. Most notable of these are those which define procedures for measuring the neutron environment and which complement E185 on surveillance and the evaluation or amelioration of radiation effects. Table 2 lists many of these secondary standards which are important. In addition, ASME has contributed to the area of in- service inspection with Section XI of the Boiler and Pressure Vessel Code. In spite of the fact that several well developed codes, standards, and guides are published and in practice in the USA to support reactor pressure vessel reliability, there remain important research tasks and opportunities for improvement of these documents. Advances to support the needed improvements in these national documents (and nuclear technology) can be projected for each of the documents listed in Table 1. These are summarized in Table 3.

TABLE 1

MAJOR USA CODES, GUIDES, AND STANDARDS AFFECTING REACTOR VESSEL SURVEILLANCE AND INTEGRITY

Document No. Title ASTM E-185 (1963)* Surveillance Tests for Nuclear Reactor Vessels

ASME Sect.lII, APP.G (1972)* Protection Against Nonductile Failure

AEC 10 CFR 50, APP.G (1973)* Fracture Toughness Requirements

AEC 10 CFR 50, APP.H (1973)* Reactor Vessel Material Surveil- lance Program Requirements NRC Reg. Guide 1.99 (1975)* Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials Date first issued.

5 TABLE 2

USA STANDARDS OF VALUE IN REACTOR SURVEILLANCE

(from 1975 Annual Book of ASTM Standards, Part 45, Nuclear Standards)

Standard Title

E-170-63 (1968) Definition of Terms Relating to Dosimetry E-181-62 (1968) Analysis of Radioisotopes E-184-62 (1968) Reco Practice for Effect of High-Energy Radiation on the Mechanical Properties of Metallic Materials

E-261-70 Measuring Neutron Flux by Radioactivation Tech- niques

E-262-70 Measuring Thermal Neutron Flux by Radioactiva- tion Techniques

E-263-70 Measuring Fast-Neutron Flux by Radioactivation of Iron

E-264-70 Measuring Fast-Neutron Flux by Radioactivation of Nickel

E-265-70 Measuring Fast-Neutron Flux by Radioactivation of Sulfur

E-266-70 Measuring Fast-Neutron Flux by Radioactivation of Aluminum

E-343-72 Test for Fast-Neutron Flux by Analysis of Molybdenum-99 Activity from Uranium-238 Fission

E-393-73 Measuring Fast Neutron Flux for Analysis of Barium-140 Produced by Uranium-238 Fission

E-418-73 Fast-Neutron Flux Measurements by Track-Etch Technique

E-419-73 Guide for Selection of Neutron Activation Dector Materials

E-481-73T Measuring Neutron Flux Density by Radioactiva- tion of Cobalt and Silver

F-590-74 Reco Guide for In-Service Annealing of Water Cooled Nuclear Reactor Vessels

6 TABLE 3

SOME REFINEMENTS DESIRED IN KEY NATIONAL CODES, STANDARDS, AND GUIDES

Document Refinement Desired

ASTM E-185 Materials selection criteria, shelf analysis; define dosimetry approaches and fracture specimen choice.

ASME Sect.III, APP.G More large section tests to validate KIR curve for other steels.

AEC 10 CFR 50, APP.G K I curve for other steels; for irradiated steels.

AEC 10 CFR 40, APP.H Better guidance for materials selection and dosimetry analysis.

NRC Reg. Guide 1.99 Better statistical base for both steels and fluences to define both at AT and AE. Reward for low sensitivity.

USA REACTOR SURVE ILLANCE RESULTS - A SUMMARY

The scope of this paper does not permit a detailed review of each USA reactor which should have produced surveillance data by this time, but some general comments are provided for the several generations of reactors in the list. A group of twenty-seven USA reactors have operated long enough to have produced surveillance results but many of these programs have not yet been analyzed or have not been released for public review. The list of twenty-seven is shown in Table 4. These may be divided in two ways; by stage of development (prototype, first generation, second generation) and by type (BWR, PWR, LGR).

For our purposes it is probably best to minimize cataloging by type and concentrate on status of development since general conclusions can be made best by the latter. The Hanford-N reactor, a light water cooled, graphite moderated reactor uses pressure tubes rather than a pressure vessel and hence should be dismissed. from this discussion. Most of the early BWRs had vessels similar to those of PWRs in type of steel and in projected fluences on the vessel so these can be treated together. The later, larger BWRs were designed for lower fluences and must be so reported. The later PWRs are all similar in the factors which affect surveillance interpretation so it is possible to discuss together in general terms the early group of both types and the later group of PWRs. The earliest prototypical reactors - Dresden I, Big Rock Point, Humboldt Bay 3, and LaCrosse (BWRs) and Yankee-Rowe (PWR) - were quite similar in lifetime fluences determined from initial surveillance except for Humboldt Bay and laCrosse which had lower projected

7 TABLE 4

Twenty-Seven USA Reactors for which Surveillance Results Should Be Available (May 1976)

Commercial Net Reactor Generator Reactor Operation hlWe Type Supplier Supplier Dresden 1 8/60 200 BWR GE GE Yankee 6/61 175 PWR W W Indian Point 1 10/62 265 PWR B&W W Big Rock Point 12/62 70 BWR. GE GE Humboldt Bay 3 8/63 68 BWR GE GE Hanford-N 7/66 860 LGR GE GE San Onofre 1 1/68 430 PWR W W Haddam Neck 1/68 575 PWR W W La Crosse 9/69 48 BWR Allis Allis Oyster Creek 1 12/69 640 BWR GE GE Nine Mile Point 1 12/69 610 BWR GE GE Robert E. Ginna 3/70 490 PWR W W Dresden 2 8/70 800 BWR GE GE Point Beach 1 12/70 497 PWR W W Millstone 1 12/70 652 BWR GE GE Robinson 2 3/71 665 PWR W W 1 Monticello 7/71 548 BWR GE GE Dresden 3 10/71 800 BWR GE GE Palisades 12/71 700 PWR C-E W Turkey Point 3 7/72 725 PWR W W Quad-Cities 1 8/72 800 BWR GE GE Quad-Cities 2 10/72 800 BWR GE GE Point Beach 2 10/72 497 PWR W W Vermont Yankee 11/72 514 BWR GE GE. Maine Yankee 12/72 790 PWR C-E W Pilgrim 1 12/72 670 BWR GE GE iOconee 1 7/73 887 PWR B&W GE I'I------ - -

8 lifetimes - 20 years versus 30 or 40 years. Lifetime fluences were found to be generally from 1 to 3 x 10'1 n/cm" (>1 MeV) and steel sensitivity was in the mid-range of that found lor steels studied in research programs. In every case some modification of the operating schedule to make provision for irradiation embrittlement is needed. In some cases it may be desirable to shorten the lifetime. The major weaknesses of these early programs were in capsule design, neutron dosimetry, materials selection, and fracture evaluation knowledge. The positive lessons learned from ihese negatives have had major impact on later surveillance programs and even on reactor design and on pressure vessel materials development.

The second group of USA reactors by stage of development are the intermediate power level PWRs, including San Onofre, Haddam Neck, Ginna, and Point Beach 1 and 2, which began to produce power in the late 1960s. The surveillance results offer two warnings: (a)relatively.high fluences on the vessel (3 to 4xl0 -), and (b) weld metal and weld heat affected zone materials sensitive to radiation and some of low shelf toughness after irradiation.

The next group of reactors to be assessed are the more standard large (700 to 1000 MWe) PWRs such as Palisades, Turkey Point 3, and Maine Yankee. While we do not have results on all three of these, data suggest that the same problems identified by surveillance of the intermediate size reactors apply to 1he larger ones. This reflects the procurement of vessels for these reactors, which first produced power in the early 1970s, before the major composition effects of copper and phosphorus and the shelf degradation of sulfur was widely publicized by NRL in 1967-1968. It is clear that for these older plants, vessel surveillance plus advances in the compltmentary technologies of steel fracture, neutron dosimetry, and reactor design will require great attention and diligence in the years just ahead of us.

SUMMARY AND CONCLUSIONS

This overview is too brief for precise statements of conclusions. Further the generalizations above are inappropriate for application to individual reactors but do provide a basis for a general look at surveillance as a complimentary effort to research studies of radiation embrittlement. It is fair to say in summary, I believe, that research data have provided the background necessary for formulating and interpreting results of reactor vessel surveillance programs. Furthermore, it must be clear to all by now that much remains to be done in order to optimize the results of such programs, since the next 20-30 years for several early light water reactors, will require extreme vigilance to assure no catastrophy traceable to a reactor vessel failure resulting, even partially, from radiation embrittlement.

In order to optimize the application of future surveillance results several tasks require advancement. These include, in the authors schedule of priorities,the following:

1. Promote the advancement of limits on residual element content to full status of an enforced national code. (This applies both to elements affecting radiation sensitivity and shelf toughness.)

2. Better understanding of the full section fracture potential for steels ol low upper shelf toughness.

9 3. Advance criteria for describing neutron exposure in surveillance for selecting lifetime peak fluences on the vessel to status of mandatory code or standard.

4. Develop procedures for improved surveillance specimens based on quantitative fracture mechanics procedures. At the same time by correlation or other means advance our understanding of Charpy V-notch toughness data which will remain the corner- stone of surveillance for many years.

5. Catalog all surveillance and research data in terms of composition and embrittlement.

6. Develop fuller understanding of the implications of the gra- dient in toughness through a reactor vessel wall. Advance criteria for reducing overconservat:isms which may be attrib- uted to current methods of analysis.

7. Assure criteria for new surveillance programs which enforce selection of capsule location at peak fluence location and selection of vessel component that is most sensitive.

10 SURVEILLANCE PROGRAMVIES PREPARED AND CARRIED OUT DURING

PRODUCTION AND EXPLOITATION OF THE A-1 NUCLEAR REACTOR PRESSURE VESSEL

Milan Brumovsk.,a Radislav Filip Skoda Works, Nuclear Power Plants Division Research and Development Centre P 1 z e n , dSSR

Jir dcervaek, Miroslav Vacek Nuclear Research Institute A e E, SSR

ABSTRACT

The first Czechoslovak nuclear reactor A-l is a !r'IGCR-type reactor working with the thermal neutrbns. Ist nominal output is 150 :Ue, The reactor pressure vessel is fabricated from mild structur[4 steel and is therefore characterized with some specific parameters such as:

- relatively low operational temperature (cca.1500 C), - l1rge vessel diameter (cca 5100 mm) resulting in the follovin specialities in technology: - the vessel rings are electroslag welded from segments ever. in the core area, - the rings welded into segments were mrnual-arc welded (sulte weld) at the site to make the whole body of pressure vesj.'c which was annealed, directly in the pit, in a special clcctric furnace to relieve stress, - relatively high ratio of fast neutrons to photons fluxes.falling on the pressure vessel wall (cca 1:100).

Regarding the above and the fact that the projects of the pressure vessel and of operational checks (expecially of surveillance sp;ecimens) had ,boen prepared at time when fracture mechanics had been far from reaching present state (the pressure vessel was finished in 1968), the suveillance snpcimrens project was broadly worked out. Its main purpose was to verify the lifetime of the pressure vessel itself Cplanned at least for 20 years), however there was an auxiliary research aspects, too. At present methods are being sought how to modernize the existing and fixed surveillance prograsnr, -to make use of the latest knowledge in fracture mechanics.

11 1. iPre:ssire Vessel Chraracteristics. The A-1 reactor pressure vessel was described in detail previously 11 . Nevertheless it may be of use to present here the following basic technical data: The diameter of the pressure vessel is 5100 irs, the thickness of the cylindrical part 150 mm, the nominal pressure is 6,4 MPa. The material chosen for the vessel is the 6SN 13030 Ni modified nnd Al+Ti treated mild structural steel. The chemical composition ;,nd mechanical properties of this steel are summarized in Tab.l. The steel was used after being air-normalized and tempered with subsequent annealing at furnace to relieve stress. Whereas the rings were electroslag welded from four segments, manual-arc circumferential welds to weld the vessel sections together were used. Besides ascertaining the basic mechanical properties a wide-spread research was made on brittle fracture resistance based both on crack arresting temperature approach (including size effect) and linear fracture mechanics (fracture toughness KC). The results of this study are given, for example, in /1,2,3/ . The calculated dose of fast neutrons (with energies above 1 MeV) striking the wall of pressure vessel in the core area is at least 3.1022 nm2. At the same time, according to both calculation and experiments, the dose of fast photons (also with energies above 1 MeV) is approximately 100 times as high, which makes a substantial differenct as compared to conditions in experimental reactors.

2. Philosophy of the Surveiflence S.ecimens PoTersmJne.

Two following approaches to the safety of nuclear reactor pressure vessels are being used today: - a temperature approach characterized mainly by brittle crack arrest temperature, CAT (sometimes also by nil ductility temperature,NDT). - an energy approach characterized mainly by fracture toughness,KiC.

Other approaches, i.e. a deformation one (critical opening displacement,COD) and an energy one based on the critical value of J-integral are not yet sufficiently worked out.

12 On account of historical reasons during the projection and n'niufcc - ture of the A-1 reactor pressure vessel, the concept of cracke'rres- ting temperature was used predominrm tly. It is somewhat more conservative, which to some extent may compensate for the increased dcanger of star ing and prop agating brittle failure caused by the high energy accumnulated in gaseous coolant. This approach is .lso .embodied in a standard /4/ at present applicable to nuclear aci- litics in CSSR. This approach does not consider the condi tionS of initiating a failure, i.e. the propagation .nnd growth of flawvs, but a temperature is looked for at which , under a given stress, the brittle crack, if any, stops propagating. If the operation-1 temperature is higher than the temperature established in this way, a catastrophical pressure vessel failure cannot take place (explosion). Only untightness may occur caused by subcritical fatique flaw growth.

The essential relation is Tworking ? CATQ + ( T damage) + T + ATE (1)

where CAT° is the startingscrack-arrest temperature ascertained on' nonirradiated material of a given thickness, subject to given stress. ( 4 Tdamage) is the sum. of changes of this temperature due to reactor operation (it includes effects due to irradiation, aging, low-cycle damage and the like) i.e. CAT,

A T Iis the safety factor (with regard to transition states), we put it + 30 ° for given steel,

4 TE is the crack arresting temperature increase due to accumulated energy (see /6 )

For the material chosen (the 6SN 13030 steel is practically non-aging; in the region of smooth rings near the core it is subject only to nominal stress well under the material yield point, i.e. in the elastic zone damage due to low cycle fatigue or deformation aging is negligi- ble) the only variable in the course of reactor operation is radiation aging which isshown by radiation embrittlemeht, this is dT. = A dnmatgeo ) The other approach, based on linear fracture toughness compares favou- rably with the first one inthat it is less conservative, ns it consi- ders primarily the conditions of starting brittle failure. Another advantage is the possibility to establish the critical defect size in

13 the material. FHowever periodic inspections must be made simultare- ously in order to check integrity of pressure vessel material. Thus the surveillance specimens programme cannot be used by itself. A costly and demanding programme to detect flaws during operation must be added. Sometimes it may however raise essentially operational safety. But this approach modified by NDT is incorporated in standards /5/ ,used in USA and West European countries.

The basic relation used is

KIC -= . ( 7 a c )1/2 (2) where t is the coefficient characterizing shape of flaw and body

' is the nominal stress

a c is the critical size of a flow. Fracture toughness KIC depends primarily upon the state of material and testing temperature. Operation-induced changes can be expressed as follows: KiC (t r O,T) = KIC (t = O,T - Tdmaged (3) where t is tine. T is the testing temperature. E( T'damage) stands for the shift of KIC-T relation due to operation, i.e. A TKi C.

The changes l( (A Tama) being known, it is possible to determine the fracture toughness value in a time instant in question and thus to establish the critical flaw size. From the results of flaw-detection checking it may be estimated whether the defects are admissible or not. Hence, using this approach allows to avoid any pressure vessel damage (i.e. even untightness can be excluded). However the stress field must be known in every instant of time, mainly during transition regimes, a prerequisite which is not always fulfilled exactly enough.

