reactor pressure vessel surveillance

IAEA-202

REACTOR PRESSURE VESSEL SURVEILLANCEPROCEEDINGS OF A TECHNICAL COMMITTEE MEETING

ORGANIZED BY THEINTERNATIONAL ATOMIC ENERGY AGENCY

WITHIN THE FRAMEWORK OF THEINTERNATIONAL 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, CorrespondingMember of the Academy of Science,CSSR.

Dr. Len Steele, Naval ResearchLaboratory, U.S.A.

Ing. S. Havel, Nuclear ResearchInstitute, 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 AustriaNovember 1977

PLEASE BE AWARE THATALL 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

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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 nationalprogrammes 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. SteeleNaval Research Laboratory

ABSTRACT

The history of the study of radiation embrittlement ofreactor pressure vessel steels has gone through three stagesin the USA -

1. A scientific curiosity,2. Empirical or laboratory evaluation of typical

steels, and3. Integration of the scientific and empirical to

advance status and evolve standard techniques.

The current stage is one in which surveillance data complimentsthe laboratory studies which characterized Stage 3. The earlyUSA surveillance programs were generally analyzed by the samepeople who were the primary laboratory investigators. An effortmust be made to continue this type of collaboration as a usefultwo-way learning procedure though it will become more and moredifficult as nuclear power is broadly commercialized. The cur-rent status of both types of USA programs will be presented toencourage the most advantageous use of data from both sources.

At this time about 25 USA nuclear power reactors have operatedlong enough to have provided initial surveillance or dosimetryresults. An effort will be made to summarize the general statusof these in order to:

1. Provide complimentary data to laboratory studies.

2. Assess directions in handling the problems ofradiation embrittlement.

3. Note lessons learned for improving surveillanceefforts in the future.

4. Identify possible research tasks for the future tosupport in-service surveillance and other measures.

5. Justify facts advancing surveillance requirementsto status of national codes and standards.

6. Justify facts requiring changes in current nationalcodes and standards.

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

INTRODUCTION

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

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and reactor design came only in the mid-to-late 1960s. For thisreason, though there were doubts enough to begin surveillance pro-gram planning, it was well after 1970 before the research datasignificantly affected design of new nuclear plants through limitson composition and on properties of pressure vessel steels.

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

The growing volume of surveillance data available in the 1970sadds a new dimension complementing greatly the available researchdata. This is particularly important as the rules or guides of theU. S. Nuclear Regulatory Commission have become more definitive inrecent years.

CRITICAL ELEMENTS OF SURVEILLANCE

The critical elements of surveillance include:

1. The fracture behavior of steels used to construct pressurevessels,

2. The influence of neutron radiation on fracture performanceof 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 tominimize influence of these factors).

All of these factors have been considered in the application of vesselsurveillance though there is room for improvement in each.

SUMMARY OF STATUS FROM RESEARCH VIEWPOINT

The fracture of steels has been the subject of much researchand technology in recent years and these advancements impact directlyon the question of reactor vessel integrity. On the positive side,empirical studies clarified the point of transition called the nilductility transition temperature and the influence of temperatureabove this point and steel thickness as well. Advancements in theunderstanding of linear elastic fracture mechanics advanced to thebenefit of quantitative analysis of flawed steel structures and,when full thickness tests are included, provides the basis for fullvessel criteria for fracture prevention. This is not to say all theessential background has been completed, but rather that the rightway has been pointed out. The greatest remaining need is to under-stand fracture in the ductile portion of the transition curve, moresystematic full section tests for all vessel components, and bettersmall specimen representation of full section.

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The influence of neutron radiation on steel fracture has beenadvanced through years of research studies to a point of relativematurity but has been limited largely by complexity of conductingfull section irradiated tests and by lack of the opportunity for astatistically based irradiation study on real reactor vessel steels.The greatest accomplishment has been the verification of residualelement effects (copper and phosphorus especially) and the advance-ment of this factor to the point of standardization limitations aswell as to inclusion in regulatory guides. The latter factors havemajor influences on surveillance implications; putting a criticalview on older reactors and possibly providing the impetus for im-proved future steels and hence reduced needs for surveillance forreactors constructed in the future.

Advances in the techniques for measurement of vessel radiationexposure have been great in support of research experiments. Majorstudies were necessary to validate irradiation embrittlement studiesconducted in research or testing reactors for projection to the powerreactor condition. Critical aspects were to measure neutron flux,fluence and spectrum and to relate these to vessel steel responsethereto. In addition, reflecting the relatively mature status of ourresearch knowledge, in these areas, a series of standards have beenpublished in support of their application.

The relative response of individual steels or steel componentsis especially critical and is related to the important factor ofsteel composition and its influence on radiation embrittlement whichwas noted above. Experience in research programs indicates high sen-sitivity in some weld metals containing high copper levels. Not allof this can be assigned to the copper level. A more systematic studyof various steels typical of those in service is needed to aid in theprojection of embrittlement and to guide establishment of regulatorycriteria.

The implications of radiation embrittlement to reactor operationhas been limited to projections based upon research and design data.Nevertheless, conservative application of these data provides a tech-nique for minimizing the potential effects by controlling vessel tem-perature and applied stress during normal startups and shutdowns. Thistechnique is especially applicable to the pressurized water reactorwhere pump heat may be used to gradually increase vessel temperaturebefore initiating nuclear power. This technique has little value forminimizing the effects of low energy ductile upper shelf however, aproblem which is probably the most critical if we are to assure vesselreliability. The fullest implication of radiation embrittlement toreactor operation must await pertinent surveillance results.

SUMMARY OF STATUS FROM SURVEILLANCE VIEWPOINT

While surveillance is not designed to provide research data, suchresults can contribute significantly to the fullest understanding ofradiation embrittlement of vessel steels. Combining research and sur-veillance results permits an assessment of the status of knowledgeas well as future needs.

The fracture behavior of vessel steels as evaluated for surveil-lance has been based largely on small Charpy V-notch specimens. Thus,the projection of effects in terms of vessel fracture potential hasbeen based on extrapolation from large unirradiated tests, on linearelastic fracture mechanics evaluations, and on radiation produced

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changes based on small surveillance specimens. This three-way analysisis useful but is not adequate for the direct quantitative evaluationdesired. The ideal would be an evaluation of the fracture behaviorof the poorest component steel in a vessel - an irradiated K curvefor that material. The best hope for reaching this goal involvesexperiments on large test specimens coupled with an acceptable quan-titative analysis based upon a smaller specimen which has been corre-lated with the larger specimens over the transition and upper shelfregions.

For the ultimate test of neutron radiation on steel fracture,large irradiated fracture mechanics specimens must be tested forAT (irradiation induced change) and for low shelf (radiation inducedshelf drop) conditions. Acceptable small correlation specimens whichdescribe the heavy section results must be found for surveillance ifdirect fracture performance of the vessel is to be defined from sur-veillance.

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

One of the major contributions of surveillance has been indefining the specific response of various vessel steels or componentsof steels to service radiation. Generally, such results have vali-dated the research results relative to composition (copper and phos-phorus) effects but on the actual steels used in vessel construction.The most startling observation has been on welds of vessels constructedbefore about 1970 wherein copper was used as a coating on weld rodsand therefore were high in copper and sensitive to radiation embrittle-ment. Such results are clear from initial surveillance data fromplants such as Maine Yankee, which contained high copper welds and arelated high level of embrittlement, a major vote for standards tocontrol copper in vessel steels. By contrast, where steel compositionhas been controlled (especially copper, phosphorus, and sulfur) as inthe plate and forgings of later USA reactors, the sensitivity to radia-tion and the upper shelf toughness are superior to all earlier steelsand hence offer the utlimate answer for vessel surveillance - improvesteels to the point where surveillance is no longer needed or is minimalat most. The meshing of projections from both research and surveillancein this case represents one of the most conclusive and positive resultsof the whole area of steel embrittlement study.

The application of surveillance results have been used in manycases to establish operating guidelines for reactor startup andshutdown. It is believed that such guides, which often are appliedroutinely while major changes in vessel toughness may be years away,would not have been so applied if it were not for a well developedbody of data on radiation embrittlement from research experimentsobtained in years of study.

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APPLICABLE NATIONAL CODES, REGULATORY GUIDES, AND STANDARDS

Because of the relative maturity of this whole area of study asnoted above and its importance to reactor safety, a series of majornational codes, guides, and standards have been issued. For use ofreference 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 whichrelate to the question of radiation embrittlement and surveillancein a secondary way. Most notable of these are those which defineprocedures for measuring the neutron environment and which complementE185 on surveillance and the evaluation or amelioration of radiationeffects. Table 2 lists many of these secondary standards which areimportant. In addition, ASME has contributed to the area of in-service inspection with Section XI of the Boiler and Pressure VesselCode.

In spite of the fact that several well developed codes, standards,and guides are published and in practice in the USA to support reactorpressure vessel reliability, there remain important research tasks andopportunities for improvement of these documents. Advances to supportthe needed improvements in these national documents (and nuclear tech-nology) 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)*

ASME Sect.lII, APP.G (1972)*

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

AEC 10 CFR 50, APP.H (1973)*

NRC Reg. Guide 1.99 (1975)*

Surveillance Tests for NuclearReactor Vessels

Protection Against NonductileFailure

Fracture Toughness Requirements

Reactor Vessel Material Surveil-lance Program Requirements

Effects of Residual Elements onPredicted Radiation Damage toReactor Vessel Materials

Date first issued.

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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 Radiationon 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 Radioactivationof Iron

E-264-70 Measuring Fast-Neutron Flux by Radioactivationof Nickel

E-265-70 Measuring Fast-Neutron Flux by Radioactivationof Sulfur

E-266-70 Measuring Fast-Neutron Flux by Radioactivationof Aluminum

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

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

E-418-73 Fast-Neutron Flux Measurements by Track-EtchTechnique

E-419-73 Guide for Selection of Neutron ActivationDector 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 WaterCooled Nuclear Reactor Vessels

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TABLE 3

SOME REFINEMENTS DESIRED IN KEY

NATIONAL CODES, STANDARDS, AND GUIDES

Document Refinement Desired

ASTM E-185 Materials selection criteria,shelf analysis; define dosimetryapproaches and fracture specimenchoice.

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

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

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

NRC Reg. Guide 1.99 Better statistical base for bothsteels and fluences to define bothat AT and AE. Reward for lowsensitivity.

USA REACTOR SURVE ILLANCE RESULTS - A SUMMARY

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

For our purposes it is probably best to minimize catalogingby type and concentrate on status of development since generalconclusions can be made best by the latter. The Hanford-N reactor,a light water cooled, graphite moderated reactor uses pressuretubes rather than a pressure vessel and hence should be dismissed.from this discussion. Most of the early BWRs had vessels similarto those of PWRs in type of steel and in projected fluences on thevessel so these can be treated together. The later, larger BWRswere designed for lower fluences and must be so reported. Thelater PWRs are all similar in the factors which affect surveillanceinterpretation so it is possible to discuss together in generalterms 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) - werequite similar in lifetime fluences determined from initial surveil-lance except for Humboldt Bay and laCrosse which had lower projected

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1

TABLE 4

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

Commercial Net Reactor GeneratorReactor 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


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

Monticello 7/71 548 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 GEI'I-- - - - - -- --- � - -

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lifetimes - 20 years versus 30 or 40 years. Lifetime fluences werefound to be generally from 1 to 3 x 10'1 n/cm" (>1 MeV) and steelsensitivity was in the mid-range of that found lor steels studied in

research programs. In every case some modification of the operatingschedule 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, neutrondosimetry, materials selection, and fracture evaluation knowledge.The positive lessons learned from ihese negatives have had majorimpact 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 thelate 1960s. The surveillance results offer two warnings: (a)rela-

tively.high fluences on the vessel (3 to 4xl0 -), and (b) weld metaland weld heat affected zone materials sensitive to radiation and someof 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 theintermediate 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 ofcopper and phosphorus and the shelf degradation of sulfur was widelypublicized by NRL in 1967-1968. It is clear that for these olderplants, vessel surveillance plus advances in the compltmentary tech-

nologies of steel fracture, neutron dosimetry, and reactor designwill require great attention and diligence in the years just aheadof us.

SUMMARY AND CONCLUSIONS

This overview is too brief for precise statements of conclusions.Further the generalizations above are inappropriate for applicationto individual reactors but do provide a basis for a general look atsurveillance as a complimentary effort to research studies of radia-tion embrittlement. It is fair to say in summary, I believe, thatresearch data have provided the background necessary for formulatingand interpreting results of reactor vessel surveillance programs.Furthermore, it must be clear to all by now that much remains to be donein order to optimize the results of such programs, since the next20-30 years for several early light water reactors, will requireextreme vigilance to assure no catastrophy traceable to a reactorvessel failure resulting, even partially, from radiation embrittlement.

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

1. Promote the advancement of limits on residual element content

to full status of an enforced national code. (This appliesboth to elements affecting radiation sensitivity and shelftoughness.)

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

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3. Advance criteria for describing neutron exposure in surveil-lance for selecting lifetime peak fluences on the vessel tostatus of mandatory code or standard.

4. Develop procedures for improved surveillance specimens basedon quantitative fracture mechanics procedures. At the sametime by correlation or other means advance our understandingof 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 composi-tion and embrittlement.

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

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

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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;eci-

mens) 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.

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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 followingbasic 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 sub-

sequent annealing at furnace to relieve stress.

Whereas the rings were electroslag welded from four segments,

manual-arc circumferential welds to weld the vessel sections to-

gether 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 substan-tial 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 displace-

ment,COD) and an energy one based on the critical value of J-integral

are not yet sufficiently worked out.

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On account of historical reasons during the projection and n'niufcc -

ture of the A-1 reactor pressure vessel, the concept of crack e'rres-ting temperature was used predominrm tly. It is somewhat moreconservative, which to some extent may compensate for the increaseddcanger of star ing and prop agating brittle failure caused by thehigh 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 ofinitiating a failure, i.e. the propagation .nnd growth of flawvs, buta temperature is looked for at which , under a given stress, thebrittle crack, if any, stops propagating. If the operation-1 tempe-rature is higher than the temperature established in this way, acatastrophical 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 toreactor operation (it includes effects due to irradia-tion, aging, low-cycle damage and the like) i.e. CAT,

A T I is the safety factor (with regard to transition states), weput it + 30 ° for given steel,

4 TE is the crack arresting temperature increase due to accumulatedenergy (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 tonominal stress well under the material yield point, i.e. in the elasticzone damage due to low cycle fatigue or deformation aging is negligi-ble) the only variable in the course of reactor operation is radia-tion aging which isshown by radiation embrittlemeht, this is dT. =

A o dnmatge )

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. Anotheradvantage 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. Thusthe 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 operationalsafety. But this approach modified by NDT is incorporated instandards /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 is the critical size of a flow.c

Fracture toughness KIC depends primarily upon the state of materialand testing temperature. Operation-induced changes can be expressed

as follows:

KiC (t r O,T) = KIC (t = O,T - d Tdmage (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 determinethe fracture toughness value in a time instant in question and thusto establish the critical flaw size. From the results of flaw-detectionchecking 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 TK (4)

and

AK iC r(a Tdamae ) TK (5)

where A TK is the change of transition temperature caused bymaterial damage in the. course of operation and determined bynotch toughness tests onr Charpy-V specimens.

If the two relations (4) and (5) hold true, then the followingrelation must be valid:

CAT = TKIC ( Tdm e ) (6)

To assess the validityof relation (4) only incomplete datn areavailable so far, because of difficulties we are coming across inirradiating specimens of large thickness at the crack-arrest tem-perature. 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 withspecimens having various dimensions and thicknesses and performedwith different loading rates (d CAT and A TK by means of dynamicand A TKIC by means of static tests). Yet, all the analyses medeso far are based on the validity of this relation and it is alsoa basis on which to Set up programmes of surveillance specimens.

At present both approaches are practically used for analysingsafety and service life of the A-1 reactor pressure vessel. Thecrack-arrest temperature still remains a prim.ry criterion, becauseof the fact that periodic inspections were not considered in theoriginal project (these are included additionally where needed inaccessible areas - see 181). With regard to the high value of accumu-lated energy and consequently to factor A TE in relation (1) andrelatively low operational temperature of the vessel, we obtain, inconnection with the supposed transition temperature shift due tooperation, the requested working temperature close to or even higherthan the operational temperature. For this reason the CAT approach doe.not ensure a perfect safety against failure and an analysis based onlinear fracture mechanics must be made and critical flaw sizes esta-blished. The process is shown in t8) in detail.

15

It means, that the surveillance specimens program employs test sa7mplesmaking 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 proper-

ties takes place (ductility, contraction) and thus the results obtai-

ned 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 pre-

vent 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 iden-

tical 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 heatsmay vary from one another, it was necessary to carry oit supple-

ment reference test.

16

The overall programme of surveillance specimens is thereforedivided into two parts:

1. Specific resistance tests2. 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 indi-vidual parts of the. programme is shown in Tab.2.

2.1. Specific resistance tests

These tests were carried out before starting up the reactor onpower operation. The recommendation from (11l could not be usedbecause our conditions are quite different: lower irradiatingtemperatures when the contents of interstatical atoms (C,N) playsthe 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 thecylindrical section of pressure vessel in the core area in terms ofradiation damageresistance. The samples were taken direct from wel-ding specimens in individual rings. Irradiation was carried out inthe WWR-S experimental reactor at Nuclear Research Institute, Ae2under the conditions simulating 20 years of service so that we canspeak of accelerated irradiation tests. Irradiation temperature wasapprox. 85°C (121. The experiment was to' determine which of thematerials studied had the minimal radiation damage resistance andwhat is its difference as compared with the heat chosen for in-serting into the reactor within the programme of long-term stability

(to insert the least resistant charge into the reactor was notpossible because of lack of material needed to make specimens forthis vast programme). Industrial materials (basic material, weldmetal and heat affected zone) were evaluated in terms of strengthproperties (primarily yield point) and notch toughness (transitiontemperature).

The following weld joints were selected for irradiation: (eachtime the two basic metals, the two heat-aCfected zones at a distan-ce 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 - 36M 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 tempera-

ture, 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 ] MP1 / 2 9A *- , 72 7 [MPaWM - 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%) bet-

ween basic material and weld metal' and the overall error is also rela-

tively 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

HAZ - ES - B1/2 =35 T 2,6 1 0C

- 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 ratioof constants A1/2 : B 1 /2 is constant for identical.materials. The fol-lowing values were determined for the tests.given:

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

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

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

19

was practically no difference in radiation damage of all the materialstested.

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 toradiation 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 referen-ce heat M4548 End its weld joints and consists of the followingtests (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 endwelding joints located on the vessel and irradiated within the speci-fic resistance programme. At the same time they were to set founda-tions for preliminary determining lifetime of the A- reactor pressurevessel.

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 speci-

mens 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 materialsranged 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 pre-

viously determined value range of 95% reliability. It did not becomeevident that some weld joints or joints location were more sensibleto radiation embrittlement than other ones. Hence, in further consi-

derations 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 tem-

perature (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. Theresults are summarized in Fig.4; they are designed BM-T with irradia-

tion temperature.

The results show that after irradiation at temperatures under 180°Cappreciable radiation damage recovery accurs. Results of the irradia-

tion 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 4548and 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 pres-tress 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, II-3, III-4, III-5, III-6;- IV-1.

In each set the specimens are arranged height-wise and in individualchains in such a way that, they could make a group in order to,hr.vea no:ded number of specimnon for one tranaition curve from onematerial by one type of test. Here it is assumed that set I will bewithdrawn first (approx.after five-year operation) and will be usedto the preliminary comparison of forecast and actual changes inmaterial properties and neutron doses,.Sets II and III will be usedto assess lifetime of the pressure vessel with more precision follo-wing 10 and 15 years of service. Set IV will be used to improveassessment of residual lifetime pt the end of operation, i.e. after20 years.

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

203. Size-effect Factor.

The crack arresting temperature - CAT -'depends not only upon thestatus of material but also upon specimen thickness - the latterdependance is known as the size-effect. Previous work carried out

at SKODA-ZVJE work /1,2,3/ on the ZZ 8000 machine has shown thatincrease in specimen thickness by 50 mm leads to increase in crack-arresting temperature by about + 10°C, which represents an increaseof at least + 30°C when going from 10 mm(the thickness of standardspecimens) to a thickness of 150 mm.