By comparing the two approaches we can see that the change of characteristic temperature is a common parameter for both the CAT -

( Tdamag e ) and KC (Tdamage)

14 With respect to the possibilities of surveillance programmes, the following problems arise in precise application of either approach: - is it true that

CAT = £ ( Tdamage).2 KT (4) and AK iC r(a Tdamae ) TK (5) where A TK is the change of transition temperature caused by material damage in the. course of operation and determined by notch toughness tests onr Charpy-V specimens. If the two relations (4) and (5) hold true, then the following relation must be valid:

CAT = TKIC ( Tdm e ) (6)

To assess the validityof relation (4) only incomplete datn are available so far, because of difficulties we are coming across in irradiating specimens of large thickness at the crack-arrest temperature. A number of data for relation (5) have been.attained recently in the frame of HSSTP t7).The validity of relation (6) is not direct apparent because it includes several tests with specimens having various dimensions and thicknesses and performed with different loading rates (d CAT and A TK by means of dynamic and A TKIC by means of static tests). Yet, all the analyses mede so far are based on the validity of this relation and it is also a basis on which to Set up programmes of surveillance specimens.

At present both approaches are practically used for analysing safety and service life of the A-1 reactor pressure vessel. The crack-arrest temperature still remains a prim.ry criterion, because of the fact that periodic inspections were not considered in the original project (these are included additionally where needed in accessible areas - see 181). With regard to the high value of accumulated energy and consequently to factor A TE in relation (1) and relatively low operational temperature of the vessel, we obtain, in connection with the supposed transition temperature shift due to operation, the requested working temperature close to or even higher than the operational temperature. For this reason the CAT approach doe. not ensure a perfect safety against failure and an analysis based on linear fracture mechanics must be made and critical flaw sizes established. The process is shown in t8) in detail.

15 It means, that the surveillance specimens program employs test sa7mples making it possible to ascertain the shift of respective temporature dependances - CAT or KIC or others (static bending etc.) In addition, in order to better analyse the pressure vessel state, it is .convenient to use a certain number of specimens for static tension. In most cases an increase in strength properties (yield point, strength) occurs but at the same time a decrease in plastic properties takes place (ductility, contraction) and thus the results obtained this way may complete both a general opinion and'the Fracture Analysis Diagram.

2. Surveillance Test Programme of the A-1 Reactor'Pressure Vessel. As _-referred previously 11,2,8,9/ , the surveillance specimens programme has, some specific features. Since the structure of the pressure vessel and internal parts prevent inserting large containers with test specimens (the gap between the'inner side of the pressure vessel and the heat shieldingbeing cca 80 mm) cylindrical containers are used containing one to six test specimens according to their types and sizes (see Fig.l). These containers are connected to one another to make chains 7750 mm long (see Fig.2). Some of them ensures that specimens are irradiated under the same stress as that in the pressure vessel wall /101 .

Sixteen chains are placed in the reactor vessel nltogether, always in groups of four turned at 900 around the circumference of the pressure vessel. Apart from these containers a certain number of semi-products for test specimens are also placed in the pressure vessel that are welded together to form bars attached to the inner wall of the vessel. Since the number of containers which can be put into areas of identical neutron fluxes is very limited, it was not possible to put in- to the reactor samples from all the heats and weld' joints used for reactor vessel in the vicinity of the core. A special characte- ristical heat was therefore chosen on which weld joints were made using the same technique as for actual pressure vessel. Testing specimens were made of it and introduced into the reactor. As radiation damage resistance of individual weld joints or heats may vary from one another, it was necessary to carry oit supple- ment reference test.

16 The overall programme of surveillance specimens is therefore divided into two parts: 1. Specific resistance tests 2. Long-term resistance tests.

Furthermore, work to Verify size-effect'in radiation embrittlement, i.e. relation (4) and (6) resp.. was also done. The scope of individual parts of the. programme is shown in Tab.2.

2.1. Specific resistance tests

These tests were carried out before starting up the reactor on power operation. The recommendation from (11l could not be used because our conditions are quite different: lower irradiating temperatures when the contents of interstatical atoms (C,N) plays the main part and another type of steel used.

These tests served for determination of the differences among the individual heats or weld joints used for the manufacture of the cylindrical section of pressure vessel in the core area in terms of radiation damageresistance. The samples were taken direct from welding specimens in individual rings. Irradiation was carried out in the WWR-S experimental reactor at Nuclear Research Institute, Ae2 under the conditions simulating 20 years of service so that we can speak of accelerated irradiation tests. Irradiation temperature was approx. 85°C (121. The experiment was to' determine which of the materials studied had the minimal radiation damage resistance and what is its difference as compared with the heat chosen for inserting into the reactor within the programme of long-term stability (to insert the least resistant charge into the reactor was not possible because of lack of material needed to make specimens for this vast programme). Industrial materials (basic material, weld metal and heat affected zone) were evaluated in terms of strength properties (primarily yield point) and notch toughness (transition temperature). The following weld joints were selected for irradiation: (each time the two basic metals, the two heat-aCfected zones at a distance of + 4mmm from sharp boundary'to basic material and weld metal):

17 ring V: ring VI: heat No weld No heat No weld No

M 2112 A -~^^24 M 9469 A - 36 M 4576 Z M 0097 Z M 2456 Z 2 M 9479 A M 4535 A 26 M 73Z 3

ring IV weld No. 112

ring V weld No. 113

ring VI weld No. 114

From these electroslag welds are : 24,26,36,38; Manual-arc welds are: 112,112,114

1280 miniature impact specimens (70 transition curves) and 620 miniature tensile specimens were evaluated altogether. Fast electron doses (of energies above 1 MeV) ranged from 0,7 to 5,1022 n/m2. All the values of the transition temperature shift founded are sumnarized in Fig.3. The following notation is used in this diagram: ES - electroslag welding joint MA - manual arc welding joint BM - Base metal HAZ - heat affected zone WM - weld metal

Besides the test results of the specific resistance programme, the results of radiation stability tests carried out previously in the same reactor designated 6SN 13030 are also presented. As an outcome resulting from the tests performed a band of values may be gained co- vering approx. 95% of all results.

The changes in basic mechanical properties - yield point and transition temperature - can be also represented as follows:

Z40,2 = A1/ 2 . (0t . 10-22) 1/2 (7)

TK = B1/2 . (0t. 1022) 1/2 (8) where A,B are material constants (depending on irradiation temperature, too) 0t is the neutron dose in n.m .

18 After evaluating the individual regions of weld Joints according to equation (7) and (8) we have the following results:

BM - ES - A1/2 = 76 7 [MPal

- MA - A1 / 2 78 6 [ MPa ] MP 1 / 2 9A *- , 72 7 [MPa WM - ES - Ag/2 =65 + 9 [MPa

- MAx/ A1/ 2 = 56 7 [MPa]

(x/ - lack of specimens). Results comparison shows there is a slight difference (cca 20%) between basic material and weld metal' and the overall error is also relatively low (10%) which confirms a good reproducibility of results.

Similarly, with the transition temperature ch-nge we have:

BM - ES - B 1/2 = 28 + 1,9 I°C) = + - MA - B 2 . 33 2,9 L°C1 0 HAZ - ES - B1/2 =35 T 2,6 1 C

- m - B1/ = 30 1,7 [°C] - WM -ES - B 1/2 = 29 2,4 [° 1 - MA- B/2 = (17)

In contrast to the results of tensile tests no systematic differerce in individual locations and types of welds were found out for the transition temperature changes. A reason for that may be a somewhat other kindtmicrostructure between basic material and weld metal end consequently a different relation between microstructure (grain size) and testing cross-section, the letter being very small for tensile tests (0 2mm). By comparing equations (7) and (8) it may be 'assumed that the ratio of constants A1/2 : B 1 /2 is constant for identical.materials. The following values were determined for the tests.given:

= BM - ES - A1 /2 : B1 2 2,9 0,5 [MPa.00Cll

- MA - A1/ : B1/2 25 ± 0,4 Mpa.o°C-l 2, + ,50MPa. C1 2 [ Ma'°'I 2,7 + 0,5o-'JPa.°C'3 1 WM - ES - MA1 /2: B/2 ==[MPa.°OC- 2,3 0,4

The assumption of the constant ratio Al/2 : B1/2 for material given and irradiation conditions was thus confirmed and showed again there

19 was practically no difference in radiation damage of all the materials tested. Moreover, on the basis of static tension tests we can Judge of change (increase) in the transition temperature from change in yield point, Which may be important in determining with more precision the pressure vessel lifetime by surveillance specimen tests. It has thus been shown that anyweld joint 'or its portion placed in the pressure vessel in the vicinity of reactor core in rings V and VI (maximum neutron fluxes) does not exhibit reduced resistance to radiation damage, mainly to embrittlement. The drawn range of values of 95% reliability can be taken as a basis for further comparison and analysis. This range also includes tests performed previously with the same steel.

2.2. Long-Term Resistance.

The long-term resistance programme is somewhat broader and more. complicated. It covers specimens manufactured just from one reference heat M4548 End its weld joints and consists of the following tests (see Tab.2):. - static tensile tests, - impact notch toughness tests with specimens of "hot laboratory" - 0 5,3, Mesnager and Charpy-V types, - static bending tests with V-notch specimens. Transition temperature shift is defined according to 141 by a criterion called 1/3 decrease in maximum load.

The programme' itself is again divided into two parts - 1. accelerated irradiation tests - 2. surveillance specimens tests placed in the reactor.

2.2.1. Accelerated Irradiation Tests. These tests were aimed at finding relation to individual heats end welding joints located on the vessel and irradiated within the specific resistance programme. At the same time they were to set founda- tions for preliminary determining lifetime of the A- reactor pressure vessel.

The following specimens were irradiated: BM - miniature tensile specimens and impact specimens, WM - impact specimens, HAZ- impact specimens. 20 Test specimens were taken both from manual-arc and electroslag welds. Meshager (R2), Charpy (RV) and mini-impact specimens (0 5/3) types of samples were used for notch toughness teats. The Charpy-V type specimens were most convenient giving the best reproducibility of resulst.

Irradiation was also carried out on VMWR-S reactor in waterproof cases, irradiation temperature was about 850C. Doses for various materials ranged between 2,1 and 9,1.1022 n.m 2.

All the results obtained are shown in Fig.4 (notation is the same as in Fig.3) /13/ .

It may be seen that all the results fall practically into the previously determined value range of 95% reliability. It did not become evident that some weld joints or joints location were more sensible to radiation embrittlement than other ones. Hence, in further considerations we may take into account the upper enveloping boundary of transition temperature changes that reaches the following values approximately: doses 6t : 0,5.10 2 ,1,0.1022 3.1022 10.1022 rn.mJ- shift TK: + 30 + 45 + 75 + 105 °C

In addition, the effect of irradiation temperature was studied, too. The accelerated irradiation was performed at a somewhat lower temperature (cca 85°C) as compared with the operational temperature .(cca 150°C). Besides irradiating in non-heated cases irradiation in apecial rigs at temperatures 150,187/ and 201°C was conducted. The results are summarized in Fig.4; they are designed BM-T with irradiation temperature.

The results show that after irradiation at temperatures under 180°C appreciable radiation damage recovery accurs. Results of the irradiation at 150°C indicated that choise .of the irradiation temperature for most of the test temperatures (85°C) was justified - the results are free of errors resulting from different irradiation temperatures.

2 2.2. Surveillance Specimens.

These test specimens were also manufactured from reference heat M 4548 and its welding joints. A survey of the types of tests is given in Tab.2o

21 In parallel to the customary test specimens special ones are also inserted in containers, in which they continue to be kept under a tensile prestress corresponding to that in the pressure vessel wall /10 . The aim of these tests was to establish the effect of long- term prestressing on change in mechanical properties (tensile prestress might accelerate difusion processes and 'so influence the resulting radiation damage). Some of the preliminary'results of accelerated irradiation tests were already shown /9/ .

The total number of 16 case chains with test specimens is divided into four sets: - I-1, I-2, I-3; - II-1, IIf-2, II-3, II-4, II-5, IT-6; -III-1, III-2,III-4, II-3, III-5, III-6; - IV-1.

In each set the specimens are arranged height-wise and in individual chains in such a way that, they could make a group in order to,hr.ve a no:ded number of specimnon for one tranaition curve from one material by one type of test. Here it is assumed that set I will be withdrawn first (approx.after five-year operation) and will be used to the preliminary comparison of forecast and actual changes in material properties and neutron doses,.Sets II and III will be used to assess lifetime of the pressure vessel with more precision following 10 and 15 years of service. Set IV will be used to improve assessment of residual lifetime pt the end of operation, i.e. after 20 years.

Every container comprises also indicators of neutron flux (Cu) and irradiation temperatures (powder diamond).

203. Size-effect Factor.

The crack arresting temperature - CAT -'depends not only upon the status of material but also upon specimen thickness - the latter dependance is known as the size-effect. Previous work carried out at SKODA-ZVJE work /1,2,3/ on the ZZ 8000 machine has shown that increase in specimen thickness by 50 mm leads to increase in crack- arresting temperature by about + 10°C, which represents an increase of at least + 30°C when going from 10 mm(the thickness of standard specimens) to a thickness of 150 mm. 22 The validity of relation (4) remains to be answered, i.e. whether it is influenced by the size-effect.

To verify this fact a series of tests was performed both on the ZZ 8000 machine and on small specimens. 'As the irradiation and evaluation cannot be made on specimens of actual thickness 'nother technique had to be used to imitate radiation damnge. A method of artificial mechanical ageing was chosen which provokes practically the some outside damage effects - hardening (increase. in yield point and strength) and embrittlement (increase in transition temperature). Artificial mechanical ageing (3; 6,5; 7,5 and 8,5% of plastic strain)was performed direct on the machine with test specimens of actual thickness (150 mm), followed by a dwell at 2500 C. After cooling down a test was made to determine crack arresting temperatur, by ESSOmethod (i.e. dynamically) at a rated stress of about 150 MPa. After carrying out the tests a new specimens were manufactured from the remainders of the test specimens for notch toughness test of Cherpy-V type and tests were performed. The results of tests are summarized in Fig.5. 114/ . On the x-axis the transition tenperatures are plotted as determined on - 2 RV type bars by a criterion 35 J.cm , i.e. T5 . On the y-axis the crack-arrest- temperatures are plotted. By connecting the respective points, a dependence of the change in transition temperature f'om notch toughness tests ( TSv) upon the change of the crack-arrest temperature ( CAT) has been plotted. The results indicate that both the changes, obtained from different methods, are in a very good consistency - the change of crack ar3esting temperature is somewhat less than that of transition temperature, which may be caused by slightly uneven deformation distribution along the body thickness. But the validity of relation (4) may be considered to be proved experimentally: no size effect was found out in the transition temperature shift determined by dynamic tests. Similar work to verify relation (5) is being performed.

3. Lifetime Evnluation by Surveillance Programme.

Lifetime evaluation of the A-1 reactor pressure vessel is being done basically by methods given in Chapter 2. The analysis is being conducted in two steps: first, from the point of view of CAT, secondly from the standpoint of critical flaw size. (esulting comparison of both results is used to improve lifetime assessment and also to pick out locations for operation inspections and to set accuracy demands. 23 3.1. Tempernture Approach.

This approach consists in estimating two factors: - crack-arrest temperature, i.e. Fracture Analysis Diagram, - transition temperature shift due to operation (irradiation).

This approach is-given in Tab.3e Bearing in mind the above consideration it is assumed that ( Tdamage) ATir

The following ATir relations are being determined during analysis:

bTir mat - shift depending on material (i.e.on heats,weld joints) 4Tirshape-shift depending on specimenshape and test type (notch toughness, static bend, Charpy-V, Mesnager.etc) aTir,spec -shift due to the difference of neutron spectra in irradiated location on experimental reactor and in A-1 pressure vessel.wall.