22

The validity of relation (4) remains to be answered, i.e. whetherit is influenced by the size-effect.

To verify this fact a series of tests was performed both on theZZ 8000 machine and on small specimens. 'As the irradiation andevaluation cannot be made on specimens of actual thickness 'nothertechnique had to be used to imitate radiation damnge. A method ofartificial mechanical ageing was chosen which provokes practicallythe some outside damage effects - hardening (increase. in yieldpoint and strength) and embrittlement (increase in transitiontemperature). Artificial mechanical ageing (3; 6,5; 7,5 and 8,5%of plastic strain)was performed direct on the machine with testspecimens of actual thickness (150 mm), followed by a dwell at2500 C. After cooling down a test was made to determine crackarresting temperatur, by ESSOmethod (i.e. dynamically) at a ratedstress of about 150 MPa. After carrying out the tests a new speci-mens were manufactured from the remainders of the test specimensfor notch toughness test of Cherpy-V type and tests were performed.The results of tests are summarized in Fig.5. 114/ . On thex-axis the transition tenperatures are plotted as determined onRV type bars by a criterion 35 J.cm- 2, i.e. T5 . On the y-axisthe crack-arrest- temperatures are plotted. By connecting therespective points, a dependence of the change in transition tempe-rature f'om notch toughness tests ( TSv) upon the change of thecrack-arrest temperature ( CAT) has been plotted. The resultsindicate that both the changes, obtained from different methods,are in a very good consistency - the change of crack ar3estingtemperature is somewhat less than that of transition temperature,which may be caused by slightly uneven deformation distributionalong the body thickness. But the validity of relation (4) may beconsidered to be proved experimentally: no size effect was foundout 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 donebasically by methods given in Chapter 2. The analysis is beingconducted in two steps: first, from the point of view of CAT,secondly from the standpoint of critical flaw size. (esultingcomparison of both results is used to improve lifetime assessmentand also to pick out locations for operation inspections and toset 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 consi-deration 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 (notchtoughness, static bend, Charpy-V, Mesnager.etc)

aTir,spec -shift due to the difference of neutron spectra in irra-diated location on experimental reactor and in A-1pressure vessel.wall.

Whereas the first two relations are being determined experimentally,the last one (spectrum effect) must be calculated. On the basis ofcalculation and preliminary results a correction to allow for thesedifferences can be done. With neutron doses of energies above

1 MeV radiation damage tomaterial of the A-1 reactor pressurevessel is expected 20% higher :than it was found out at acceleratedtest in VWIR-S reactor.

2o2. Fracture Touehness Approach.

Within this .nralysis the assessment of critical flaw size forpressure vessel material is being made by methods of linear

fracture mechuinics.. Knowing fracture mechanics parameters, rateof 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) maybe forecast after a given operational period in relation toservice time (t), number of operation cycles (N) and operationaltemperature (T). A scheme .of this assessment is given in Tab.4.

To estimate the irradiation-induced change in mechanical proper-ties, i.e. (KiC) = f (0t, T), relation (3) is used and conse-quently the some transition temperature shifts 4 Tir =ACAT asin 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 surveil-lance specimens fabricated from the A-1 reactor pressure vessel

material were accomplished. Most valuable .data on pressure vesselmaterial behaviour (basic material and weld joints) in the courseof 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.partof the A-1 reactor pressure vessel (in the core region) hasbeen found out.

2/ Good agreement of radiation damage resistance of materials usedfor actual pressure vessel with materials manufactured withinresearch and development work has been reached.

3/ Yield point increase - transition temperature increaserelationship 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 methoddoes 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 tempera-ture 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 mindthat the surveillance programme in question is 'not a standardone 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-1Reactor", IAEA Symposium on Performance of Nuclear PowerReactor Components, Prague,November 10-14,1969,IAEAPublication 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 CzechoslovakNuclear Power Stations", IAEA Panel on Recurring Inspectionsof Nuclear Reactor Steel Pressure Vessels,Plzen,1966,IAEAPublication STI/PUB/81, p.73. Vienna 1969

/3/ Brumovsky M.,K6lna K. Vacek M. - "Safety of Reactor PressureVessels from the Standpoint of Brittle Fracture", 4th UnitedNations 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 forInservice Inspection of Nuclear Power Plant Components

/6/ Brumovsky M.,KAlna K.,ltepgnek S.,Urban A.- "Study of CrackInitiation-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 ofIrradiated Materials", Quarterly Progress Report on ReactorSafety Programs Sponsored by the NRC Division of ReactorSafety Research for April-June 1975, II.Heavy-Section SteelTechnology Program, ORNL-TM-5021,Vol.II,pp.24 - 32

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

/9/ Brumovsky M.- "Radiation Damage and Surveillance Programs forCzechoslovak Reactor Steel Pressure Vessels", IAEA TechnicalReport No.IAEa-117 on "Development of Advanced Reactor PressureVessel Materials", Paper No.6, Vienna 1970

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

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

12/ Vacek M. - :Radiation Damage of the A-% Reactor Pressure VesselWelding Joints after Irradiation at 85 C (in Czech), NuclearResearch Institute, Report UJV 3621 - M, Re2 1975

26

/13/

/14/

Vacek M., terv6aek J.,Havel S.,ChamrAd B.,Pav T. - "RadiationEmbrittlement of Surveillance Specimens from the A-1 ReactorPressure Vessel after Accelerated Irradiation, (In Czech),Nuclear Research Institute, Report tJV 3807-M,Ret 1976

Brumovsky M, - "Size Effect in Irradiation Embrittlement ofSteels", ASTM 8th International Symposium on the Effects ofRadiation on Structural Materials, St.Louis, May 4-6,1976

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27

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of pre-stressed surveillance

specimens

31

AFUTROM fI LX WON THE' M/S/DPRiESSURL ' 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

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33

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34

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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.

37

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38

MATERIAL SURVEILLANCE PROGRAMME

OF

PRESSURE VESSEL STEELS IN INDIA

K. S. Sivaramakrishnan

Radiometallurgy Section, Metallurgy Group, Bhabha Atomic ResearchCentre, 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 al-

so manifested in the shift in the Nil Ductility Transition tem-

perature 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 dur-

ing 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 ten-

sile 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 accelera-

ted neutron flux positions compared to the pressure vessel wall

(Fig. 1). In addition, thermal control specimens are kept at lo-

cations 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 compo-

sition and heat treatment is given in Table 2. Charpy V-notch and

tensile samples from this material have been prepared to be irra-

diated 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 fabri-

cated 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 flux232

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 pro-

perties 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

1 0.20 1.27 0.21 0.0:

2 0.20 1.32 0.18 0.2;

Heat Treatment :

a. Austenitize : 1725 - 1775 (

b. Water spray to : 500°F

o. Temper : 1200 - 1250'

d. Air cool

16

)

S

0.036'

0.023

Mo

0.47

0.46

'F

°F for 10 hrs.

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

42

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46

WESTINGHOUSE NUCLEAR EUROPE REACTOR VESSEL SURVEILLANCE PROGRAM

T. R. Mager

Westinghouse Nuclear EuropeBrussels, 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 availa-

ble 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 regula-

tory 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 102

n/cm the reactor vessel materials studied exhibited fracture

toughness sufficiently high for continued safe operation of

the nuclear power plants.

INPTDUCTION

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, post irradiation data are available from six of these

plants. A list of the ten plants are given in Table 1.

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 compos-

ition 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.

51

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52

Table 2

CONTENTS OF CURRENTWESTINGHOUSE NUCLEAR EUROPE

SURVEILLANCE CAPSULES

MATERIAL

LOWER SHELL COURSE *

INTERMEDIATE SHELL COURSE *

No. OFCHARPYS

18

18

No. OF No. OFTENSILES 1/2 T - CT's

6

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

Ni 5 8

( n,p ) Mn54

( n,p }

Cu6 3 ( n,ca )

Np 2 3 7 ( nf )

Co 5 8

Co6 0

Cs137

U2 3 8 ( n,f ) Cs 1 37

ALSO THERMAL FLUX MONITORS -BARE AND CADMIUM - SHIELDED

CO-AL

53

Table 4

WESTINGHOUSE NUCLEAR EUROPESURVEILLANCE 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

54

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55

FPgure 1

CHARPY V-NOTCH IMPACT SPECIMEN

r^ -0.009

8 ° - .396 -90/ -0 0.393

., / t WS. 3

r+8 6 00.3950.393

+ . . .

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56

Figure 2

TENSILE SPECIMENi 1. !.

1.0050.995

0.251 DIA

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GAGE LENGTH

DIA "B"-- 0.395

10.393I

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1.250 REDUCED

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NOTE :

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

57

k'Flure 3

WEDGE OPENING LOADING SPECIMEN ( WOL )

06,00-20 THD CL 3 B0.375 DEEP

58

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Figure 5

REACTOR VESSEL SHOWING ARRANGEMENT OFSURVEILLANCE CAPSULES IN THE VESSEL

(LEAD FACTORS FOR THE CAPSULESARE SHOWN IN PARENTHESES)

270 oR (2.5)_ I REACTOR VESSEL

THERMAL SHIELDN (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 appli-

cation 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 ex-

perience, 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 be-

haviour 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 re-

placed 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 with-

drawal 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 pre-

cracked 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 impli-

cations of flaws, detected by inservice inspection.

- In order to enable the measurement cf some elastic-plastic

fracture parameters, the WOL specimens, provided in the sur-

veillance 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

b-

ciI-

fixedenergylevel

(RKIR

AKIR T

unirrodiated-90~

00

�y 0

irradiated

15P11Raffi

ART NDT J Temperature

or ATNDT

fixed operating temperature

unirrodiated

T - RTNDTNDTART NOT

Fig. 1 Estimation of the change of KIR at a giventemperature with n-irradiation, using impactenergy transition curves (schematic).

65

CU

I.-.1 _05

E 20

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.

67

(c.

c

o

-0.oaCL

8

00

0

0

0U.