Whereas the first two relations are being determined experimentally, the last one (spectrum effect) must be calculated. On the basis of calculation and preliminary results a correction to allow for these differences can be done. With neutron doses of energies above 1 MeV radiation damage tomaterial of the A-1 reactor pressure vessel is expected 20% higher :than it was found out at accelerated test in VWIR-S reactor.

2o2. Fracture Touehness Approach.

Within this .nralysis the assessment of critical flaw size for pressure vessel material is being made by methods of linear fracture mechuinics.. Knowing fracture mechanics parameters, rate of defect growth (da/dN), initial defect size in material (a .) and supposed changes of all these parometers in the coutrse of

operation, (irradiation) the critical flaw size a c (t,N,T) may be forecast after a given operational period in relation to service time (t), number of operation cycles (N) and operational temperature (T). A scheme .of this assessment is given in Tab.4.

To estimate the irradiation-induced change in mechanical properties, i.e. (KiC) = f (0t, T), relation (3) is used and consequently the some transition temperature shifts 4 Tir =ACAT as in section 3.1. 24 The remaining procedure is practically independent of radiation influence and is the same for both irradiated and non-irradiated portions of pressure vessel (i.e. except for the core region).

4 Conclusio n

The first two stages of the evaluation of the pr.ogramme of surveillance specimens fabricated from the A-1 reactor pressure vessel material were accomplished. Most valuable .data on pressure vessel material behaviour (basic material and weld joints) in the course of neutron irradiation have'been'attained. The essential and fundamental results obtained are as follows: 1/ Practically no difference in radiation damage resistance for individual heats and for weld joints used for the smooth.part of the A-1 reactor pressure vessel (in the core region) has been found out. 2/ Good agreement of radiation damage resistance of materials used for actual pressure vessel with materials manufactured within research and development work has been reached. 3/ Yield point increase - transition temperature increase relationship for given material has been established. 4/ It has been confirmed that in transition temperature shift (or crack-arresting temperature) induced by irradiation or ageing the size effect (or effect of specimen shape and test method does not become evident (at least with dyn3amc. tests). 5/ The results obtained serves for lifetime assessment of the A-1 reactor pressure vessel by means of crack-arrest temperature as well as fracture toughness approaches.

When evaluating the surveillance specimens programme for the A-1 reactor pressure vessel in general, one must bear in mind that the surveillance programme in question is 'not a standard one and can also serve, to p 'great extent, for research purposes.

REFERENCES

/1/ Brumovskj M.,Becka J.,Urban A.- "Experience from the Manu- facture and Testing of the Pressure Vessel for the A-1 Reactor", IAEA Symposium on Performance of Nuclear Power Reactor Components, Prague,November 10-14,1969,IAEA Publication STI/PUB/240, p.405, Vienna 1970

25 /2/ Brumovsky M.,Filip R.,Indra J.,Kdlna K.,Komarek A.- "Operational Safety of Pressure Vessels at Czechoslovak Nuclear Power Stations", IAEA Panel on Recurring Inspections of Nuclear Reactor Steel Pressure Vessels,Plzen,1966,IAEA Publication STI/PUB/81, p.73. Vienna 1969

/3/ Brumovsky M.,K6lna K. Vacek M. - "Safety of Reactor Pressure Vessels from the Standpoint of Brittle Fracture", 4th United Nations Conference on "Pecaeful Uses of Atomic Energy", Geneva 1971, Vol.3, p.265, Vienna 1972

/4/ HopM. pacueTa Ha nponHocTb azeMeHTOB peaCTopoB, naporeHepaTopoB, cocyAoB X Tpy60npOBOROB aTOMH5X OaeKeTpocTaHxMr OuWTHWX K MccAe- AOBaTejbCKMy sAepHKx peaKTOpoBD ycraHOBOK, MOCKB8, "Mewaazyp- rxq", 1973

/5/ ASME Boiler and Pressure Vessel Code, 1974 Section III - Nuclear Power Plant Components; Section XI - Rules for Inservice Inspection of Nuclear Power Plant Components

/6/ Brumovsky M.,KAlna K.,ltepgnek S.,Urban A.- "Study of Crack Initiation-Arresting Conditions in Plane Plates and Cylindri- cal Model Vessels", 3rd International Congress on Fracture, Munich 1973, Vol.II, Paper No.224

/7/ Breggren R.G.,Canonico D.A. - "Toughness Investigation of Irradiated Materials", Quarterly Progress Report on Reactor Safety Programs Sponsored by the NRC Division of Reactor Safety Research for April-June 1975, II.Heavy-Section Steel Technology Program, ORNL-TM-5021,Vol.II,pp.24 - 32

/8/ Brumovsky M.,,tBpAnek S.,Havel S.,Plantk V. - "System of Recurring Inspection in the First Czechoslovak Nuclear Power Plants", Inst.Mech.Engrs.Conference on periodic Inspection of Pressurized Components, June 1974,London,Institute of Mechanical Engineers, 1974,Paper No.C 81/74

/9/ Brumovsky M.- "Radiation Damage and Surveillance Programs for Czechoslovak Reactor Steel Pressure Vessels", IAEA Technical Report No.IAEa-117 on "Development of Advanced Reactor Pressure Vessel Materials", Paper No.6, Vienna 1970

10/ Louda J. - "Equipment for Detection of Radiation Influence on Material of Nuclear Reactor Pressure Vessels, Czechoslovak Patent No.103 844 (1961)

11/ Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,ASTM E 185 - 73

12/ Vacek M. - :Radiation Damage of the A-% Reactor Pressure Vessel Welding Joints after Irradiation at 85 C (in Czech), Nuclear Research Institute, Report UJV 3621 - M, Re2 1975 26 /13/ Vacek M., terv6aek J.,Havel S.,ChamrAd B.,Pav T. - "Radiation Embrittlement of Surveillance Specimens from the A-1 Reactor Pressure Vessel after Accelerated Irradiation, (In Czech), Nuclear Research Institute, Report tJV 3807-M,Ret 1976

/14/ Brumovsky M, - "Size Effect in Irradiation Embrittlement of Steels", ASTM 8th International Symposium on the Effects of Radiation on Structural Materials, St.Louis, May 4-6,1976

Tab.1. Chenmical coniposition and m ochanical propertios oJ' (Si 1.t030.,s) IBi,All-l:i stoel

0 iin Sij P S Mi Cr

max 1i.10 0 20 maLYx max MaX *- max o.20 1.40 0.20 . 00020 .020 .45 0.20

E^l--^fon gation. _i, ijT.. Elnation ol na

min. )n. iin. nin. 220 4"0 22.0 45.0 ......

27 ( aI I + I I I-I a) 0

CO C) 0 C) .'D- A i I I + I ! o tF N 0

P 04 _ I I _ I r0

0*. __H.sI _+ I + I

IM' " I §" ' 1 + 4 E01 rCI (pf P, .2, 0 r _ I _ .1 0 4 C 0 ------0 3 U2 I I 1+I _ + +I+ 1

s + i ' +ii t +I E-f0

ci EI · I I I Ii+ + + c ri Oa c3) - S' [I ( +- i + bD CQ-«R d ------a)) a rlt43 SI· 1 1 + + + + tcul * c V bo o U) 94 d O H z iu 0 D *1

To 10 o d (D< a) Q) wU) CQa 03 Q) O 4o Co 01 . BI)la r Q) d COo h-;kdF4 s: F4 F4 a) 4t4,m) 4-'$ wS40a - (D ci oq ? 4 i I i L

d0C d,(0

d0 ) P4 ) cad ) P pi 4)4 Q) 6 (qtU m COP

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4> -d 0

Q) i-^ . ------P "0 h0F4 P4 u~~~~~~~~e Qa 04 ~ ~ ~ 0) 4,

0.Q) FA

0- o CJ Ut G4 o

F: 040 -p ,) a0) cO FA *.4 a.> (P4 F:fd P< q(4 ,9 -p 09 .r4 CO 4,> PA 4> Cd E-4 .-4 Fc 0 0) \I . .H PA Q) 0 E4q4 C4- Q) Wm 0 o. P4 1 $4 Q) n , 0 94 0 0 n tRl )r $4 -p 0 E-4 n *r CQ N 0Q 'H4 -) C-4 0 (a ^ c '-i M^ 0) ( w P c6-p ) c $4 0) *H X!': (S -r- 01 *- - UT 0, V 0 Uh 0 C2 CP4 4-) .-k ;«0 o F3 C) :i II Hr ii g 11 ri c -.- 0 0 Q) 0 dr4-t be-rt (4 $r0I > > rl t>0 1 r-4 -d

29 4

(-1

(4 3 o r- *4 > .0 0

Pd 4

*O0, 0d F ^= 0 t- r-t * c3 o -p. Ms -f- o d H o

-o0

(/020

Fa .A

30 Fig.L. Scheme of container for irradiation of pre-stressed surveillance specimens

31 AFUTROM fILX WON THE' M/S/D PRiESSURL ' VESSEL ,'LL. .^10

Fig.2. Schematic view of the A-1 nuclear reactor, showing relative location of surveillance containers and neutron flux levels at various locations

32 ATC lO] 400

80 iai A" a 70 A A k

A a

40 70n i A .2 0 s/ 0 0 A U 20 ES MA SN _____^ A2,a050 WM o · 40 '/ HAZ A A BM C_0 12 .. L

0.5 .G 2 3 4 5 6 7 8 0o' {i t n.M, E,,> 41'Me,;

Fig.3. Results of specific resistance program

33 ATK E0CI

®-s50"c

80 1 A A

0 5-11B711 ¢o 4 13 0 ®-2o01C 50 0

___ R2 RY I 20 BM 0 C ES-WM D Il - HA A A MA-WM +4 -HAt V

6^T® ,, ___

0 1 _ -- 1__I __ I 05 0o. 0. 1x1x10 3 5 ¢ 7 8 IOx4o2e t EIh n F^h> 1MeVJ] Fig.4. Results of long-term resistance program 34 AT 5 L 0C3 0 -44 +20 T30 hi+0 +50

1 a ' I 1 4 j' I i 1 f~~~~~~~~ 440 ACAT rC] I~~~~~~I +30 450 1

CAT --J l°C3 +20 440

-4 -

+30

0 +20 i

+40 L I I i I , A i 1 I -30 -20 10 :0 4 40 +20 'C^ L "- 1

Fig.5. Comparison in transition temperatures shifts, produced by artificial mechanical aging

35 Present Status of Surveillance Tests for

Nuclear Reactor Vessels in Japan

S. Miyazono Chief of Mechanical Strength and Structure Lao. JAERI JAPAN

Abstract

In the presentation is explained the present state of surveillance for Light Water Power Reactors in Japan by comparing ASTM Designation, E185-73 and JEAC Standard, JEAC 4201-1970 which are issued by the American Society for Testing and Materials and Japan Electric Association, respectively and some future subjects will be proposed.

Surveillance tests on the nuclear powexreactors in Japan

are now being performed according with the recommended practice of

JEAC 4201 which was issued by the Japan Electric Association in 1970.

This practice is nearly the same as ASTM E 185-66 in the United States, which was revised as E 185-73 in 1973, but there are some minor diffences between them as follows;

1) In JEAC 4201 only nuclear reactor ressels are dealt with, while in ASTM E 185-66 nuclear reactor vessels and internal structural components are covered.

2) In ASTM E 185-66 only the significance of the surveillance test is described, while in JEAC the kinds of surveillance test, testing procedure and materials are specified in detail.

3) In ASTM E 185-66 tensile tests are performed at the service temperature of the components being surveyed, while in JEAC they are carried out at room temperature.

In Japan the recommended practice of JEAC 4201 shall be revised in the near future and in JEAC is advanced the preparation for revision.

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38 MATERIAL SURVEILLANCE PROGRAMME OF PRESSURE VESSEL STEELS IN INDIA

K. S. Sivaramakrishnan Radiometallurgy Section, Metallurgy Group, Bhabha Atomic Research Centre, Trombay, Bombay 400 085, India.

ABSTRACT

The surveillance programme of pressure vessel steels in India is reviewed. Details concerning Tarapur and Rajasthan nuclear power plants are presented. Tests to be carried out in the BARC Hot Laboratory Facilities on the irradiated specimens are described.

Introduction

Pressure vessel steels in nuclear power plants are exposed to an environment of neutron flux and temperature which brings about changes in the mechanical and physical properties of vessel material. These effects are manifested through an increase in tensile properties and reduction in ductility. Further it is also manifested in the shift in the Nil Ductility Transition temperature and also brings down the maximum shelf energy. Nuclear Pressure Vessels being a thick walled structure and having stress concentration sites due to presence of internal flawst brittle fracture can be of the principal forms of failure. Therefore, in addition to having control on the initial condition of the material- during fabrication and subsequent thermomechanical treatment, it becomes.necessary to monitor changes in the material occurring during reactor operations.

Surveillance Programme

In India there are two types of operating reactors viz.t the Boiling Water Reactors at Tarapur and the Pressurised Water Reac- tors at Rajasthan. We have a surveillance specimen programme insti- tuted by the General ELeetric Co,, U.S.A. for the Reactors at Tarapur. We have initiated a surveillance programme in connection with the end shield material utilised in Rajasthan reactors.

39 Materials and Specimens

Tarapur reactor pressure vessels have been fabricated from ASTM-A-302-B steel. Details with respect to composition and heat treatment are given in Table 1. Charpy V-notch impact and tensile samples prepared as per ASTI specifications for base metal, weld deposit material and heat affected zones have been placed at different locations near the vessel wall and also at some accelerated neutron flux positions compared to the pressure vessel wall (Fig. 1). In addition, thermal control specimens are kept at locations of insignificant neutron flux levels to assess the effect of temperature on these specimens. No temperature measurement monitors have been installed since it is a boiling water reactor, and as such the operating temperature can be considered essentially constant.

The material of the end shield in RAPP reators is ASTM-A-203 Grade D (3.5% Ni) low alloy steel. Details with respect to composition and heat treatment is given in Table 2. Charpy V-notch and tensile samples from this material have been prepared to be irradiated to estimate changes in properties, with different neutron fluences.

Irradiation Conditions

Tensile and impact speciments are encapsulated in thin-walled aluminum tubes (Flgs. 2 & 3) and placed inside a specially fabricated basket (Fig. 4) for irradiation in Tarapur reactors. Copper, nickel and iron monitors incorporated inside the impact specimen capsules give data regarding the integrated flux values experienced by the test samples.

The end shield material specimens of RAPP reactors are encased in aluminum capsules (Fig. 5). For measurement of integrated flux 232 values each capsule is provided with a TH foil sandwiched between Gadolinium foils. Provision has been made for measuring the irra- dation temperature. The approximate fast flux that will be encountered by the samples can be given on 1012 n/cm sec. (Energy 1 Mev).

40 Test Procedures

The following tests are proposed to be carried out in the BARC Hot Laboratory Facilities.

Tensile tests will be carried out to assess the changes in the values of:

1. Tensile strength 2. Yield strength 3. Uniform elongation 4. Total elongation 5. Reduction of area

Impact tests will be carried out to obtain the following:

1. The entire plot of the impact curve 2. Change in 30 ft. pound transition temperature 3. Examination of fractured surfaces

Metallography will be carried out to assess the changes in the structure of material.

The various tests to be performed on the irradiated specimens will provide information regarding the changes in the various properties and 'also will provide information on the condition of the Pressure Vessels and-the end shield materials. Farther, there are also plans for carrying out other type of studies viz.9 fracture mechanics, etc. to understand the behaviour of these materials.