004-a LLc:

N

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0 \

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68

0 10 withdrowol time in yeors -

ARTNDT IwNDT ptoition of surveillance specimens

&- I.0 10 20 30 40

vessel life in years, time of information -

Fig. 5 Projected increase in transition temperature for vesselwall and position of surveillance specimens, showingaccelerated information.

(schematic, based on ASME sec 71, Fig. A - 4400-1)

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).

Results show that after irradiation upto 8.5 x 1020 n/cm2 (> 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 vesselmaterial (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/cm2 (> 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 powerper unit

1962 MOL (BR3) 11

1975 Doel I 393

1975 Tihange I 870*

1975 Doel II 3931979 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

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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

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

77

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78

"A UTILITY REVIEW OF IRRADIATION SURVEILLANCEPROGRAMS AND INDUSTRY RESPONSIBILITIES"

Thomas D. Keenan PEManager

Plant Engineering DepartmentYankee Atomic Electric Company

Nuclear Services DivisionWestboro, Massachusetts USA

ABSTRACT

Yankee Atomic Electric Company owns and operates one of theworld's first nuclear power plants. In addition, its Nuclear ServicesDivision has behind it many years of experience in supervisingconstruction, operation, and maintenance of four nuclear power plants,three pressurized light water and one boiling water reactor. As amember of this company, the author has had extensive experience indealing with the major nuclear steam system suppliers, architectengineering firms, and regulatory agencies. Based on this extensiveand varied experience, he has taken a critical view of the manner inwhich the nuclear industry deals with the irradiation surveillance programsand its problems, as seen from the nuclear utility side.

In his paper, the author reviews the safety implications inherentin the subject and discusses many of the present and future technicalissues, as well as the economic considerations and future implicationswhich affect utility decisions in this area.

Finally, the author provides critical commentary on the attitudes ofnuclear utilities and the nuclear plant vendors,'which must change, in hisview, 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 consolidatedapproach 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 embrittle-

ment 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 represents1.46 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

for reactor coolant

mo opera- 13 01

0 00

o - I 0 0

100 20 300 400 500 600

.ndicted Reactr Coon Temperature F u

Relationship for Steady - State Operation

85

4- -/, ,- ~

ru pump operation

0 100 200 300 400 500 600

Indicated Reactor Coolant Temperature ~F )

Fig. 1: Yankee Rowe Indicated Pressure TemperatureRelationship 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 surveillanceprograms

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 surveil-

lance program for the A-1 pressure vessel, the following one para-

meter 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 sur-

veillance specimens programs :

I. Materials

/a/ Base material - all heats used in the pressure vessel near

active core, at least receiving neutron doses higher than2? -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 maxi-

mum 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 speci-

mens 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 pre-

pared 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 app-

ropriate 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 RTNDTNDT

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 pro-

perties (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 con-

sidered 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

Li 1-4 -^--_ t<^-

'J~~H ~ ~1 __L__, SheEu-i;.l cn -ci alni uii U ullcn l Il,:fhcr RT',ith Lo.r RT ,F; wit

,Cum. u ( Cu&I'Z » < 1' Ht1hrCCku&Ph Cu% & uP1`r

rrquLrimicnt I |Ionr qui iiln; I ;ll n [N>e N incqu; nt(

PJ 1 P% C I _u.

Ul:llt c al t Mc1 U, ¢ s l';ti.'in |Ul t N.^illl t" .s , *RTt....with I1 ol.cst . i l nhloh Ilt. t l h II itl lfthcsl | i;hc l RT >.1r dC.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',\:1 uvpcratlng 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 appli-

cable, the same heat or filler material, and the same production welding con-

ditions 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 rep-

resentative 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

,[.1 d $ 1-FZ .-- - - - 1 - _ -

CL of WeldI IG. I, irtcimrn Orinlati in 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 approx-

imately 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 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 modi-

fied 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 ref-

erence 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 with-

drawal

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 speci-

mens 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 predis-

position 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 ge-

neral 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 Gto 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 i renments

Peculiar in the Federal Republic of Germany is the dis-

cussion 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 sur-

face 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 calan-

dria tubes wall appears to be the limiting factor of heavy water

calandria service life and therefore the evaluation of wall thick-

ness 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 fundamen-

tal data for calculation of the instantaneous corrosion rate,

idr = 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 instan-

taneous 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, rep-

resents the wall thickness corrosion decrease for the case of hy-

drodynamically 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 measur-

ing 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 pos-

sibility of repetition of wall thickness decrease evaluation, re-

placement 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 re-

actor 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 sig-

nal 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 measure-

ments without refilling or renewing of coupling agent.

The accuracy of measurements was verified on the models of ca-

landria tubes of graded thickness.

2.2.2 Experimental results

Measurements were realized on eight selected tubes in two re-

gions. 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 pur-

pose 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). Inthe table are also written down the mean distance (h) between con-

tainer and the upper surface of dibiding plate of heavy water calan-

dria"'and estimated neutron dose (0 t) absorbed by surveillance spe-

cimens 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 re-

moved 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 reinforce-

ment 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 (ap-

proximately at 20o at 0,2 yield strength and at 10% at ulti-

mate 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 spe-

cimens from containers number 5A7, 5All and 5A10 with high

neutron doses and highest corrosion losses the fracture cha-

racter is already totally smooth and slanting. Predominant

part of fracture surface of all specimens has pitting struc-

turet characterizing intragranular plastic failure.

Increase of strength properties influenced by reactor opera-

tion 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 ul-

timate tensile strength).

Uniform elongation at higher testing temperatures decreases

(up to 50% of original value at 230°C test); total elonga-

tion 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 approxi-

mately 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 tempera-

ture (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 speci-

mens (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 ob-

jectivity of ultrasonic method, used for wall thickness measure-

ments.

(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).

121

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Scope and results of the Reactor Vessel Radiation Surveillance Program of the

Nuclear Power Plant Beznau I

E. Sandona and P.'P1Uss

Nordostschweizerische Kraftwerke AG., Switzerland

Nuclear Power Plant Beznau

ABSTRACT

The Nuclear Power Plant Beznau I owned and operated by the Nordostschweizerische

Kraftwerke AG (NOK) is equiped with a NSS-System supplied by Westinghouse. First

criticality was reached end of 1969. SFAC, Le Creusot/France manufactured the-

entire Reactor Vessel out of Type 1,2 MD07 Steel, about equivalent to SA 508,

C1. 2 and SA 533, Gr. B qualities.

The Reactor Vessel Radiation Surveillance Program was elaborated by Westinghouse

and approved by the Swiss Authority ASK. Purpose and scope of this program is to

obtain information on the effect of radiation on the Reactor Vessel material under

actual operating conditions. Pre-Irradiation Testing was also performed by the

NSSS supplier. The ductile-to-brittle transition temperature increase due to

radiation can be monitored by a surveillance program which consists of periodically

checking irratiated Charpy V-notch impact specimens. The results will allow to

establish new pressure-temperature limits of the vessel during startup and cooldown

of the plant. In addition to the transition temperature approach, a fracture

mechanics approach utilizing WOL-specimens is used to evaluate the effects of

radiation on the fracture toughness of the material. Furthermore the test capsules

contain tensile specimens to determine the mechanical properties as well as dosimeters

used to measure the integrated flux at specific neutron energy levels and low melting

point eutectic alloy thermal monitors. Test specimens were machined from the vessel

shell courses adjacent to the core region, weld metal and heat affected zone metal.

In addition, correlation monitors made from SA 302, Gr. B material obtained through

Subcommittee II of ASTM Committee E 10 on Radioisotopes and Radiation Effects are

inserted in the capsules.

A total of originally six test capsules are located in the reactor between the

thermal shield and the vessel wall. These capsules have to be removed from the

reactor during normal refueling periods. It is intended by the plant operator

125

to modify the originalremoval schedule as proposed by Westinghouse in order

to combine the surveillance programs of the Beznau I and the Beznau II plant.

The latter is a sister plant of Beznau I.

Two capsules have been removed up to now and were transferred to a post-irradiation

test facility for disassembly and testing. The Swiss Federal Institute of Reactor

Research has been commissioned to do the work. The integrated flux (t 1 MeV) was

found to be 2,77 x 1018 nvt after 413 days (end of 1st core cycle) and 5,72 x 1018

nvt after 971 days (end of 3rd core cycle). The largest shift of the NDT-Temperatur

occured in the 1,2 1DD07 and SA 302, Gr. B base metal, namely 60°C after 5,72 x

108 nvt. Both test performance and results will be discussed in the full length

paper to be presented during the meeting.

The Nuclear Power Plant Beznau I is owned and operated by the Nordostschweizerische

Kraftwerke AG (NOK). It is a so called second generation Pressurized Water Reactor

(PWR) Plant with a net electrical output of 350 MW. First criticality was reached

end of 1969. The Nuclear Steam Supply System (NSSS) has been designed and supplied

by Westinghouse. The system operates at 155 bar and 572 OF (300°C) average Tempera-

ture.

The consortium Brown Boveri-Westinghouse, responsible for the turnkey contract,

subcontracted the Societe des Forges et Ateliers du Creusot (SFAC), le Creusot/

France to manufacture the entire reactor vessel, while the reactor internals have

been supplied from the United States. A schematic view of the reactor vessel is

shown in Fig. 1.1. The total weight (vessel and head) is 195 Mp. Other pertinent

measurements are: Inside Dia. (core region) 3337 mm, Wall Thickness (core region)

166 mm, Wall Thickness (flange region) 180 mm. A Manganese-Molybdenum-Nickel steel

in quenched and tempered condition, SFAC type 1,2 MD 07, was selected for the con-

struction of the reactor vessel.

This type of steel is comparable with A508, C1. 3 (forgings) and A533, Gr. C

(steel plates). A comparison in chemical compositions of the 1,2 MD 07 steel

with some typical steels used for reactor vessels is given in Table 1. The

specified minimum values for the mechanical properties of the 1,2 MD 07 steel

are as follows:

126

Tensile strength 549 N/Mrr

'Yield strength 343 N/mm2

Llongation 18 %

Reduction of area 38 %

Cy-Energy 5,2 mkpcm 2 at -12 °C

(average of 3 specimens)

4,2 mkpcm 2 at -12 °C

(each individual spec.)