41 Table 1

MATERIAL FOR B.W.R.s AT TARAPUR

Composition (%) :

Type C Mn Si P S Mo

1 0.20 1.27 0.21 0.0:16 0.036' 0.47

2 0.20 1.32 0.18 0.2;) 0.023 0.46

Heat Treatment : a. Austenitize : 1725 - 1775 ('F b. Water spray to : 500°F o. Temper : 1200 - 1250'°F for 10 hrs. d. Air cool

Table 2

MATERIAL FOR R.A.P.P. REACTORS AT RAJASTHAN

Composition (%) :

Type C Mn Si P S Ni Co

1 0.13 0.46 0.21 0.013 0.017 3.5 0.038

Heat Treatment : a. Normalised 1500/1550°F, held for 1 hr./inch, air cooled b. SR 1150/12000F, held for 1 hr./inch, furnace cooled to 600 F

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46 WESTINGHOUSE NUCLEAR EUROPE REACTOR VESSEL SURVEILLANCE PROGRAM

T. R. Mager Westinghouse Nuclear Europe Brussels, Belgium

ABSTRACT

Currently, ten nuclear power plants are operating in Europe with reactor vessel radiation surveillance programs designed by Westinghouse. Of these ten plants, four are in Belgium, two in Switzerland and one each in Sweden, Italy, France and Spain. To date, postirradiation data are available from six of these plants. As a minimum, Westinghouse Nuclear Europe reactor vessel surveillance program is based on ASTM E-185, Recommended Practice for Surveillance Tests on Structure Materials in Nuclear Reactors. In addition to the basic requirements of ASTM E-185, Westing- house encapsulates fracture mechanics specimens to provide a quantitative assessment of the irradiation.

The purpose of the surveillance program is to monitor the effect of neutron radiation and other environmental factors on the vessel materials during operational conditions over the life of the plant. Westinghouse's basic philosophy as to reactor vessel material radiation can be summarized as follows : a) Sufficient data are provided to assess the margin for continued safe operation of the plant. b) Sufficient data are provided to set the heatup-cooldown limitation curves. c> Data are provided to perform a quantitative assessment of reactor vessel integrity. d) Sufficient capsules are provided to develop trend in the irradiation damage and provide sufficient samples for annealing if required.

47 The available data are evaluated in terms of current regulatory rules and guides, and copper-fluence trend curves.

In addition, data from accelerated irradiation test programs are reviewed in terms of post-irradiation annealing parameter. The results to date indicate tha't after a fluence of 3 x 10 2 n/cm the reactor vessel materials studied exhibited fracture toughness sufficiently high for continued safe operation of the nuclear power plants.

INPTDUCTION

The purpose of the surveillance program is to monitor the effect of neutron radiation and other environmental factors on the reactor vessel materials during operational conditions over the life of the plant. Westinghouse's basic philosophy as to reactor vessel material radiation surveillance programs can be summarized as follows a) Sufficient data are provided to assess the margin for continued safe operation of the plant. b) Sufficient data are provided to set the heatup-cooldown limitation curves. c) Data are provided to perform a quantitative assessment of reactor vessel integrity. d) Sufficient capsules are provided to develop trends in the irradiation damage and provide sufficient samples for annealing if required.

SCOPE

As a minimum, Westinghouse reactor vessel surveillance program is based on ASTM E185, Reccimended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors.

48 Currently, six capsules are inserted into each Westinghouse nuclear reactor between the core and the pressure vessel wall. Previously, six (2loop plants) or eight (3 and 4 loop plants) capsules were inserted in each reactor vessel. The capsule consists of welded tight fitting stainless steel enclosure halves to prevent corrosion and to ensure good thermal conductivity. The capsules are contained in specimen guide tubes attached to the thermal shield or thermal pads (depending on the plant vintage). Current plants use the thermal pad concept.

Each of the surveillance capsules contain Charpy-V-notch specimens, tensile specimens and IX-WOL or ET-CT specimens (current plants utilize %T-CT specimens) machined from materials representative of that from which the reactor vessel was fabricated. The representative materials include base metal from the core region shell courses, associated weld and HAZ (heat-affected-zone) material. Charpy-V-notch impact specimens fabricated from a'well documented heat of steel as to irradiation damage are also in each capsules as correlation monitor material. As an example, a typical Westinghouse Nuclear Europe capsule contains the following :

Material No.of Chrpys No.of Tensiles No.of ;T-CT

Lower Shell Course 18 3 6 Intermediate Shell Course 18 3 .6 Weld Metal 18 3 6 Heat-Affected-Zone 18 - - Correlation Monitor 8 - -

The various specimen types are shown in Figures 1 to 4.

To effect a correlation between fast neutron (E>lMeV)exposure and the radiation- induced property changes observed in the test specimens, a number of fast neutron flux monitors are included as an integral part of the Reactor Vessel Surveillance Program. -In particular, the surveillance capsules contain detectors employing the following reactions.

Fe54 Sn,p) Mn54

Ni58 (n,p) Co58

Cu63 (n,c) Co60

Np237 (n,f) Cs137

U238 (n,f) Cs137 49 The capsules contain two low melting point eutectic alloys to define more accurately the temperature attained by the test specimens during irradiation. The thermal monitors are sealed in Pyrex tubes and are of the following composition and melting point :

97.5% Pb, 2.5 % Ag (579°F melting point) 97.5% Pb, 1.75% Ag (590°F melting point)

As part of the basic Nuclear Steam Supply System package, Westinghouse includes pre-irradiation testing of the materials encapsulated in the test capsules and issues a report to the given utility documenting the surveillance program.

As a rule, the post-irradiation evaluation of the test capsules is not included in the basic package, However, Westinghouse provides this service upon request.

Capsule Removal

Specimen capsules are only removed from the reactor during normal refueling periods. The first capsule is normally removed at the end of the first core cycle. The second, third and fourth capsules are removed at approximately maximum exposure representative of i, ½ and i of service life. The two remaining capsules are for standby. The removal schedule meets the intent of 10 CFR Part 50 Appendix H.

RESULTS

To date, post irradiation data are available from the Chooz, Trino-Vercellese, Beznau No. I, Beznau No,2 and Jose Cabrera - Zorita plants. Of course, post- irradiation data are available from approximately ten PWR plants in the USA and trends are developing, Because I expect to hear detailed reports from the owner-utility of the above plants at this meeting I will only summarize the results to date. The results are sunmarized in Table 5.

Review of-the data summarized in Table 5, as well as data from US surveillance and accelerated irradiation programs indicate various trends in irradiation effects to reactor vessel materials. These trends can be summarized as follows: a) The majority of the decrease in the upper shelf impact energy occurs between approximately 1 x 1018 and 8 x 1018 n/cm2 .

50 b) Copper content has a significant effect on decrease in upper shelf impact energy. Knowing the copper content and pre-irradiation upper shelf impact energy, one can use the following factors to estimate post-irradiation upper shelf impact energy.

1. Pre-irradiation upper shelf impact energy greater than 120 ft-lb - decrease in shelf 2 ft-lb per 0.01 per cent copper.

2. Pre-irradiation upper shelf impact energy 60 to 120 ft-lb - decrease in shelf 1 ft-lb per 0.01 per cent copper.

3. Pre-irradiation upper shelf impact energy 30 to 60 ft-lb - decrease in shelf 0.5 ft-lb per 0.01 per cent copper. c) The pre-irradiation A Cv-shelf between longitudial (RW) and transverse (WR) oriented specimens will be maintained in the post-irradiation condition.

d) The post-irradiation increase in A NDT or A RIND T can be predicted from the trends curves given in Figure 5.

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52 Table 2 CONTENTS OF CURRENT WESTINGHOUSE NUCLEAR EUROPE SURVEILLANCE CAPSULES

No. OF No. OF No. OF MATERIAL CHARPYS TENSILES 1/2 T - CT's

LOWER SHELL COURSE * 18 6

INTERMEDIATE SHELL COURSE * 18 3 6

WELD METAL 18 3 6

HEAT-AFFECTED-ZONE 18

*SPECIMENS ORIENTED IN THE TRANSVERSE DIRECTION (WR)

Table 3 WESTINGHOUSE NUCLEAR EUROPE SURVEILLANCE CAPSULE - DOSIMETRY

Fe5 4 ( n,p ) Mn54

5 8 5 8 Ni ( n,p } Co

Cu6 3 ( n,ca ) Co6 0

137 Np 2 3 7 ( nf ) Cs

1 U2 3 8 ( n,f ) Cs 37

ALSO THERMAL FLUX MONITORS - BARE AND CADMIUM - SHIELDED CO-AL

53 Table 4 WESTINGHOUSE NUCLEAR EUROPE SURVEILLANCE CAPSULE THERMAL MONITORS

2.5% Ag, 97.5 % Pb MELTING POINT 579°F

1.75% Ag, 0.75% Sn, 97.5 % Pb MELTING POINT 590°F

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55 FPgure 1 CHARPY V-NOTCH IMPACT SPECIMEN

r^ -0.009

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0.395 0.393 + . . .

2.105

56 Figure 2 TENSILE SPECIMEN

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0.251 DIA

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NOTE : "B" DIA IS TO BE ACTUAL "A" DIA + 0.002 TO 0.005 TAPERING TO "A" AT THE CENTER

57 k'Flure 3 WEDGE OPENING LOADING SPECIMEN ( WOL )

06,00-20 THD CL 3 B 0.375 DEEP

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270 o R (2.5)_ I REACTOR VESSEL

THERMAL SHIELD N (1.4)

CAPSULE (TYP)

60 REACTOR VESSEL SURVEILLANCE

PRESENT PRACTICE AND FUTURE TRENDS IN SWITZERLAND

G. Prantl*, T. Varga**, D.H. Njo**

Engineering Division, Fed. Inst. for Reactor Research (EIR), WOrenlingen.

** Laboratory for Material Behaviour, R. and D. Dep., Gebr. Sulzer AG, Winterthur. ***Nucl. Safety Division (ASK), Fed. Office of Energy, Wurenlingen.

Abstract

- Surveillance program for the existing LWR-plants.

- Problems arising from the use of conventional impact test specimens (V-notch).

- Attempt, to replace these by improved specimens, modelling actual flaws more closely.

- Choice of specimens in order to get material data, enabling the application of elastic and eventually elastic-plastic fracture mechanics methods. (precracked ISO-V;three point bend specimens)

- Withdrawal schedule tailored to the individual conditions in a LWR-vessel. (accelerated irradiation of the specimens with respect to the wall, projected properties of material over the life of the reactor).

The purpose of the surveillance program is, to monitor the change of material properties with n-irradiation in the reactor environment. It has to provide the material data necessary fcr an evaluation of the safety margin against brittle fracture well to the end of the life. These data shall be suited, to estimate the significance of detected flaws in a quantitative manner.

61 In accord with the still used transition temperature approach to the prevention of brittle fracture, Charpy-V type impact specimens are inserted in the existing reactors (one BWR, two PWR's) for monitoring the influence of n-irradiation on the impact properties of the material in the belt line region. These specimens are supplemented by tensile specimens for comprehending the conventional mechanical properties and by a few lin-WOL specimens, to be tested in the temperature range, where LEFM is applicable. From the Charpy-V. test results only an indirect evaluation of the change of the fracture toughness can be made, using the RTNDT-shift in connection with the KIR-curve from the ASME code. This correlation is based on the assumption of validity of this concept for the material in question in the unirradiated as well as the irradiated state (fig. 1). The small WOL-specimens incorporated by some manufacturers (e.g. Westinghouse, Beznau) are not suited for a verification of the KIR curve. They have to be tested in a temperature range, which is not at all relevant to the operating conditions of the vessel. The calculation of permissible stresses (or int. pressure) in the presence of postulated or even detected flaws is therefore not possible, using actual materials properties. In certain cases the application of this philosophy might result in an overconservative definition of the minimum operating temperature and therefore in a severe operational drawback.

Apart from the discussed conceptual problems it is sometimes rather difficult, to uniquely define a shift in transition temperature from the Charpy-V curves due to the experimental scatter inherent in the transition range of impact testing (fig. 2). This scatter is often even enhanced by n-irradiation.

For those reasons, an attempt was made, to replace the common impact specimen by others, giving a better indication of the change of the fracture properties. A boundary condition for this is of course the space available in the irradiation capsules within the vessel.

62 The small impact specimen with a fatigue precrack instead of the V-notch offers a relatively cheap opportunity to measure a kind of dynamic toughness value at the same requirement for irradiation space. Due to the higher strain rate, compared with a static test, a valid KICd can be measured at relatively high temperatures. The necessary instrumentation of the impact testing machine is already developed. Moreover, according to our experience, the scatter in the energy to fracture versus temperature curve tends to be much smaller, and the transition range is much narrower, when precracked specimens are tested. The proposed dimensions of the starter notch and the precrack, which is of course applied before insertion into the vessel, are shown in fig. 3. Fig. 4 compares the transition curves of a two percent- chromium steel, measured with notched and precracked specimens, and reveals the typical differences. For reasons of comparison with the vast amount of available data on the irradiation behaviour of steels, Charpy-V(ISO-V) specimens should be maintained in the surveillance programs during a transition period of a few years from now.

The lin-WOL specimens, that have been mentioned above, are replaced by three point bend specimens of 25 by 25 mm cross section and approximately 110 mm length. Their main advantage is the possibility to measure an elastic-plastic fracture parameter, such as COD, either in a static or in an impact test, using the well-known and developed instrumentation. This may become very important for the judgement of the safety of the vessel belt line region, if inservice inspection indicates flaws. It should be mentioned, that bath the proposed specimens fully correspond to ASTM standard E399-74, as far as dimensions and precracking conditions are concerned.

The withdrawal schedule must be tailored to various parameters of the individual plant. This is illustrated in fig. 5. The relatively steep increase in ATNDT during the early years of the life, together with the higher n-flux at the irradiation position as compared to the one at the vessel wall, dictates the optimum times for withdrawal. These should be co-ordinated with

63 the anticipated refuelling shut down periods and the inservice inspection schedule. Fig. 5 demonstrates schematically the withdrawal schedules, using 4 irradiation capsules and an assumed flux ratio of 5. It must be kept in mind, that the high flux ratio (10-13) inherent in the design of some modern PWR's,may lead to inconveniently short withdrawal periods during the early plant life. In addition,it creates uncertainty with respect to the still not fully understood flux rate effect. For the latter reason in such reactors the insertion of a limited number of specimens at a lower flux position (acceleration factor 1 to 3) is required. This can only be done in an area of steep n-flux gradients. There, the exact location of the flux monitors relative to the anticipated fracture area of the specimens is extremely important.

Conclusions

- Charpy-V specimens (ISO-V) are supplemented by fatigue precracked specimens of the same overall size. These are impact tested and provide, apart from a better defined transition range, a limited quantitative knowledge of the change of fracture toughness with irradiation. In this case instrumentation of the pendulum is necessary, appropriate to record load versus time (or deflection) during the impact test. This permits application of LEFM-methods for the evaluation of the implications of flaws, detected by inservice inspection. - In order to enable the measurement cf some elastic-plastic fracture parameters, the WOL specimens, provided in the surveillance programs by certain manufacturers, are replaced by three point bend specimens of the same cross section. These can be tested either statically or dynamically. - Tensile specimens are kept for determination of the conventional mechanical properties, in particular yield points. - The number of specimens required, is determined by the different material-,used in the belt line region of the vessel. Generally, base material, weld material and HAZ have to be incorporated in the program.

- The withdrawal schedule must take into account the individual situation of the vessel in question.

64 unirrodiated

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ci irradiated I-

fixed y 0 15P11 energy level Raffi

ART NDT J Temperature or ATNDT

KIR (R fixed operating temperature

unirrodiated AKIR T

T - RTNDTNDT ART NOT

Fig. 1 Estimation of the change of KIR at a given temperature with n-irradiation, using impact energy transition curves (schematic).

65 CU E 20

I.-.1 _05 0

5 -

-150 -100 -50 ilDT0 50 100 Temperature t 'C 3 ---

Fig. 2 Demonstration of experimental scatter in impact testing. Steel: 1,2 MD07; KKB forged ring [after: Ullrich and Sandona 1974 1

66 45` 0,5 to

Ibly " 3)

Geometry of starter notch for fatigue precracked impact specimens.

.I* .0¢

Dimensions of three point bend specimens for COD - measurement.

Fig. 3 Geometry of proposed specimens. Dimensions in mm.