The Radiation Surveillance Program for the Beznau I reactor vessel [13 was

elaborated by Westinghouse and approved by the Swiss Authority ASK. This program

is based on ASTM-E 185, Surveillance Tests for Nuclear Reactor Vessels. Objective

of the program is to obtain information on the effect of radiation on the reactor

vessel material under actual operating conditions. Pre-irnadiation testing was

performed by the NSSS supplier in order to obtain sufficient base-line informa-

tion.

The ductile-to-brittle transition temperature increase due to radiation can be moni-

tored by periodically checking irradiated Charpy-V-notch impact specimens. Such

results will allow to establish new pressure-temperature limits of the vessel during

start-up and cooldown of the plant..

In addition to the transition temperature approach, a fracture mechanics approach

is used to evaluate the effects of radiation on the fracture toughness of the mate-

rial. Fracture toughness test procedure involves the tension testing of notched

wedge opening loading (WOL) specimens which have been precracked in fatigue. The

KIc value is calculated by an equation which has been established on the basis

of elastic stress analysis of fracture mechanics specimens,

The mechanical properties of the material under irradiation is monitored by

testing tensile specimens.

Neutron dosimeters of Copper, pure Nickel, Aluminium-Cobalt wire, Neptunium-237,

Uranium-238 and Iron will be used to measure the actual neutron environment

along the capsules.

127

Thermal monitors (2.5% Ag, 97.5% Pb, melting point 579 OF -- 1.75 % Ag, 0.75 % Sn,

97.5 % Pb, melting Point 590 OF) sealed i,, pyrex tubes allow to define mote accu-

rately the temperature attained by the test specimens during irradiation.

In order to implement the general scope described above during the lifetime of

the plant six material test capsules are located in the Beznau I reactor between

the thermal shield and the vessel wall. The disposition and the numbering of the

capsules is shown in Fig. 1.2. The test capsules are contained in baskets atta-

ched to the thermal shield (Fig. 1.3). These capsules have to be removed from the

reactor during normal refueling periods. The schedule for removal as recommended

by Westinghouse is as follows:

Capsule: V Removal time: End of 1st core cycle

R End of 2nd core cycle

S 5 years

N 10 years

T Extra tapsule (spare)

P Extra capsule (spare)

It is intended by the plant operator NOK to modify the recommended removal schedule

in order to combine the surveillance programs of the Beznau I and the Beznau II

plant which is a duplicate of unit I. A corresponding proposal will be submitted

to the authorities for review and comment by the end of this year.

The inventory of each capsule is pictured in Table 2. Test specimens as listed

were machined from the vessel shell courses "C" and "D" adjacent to the core

region, weld metal and heat affected zone metal. In addition, correlation monitors

made from A302, Gr. B material obtained through Subcommittee II of ASTM Committee

E 10 on Radioisotopes and Radiation Effects are inserted in the capsules. The

heat treatment condition of the test material is as stated in Table 3.

Capsule V was removed from the core during the refueling shutdown in June 1971

(end of 1st core cycle) after 413 days of irradiation. Neutron flux and fluenlce

for energy levels>l MeV have been imasured using Fe-Dosimeters (Fe54 (n,p) Mn54).

The neutron flux was found to be 7,76x10 /cm s and the fluence 2,77x1018 /cm

These values have been checked using Ni-Dosimeters (Ni58 (n,p) Co58 ) and Cu-Dosi-

meters (Cu63 (n, C) Co60 ). The coincidence was appropriate.

128

In 1974 (end of 3rd core cycle) after 971 days of irradiation capsule R has Deen

withdrawn. The neutron flux and fluence have been again calculated by means of radio-

activation of Iron, Nickel and Copper dosimeters. The obtained values are

6,82x1010 n/cm2s for the flux and 5,72x1018 n/cm2 for the fluence.

The Swiss Federal Institute of Reactor Research has been commissioned for transpor-

tation and disassembly of the capsule and to test all the specimens in their post-

irradiation test facility. The test specifications for these orders have been

worked - out by NOK [2] , [3]

The test results are compiled in three reports [41 , [51 , 6 . A summary

of the findings is given below.

Charpy-V-notch impact tests were performed on specimens from each of the shell

forgings, the weld and heat affected zone metal and the correlation monitor material

to obtain full transition curves. The curves of energy versus temperature are plot-

ted in Fig. 2. The notch ductility of the 1,2 MD 07 base metal is markedly superior

to those of the A302, Gr. B steel, both before and after irradiation. The best

values have been obtained from material machined from the heat affected zone and

the weld metal.

The Charpy-V-notch temperature supposed to correspond to the nil-ductility transition

temperature, NDTT (ASTM-E 203, Conducting Drop-Weight test to determine nil-ductili-

ty transition temperature of Ferritic Steels) is the temperature at 30 ft-lb for

the 1,2 MD 07 steel. This relationship has been listed in the ASME, B & PV Code,

Section III, Nuclear Power Plant Components. However, dropweight tests performed

on unirradiated 1,2 MD 07 material for the Beznau I reactor vessel revealed extreme

deviations.from the presumed relationship. The following values may illustrate

this statement:

Material C Impact Test Drop-weight Test1,2 MD 07 TVat 30 ft-lb NDTT

Shell Course "C" -44 °F (-42 °C) 30 °F (-1 °C)

Shell Course "D" -22 OF (-30 C) 20 oF (-7 °C).. . . . . . . . _j.~~~~~~~~~

129

Based on this siLuation the most conservative approach was taken to determine

the minimum pressurization temperature (beginning of life) for the NSSS.

Tm n = NDTT (Forging "C") + 60 °F = 90 °F (32 °C)

This value will be used over the lifetime of the plant as a reference to establish

new pressure-temperature limits for the reactor vessel.

The tendency in temperature shift &TCv at the Cv

embrittlement is also shown in Fig. 2. The exact

following table.

30 ft-ib level due to radiation

values may be taken from the

MateriaTotal ShiftAT Total Shift To Material Total iftTCv „2 Irr 5,72x028 cv 2

Irr. 2,77x10 /cm2 Irr. 5,72x1018 /cm

Shell Course "C" 72 F (40 °C) 108 F (60 'C)

Shell Course "D" 45 OF (25 °C) 63 °F (35 °C)

Heat aff. zone 45 OF (25 °C) 81 F (45 °C)

Weld metal 36 OF (20 °C) 36 F (20 °C)

A302, Gr. B 45 OF (25 °C) 108 F (60 °C)

A comparison of the neutron embrittlement sensitivity of the Beznau I 1,2 MD 07

steel and two typical Reactor Pressure Vessel Steels is shown in Fig. 4.2.

In addition to the discussed results, fracture appearance (% shear) , lateral

expansion and root notch contraction have been measured.

Tensile tests have been performed at room temperature (RT), approximate opera-

ting temperature of the reactor (297 °C) and an intermediate temperature (150°C).

Ultimate tensile strength, 0.2 % Yield strength, Total elongation (gage length

1 inch) and Reduction in area were measured. The results (RT and 297 °C) are

summarized in Fig. 3. All values are within the expected range.

A fracture toughness test program was performed on unirradiated specimens at

different temperatures. The resulting KIc versus T ( OF, °C) curve for the 1,2 MD 07

steel is compared with A533, Gr. B, C1. 2 material in Fig. 4.1. The fracture

toughness values of the 1,2 MD 07 look favourable to those of the A533 steel.

130

The WOL specimens for each of the irradiated materials have been tested at a

temperature based on the transition temperature shift obtained from the associa-

ted Charpy impact specimens. The selected test temperature was -200 OF (-129 °C)

plus the transition temperature shiftATcv. For the results see Fig. 4.1.

LIST OF REFERENCES******************

I1Aj NOK, Reactor Vessel Radiation Surveillance ProgramWestinghouse Electric Corporation, APDJune 1968

[271 Spezifikation der Nordostschweiz. Kraftwerke AG fir die Abwicklungder Nachbestrahlungsuntersuchunyen an Materialproben der Peaktor-DruckgefasseKapsel V, Rev. 0, 23.4.71

31 Spezifikation der Nordostschweiz. Kraftwerke AG fUr die Abwicklungder Nachbestrahlungsuntersuchungen an Materialproben der Reaktor-DruckgefasseKapsel R, Rev. 2, Sept. i4

|4J Eidg. Institut fUr ReaktorforschungPrUfbericht PB-ME-73/9.Nachbestrahlungsuntersuchungen an NOK-Reaktordruckgefass-Material Beznau I

[51 ~ Eidg. Institut fur ReaktorforschungPrUfbericht PB-ME-75/02Nachbestrahlungsuntersuchungen an NOK-Reaktor-druckgefass-Material Beznau I, Kapsel V.Ermittlung der Neutronenfluenz sowie Lateral-Expansionund Root-Notch-Contraction

36J 1Eidg. Institut fur ReaktorforschungPrUfbericht PP-ME-75/03Nachbestrahlungsuntersuchungen an NOK-Reaktor-druckgefass-Material der Kernkraftwerke Beznau 1/II,Kapsel R

131

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Table 3

HEAT TREATMENT CONDITION************************

1 1.2 MD 07 Surveillance Material

ForgingC:

Forging _D

Heated at 925 0C- 4 hrs -Tempered at 650 C - 4 hrsFive stress reliefs at '50(15 hrs total)Stress relieved at 600°C Stress relieved at 6000C -

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138

BRITTLENESS

Presupposition (criteria) for reactorvessel brittle fracture

E. Bazant, BBR Mannheim

ABSTRACT

Safety approaches and criteria related to reactor pressure vessels

as well as the applicability of fracture mechanics concepts are discussed.

2. Summary

Brittleness, the subject of this report, is the prerequisite

for brittle fracture of a reactor vessel (RV). Brittleness is

not actually a material property; rather it is a tendency to-

wards fracture without showing any plastic deformation. Thereare two kinds of brittleness: one is dependent on the material,

the other one depends on the load conditions.

The report begins with a description of the brittle/ductile

temperature concept, a method to determine the transition

temperature of the impact test.

The section "fracture mechanics" concludes that brittle

fracture of crack-free material is not applicable to the

RV. The fracture analysis diagram, however, is a means todescribebrittle fracture as a problem of crack growth-and

to specify a relation between such variables as crack size,

stress, temperature, and type of failure.

Following an outline of the various brittle fracture test

methods, Porse's fracture criterion is applied to include

brittleness into the RV design. The report finally givesan account of the fracture criteria relevant for RV operation.

139

3. Brittleness, prerequisite for brittle fracture

The RV of a nuclear reactor might have hidden material dis-

continuities. These discontinuities could perhaps result in

brittle fracture. The prerequisite for brittle fracture is

brittleness. Brittleness is not just another material property,

it rather is the tendency towards fracture without noticeable

plastic deformation. D. Radaj in (1) gives, among other things

a brief outline on brittle fracture.

In addition to the material, brittleness does also depend on

the load conditions: consistent with the load conditions the

material will be either ductile (tough) or nonductile (brittle).

The following load conditions promote brittle fracture:

- the spacial stress condition

- the stress level

- the dimensions of the area where these stresses

are acting

The three-directional stress condition occurs preferably in

front of the crack tip subsequent to a local plastic defor-

mation along with material contraction normal to the plate

surface when the plate is subjected to uni-directional tension.

The three-directional stress condition does not necessarily

imply the presence of a crack, as in case of welding-induced

residual stress. If subjected to similar material conditions,

large components will show a higher degree of brittleness

than smaller samples. A reference to this geometric influence

on brittleness is made by M.E. Shank in (2). He conducted a

series of charpy tests, using test specimens of varying size

but of identical metallurgical properties and of similar

geometrical properties (incl. the notch). At a certain tem-

perature and loading velocity the largest specimens' were

completely brittle whilst the smallest specimens remained

entirely ductile.

In case of ferrits, brittleness increases as the load velo-

city increases (impact embrittlement). A sudden drop of

ductility can be observed when the temperature is lowered

(temperature embrittlement).

140

3.1 The brittle/ductile temperature concept

For ferritic (body-centered cubic-b.c.c.) steel the

Charpy-notch-impact test-data-temperature curve is as

outlined in diagram Fig. 516. This curve gives a good

representation of the sudden drop in ductility. Fractures

occuring in the upper shelf region are characterized by a

large percentage of plastically deformed fracture appereance

(shear or fibrous fractures). This percentage decreases

in the transitional temperature range (where test results

vary widely) (mixed fractures) and is eliminated altogether

in the lower shelf region. There, fractures are practically

100-percent pure brittle fractures (cleavage fractures).

Given a sufficient number of plotted test points it is

possible (according to K. Heckel (3)) to establish a mean

curve in the transitional temperature range and to determine

a characteristic temperature by defining that temperature

as transitional temperature at which the Charpy-notched-

impact-value falls below a certain preselected value Tu.

There are some other criteria for determining the transi-

tional temperature (see Radaj (1)), some of which are based

on conventions rather than physical criteria (see Fig.. 517).

Material-induced brittleness depends on the material's

chemistry and its structure; structural changes during

stress relief treatments for instance may cause stress-

relief embrittleument.

3.2 The. technical fracture mechanics

Technical fracture mechanics deals with the individual

phases of a technical fracture, i.e. crack initiation,

crack propagation, and crack arrest on the basis of

continuum mechanics. The objective is to verify that

test results of small samples are also applicable to

the large structure.

The approach of technical fracture mechanics is a

macroscopic one: the material is regarded as a continuumwith homogeneous and isotropic behavior also in the

141

micro area. Crack propagation is regarded as a continuous

progress.

The macroscopic approach of technical fracture mechanics

concentrates on the technically most essential aspect of

the fracture process. The microscopic approach of metal

physics ona the other hand deals with that part of the

fracture process which is characterized by discontinuous

material separation in the inhomogenous and anisotropic

range of cristallites.

According to Radaj, for instance, continuum fracture

mechanics is the fracture mechanics of continuo which

are entirely free from cracks or notches. Continuum

fracture mechanics includes the yielding, fracture and

failure accumulation hypothesis. How essential continuum

fracture mechanics is may be gathered from the fact that

brittle fracture can'occur with materials that are free

from any cracks; i.e crack fracture mechanics and notch

fracture mechanics alone do not sufficiently cover the

whole spectrum of the brittle fracture problem.

The mechanical properties of materials which are unaffected

by cracks, i.e. yield stress, tensile strength, and true

stress at fracture, are dependent on temperature. Fig. 519

shows these forcing.functions. Crack-free materials which

are subjected to tensile testing at temperatures below a

very low nil ductility transition temperature (no flaws)

(<-200°C) (4) will show material separation (cleavage

fracture) at the yielding point. The stress at fracture

of crack-free material (true stress at fracture) at this

nil ductility transition temperature (no flaws) equals

the yield point. Below this nil ductility transitions

temperature (no flaws) the yield stress is identical with

with the tensile strength, indentical with the true stress

at fracture, and the elongation is zero (from the macroscopic

point of view).

Brittle fracture occurs with crack-free material if the

true stress at fracture falls below the yield point. Since,

142

however, for crack-free material at temperatures above

this nil ductility transition temperature (no flaws) yield

stress is lower than the fracture stress without flaw

the material will be deformed before it breaks.

For a reactor vessel, brittle fracture of crack-free

material is not relevant, as the RV is operated at tempe-

ratures far above this nil ductility transition temperature

(no flaws). For mild steel this nil ductility transition

temperature (no flaws) takes place at about -156°C (acc.

to (5)).

3.3 The fracture analysis diagram (FAD)

Since 1950 brittle fracture tests have been conducted in

the U.S.A. by Pellini who subjected plates to transverse

bending stresses and used over-size Charpy-notch test

specimens ("type specimens"). In these studies the con-

ditions of crack propagation and arrest were analysed

(assuming cracks of various sizes in either materials

under elastic stresses or plastically deformed materials

at varying temperatures) which were then identified by

means of critical stresses. From these, the maximum tem-

peratures were derived which characterize the respective

different kinds of crack growth. The brittle fracture

problem is regarded as a crack expansion problem.

For the first time an approximate, quantitative-relation

was specified - known as Pellini's and Puzak's fracture

analysis diagram (FAD) (Fig. 518 and 514), which indicates

the forcing functions between crack size, critical

stress, temperature, and type of failure (e.g. brittle).

Later, Porse included brittleness into the design.

The nil-ductility transition temperature (NDT-T) in the

brittle fracture diagram is defined as the highest tem-

perature at which a small flaw in that part of a plate

which is subjected to elastical stresses close to the

yield limit could result in brittle fracture.

143

The highest temperature at which a large crack in elastically

stressed parts of the plate could result in brittle fracture

is called Fracture Transition Elastic Temperature (FTE-

Temperature). In more general terms, the FTE-temperature

is defined as the temperature below which a brittle

fracture will run through parts subjected to elastical

stresses.

Point K (in Fig. 518) indicates the intersection of the

CAT and tensile strength curves, i.e. in terms of ability

to withstand loads the material behaves as if it were free

from cracks. At this point K the so-calles Fracture

Transition Plastic (FTP) temperature is reached which

is defined as the temperature at which the first brittle

fracture symptoms occur and above which only shear

fractures are possible (5).

For mild steel the temperature range between IDT and FTP

temperatures (i.e. the range where sudden growth of a

minor crack is possible up to the yield point whilst a

major crack cannot continue to grow unsteadily at tensile-

strength stress levels) is about 66°C.

Experience shows that the FTE-temperature tends to be

60F (33°C) higher than the NDT-temperature. Water-cooled

reactors are never subjected to temperatures below the

FTE-temperature.

The unsteady (spontaneous) rupture of elastically loaded

parts of the reactor vessel with its disastrous conse-

quences (i.e. crack growth rates of some thousend feet

per sec.) is not possible at temperatures above FTE. The

FAD is applicable to carbon and low-allow steels which

feature a rather sudden transition from nil-ductility to

ductility within a narrow temperature range above NDT-

temperature and whose thickness is less than 76mm.

144

3.3.1 The influence of wall thickness

The influence of wall thickness is of mechanical nature only

and has no metallurgical reasons. Wall thickness effects

are taken into account by the FAD which specifies a 70°F

increase of FTE temperature for wall thickness ranging

between 6 to 12 in..

This means that the FTE temperature as in Fig. 522 is now

exceeding the NDT temperature by 60F + 70F = 130F. Fractures

occuring in thick parts (> 3 in.) subjected to temperatures

about NDT + 200F will be of the ductile type and the uppershelf of- the transition curve will be reached.

3.4 Brittle fracture test methods

Originally, NDT, FTE and FTP temperatures had been deter-

mined on the basis of the Explosion Bulge Test (EBT) (6).

However, the EBT acc. to (1) does only give rather rough

approximates of these temperatures.

An improved variation of the EBT is the Drop Weight Test

(DWT). Although the DWT can determine the NDT temperature

only, it is advantageous in so far as it can be-used to

test both non-welded materials or materials from the heat

affected zone and welded materials.

The DWT defines the NDT temperature as the lowest temperature

at which the plate material is able to inhibit the growth

of a small crack.

According to Pellini and Puzal (in (7) and (8)) the NDT-T

is the temperature below which the steel will not be de-

formed before it breaks.

In (1), Radaj points out that the NDT, FTE and FTP tem-

peratures have been thoroughly compared to the ISO-V

transition temperature (9) and the DVM notched-impact

test (10).

According to (1), the Explosion Tear Test (ETT) and the

Drop Weight Tear Test (DWTT) are test methods which Pellini

et al. have developed for materials with little or no

145

temperature embrittlement. Acc. to (1), temperature

embrittlement occurs when a decrease of temperature

-causes a sudden drop of ductility. ETT and DWTT are

also.applicable for thick-walled materials whose thoughness

varies across the thickness. Another feature of these

tests is the fact that they can be used to determine

the lower region of the FAD's CAT curve of materials

which are prone to temperature embrittlement.

The Crack-Arrest Temperature Curve (CAT curve) defines

the temperatures at which long cracks can still be ar-

rested in a material subjected to various elastical

stresses. In other words, the CAT curve indicates the

limit of acceptable stresses below FTE temperature.