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69 PWR PRESSURE VESSEL SURVEILLANCE PROGRAMME IN BELGIUM

Ph. Van Asbroeck

C.E.N./S.C.K. Metallurgy Department

Abstract

It is scheduled that in 1983, nine nuclear reactors located in Belgium, will produce about 6500 MWe. These reactors are of the PWR type with Mn-Mo-Ni low alloy steel pressure vessel. Tension, Charpy and WOL IX samples are machined out the pressure vessel material, in the weld and in the heat affected zone. These samples together with temperature and flux monitorsare enclosed in capsules which are introduced in the reactor. These capsules are periodically withdrawn from the reactor and the samples are tested in order to determine fracture Mechanics characteristics (NDT ; KIC). 2 Results show that after irradiation upto 8.5 x 1020 n/cm (> 1 MeV) at 295 + 20°C, A 302 B NDT shift at 4.15 kgm'may attain 4300C at a 0.01 significance level and 250°C after 2.2 x 1020 n/cm2 most of this last damage can be relieved after 72 h at 400°C.

I. Introduction

Four PWR nuclear power plants are already working, they produce a maximum total net power of 1667 MWe ; it is scheduled that in 1983, five additional units will raise the power to about 6500 MWe (Table 1). This increase corresponds to a mean GNP growth rate of 0.04 and to an annual electricity consumption growth rate between 0.054 and 0.077, discount rate being 0.042 (1). All these reactors are Westinghouse licenced, primary circuit and pressure vessel data are given in table 2 (2 to 6). Chemical composition of the pressure vessel steels is indicated in table 3.

2. Pressure vessel surveillance ...... ~~~~~~~~~~~~~~

Charpy V notch, tension (ratio reduced section to diameter equal to 5) and WOL IX (25.4 x 25.4 x 36.5 mm) samples are machined out of pressure vessel material (in the base material, the weld and the heat affected zone) and out of reference material (SA 533 grade B class 1, supplied by the ASTME1O commitee)(4).

71 The base material samples are machined out of two coupons after heat treatment and before welding at 1/4 coupon thickness. Regarding the heat affected zone and weld samples, they will be taken after stress relieving at respectively 1/4 coupon thickness and in any positions excluding the weld root or any zone with puckering. The orientation of the different test samples is described in table 4. Furthermore some material in the form of a parallelipede is cut out the base material (25.4 x 25.4 x 91.5 and 25.4 x 25.4 x 55 mm), le,;gth being parallel to the circular fibration. These samples, together with temperature and flux monitors are enclosed in capsules. which are then introduced in the reactor. Two eutectic alloys are used as temperature monitors : Pb-2.5Ag and Pb-1.75Ag-0.75Sn with a melting point of respectively 304 and 310°C. The neutron fluences are measured at different energy levels with the dosimeters described in table 5. Capsules with composition as given in table 6 are introduced in different positions of the reactor where up to 1.8 time the fluence on the pressure vessel is reached. These capsules will be withdrawn from the reactor after about 1.5; 4.5 ; 9.5 and 18 years. Reserve capsules may also be taken, their withdrawn being dependent of the results of the other ones. Charpy testing is made according to (7) in order to determine NDT shifts. Tension testing is carried out according to (8)(9) at room temperature, 150, 300°C before irradiation and at reactor working temperature. Some tests will be done in order to determine the 0.2% yield strength which is used for the WOL test. WOL IX testing are done in order to determine the KIC value in function of the temperature.

3. Results

A 302 B pressure vessel steel specimens were irradiated up to a fast fluence of 8.5 x 1020n/cm 2 (> 1 MeV) at 295 + 20°C in the BR3 reactor (10). Tensile and impact tests, hardness measurements and micrographical examination were performed after irradiation. Results agree with the observed trend in NDT shift, tensile or hardness-increase (Fig.1). Furthermore annealing of some specimens irradiated up to 2.2 x 1020 n/cm2 shows that most of the damage can be relieved after 72 h at 400°C.

72 REFERENCES

(1) A.JAUMOTTE ; J.HOSTE Rapport final de la commission d'evaluation en matiere d'energie nucleaire Royaume de Belgique ; ministere des affaires economiques, mars 1976.

(2) M.POTEMANS De eerste belgische nukleaire centrale BR3. Technisch-Wetenschappelijk tijdschrift 32, (7) 1963

(3) La centrale nucleaire de Tihange S.E.M.O. Bruxelles, Belgique

(4) M.DUBOURG ; JR QUERO Tihange, cuve de reacteur Framatome 1971

(5) Kerncentrale Doel Trabel, Brussel, Belgie

(6) L.LAURENT Het reaktorvat en de inwendige strukturen Kerncentrale Doel 1972

(7) ISO 148

(8) NBN 117 03 ; (NFA 03 151)

(9) ISO/R 205

(10) P.DE MEESTER, Ph.VAN ASBROECK Neutron embrittlement and amage relief of the BR3/Vulcain pressure-vessel steel IAEA symposium on radiation damage in reactors materials, Wien June 1969.

73 Table I Belgian nuclear power plants programme (1)

Start up Localizat:on Electrical power per unit 1962 MOL (BR3) 11 1975 Doel I 393 1975 Tihange I 870* 1975 Doel II 393 1979 Doel III 930 1980 Tihange II 930 1981 Site A.I. 1000 1982 Site B.I. 1000 1983 Site A.II. 1000

0.50 of the power is exported to France

Table 2 Primary loop and pressu:e vessel data

Reactor Primary loop Pressure vessel

water temperature (°C) Material Max. pressure-- thicknes (bar) inlet outlet (mm)

Mol (BR3) 140 255 270 A302 B 115

Tihange I 155 284 323 SFAC 1,2 MD 07 200

Doel I, II 157 287 317 Soudotenax 56 180

74 ----- W-4

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76 Table 5

Neutron dosimetry

- copper - nickel - aluminium - 0.15 cobalt - aluminium - cobalt cladded with cadmium

- IrQn

- uranium 238 - neptunium 237

Table 6 Capsule composition

Test specimens Tihange Doel I, II

Charpy spec. 32 40 Tension spec. 3 9 WOL spec. 4 5

...... -.. ...~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

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C)lI- C. Co 0 C.

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4 §s-- t ,-, cl \ i

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i , i A

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I I

78 "A UTILITY REVIEW OF IRRADIATION SURVEILLANCE PROGRAMS AND INDUSTRY RESPONSIBILITIES"

Thomas D. Keenan PE Manager Plant Engineering Department Yankee Atomic Electric Company Nuclear Services Division Westboro, Massachusetts USA

ABSTRACT

Yankee Atomic Electric Company owns and operates one of the world's first nuclear power plants. In addition, its Nuclear Services Division has behind it many years of experience in supervising construction, operation, and maintenance of four nuclear power plants, three pressurized light water and one boiling water reactor. As a member of this company, the author has had extensive experience in dealing with the major nuclear steam system suppliers, architect engineering firms, and regulatory agencies. Based on this extensive and varied experience, he has taken a critical view of the manner in which the nuclear industry deals with the irradiation surveillance programs and its problems, as seen from the nuclear utility side.

In his paper, the author reviews the safety implications inherent in the subject and discusses many of the present and future technical issues, as well as the economic considerations and future implications which affect utility decisions in this area.

Finally, the author provides critical commentary on the attitudes of nuclear utilities and the nuclear plant vendors,'which must change, in his view, in order to enable the industry to achieve its end:well-designod, cost effective nuclear plants which maximize on-line time. He closes with a brief discussion of .hat is perceived to be a key fault - no consolidated approach to a very complex problem.

I. Introduction

The advent of commercial nuclear power plants has required the

development of new technologies, and yet for all our apparent sophistica-

tion and; knowledge, it seems that we sometimes forget to apply some of

the basic principles of "good engineering practice." Using previously

developed data instead of developing it over again and reaching the

same conclusion is one example of this philosophy. Failure to apply

these types of basic principles does not make a nuclear plant "more"

or "less" safe, it usually results in higher costs. This is one area

79 where significant gains can and must be made if we are to keep nuclear

power a competitive industry.

There are many ways to accomplish "good engineering practice".

Mutual exchange of information is one of the cornerstones; another,

is to defend what is correct and pursue the point even with regulatory

agencies, because no one is infallible and they too can be in error.

It is the intent of this paper to discuss Reactor Surveillance

Programs and the views of Yankee Atomic Electric Company Nuclear Services

Division, as a utility industry representative, on that subject.

The Yankee organization has behind it almost 20 years of experience in

supervising construction, operation, and maintenance of four nuclear

power plants in New England, three pressurized water reactors and one

boiling water reactor. We presently are undergoing licensing and

construction of four additional pressurized water reactors. Our

involvement with these projects has brought us into technical and

administrative contact with three nuclear steam suppliers and three

architect-engineering firms, as well as the governing regulatory

agencies. This contract has enabled us to gain valuable experience and

insight regarding many technical issues facing the nuclear industry.

Ensuring the fracture safe perfor;m-nce of light water reactor

pressure vessels (RPV) is a complex issue involving numerous activities;

the definition of mechanical properties and tests of fracture behavior

in vessel steels, methods development for application of data to RPV

fracture analysis during service lifetime; quantification of irradiation

effects and other in-service related phenomena, and finally monitoring of

vessel in-service conditions to assess changes, through the use of

properly selected parametric measurement tools (Charpy specimens, etc.).

The historical applied engineering approach to assessing irradiation

embrittlement has been based on observation of changes in the impact

properties of prototype samples and actual RPV steels during irradiation

in test facilities or power reactors. Accelerated irradiation is done

80 to provide end of lifetime information early in a vessel life. Trend

bands for impact properties, as a function of irradiation are established

for various material characteristics such as chemical composition, micro-

structure, thickness, etc.

The unit of irradiation exposure generally used to permit data

comparison is time integrated flux or fluence (n/cm ) of neutrons above

1 Mev.

II. Safety Implications of Surveillance Programs

Sufficient information has been generated regarding radiation

embrittlement to firmly establish the need for understanding the technical

aspects of the subject and properly addressing resolution of those issues

over RPV lifetime. It should be obvious that the maintenance of RPV

integrity requires us all to fully address the problem and take the

steps necessary to minimize its effects over a vessel lifetime. The

fact that radiation embrittlement can become a limiting condition for

continued operation must be recognized, accepted, and dealt with

effectively. Unfortunately, it has been a frequent observation that this

highly relevant subject is considered "last" or of no immediate concern

because, in reality, its effects will not really be felt until the

latter stages of RPV lifetime. Although its impact is a late-in-life

issue, it must be considered in the beginning-of-life to preclude

loss of options later in life, and with that reduced operating lifetime.

Any potential problem which could ultimately lead to reactor shutdown

for either annealing in-situ or decommissioning must be considered an

extremely important, safety-related subject worthy of significant

attention by the industry.

III. Technical Issues

In light of the above discussion relative to the safety implications

of surveillance programs, it becomes increasingly difficult to understand

the continuous proliferation of the "proprietary umbrella" thrown up

by nuclear steam supply system vendors and other industry members

81 around this subject and other current problems the nuclear industry faces.

We all admit that our industry is still in its infancy and we have a long way to go before we are really established. Witness the present controversy in the United States regarding reactor safety and the consideration being given moratorium and "slow-down-development" bills in legislatures.

Inhibiting the exchange.of detailed information has the cumulative effect of slowing down development or problem solutions, reducing efficiency, wasting money because of duplication of effort, and providing anti-nuclear forces with fuel for their suppression-of- information position.

Is this the image a developing industry wants to have? Can we afford the proprietary luxury at this stage? The answer to both questions is "no" and we therefore must fight the "proprietary" issue whenever and wherever we can.

Another disconcerting aspect of this issue was mentioned briefly before. That is, our technical common sense is sometimes overcome by our lack of perspective and we relegate'radiation embrittlement problems to last place because they are 30 years away from becoming a reality. This loss-of-perspective was made vividly clear several years ago when our company was negotiating for a new light water reactor project with several of the major U.S. nuclear steam supply system vendors. The philosophy expressed by these technological experts on the subject of reactor vessel surveillance programs was that all of the major issues had been settled and standardization had been achieved.

The issue has arisen over a discussion concerning the number of surveillance capsules to be included with the reactor vessel. At that time, there was some preliminary work going on in regard to copper-phosphorus embrittlement characteristics. In addition, the Naval Research Laboratory had done a study of neutron embrittlement predictive-codes and the lack 82 of agreement was obvious. Fracture mechanics was still on the horizon.

How anyone technically knowledgeable in the field could have concluded that we knew all the answers is beyond comprehension. On a relative basis, excess surveillance capsules are cheap insurance and a good hedge on new developments; reactor vessel archive material falls into that category also. These aspects will be discussed in more depth later. The point is that the issue of vessel irradiation surveillance programs have historically not received recognition as extremely important, current concerns which cannot be put-off to the future. We cannot afford to make premature judgements in this area.

There are now and have been for some time, a number of unresolved problem areas or concerns which require complete resolution before standardization can be achieved and surveillance programs relegated to secondary importance. The following discussion summarizes the issue and the basis for concern.

1. Inconsistency among codes in predicted neutron exposure

The Yankee Nuclear Power Station (Rowe, Massachusetts),

one of the first commercial reactors in the world, has been

in operation for over 15 years. Its irradiation surveillance

program has contributed etensive technical data upon which

much of today's knowledge of irradiation damage to pressure vessel

steels is based.

Briefly, two elongated specimen capsules were located

between the thermal shield and pressure vessel wall, and eight

other similar capsules were located inside the thermal shield

adjacent to the shroud surrounding the core. These eight

capsules were in "accelerated" locations which meant they

received a neutron dose rate higher than that of the vessel.

This enabled us to determine the vessel end-of-lifetime

transition temperature shift relatively early in the vessel

83 lifetime; knowing vessel operating history and neutron spectrum we are able to adjust vessel operating curves accordingly for continued safe operation.

Each surveillance capsule contained Charpy V-notch and tensile specimens from an A 302-B steel fabrication test plate for the upper vessel shell course. The program results are typical of radiation damage effects, including nil-ductility transition temperature and decreased upper shelf energy. An annealing test program was carried out successfully on some of the

surveillance specimens and those results are also discussed

in a report issued by the Naval Research Laboratory.

There is one area pointed out by the results of this

study, which has direct generic application to today's plants

and those of the future. The maximum predicted fast neutron

(E > 1 Mev) exposure for the Yankee reactor vessel after 30

years of operation is 2.5 x 1019 nvt (fluence) according to

studies.conducted by the plant supplier, using diffusion theory.

The Naval Research Laboratory report discussed above

predicts a maximum service fluence of 1'.46 x 1019 nvt (E > 0.5

Mev) for the same time period using transport theory.

Figure 1 contains a steady-state pressure temperature curve

for the Yankee plant. On it are plotted the two differing

values for end of life neutron fluence. As can be seen from

these curves, there is a substantial difference between the two

values. In fact, the difference approaches 11 years of additional

operating capability for the vessel, if the lower value

(1.46 x 1019 nvt) is correct. This discrepancy is not limiting

at this early stage in the vessel's life, however, in the later

years of operation, this difference will become significant

and will have to be resolved if an extended operating life for

the plant is considered desirable. This last subject has not

84 been widely discussed in the nuclear industry. However, as part

of the licensing issue, relatively detailed plans are provided

for decommissioning and dismantling of each new plant as it comes

along. This is certainly an admirable concern, nevertheless,

it is equally worthwhile to consider how plant lifetime could

be extended. Environmentalists should like this idea since

it would tend to reduce the need for new replacement plants

30 - 40 years from now.

bosed on based on represents 1.46 xlO'9 nv 2.5x 1O'9nvt approx. II yrs. E >.5 Mev E > I Mev full power operation

o '>a .'-unirra diaoed

c) 4

- 13 01 mo opera

0 00

o - I 0 0

100 20 300 400 500 600 4--/, ,- ~

.ndicted Reactr Coon Temperatureu F

for reactor coolant ru pump operation

0 100 200 300 400 500 600 Indicated Reactor Coolant Temperature ~F )

Fig. 1: Yankee Rowe Indicated Pressure Temperature Relationship for Steady - State Operation

85 The difference in predicted neutron exposure discussed

above is primarily a result of the codes used. There are many

codes now in use and most of them provide somewhat differing

results. These differences must ultimately be resolved by

more extensive test programs and/or more sophisticated codes.

This problem will not confront the new units for some

time, particularly with the new low copper, phosphorus core

belt line region plate material, which substantially reduces

radiation sensitivity and thus embrittlement. Nevertheless,

the problem should not be forgotten and owners of nuclear power

stations should all be aware of the situation for possible

future action.

Notice in the above discussion that two different neutron energies are utilized as a basis for calculation, 1 Mev and 0.5 Mev. Is this a cause for concern? Possibly not, but there should be consistency in the calculational basis used by various experts (either 0.S or 1.0 Mev).

If the difference is causefor concern, then that concern should be quantified and addressed by us all now rather than relegating it to

future consideration and then see it become an operating or end of

lifetime penalty.

2. Projecting the surveillance capsule determined fluence to the vessel wall appears to be a calculational procedure fraught with uncertainty

There has been recent work in the area done to correct past

mistakes, such as failing to average the projected value over

the entire vessel wall from the effective line source, and

instead doing a projection of the capsule fluence only to the

wall directly behind it.

'3. Application of Fracture Mechanics Principles

This has been very valuable in helping us all achieve

a better understanding of materials behavior and yet we have

86 so far been unable to make the complete transition from the

old Charpy V-notch era to the new world of fracture mechanics.

This difficulty is of serious concern because almost all

of today's operating reactors utilize the Charpy V-notch

approach and we are now struggling to equate that data to

fracture mechanics language. This approach is complex and,

at the moment, filled with uncertainty:

a. We cannot fit fracture mechanics specimens into surveillance

capsules. Is that really true or is it merely a design

problem that needs attention!

b. Is the size effect (plane strain vs plane stress) adequately

accounted for? At one point our Regulatory Agency was

advocating a 7 F per inch embrittlement penalty on thick

sections. We view that now with disdain but it took such

unreasonable conservatism on the part of Regulatory to move

the industry to address the problem.

Our present approach is to develop a KIR curve based on

fracture mechanics concepts and index it to transition

temperature tests using drop weight-NDT and Charpy V-notch

specimens. We then measure embrittlement by monitoring

Charpy-shift rather than a fracture mechanics parameter. Is

that completely valid or are the utilities going to pay the

price 30 years from now as an operating restriction?

4. .Doesmicrostructure have an influence on radiation sensitivity?

It is not accounted for in the selection process for vessel

surveillance material discussed in our standards. If it could

be a problem, let us work on an answer now.

5. Regulatorv limits for acceptable radiation-induced shifts

A co>iAt regarding positions of our Regulatory Agency

is the apparent arbitrary selection of the 50 ft-lb or 35 mil

level for measuring the radiation-induced shift. The original

87 reason for introducing the 50/35 criteria into the code was to

ensure that a significant increase in toughness occurred

within 60°F of the RTNDT temperature. Since radiation increases

the yield strength and decreases ductility, there is a question

as to whether the 50/35 criteria represents the same index

point for irradiated material. Also, historical embrittlement

data is based on the radiation-induced shift at the 30 ft-lb

level, which is supposed to correspond to the NDT temperature.

The present regulations require that the Charpy upper shelf

energy must be greater than 75 ft-lb before irradiation and that

if radiation decreases the shelf value below 50 ft-lb, a fracture

mechanics analysis must be used to justify continued operation

of the reactor. The 50 ft-lb energy requirement is to avoid

a low energy tear fracture. It is not clear that a fracture

mechanics analysis based on initiation toughness can handle

this problem. There is evidence that the primary effect of

radiation in the upper shelf region is to decrease the resistance

to crack propagation. Thus, crack arrest considerations may

be more pertinent.

This whole issue of mixing fracture mechanics and the NDT

Charpy V-notch approach is of concern in that it may prove to be

a future problem. Positive action should be taken now to fully

assess this issue.

6. Accelerated capsules are an integral part of many surveillance programs

Is there a substantive difference in resultant embrittlement

as a function of dose rate? Some data indicates that total

embrittlement increases with dose rate for certain steels.

If this can be quantified as a real conservatism, it could be very helpful for obvious reasons.

88 For the above reasons, when a vendor states that his vessel irradiation

surveillance program is right on top of all the issues and we shouldn't

worry about it - we do worry because obviously he should be and is not.

The conclusion to be drawn is that he doesn't understand the subject

very well!

IV. Economic Considerations and Future Implications

It was said before and it bears reiteration. Vessel surveillance

programs are cheap insurance if they have built into them the flexibility

to be modified in the light of new knowledge. What is flexibility?

It means extra installed surveillance capsules; enough to conduct

a program for 40 years, two capsules to permit assessment of the annealing

and finally several more to check re-embrittlement of the steel

and verify that the rate of re-embrittlement is unchanged.

Flexibility also means adequate vessel archive material from various

locations in the vessel.

The real economic considerations are not how much it costs to test

a surveillance capsule, but rather:

a. How can we utilize or modify a surveillance program to minimize its

influence on vessel operating parameters (heatup, cooldown, rates,

pressurization curve).

b. At the end of its present lifetime will the surveillance program

permit annealing and resumption of extended operation?

It is difficult to believe that a large present day commercial

plant will be arbitrarily written off at the end of its design lifetime. The fatigue analyses are all extremely conservative and everyone

is aware that many of the design lifetime limitations are arbitrary.

Reanalysis can eliminate those limitations; a well though out, flexible

(previously described) vessel irradiation surveillance program can

support extended operation beyond present design lifetime. A poorly

designed, save-a-dollar-today surveillance program cannot do that.

89 The issue of potentially extended vessel lifetime has merit and

deserves serious attention by the nuclear industry.

V. Industry Responsibilities

A. Nuclear Utility Companies

We have an obligation to ourselves to develop expertise

within our own organizations to deal with the many complex issues

of our growing technology. No organization can or will function

with the utility viewpoint in mind except a utility,

Each organization works for its own benefit and it is mandatory

that a utility oversee that work to ensure efficiency and effective

cost-control. In addition, it must be realized that the utility

must live with decisions made on its project for the full plant

lifetime. The utility bears the cost of maintenance problems due

to poor design and therefore should be active in the decision making

processes during design.

Reactor vessel surveillance programs should be reviewed in

more depth than is presently the case. As new rules are established

for evaluating surveillance data, the originally created surveillance

program may have to be changed to accommodate these new rules. A

flexible program can do that - and therein lies an indictment of

reactor vendors for insensitivity to the ever changing licensing

process and for premature judgments concerning what is adequate in

this area based on a "We have all the answers" attitude. It is

the responsibility of the utility arm of this industry to see

through the facade and force the necessary reevaluations. This

requires technical knowledge, not merely project management. Reactor

vessel irradiation surveillance programs are one area where utility

expertise is needed.

Utilities must adopt a more aggressive participatory role

now! Along with developing this new attitude, we must also learn

to shed the old attitude of many separate utilities disinterested

90 in what the other company does. The time has come to recognize that

the separate companies are all interrelated such that the decisions

and actions of one utility do have an impact on the other utilities.

What has been lacking, in many instances, is the recognition of

that fact and then altering one's actions as a result of that

knowledge.

Along with increased awareness of interrelationships comes

responsibility to act for the good of the industry as a whole. An

example might be offering a surveillance capsule location in a large

power reactor to the research arm of our industry for their use.

This could be done after a presently installed capsule is removed.

The potential information gain would obviously benefit the entire

industry.

A nuclear utility in New England has made such an offer -

unfortunately, the response has been disappointing to date. It

is hoped that the offer will ultimately be accepted.

As mentioned previously, surveillance programs are cheap

insurance, and what that really means is additional money spent in the beginning of a project to ensure adequate surveillance

capsules, can save far more money in the future. That philosophy

is somewhat akin to not necessarily buying a component from the

lowest bidder because the money saved there may be spent several

times over in maintenance costs. Experience and is}:u nt must be

exercised.

B. Reactor Vessel Vendors

There are several areas wherein vendor attitudes must change

if we are to bring this industry to realize its full potential.

1. It is time for recognition of the fact that a Nuclear

Utility can have expertise of sufficient depth to critically

review the positions of the major vendors on many major

issues. Refusal to acknowledge this FACT can only cause

91 needless friction and impede our mutually agreed objective -

a well designed, cost-effective nuclear station which maximizes

plant availability.

2. Responsiveness and sensitivity to changing regulatory positions

is required in concert with responsible utility support, not

independent of them. Becoming less authoritarian and more

consultant oriented is desirable. It would be well to remember

that without the nuclear utility and its plants, there would

be no commercial industry. Technology development is not an

end in itself but is necessary and desirable for what its

application can bring us.

VI. Joint Industry Goals

Ultimately we must all work together to solve our technical problems.

The major difficulty in the area of reactor vessel surveillance programs

appears to be the lack of some unifying organization to focus attention

on the key problems and provide the needed enforcement strength. Efforts

to educate the reactor vendors and operators have been primarily borne

by the Naval Research Laboratory. It is doubtful whether a majority

of reactor operators fully understand the implication of surveillance

programs. There is no on-going, organized effort to collect, analyze

and disperse surveillance data. It is becoming apparent that some

surveillance related problems will only be solved by some forced

coordination among divergent parties.

-There are obviously several solutions, including creation of an

independent organization or utilization of an existing one to be

that co6rdinating and unifying force.

One potential major benefit from coordination and unification of

data and programs could be the beginning of standardization. By

that is meant a reduction of the need for each reactor vessel to have

a totally independent surveillance program including capsules. A

92 check capsule to verify the identical nature of the vessels, fluence,

and shift would most probably always be necessary, as well as several

for contingency (annealing, etc.). Archive material could then

provide all the necessary program flexibility. We are still a long

way from standardization but it is ultimately a desirable end.

VII. Conclusion

Our industry has come a long way in a very short time. We have

the talent and the dedication within each of our organizations to

solve all of our problems and provide a mature technology which can

serve mankind. I hope the comments and oft-times blunt criticisms

expressed herein are taken in the spirit with which they are given.

That is, to help us all, in some small way, reach the objective

stated above.

93 CONTRIBUTION TO THE QUESTION OF SURVEILLANCE PROGRAMS FOR NUCLEAR REACTOR PRESSURE VESSELS

Milan Brumovsky, KODA Works, Nuclear Power Plants Division, Research and Development Centre, PlzeA, C.S.S.R.

ABSTRACT

Results of the evaluation of surveillance specimens and in-service inspection of tubes of A-1 reactor heavy water Calandria are presented. Results obtained are in good agreement and prove the objectivity of ultrasonic method used for wall thickness measurements. High radiation stability of Al-Mg-Si alloy was proved.

Two following approaches to the safety of nuclear reactor pressure vessels are being used today:

- a temperature approach characterized mainly by brittle Crack Arrest Temperature, CAT / sometimes also by Nil Ductility Temperature, NDT /,

- an energy approach characterized by fracture toughness, KIC.

Other approaches, i.e. a deformation one / critical Crack Opening Displacement, COD/ and an energy one based on the critical value of J-integral are not yet sufficiently worked out.

As it was shown in our previous paper, dealing with surveillance program for the A-1 pressure vessel, the following one parameter is common for both approaches: AT - transition temperature shift, caused by irradiation, ageing and other degradating influences. Postulating / and confirmed / that no size effect appears in these degradations of material, it is possible to choice the most suitable, simple and cheap method and type of specimens for its measurement. In most cases the test of dynamic notch toughness is used, mostly in Charpy-V-notch type specimens.

With resnect to the existence of active core placed in the cylindrical part of pressure vessel, the last one can be divided into two Darts:

95 - smooth cylindrical part, characterized by high neutron flux, practical with no stress concentrations and that's why with minimal crack growth during operation / and also with minimum welding joints/,

other part of vessel, characterized by absence of neutron flux, but with high local stress concentrations serving as a source of low-cycle fattgue of material and subcritical crack growth. Welding joints are widely used and for some type of steels their temperature ageing / connected with low-cycle damage fatigue / can appear.

To secure high reliability of pressure vessel service and to approve its life-time, it is necessary to know very precisely and certainly two parameters of fracture mechanics:

- transition temperature shift, AT, and crack growth, aa.

While the crack growth measurement is the object of in-service and periodical defettoscopic control, transition temperature shift is finding out with the use of surveillance specimens program.

Taking into account these facts, knowledge about radiation damage and other degradation processes of materials during their service, it is possible to compound the main principles for surveillance specimens programs :

I. Materials /a/ Base material - all heats used in the pressure vessel near active core, at least receiving neutron doses higher than 2? -2 lx 102 n-m , E > 1 NeV /b/ ..Welding metal - from welding joints of the same part of pressure vessel, if received the same dose as in /a/, /c/ Heat affected zone - similarly to /a/ + /b/, /d/ Reference base material - well documented and used also for other irradiation tests

2. Specimens /a/ Charpy-V-notch specimens for dynamic notch toughness tests, at least 12 pieces per one place of welding joint, /b/ Static tensile specimens, at least 3 pieces from base and welding materials,

96 /c/ Fracture mechanics specimens - for measurements of fracture toughness, crack opening displacement etc. - optional.

3. Capsules /a/ Made from stainless steel, projected with maximum heat transfer, i.e. to secure the temperature of specimens very similar to pressure vessels, /b/ Each capsule must contain the whole set of specimens from one material at least, better from the whole welding joint.

4. Location of capsules /a/ In the beltline of pressure vessel in the place of maximum received neutron dose of pressure vessel / received neutron dose by capsule must be very close to the dose of pressure vessel / but less than twice higher than the dose of pressure vessel, /b/ Accelerated irradiation / in the reactor reflector, where neutron dose is in the range of five to ten times higher than in inside wall of pressure vessel /, /c/ Out of reactor core for the measurement of temperature ageing / for analysis of non-irradiated part of vessel /.

5. Neutron dose detectors /a/ Threshold detectors in the quantity and sortiment that permit to determine not only exact neutron doses but also the difference between neutron spectra at inside pressure vessel wall, position of normal irradiation and accelerated irradiation to be possible to make corrections on neutron spectra, /b/ Measurement of neutron flux and spectrum is necessary to carry out along the height of pressure vessel / and repeat it in all cases when large changes in operational regimes are made /.

6, Temperature detectors /a/ Set of low-temperature melting detectors, resp. also diamond, /b/ Instrumentation / if possible / in some capsules - optional, /c/ Precise recording of service temperature and regimes.

97 7. Time of withdrawal and number of capsules /a/ On inside pressure vessel wall /1/ At least three sets of base material, weld metal and heat affected zone, /2/ First withdrawal realize at maximum after 30C, of planned life-time / if it is common with the first dose measurement then no longer after two years of operation - it depends on life-time of detectors/ /3/ Last withdrawal after approximately 803 of planned life-time, /b/ Accelerated irradiation /1/ At least two or three sets of the same materials, /2/ Withdrawals must be planned in such a way that the first one will receive approximately the planned neutron dose for pressure vessel, /c/ Temperature ageing /1/ One or two sets of the same materials, /2/ First withdrawal realize after five years of service,

It is clear that these recommendations are necessary to arrange and modify with respect to the constructional and evaliation possibilities and to the present stage and knowledge of fracture mechanics and philosophy of pressure vessel reliability. In all cases it is necessary to take into account also two following instructions :

- surveillance program must be economic, i.e. only necessary number of specimens and capsules must be used, as normally it must not serve as a research program, - surveillance nrogram must be so wide and full to secure the evaluation of service life-time with high accuracy and also pressure vessel reliability even in cases with possible un- planned changes in active core and operational regimes during reactor service.