To establish a FAD it is necessary, acc. to (1), to

conduct conventional tensile tests plus ETB and DWT

or ETT and DWTT. The most reliable approach to determine

the CAT curve is the Robertson-Test (11).

In any case, the FAD can only furnish a rough baseline

for the assessment of a structure's brittle fracture

risk.

3.5 Porse's fracture criterion

By means of Porse's fracture criterion brittleness can be

taken into account in the reactor vessel design. In (12) this

criterion is determined by the design transition temperature

(DT-T) which is specified either as FTE temperature of

as NDT-T + 130 F for thick-walled components. Here, the

NDT-T depends on the temperature resulting from either

drop weight test or notched-impact test - whichever

temperature is higher. Test criterion is 30 ft-lb (= 51 J/cm2

= 41 J per ISO-V test specimen = 5,2 mkp/cm2 ).

Porse's design and operating limitations for the RV are

as follows:

146

- Above the design transition temperature (DT-T) the

allowable combined thermal stress and interior-pressure-

induced stresses equal the yield limit of the material

that is being used in the core area (base material of

forgings, material of the weld area and the heat

affected zone).

- At DT-T the allowable stress is 20% of the yield point.

This limit is down to 10% of the yield point for the

temperature: DT-T -200F; the percentage remains un-

changed for lower temperatures (Fig. 521).

A safety margin exists in so far as allowed stresses are

markedly below the material's yield point for temperatures

which fall below the DT-T (FTE).

Since knowledge on crack growth was rather limited at the

time, one proceeded from the assumption of residual stresses

and a stress concentration factor of 4 at the crack tip. In

the American Welding Handbook (13) residual stresses are

assumed to be 20% of the stress at yield point. This stress

bairer as defined by Porse is based on studies conducted

by Robertson on crack growth as well as on studies by

Kihara and Masubushi regarding the effect of residual

stress on brittle fracture (14).

3.6 The fracture criterion during operation

In the course of RV life the mechanical properties of the

materials in the RV's belt line area are subject to changes

due to neutron embrittlement. The changes are indicated by:

- increase of the ductile/brittle transition temperature,

- decrease of the work energy absorbed by the notched-impact

test specimen, (decrease of upper shelf energy),

- increase of the yield limit.

147

The actual properties of the materials located in the belt

line area are determined by means of the surveillance program.

In this program complete sets of Charpy-V-notch specimens

and tensile test specimens are subjected to the neutron flux

of normal operating conditions.

As drop weight test specimens are too large to be used in the

surveillance program, the increase of the transition tempera-

ture ATT is determined by means of Charpy-V-notch test onthe basis of the 30 ft-lb criterion.

This means that the Design Transition Temperature (DT-T) for

the irradiated condition equals the DT-T of the unirradiated

condition plus A TT.

The temperature change is shown in Porse's diagram (Fig. 520).

The result is an allowable corridor and a qualified caution

zone in the stress/temperature range below DT-T.

3.6.1 Inclusion of material toughness

Linear elastic fracture mechanics only can fully include

the material toughness into the design calculation. It com-

bines stress and defect size by specifying equations for

stress intensity factor at the end of a sharp.crack.

Here, the fracture toughness (= critical stress intensity

factor) is introduced as a material characteristic. It is

determined in either statical (KIC '^ crack initiation,

KIa- crack arrest) or dynamical (Kid impact strength)

tests,

KiC (Fig. 515) is markedly higher than Kiaand Kid which is

why only the later two provide the basis for the reference

curve KIR of the steels SA-533 B-I, SA-508-2, and SA-508-3

(Rig. 523) (15).

By relating the measured KIa and KId values to the NDT

temperature one can establish an envelope curve which com-

prises all known variables (Fig. 523).148

The NDT-T will become the reference temperature RTNDT acc.to definition if the impact value of the same material isat least 67 J per ISO-V specimen (8,5 mkp/cm2) and a

lateral expansion of ._ 0.9 mm when subjected to a tempera-

ture 33°C above NDT-T.

From this together with the shift of brittle/ductile transi-

tion temperature that has been observed with irradiated

Charpy test specimens, one can derive the neutron irradiation-

adjusted reference temperature.

3.6.2 Neutron Irradiation-induced changes of material

properties according to Palme

Certain mechanical properties of reactor vessel steels willchange when these steels are subjected to neutron irradiation.

A particularly obvious change is the increase of the transi-

tion temperature of the Charpy-V test temperature curve (16).

Quantitative relations are shown in Fig. 1 where the AT in-crease of the transition temperature is plotted to the

neutron flux. The parameter for the group of curves is the

Cu contents (the P contents has been taken into account).

The table at the bottom of Fig. 1 specifies the maximumAT increase that has to be expected for certain Cu con-tents at preselected neutron flux ratios.

3.6.3 Combination of the reference temperature concept with

linear elastical fracture mechanics

The shift of temperatures in Fig. 2 and 3 serves to deter-mine the change of the KIR curves (reference fracture tough-ness curves) as compared to the unirradiated condition

(On/cm2 ).

Consistent with ASME Code, Section III, Appendix G (17)provisions shall be made to prevent brittle fractures ofthe RV.

149

The curve of the reference stress intensity factor (reference

fracture toughness) plotted to the temperature is the lower

limit curve of all known KIo, KId and Kia values of the

materials A 553 Gr. B Cl. 1 and of A-508 steels. Depending

on neutron flux and the steel's Cu and P contents the lower

limiting curve of the reference stress intensity factor

will be shifted to higher temperatures. The basis for this

are measurements taken at irradiated Charpy-V specimens.

Figs. 2 and 3 show clearly that the calculated stress inten-

sity factors are always less than the reference fracture

toughness values if the neutron flux is, for instance,

5 x 1019n/cm2 > 1MeV and if the contents of both Cu and

P is more or less limited. Combining the reference tempe-

rature concept with linear elastical fracture mechanics

is a safeguard that operation as well as startup and shut-

down of a reactor will be performed within a range that

ensures sufficient material toughness even in the presence

of a defect whose depth is 1/4 of wall thickness and whose

length is 1,5 the wall thickness. In addition, there is

a sufficient safety margin be ween said range and the

brittle-fracture-risk buffer zone which (i.e. the buffer

zone) is growing due to neutron irradiation during RV

life.

(1) Radaj D., Festigkeitsnachweise, Tell I,Grundverfahren, DVS, DUsseldorf p.. 100/1,p. 108/9, p. 110, p. 94Tell II Sonderverfahren p. 172

(2) M.E.Shank, Control of Steel Construction toavoid Brittle FailureWRC 1957, p. 11, 28

(3) K.Heckel, EinfUhrung in die technische Anwendungder Bruchmechanik,p. 13 Carl Hauser Verlag, MUnchen 1970

(4) Pellini, W.S., Evolution of engineering principlesfor fracture-safe design of steel structures,p. 5 NRL Report 6957

(5) Tetelman, McEvily, Bruchverhalten technischerWerkstoffe, p. 82, p. 211,Verlag Stahleisen 1971, DUsseldorf.

150

(6) Pellini, Goode, Puzak, Lange, Huber

Review of Concepts and Status of Procedures forFracture Safe Design of Complex Weld StructuresInvolving Metals of low to Ultra-High StrengthLevelsUS-NRL 6300 p. 1/84, 1965

(7) Puzak, Pellini

Evaluation of the Significance of Charpy-V-Testsfor Ouenched and Tempered SteelsAWS Journal, 1956, p. 275/297

(8) Fracture Analysis Diagramm for the Fracture SafeEngineering Design of Steel StructuresUS-NRL 5920, 1963

(9) ASTM E 23-64Notched Bar Impact Testing of Metallic Materials

(10) DIN 50 115 E (1970) Kerbrchlagb)iegevcrsuch

(11) Robertson T.S

Brittle Fracture of Mild Steel Engineering172 (1951) Oct. 5, p. 445/48

(12) Porse L.

Reactor Ve;:el Desigr consJdfe' Jnf R ad ;tJon EffectsJournal of B3asic Enigineering., D)ec. 1964

(13) Welding Handbook, Section I, American Welding Society,1957, p. 5.2/4.

(14) Klhara H., Masubushi K.

Effect of Residual Stress on Brittle FractureWelding Journal, Vol. 38, April 1959, WeldlingResearch Supplement, p. 159.

(15) PVRC-Bull. No. 175, August 1972

(16) H.S. Palme

Radiation embrittlement sensitivity of reactorpressure Vessel steels BAW-10056, March 1973

(17) ASME, Section III 1974, Division 1 - Subsection NA,Nonmandatory Appendix G

151

It;IL upper shelf region

c 10-_ _- -_ shear or fibrousov /1 <fractures

0. I^^^ / I I

I8 . J_-- I1

H6 C---- transition rangeCO _ l § mixed fractures

0

h 2.----- -- lower shelf region* -I 1 I' "cleavage fratures

ax ___ _u0 I

'._. ,I,_ _

-60 -4.0 -20 ±0 +20 +40 -+60+CO CTa

test temperature

Fig. 1 Charpy-V-notch-data as a fA:ct[ion of tempera-ture for ferritic steels (di-_gram) as perHeckel 19-70

TO for akmax/2 withakmax Charpy- V notch upper sheTf--value

Ti for ak = 35, 50 or. 70 mN/cm2 ,

T for 50/50 distribiitiori of'cistalline and dull fracturesurfaces (mixed-fracture temperature)

Tijfo0 100% dull fracture surface

Tufor alc = (ako + akiloo)/2 with!ko and kt100 being the Charpy-V notchdata atO ,and, 100% dull fracture surface

I Fig. 2 Criteria to determine the transition tempera-ture TU according to Radaj, 1974

152

NRL REPORT 6957

bTI~S 1 D00UCTILITY

50 - M

LOWER YIELD FULL

Q ,^ ^'',^-"; ''--'.':;'-;-- ;SMOOTH BODY CRACKED BODY) \^ ,' .-; ;TRPANSITION BRITTL FRAACTURE

'"':

.I :,.V r;';-'->i»- PINITRANSITION

TEMPERATURE BRITTLE

Comparison of transition temperature ranges defined by tensile and dynamic

fracture tests for a typical structural mild steel, The highest possible transition

temperature range is established by increases in dynamic fracture toughness whichpreclude, the development of unstable fracture. All transitions related to flaw size

,t .V'S^^V.W^-'-i,^ ~:~ · LIMIT

-400 -350 -300 -250 -200 -150 -100 -50 0 50 100 150 200 (*F)11 I I . . t I. , , ..I. . .1

-250 -200 -150 -100 -50 0 50 100 CC)TEMPERATURE

Comparison of transition temperature ranges defined by tensile and dynamicfracture tests for a typical structural mild steel, The highest possible transitiontemperature range is established by increases in dynamic fracture toughness whichpreclude, the development of unstable fracture. All transitions related to flaw sizeand loading rate aspects must be below this limiting transition temperature range.