Surveillance specimens programs must he planned since the first stages of reactor proiect and must be one of the inevitable parts of recurring .inspections of reactor pressure vessels as results received from this program are very important for the analysis of pressure vessel reliability.

98 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM

A.N.C.C. - I T A L Y

ABSTRACT

A draft of the Reactor vessel surveillance programme in Italy is presented. It covers the scope of the programme, its criteria, material selection procedures, test specimens as well as their types, orientation and location, withdrawal schedule, flux measurements, description of test results.

Scope

The purpose of the material surveillance program required by this document is to monitor the changes in the fracture toughness properties of the reactor vessel steels in the belt-line region of light water-cooled nuclear power reactors.

The data of fracture toughness are obtained from specimens located in special holders inside the pressure vessel in selected areas; these specimens are withdrawn from the vessel at specified times.

The requirements of this document provide adequate margins of safety during any condition of normal operation to which the pressure vessel may be subjected over its service lifetime and apply to the following materials:

1) Carbon and low-alloy ferritic steel plate and forgings of ASTM specifications; 2) Welds and weld heat-affected zones in the material specified in 1).

Surveillance program criteria a) No material surveillance program is required for reactor vessels for which the manufacturer and the owner declare that the peak neutron fluence (E > 1 MeV) at the end of the design life of the vessel will not exceed 1017 n/cm2 . b) Reactor vessels which do not meet the conditions of a) shall have their belt-line regions monitored by a surveillance program prepared by the manufacturer according to the following paragraphs.

99 To determine the minimum number of test specimens, this document defines two cases.

First case: Where both the predicted increase in transition temperature of the reactor vessel is 38C or less and the calculated peak neutron fluence (E > 1 MeV') of the reactor vessel is 5.1018 n/cm2 or less.

Second case: Where both the predicted increase in transition temperature of the reactor vessel steel is greater than '8C or where the calculated peak neutron fluence (E > 1 MeV) of the reactor vessel is greater than 5.1018 n/cm.

To determine the withdrawal schedules of surveillance capsules this document establishes the following Groups:

Group A) Reactor vessels for which it can be conservatively demonstrated by experimental data and tests on comparable vessel steels, making appropriate allowances for all uncertainless in the measurements, that the adjusted reference temperature will non exceed 38C at the end of the service lifetime of the reactor vessel.

Group B) Reactor vessels which do not meet the conditions of Group A but for which it can be conservatively demonstrated by experimental data and tests performed on comparable vessel steels that the sdiJuted reference temperature will not exceed 93C at the end of the service lifetime of the reactor vessel.

Group C) Reactor vessels which do not meet the conditions of Group A nor Group B.

The apourtenance Group of the reactor vessel shall be declared by the power plant owner.

Surveillance material selection procedures

The selection of the base metal and weld metal from the reactor irradiated region that is best suited for surveillance monitoring will be determined for 1 case from a consideration of initial RTNDT temperatures and for ? case from 'oint considerations of initial RTNDT temneratures and residual element contents (Cu,P). In both cases the weld heat-affected zone to be monitored will correspond to the base metal choice. Materials that exhibit a C unper shelf energy level of 10,5 kgf.m or less i.n the orientation of interest shall be treated as a special situation.

100 ase metal exhibiting differences in initial RTT temperature of ?6C or less shall be considered equivalent: weld metals exhibiting difference

in initial HTNDT temperature of 16C or less shall be considered equivalent. 1S], s'. i mnetals (or weld metals) having differences in copper content of 0,03 weight % or less and differences in phosphorus content of 0,003 weight % or less shall be considered equivalent.

First case. The base metal to be selected shall have the highest pre-service

RTNDT temperature of those base metals located in the irradiated region. When base metals are equivalent with respect to initial RTMDT temperature, the base metal with the highest copper content shall be selected. The same selection procedures will be applied separately for weld metal selection.

Second case: a) Equivalent initial RTNDT temperatures When the base metals have equivalent copper and phosphorus contents, select the base metal with the lowest C upper shelf energy level in the orientation of interest. When copper contents are equivalent, but phosphorus contents vary by more than 0,003 weight %, select the base metal with the highest phosphorus content. When phosphorus contents are equivalent, but copper contents vary by more than 0,003 weight %, select the base metal with the highest copper content. When neither the copper contents nor the phosphorus contents are equivalent, select the base metal with the highest copper content. If applicable, the same selection procedures will be applied separately for weld metal selection. b) Non-equivalent initial RTNDT temperatures with higher initial NDT temperatures exhibited by the higher copper, phosphorus content materials. Apply selection plan a) to the higher copper, phosphorus content base metals (or weld metals) as a group. c) Non-equivalent initial RTNDT temperatures with lower initial RTNDT temperatures exhibited,.by the higher copper, phosphorus content materials. If initial RTND NDTT temperatures of the higher copper, phosphorus content base metals are not more than 16C below the initial RTNDT temperatures of the lower copper, phosphorus content base metals as a group, apply selection plant a) to the higher copper, phosphorus content base metals. If the initial RTNDT temperatures of the higher copper, phosphorus content base metals are more than 16C below the initial RTNDT temperatures of the lower copper, phosphorus content base metals as a group, select the base metal whose properties (RTNDT C upper shelf energy) during service would first appear to limit the vessel operating lifetime. If applicable, the same selection procedures will be applied separately for weld metal selection.

101 Special situation Charpy V upper shelf energy levels are 10,5 kgf.m or less. The radiation-induced reduction in Cv upper shelf energy level can, in some cases, become a limit factor in advance of the radiation-induced elevation of RTNDT temperature. This occurrence depends on certain combinations of initial properties (Cv upper shelf energy level, RTNDT temperature, yield strength) along with reactor primary system operating characteristics and limitations. Accordingly the presence of base metal in the irradiated region which has preservice shelf energy levels of 10,5 kgf.m or less in the orientation of interest shall be considered a special case. In such a case, the base metal best suited for surveillance monitoring shall be established through a comprehensive evaluation of preservice properties of all base metals in the irradiated region to identify that material whose properties (RT or shelf energy level) during service would first appear to limit the vessel operating lifetime. If applicable, the same evaluation would be applied for weld metal selection.

Material selection flow diagram

Figure 1 should be used as a gdide for the selection of base metal from the reactor irradiated region for surveillance monitoring.

: ore Rc .,ion

1 Li -4 -^--_ t<^- 'J~~H~~1 __L__, Eu-i;.l cn -ci U alniullcn uii l Il,:fhcr RT',ith Lo.r RT ,F; wit She ,Cum.u ( Cu&I'Z » < 1' Ht1hrCCku&Ph &Cu% uP1`r

rrquLrimicntI |Ionr qui iiln; I ;ll n [N>eN incqu;nt( PJ 1 P% C I _u.

Ul:llt c al U,s t Mc1 ¢l';ti.'inN.^illl |Ul t t" .s , *RTt.... withI1 ol.cst . i l nhlohIlt. t l itlh lfthcslII | RT r i;hc l 1>. d C.Sll cL j C. Shlclf Cu%_I oI Loc.t C. Shed

ilh llithes Cu% , Nolt.-Sclrct natctial whols propcrlics (IRTIo. C, uplpe shclf encr.)) during scrxicc sould fitf uppFar to liiil Ihe ',\:1uvpcratlng lifetime. FIG. 4 Sunillanrn Mhanilt Srlctlion Proceduret

Tests specimens

Test specimens shall be prepared from the actual materials used in fabricating the irradiated region of the reactor vessel. Samples shall represent a minimum of one heat of the base metal and one butt weld and one weld heat-affected zone if a weld occurs in the irradiated region.

102 Materials used to prepare test specimens shall he taken directly from excess material and welds in the vessel shell course following completion of the production longitudinal weld joint and subjected to a heat treatment that produces metallurgical effects equivalent to those produced in the vessel material throughout its fabrication process.

Where seamless shell forgings are used, or where the same welding process is used for longitudinal and circumferential welds in plates, the test specimens may be taken from a separate weldment provided. that such a weldment is prepared using excess material from the shell forging or plates, as applicable, the same heat or filler material, and the same production welding conditions as those used in joining the corresponding shell materials. A minimum test program shall consist of specimens taken from the following locations:

1) base metal of one heat used in the irradiated region, 2) weld metal, fully representative of the fabrication procedure used for a weld in the irradiated region and. the same type of flux, and filler metal, ') the heat-affected zone associated with the b:se metal noted above.

Representative test coupon to provide two additional sets of test specimens of the base metal, weld and heat-affected zone shall be retained with full documentation and identification.

The test material shall follow a fabrication history fully representative of that used for the material in the irradiated region of the reactor vessel.

The chemical composition required by the material specifications for the test materials shall by obtained and include phosphorus, sulfur, copper and vanadium.

Type. orientation and location of specimens

Tension specimens shall be of the type, size and shape described in ASTM F 8.

Charpy V notch impact specimens corresponding to the type A specimen described in ASTMI A 370 (fig. 2) shall be employed.

Both irradiated and unirradiated types of specimens shall be of the same size and shape. (For tension specimens, only the gage section need be of the same size and shape).

103 Fusion - ' Line 'n- CL of Notch ,[.1d $ 1-FZ.- - - - - 1 - _ -

CL of Weld I IG. I, irtcimrn Orinlati in Mt!d 2nd IHA.

For both tension and impact specimens from base metal, the major axis of the specimen shall be machined normal to the principal rolling direction for plates and normal to the major working direction for forgings.

The length of the notch of the Charpy V'impact specimen shall be normal to the surface of the material. The orientations of the impact and tension specimens with respect to the weld are shown in fig. 2.

Weld metal tension specimens may be oriented in the same direction as the Charpy specimens provided that the gage length xonsists of all weld metal.

'No specimens are to be removed within 13 mm of the root or the surfaces of the welds.

Sections of the weldment shall be etched to define the weld heat affected zone.

The impact specimens from the weld heat-affected zones shall have their notch roots in the heat-affected zone at a standard distance of approximately 0,5mm from the fusion line.

Specimens representing the base metal (tension and impact) and the weld heat-affected zone shall be removed from the quarter thickness location. 104 The minimum number of test specimens, respectively for the 1 and 2 case and for each capsule, shall be as follows:

MERITAL 1 CASE ' 2 CAS CHIRPY CHARPY TENSION

Base metal 12 1? 2

Weld metal 12 12 2

Heat-affected zone 12 12

At least 15 Charpy impact specimens shall be used to establish an unirradiated transition curve for each material.

For 2 case, three tension test specimens shall be used to establish unirradiated tensile properties.

Withdrawal schedule

Specimens shall be irradiated at a location in the reactor that duplicates as closely as possible the neutron flux spectrum, temperature history and maximum accumulated neutron fluence experienced by the reactor vessel.

Surveillance capsules containing the surveillance specimens shall be located near but not attached to the inside vessel wall in the beltline region, so that the neutron flux received by the specimens is at least as high but no more than three times as high as that received by the vessel inner surface, and the thermal environment is as close as practical to that of the vessel inner surface.

The design and location of the capsules shall permit insertion of replacement capsules.

Accelerated irradiation capsules, for which the calculated neutron flux will exceed three times the calculated maximum neutron flux at the inside wall of the vessel, may be used - for information only - in addition to the required number of surveillance capsules.

105 Flux measurements

Dosimeters with the vessel wall specimens shall be employed to measure the neutron fluence.

Where accelerated irradiation specimens are used, dosimeters with the test specimens and dosimeters either in a separate flux monitor capsule adjacent to the vessel wall or in a vessel wall capsule shall be employed.

Calculation methods employed to predict the radiation exposure of the reactor vessel from the data revealed by the surveillance specimen dosimeters and flux monitor dosimeters (where used) and an estimate of the accuracy of the calculations shal. be recorded.

To prevent deterioration of the surface of the specimens during test, the specimens should be maintaned in an inert environment within a corrosion- resistent capsule.

Care shall be taken to ensure that the reactor vessel and specimen temperatures are similar.

Knowledge of the specimen temperature as well as the reactor vessel temperature during irradiation are required. The temperature history of the specimens shall duplicate as closely as possible the temperature experienced by the reactor vessel. Small pieces of low melting point elements or alloys may be inserted into the actual test specimens (in drilled holes at positions away from the failure sections) or in capsule filler pieces adjacent to the specimens to monitor the maximum temperature experienced by the specimens. Surveillance capsules shall be sufficiently rigid to prevent damage to the capsules by coolant pressure or coolant flow thus hindering specimen removal or causing inadvertent deformation of the specimens.

The design of the capsule shall permit the insertion of replacement capsules into the reactor at a later time in the lifetime of the vessel.

The require number of surveillance capsules and their withdrawal schedules are as follows: a) For reactor vessels which meet the conditions of Group A, at least three surveillance capsules shall be provided for subsequent withdrawal as follows:

106 Withdrawal schedule First capsule - One-fourth service life Second capsule - Three-fourths service life Third capsule - Standby

In the event that the surveillance specimens exhibit, at one-quarter of the vessel's service life, a shift of the reference temperature greater than originally predicted for similar material as recorded in the applicable technical specification, the remaining withdrawal schedule shall be modified as follows:

Revised withdrawal schedule Second capsule - One-half service life Third capsule - Standby b) For reactor vessels which meet the conditions of Group B, at least four surveillance capsules shall be provided for the subsequent withdrawal as follows:

Withdrawal schedule First capsule - At the time when the predicted shift of the adjusted reference tpmDerature is approximately 28C or at one-fourth service life, whichever is earlier Second capsule - At approximately one-half of the time interval between first and third capsule withdrawal Third capsule - Three-fourths service life Fourth capsule - Standby

c) For reactor vessels which meet the conditions of Group C, at least five surveillance capsules shall be provided for subsequent withdrawal as follows:

Withdrawal schedule First capsule - At the time when the predicted shift of the adjusted reference temperature is approximately 28C or at one-fourth service life, whichever is earlier Second and third capsules - At approximately one-third and two-thirds of the time interval between first and fourth capsule withdrawal Fourth capsule - Standby d) Provisions shall also be made for additional surveillance tests to monitor the effects of annealing and subsequent irradiation.

107 e) Withdrawal schedules may be modified to coincide with those refueling outapes or plant shutdowns most closely approaching the withdrawal schedule. f) If accelerated irradiation capsules are employed in addition to the minimum reruired number of surveillance capsules, the withdrawal schedule may be modified, taking account the test results obtained from testing of the specimens in the accelerated capsules. The proposed modified withdrawal schedule in such cases shall be approved on an individual case basis. g) Proposed withdrawal schedules that differ from those specified in paragraphs a) through f) shall be approved, with a technical justification therefore, on an individual case basis.

Predisposition program report

On the basis of the criteria and requirements established in previous paragraphs, the manufacturer shall present a report about the actual predisposition of surveillance program.

Measurement of neutron exposure

The neutron flux, neutron energy spectrum, and irradiation temperature of surveillance specimens and the method of determination shall be documented.

Flux dosimeters for a particular program shall be determined by referring to Method E 261. It is recommended that iron and unshielded cobalt dosimeters shall be included in every capsule regardless of their intended exposure location within the reactor.

Report of test results

Each capsule withdrawal and the results of the test shall be the subject of a summary technical report to be done by the owner.

The report shall include: a) withdrawal time; b) number and position of the capsule withdrawn; c) ' authorisation for test laboratory and calibration data for each apparatus; d) elements and considerations for the valuation of the results obtained accordinp to document NF.l.D.

108 Impact and tension tests

The owner shall present the report of the mechanical tests performed.

This report prepared by manager of the test laboratory shall include the following: a) description of the reactor vessel being surveyed and of the location of the surveillance specimens with respect tot he reactor vessel, thermal shields, and to the reactor core; b) description of the material tqsted including the metallurgical history and any deviation in this histroy from the reactor vessel metallurgical history; c) location and orientation of the test specimen in the parent material; d) data on radiation environment.

Tension tests

The report on tension tests shall include the following: a) tension specimen design; b) trade name and model of the testing machine, gripping devices, extensometer, and recording devices used in the test and in calibrating this apparatus; c) speed of testing and method of measuring the controlling testing speed; d) complete stress - strain curves; e) yield strength.or yield point and method of measurement; f) tensile strength, fracture load and fracture stress; g) uniform elongation and method of measurement; h) total elongation, and i) reduction of area.

Impact tests

The report on impact tests shall include the following: a) trade name, model of the testing machine, hammer kgf., force rating and striking velocity, and a description of the procedure used in the inspection and calibration of the testing machine.

109 Test data as follows: b) temperature of test; c) energy a'bsorbed by the specimen in breaking, reported in kgf.m; cl) percent ductile fracture; e) lateral expansion; f) Charpy 6,90 kgf.m temperature of unirradiated material and of each set of irradiated specimens along with the corresponding Charpy 6,90 kgrf.m temperature increases for these specimens, and g) energy a.,sorbed in the regioh of 100% shear fracture.

Deviations

Deviations in procedure from this mandatory practice shall. be identified and described full.y in the report.

110 COMMENTS ON REACTOR VESSEL SURVEILLANCE PROGRAMMES IN THE FEERAL REPUBLIC OF GERMANY

E. Bazant, BBR, FRG.

ABSTRACT

Some comments are presented in connection with the situation of the reactor vessel surveillance programmes in the Federal Republio of Germany. BR1's programme guidelines are mentioned.

Session I

The proposed topics for discussion in Session I are general reports reviewing national programmes on reactor vessel surveillance.

In the Federal Republic Germany (BRD) does not exist a national standard, rules or guides for the reactor vessel material surveillance program.

There are only material surveillance program (S) of Nuclear Steam Supply System suppliers as Kraftwerk- Union (KWU) and Babcock-Brown Boveri (BBR) for their nuclear power stations.

The BBR -Program at the present time is in the stage of licensing by the German licensing organisation (TUV).

The BBR-Program is designed in accordance with

- american standard ASTM E 185-73 "Surveillance Tests for Nuclear Reactor Vessels" and - state of the art given by american Babcock-Wilcox. (B&W)

The B&W practices, methods and criteria are - in compliance with the requirements of Appendix G to 10 CFR 50 "Fracture Toughness Requirements",

- in compliance with the requirements of Appendix H to 10 CFR 50, "Reactor Vessel Material Surveillance Program Requirements", 111 - in compliance with all data from research programs conducted at the Naval Research Laboratory, data from the Heavy Section Steel Technology Program and data from several american Surveillance programs.

- in accordance with Appendix G to the ASME Boiler and Pressure Vessel Code, Section III

- given in the Reactor Vessel Material Surveillance Program

Some fundamental work, given in various B&W-Reports are dropped in the american regulatory praxis.

Section III Surveillance leciu irenments

Peculiar in the Federal Republic of Germany is the discussion with people of "Burgeraktion Atomschutz Mittel- rhein e.V", the intervenor in Mulheim - Kr.lich. They claim, that the embrittlement of the vessel material would be accelerated under the influence of stress.

Resulting of these discussion, the government of Rheinland-Pfalz will release the 'eroctjon of Mill- heim-Karlich, if the safety requirement Pr. 9 is accomplished. The safety r'equirement Nr. 9 provides a sufficienL amount of surveillance specimensC Up to a certain extent the specimens shall be prestressed.

The BBC/BBR-Konsortium Nuc ear Power Station Mulheim- Karlich in this connection referred to the German Re- search Program "Component Safety", which is funded by the Federal Ministry of Science and Technology (BMFT), the industry and the public utilities. (EVU)

In this program also the influence of prestressing on the embrittlement of materials by irradiation of fast neutrons will be checked. The performance of the safety requirement and Research Program is not finally settled.

112 EVALUATION OF SURVEILLANCE SPECIMENS AND IN-SERVICE INSPECTION OF TUBES OF A-1 REACTOR HEAVY WATER CALANDRIA

P. Mrkous, M. Brumovsky, J. Prepechal

SKODA Works, Nuclear Power Plant Division, Plzen t CSSR

ABSTRACT

Basic principles of reactor pressure vessel surveillance programmes are considered, including materials, specimens, capsules, location of capsules, detectors and procedures.

1. Introduction

The reactor of the A-1 nuclear power plant is carbon dioxide cooled and heavy water moderated typeo Moderator is placed inside the aluminium alloy (type Al - MgSi) calandria. Calandria tubes made from the same material go 'through the vessel. Internal surface of them is washed by carbon dioxide of temperature t = 120 C, the external one by heavy water of temperature t = 40 - 900C. Above the heavy water level in heavy water calandria there is a gas cushion of carbon dioxide with some content of explosive mixture (up to 3%) like a product of radiolysis. Shape stability of calandria tubes wall appears to be the limiting factor of heavy water calandria service life and therefore the evaluation of wall thickness corrosion decrease and the evaluation of changes of mechanical characteristics was realized as a first phase of in-service inspection.

'2. Evaluation of calandria tubes wall thickness corrosion decrease

Methods of indirect evaluation of calandria tubes wall thickness corrosion decrease on the one hand and method of direct measuring of wall thickness by ultrasonics method on the other hand, were applied,

113 2.1 Indirect methods;

2.1.1 Beecription of methods

For the evaluation of corrosion situation in the heavy water circuit a polarizing resistance electroohemioal method was applied.

Electrode made from the same material as heavy water calandria, deeped into the corrosion medium (t = 40 - 90 C, p = 6 M'a) is polarized by direct current of constant value from an external

supply. Current density id and polarization s are fundamental data for calculation of the instantaneous corrosion rate,

id r = KT AE

where r - corrosion rate K - constant depending on electrochemical reaction only T - absolute temperature in K

id - polarizing current density

C $- polarization

Pi2- - R - polarizing resistance. id P

Constant K must be experimentally evaluated.

ElectrochemicaJl transducers (made from Al-alloy to be tested) were installed in heavy water circuit collectors before complex tests of the A-1 reactor. Instantaneous corrosion rate of indicat- ing electrode made from Al-alloy at hydrodynamically non-stablilized water flow conditions, when the laminar layer thickness approaches zero, is measured. Signals of transducers were used for estimation of corrosion near the bottom of the heavy water calandria.

Corrosion attack above the bottom of the heavy water calandria was evaluated by the use of surveillance specimens, spaced in its height. Laminar hydrodynamically stabilized flow and the existence of laminar boundary layer along calandria tubes and surveillance specimens is assumed.

114 2.1.2 Experimental results

Results of electrochemical method, by which the instantaneous corrosion rate at hydrodynamically non-stabilized water flow conditions was evaluated, show to the integral decrease of 0.35 mm/l/ during operating period. Evaluation of surveillance specimens spaced along the height of heavy water calandria, represents the wall thickness corrosion decrease for the case of hydrodynamically stabilized flow. Weight loss evaluation shows that the decrease caused by corrosion is in the range of 0.236 - 0.290 mm/2/.

2.2 Measuring of calandria tubes wall thickness by ultrasonic method

2.2.1 Description of method and equipment

For direct measuring of calandria tubes wall thickness the ultrasonic impulse method was used. It is based on the measuring of pass-time of ultrasonic impulse through the tube wall. Ultrasonic signal is transmitted into the measured point by the use of double-probe. The value required is received by subtract- ing the constant value of pass-time of impuls (through sliding rods of the double-probe and through the coupling agent) from the total period between sending and receiving of impuls by piezoelec- tric transducer. Time is digitally measured by counting impulses of suitable frequency, at which their period is equal to double- pass time through layer of thickness 0.1 mm. It enables to read off the thickness directly with accuracy to 0.1 mm.

The method mentioned was applied at reactor conditions. Ul- trasonic probe must work reliably without change of accuracy in temperature range from 20° to 120 ° C in a very strong ionization field and at pressure up to 6 MPa. In such case it is not possible to use current liquid agents for acoustic coupling and also some of known agents for dry acoustic coupling (for example plasticine). Calandria tube tested is located in depth of 12 to 16 m under the reactor hall floor and the signals have to be led additional 50 m to the measuring stand. During measuring of calandria tubes wall thickness the reactor does not operate, but it is hermetically

115 closed and internal pressure is 1 MPa at least. To secure the possibility of repetition of wall thickness decrease evaluation, replacement of the probe to the same position have to be accurate to + 1 mm. A great number of points measured requests certain degree of automation of equipment and the automatic record of data measured.

Measuring equipment developed consists of mechanical part and the block of measuring, supplying, controling and registering in- struments. Mechanical part of total length about 20 m consists of two parts# the lower of them bearing ultrasonic probe, is inserted into the reactor by the use of charging machine by the same way as fuel elements are inserted. The upper part protruding into the reactor hall serves for probe moving disposal. Signal from the probe goes through the mechanical section in steel shielded coazial cables with mineral insulation. It enables that high frequency signals can be led out of hermetically closed space at acceptable level of signal loss. Temperature and radiation resistance of the probe was reached by consistent use of inorganic constructional and insulat- ing materials. Suitable probe design secures the right coupling of its contact surface to the surface tested. Special way of dry acoustic coupling enables carrying out of several thousands measurements without refilling or renewing of coupling agent.

The accuracy of measurements was verified on the models of calandria tubes of graded thickness.

2.2.2 Experimental results

Measurements were realized on eight selected tubes in two regions. First region was situated in the lower part, the second in the central part of heavy water calandria. The length of sections tested were 200 mm and the measurements were made on twelve surface lines, equally distanced round the tube perimeter.

Wall thickness decrease measured was in the range 0 - 0.3 mm/ 3/. More detailed evaluation showed, that the transition between the points with smaller and greater corrosion loss is gentle and therefore the corrosion attack has uniform character.

116 3. Evaluation of changes of mechanical properties

For checking of changes of mechanical properties of calandria tubes material, caused by reactor operation, two special channels in peripheral part of reactor core were established and chains of containers with surveillance specimens were placed into. The purpose of surveillance specimens was twofold:

- to enable weight estimation of corrosion loss - to realize static tension test (base material and welded joints) after evaluation of corrosion loss (after corrosion layer removal).

In 1975, the containers number 5A1, 5A3, 5A5t 5A7, 5All were took out from channel A5 and according to the state of specimens placed in those containers, the corrosion loss estimation was done. Then tension tests and factography of fracture surfaces were done. For comparison, the container number 5A10 was took out too. The specimens from this container were tested by tension tests without removal of corrosion layer.

3.1 Experimental results /4/

Results of static tension tests made at temperature 20 0 and re-counted to the real section of testing specimen are in table 1. The following values were determined: 0.2 offset yield strength ( 0.2)' ultimate tensile strength ( U ), uniform elongation ( UT) total elongation ( tot)' total strain work (A). In the table are also written down the mean distance (h) between container and the upper surface of dibiding plate of heavy water calandria"'and estimated neutron dose (0 t) absorbed by surveillance specimens of mentioned container.

Table 2. shows the influence of temperature during test (there is always 3 hours delay at testing temperature). The strain rate during the test was in all cases approximately 6.67 x 10-4s 1 . Both values for base material and for welded joints are introduced. In the case of welded joints, there was not removed reinforcement and therefore in most cases the failure in base material appeared.

117 Graphic illustration of results versus distance (h) shows fig. 1. In common with mean values, the mean square errors of those values are plotted.

The analysis of results shows:

- Strength properties of the base material and of welded joints (0.2 offset yield strength, ultimate tensile strength) do not differ in fact, if the failure in base material occurs,

- The ductility of base material is approximately twice higher than the ductility of specimens with welded joints; it shows to the influence of the heat affected zone, and by reinforcement of weld metal.

- Total strain work for base material is also approximately twice higher than at welded joints (it is caused first of all by different values of elongation).

- Increase of temperature during test from 20° to 1000 resp. to 2300C causes ultimate tensile strength decrease of about 25, resp. ,IPa.

3.2 Evaluating of changes of mechanical properties

During reactor operation, observable changes of properties took place. It is first of all:

3.2.1 Base material

- Moderate increase of 0.2 offset yield strength and ultimate tensile strength in comparison with the original state (approximately at 20o at 0,2 yield strength and at 10% at ultimate tensile strength, which is more than the mean error of results).

- Decrease of both uniform and total elongation (roughly at 10%) took place simultaneously.

- Total strain work did not almost change.

118 Fracture surface character has changed. While at specimens in original state and specimens from container number 5A1, i.e. the specimens with low neutron dose and practically without corrosion loss, the fracture character is ductile with deep cup, at specimens from container number 5A3 and 5A5 with higher neutron dose, is mixed fracture observable (fracture of ductile and slanting character). At test specimens from containers number 5A7, 5All and 5A10 with high neutron doses and highest corrosion losses the fracture character is already totally smooth and slanting. Predominant part of fracture surface of all specimens has pitting struc-

ture t characterizing intragranular plastic failure.

Increase of strength properties influenced by reactor operation is observable even at temperatures 1000 and 230°C and just then the decrease with increasing testing temperature for irradiated and non-irradiated material is roughly the same (about 25 to 35 MPa at 100°C and 90 MPa at 230°C at ultimate tensile strength).

Uniform elongation at higher testing temperatures decreases (up to 50% of original value at 230°C test); total elongation is changed very little.

3.2.2 Welding joints

- Strength properties (0.2 offset yield strength and ultimate tensile strength) change (i.e. increase) corresponds approximately with changes of base material (because the fracture occurs in base material.

Elongation decreases by the irradiation.

- Total strain work decreases as well.

- Fracture surface character is the same like at the base material.

- Decrease of strength properties at higher testing temperature (1000C) is roughly the same like for the base material.

119 The corrosion layer at test specimens did not influence in fact neither the change of strength properties nor the elongation (in comparison with test specimens with removed corrosion layer). It shows that the brittle corrosion layer at test specimen does not influence unfavourably the ductility of test specimen by its cracking and flaking-off during the test (transfer of these cracks from surface oxide layer into the base material does not probably occur). By another words, removal of corrosion layer by chemical way does not observably influence mechanical properties of material. The attempt to evaluate the stress-strain diagrams from test specimens (surveillance specimens') aimed to determination of Young's modulus was done. It was shown that the Young modulus of elasticity will not decrease by the influence of neutron dose during reactor operation, but moreover it will be moderately increasing.

3.3 Evaluation of results

Results obtained prove the facts that heavy water calandria tubes material (alluminium alloy of type Al - Mg - Si) in applicated state, i*e. after stabilizing the annealing, is in fact full stable at operating temperature up to 100 C even at operating period longer than two years. Small changes of material properties can be explained by the influence of high neutron dose (r, 3.10 2 5 n.m 2 ).

4.0 Conclusions

At first phase of in-service inspection of heavy water cal- landria of the A-1 reactor the following was done:

(a) Evaluation of corrosion losses by three independent methods. Results obtained are in a good agreement and prove the objectivity of ultrasonic method, used for wall thickness measurements.

(b) Evaluation of changes of mechanical properties proved high irradiation stability of Al - Mg - Si alloy; it is in a good agreement with literature /5/.

(c) Results obtained appears to be a sufficient base for service life evaluation of the A-1 reactor heavy water calandria.

120 5.0 Literature

/1/ Beran J., Tomfk L. - Experiment R5, Report iKODA + EBO, AE 3498/Dok (in Czech), 1975

/2/ Tomfk I., Kone~ny L. - Report on evaluation of corrossion in irradiated aluminium alloy type specimens, Report EBO (in Slovak), 22.8.1975

/3/ Prepechal J., Bumbalek A. - In-service inspection of the A-1 reactor during 1975t Report KODA - ZVJE, Ae 3672/Dok (in Czech), 1976

/4/ Vacek M. et al. - Mechanical tests of surveillance specimens from the A-1 heavy water calandria tubes, Progress Report, Inst. of Nuclear Research (in Czech), 1975

/5/ Votinov S.N. et al - Study of radiation stability of calandria material for heavy water reactor, Conf. Atomic power, nuclear cycles and radiation metallurgy, Ulyanovsk, U.S.S.R., Vol. 3, p. 426, 1970 (in Russian).

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