Fig. 3 Temperature innfluence on mechanical propertiesof weldable structural steel as per Pellini,1969

153

Q(free from cracks)

II~~ 0 oF(small crack| )J/ !J

I (large cra.e ' /

lower limit of a

TD 's T'D ('NDT) (F'E) (-TP)

T----

rig. 4 Influence of ,temperature T on yield point (y ten-sile sterength6,and true st.ess at fracture eofcrack-free mnaterial. Furthermore, true sLtr.ess atTracture 64 are specified for materials withcracks of various lengths. Acc. to TetelmannMcEvily, 1971

U.S. NAVAL RESEARCH LABORATORY

TENSILESTRESS

t YIELDSTRESS0)W:

 PROPAGATE

TEMPERATURE LIMITATION)

I l iI I i nL v ---- -- ------NDT NOT + 30F

TEMP. -NDT + 60°F NDT + 1200F

7Fig. 5 Simplified fracture anaLysis diagram for NDT-T,according to Pellini and Puzak, 1963 (as perASTM E 208)i

,NDS 6 514756

154

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ANALYSIS OF MECHANICAL PROPERTY DATA OBTAINED FROMNUCLEAR PRESSURE VESSEL SURVEILLANCE CAPSULES

J. S. Perrin

Battelle Columbus DivisionBattelle Memorial Institute

Columbus, OhioUSA

ABSTRACT

A typical pressure vessel surveillance capsule examination program

provides mechanical property data from tensile, Charpy V-notch impact, and,

in some cases, fracture mechanics specimens. This data must be analyzed in

conjunction with the unirradiated baseline mechanical property data to

determine the effect of irradiation on the mechanical properties. In the

case of Charpy impact specimens, for example, irradiation typically causes

an increase in the transition temperature and a decrease in the upper shelf

energy level.

The results of the sharpy impact and other mechanical specimen

tests must be evaluated to determine if property changes are occurring in

the manner expected when the reactor was put into service. The large amount

of data'obtained from surveillance capsule examinations in recent years

enables one to make fairly good predictions.

After the changes in the mechanical properties of specimens from

a particular surveillance capsule have been experimentally determined and

evaluated, they must be related to the reactor pressure vessel. This requires

a knowledge of the neutron fluence of the surveillance capsule, and the ratio

of the surveillance capsule fluence to the pressure vessel wall fluence * This

ratio is frequently specified by.the reactor manufacturer, or can be calcu-

lated from a knowledge of the geometry and materials of the reactor components

inside the pressure vessel. A knowledge of the exact neutron fluence of

the capsule specimens and the capsule to vessel wall neutron fluence ratio

is of great importance, since inaccuracies in these numbers cause just as

serious a problem as inaccuracies in the mechanical property determinations.

A further area causing analysis difficulties is problems encountered

in recent capsule programs relating to capsule design, construction, operation,

and dismantling.

INIRODUCTION

Irradiation of a nuclear reactor pressure vessel results in

changes in the mechanical properties of the pressure vessel steel. As

163

the fluence received by the pressure vessel increases during plant life-

time, the mechanical properties continue to change. As a result, the reactor

pressure temperature heatup and cooldown curves for normal and hydrotest

operation must be periodically changed to take into account mechanical

property changes.

Commercial nuclear power plants in the United States each contain

a series of surveillance capsules. A capsule typically contains neutron

dosimeters, thermal monitors, tensile specimens, and Charpy V-notch impact

specimens. Some capsules also contain fracture mechanics specimens of either

compact tension or wedge opening loading design. The mechanical property

specimens are machined from actual pressure vessel material. Surveillance

capsules are periodically removed from a reactor and examined to determine

property changes resulting from irradiation.

There are various standards and regulations which must be followed

during examination and analysis of surveillance capsules in U.S. reactors.

Appendix G, "Fracture Toughness Requirements", and Appendix H, "Reactor Vessel

Material Surveillance Program Requirements", to 10 CFR Part 50, "Licensing

of Production and Utilization Facilities" describe U.S. Nuclear Regulatory

Commission requirements. USNRC Regulatory Guide 1.99 "Effects of Residual

Elements on Predicted Radiation Damage to Reactor Vessel Materials" supplements

Appendix G of 10 CFR Part 50. These USNRC documents reference other documents

which must be followed. These other documents include Section III, "Rules for

the Construction of Nuclear Power Reactor Components" of the American Society

of Mechanical Engineers Boiler and Pressure Vessel Code, and numerous standards

prepared by the American Society for Testing and Materials.

In this paper, the analysis of the data obtained from surveillance

capsules will be discussed. The discussion will include analysis of the Charpy

impact data. The manner in which the surveillance capsule Charpy impact data

is related to the actual pressure vessel will also be presented. In addition,

problems experienced in a number of recent capsule programs relating to capsule

removal, capsule disassembly, and specimen preparation will be reviewed.

ANALYSIS OF CHARPY IMPACT DATA

Figure 1 shows a Charpy impact curve for a weld metal material

before irradiation and after irradiation at 550 F to a fluence of

2.5 x 1018 neutrons/cm (E >1 MeV). The weld metal material is from a weld

region between pressure vessel sections composed of SA 533 Class B. The

complete curve has been shifted to higher temperatures. The 50 ft-lb

reference temperature (RTiDT) of the weld metal increased from 50 F to 300 F,

164

an increase of 250 F. In addition, the upper shelf energy level dropped

substantially, from an unirradiated value of approximately 70 ft-lb to an

irradiated value of approximately 53 ft-lb, a drop of 17 ft-lb.

The two major items of information obtained from irradiated Charpy

impact specimens are thus the RTNDT and the drop in upper shelf energy level.

The RTNDT values of the reactor materials are needed for use in determining

revised pressure-temperature operating curves. The upper shelf energy level

can be of appreciable significance, because if it drops below 50 ft-lb, the

RTNDT obviously cannot be determined.

In the analysis of Charpy impact data from a particular capsule,

the results can be compared to data obtained from other surveillance capsule

programs. Figure 2 is such a comparison. This figure is a plot of 30 ft-lb

transition temperatures for base, weld, and heat-affected zone metal from a

number of plants, and is similar to a 50 ft-lb transition temperature plot.

The trend band drawn has a very large (>100 F) range for fluences above

3 x 1018 neutrons/cm2 (E >1 MeV). Note that the weld metal transition

temperature increases tend to be greater than the base metal or heat-

affected zone metal specimens. It can also be seen that the most rapid

increase in transition temperature occurs in the lower fluence range.

Figure 3 is a plot of Charpy impact energy as a function of

temperature for heat-affected-zone material from the same surveillance

capsule shown in Figure 1. Figure 3 is included to show a difficulty

experienced in many capsule examinations. The number of irradiated specimens

of a given material may be as low as eight, as was the case in Figure 3. It

is necessary to pick test temperatures such that both the transition tempera-

ture range and the upper shelf are well defined.

Once the specimens have been tested, the next problem is to draw

the Charpy curve through the data in order to determine the RKTDT and the

upper shelf energy, the latter being of special interest if it is approaching

50 ft-lb. In Figure 3 a dashed line has been drawn through the eight

irradiated data points; it is only a line suggesting a possible curve. The

scatter among the eight points is too great to define a meaningful curve.

An ASIM task group is currently considering the question of how to best

determine Charpy impact information from a limited number of specimens.

RELATION OF CAPSULE PROPERTIES TOVESSEL WALL PROPERTIES

After the Charpy impact properties of the surveillance capsule

specimens are determined, they must be related to the changes in the pressure

165

I

-100 0 100 200 300 400TEMPERATURE, F

FIGURE 1. CHARPY IMPACT ENERGY VERSUS TEMPERATURE FORSURRY UNIT NO. 1 WELD METAL

166

200 Elk River X

cY

\\ < < ax \ 0 \ ^^ \ Trend Band For 550- 590 F

,o

1018 1013 l020

Neutron Fluence, nvt

FIGURE 2. COMPARISON OF 30 FT-LB TRANSITION TEILPERATURE SHIFT VALUESFROM VARIOUS SURVEILLANCE PROGRA4S FOR A302 GRADE B ANDA533 GRADE B PRESSURE VESSEL STEELS

167

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FIGURE 3. CHARPY IMPACT ENERGY VERSUS TEMPERATURE FOR SURRY UNIT NO. 1HEAT AFFECTED ZONE METAL

168

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75SHIFT OF50 FT-LBRTNDT F

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FLUENCE, n/cm2

_ _ _ _ _

FIGURE 4. DETERMINATION OF VESSEL RTNDT SHIFT USING SURVEILLANCECAPSULE DATA

In the example shown, the capsule received a fluenceof 6 x 1018 n/cm2 , while the vessel received 2 x 1018 n/cm2 .

vessel itself. In order to do this, the lead factor of the capsule with

respect to the pressure vessel must be known. The lead factor is defined as

the ratio of the neutron exposure of the surveillance capsule with respect to

the maximum neutron exposure of the pressure vessel.

The lead factor is sometimes specified by the reactor manufacturer.

If it is not known, it can be calculated using data such as the size, position,

dimensions, and composition of reactor internal components from the core out

to the pressure vessel wall.

Typical lead factors in PWR plants are in the range of 1.0 to 4.0.

In BWR plants, however, some capsule lead factors are substantially higher.

The lead factor of a surveillance capsule must be taken into account when

determining when capsules are to be removed. Capsules with very high lead

169

factors should obviously be examined relatively early in the life of a plant,

or the fluence of the capsule specimens will reach a level beyond the end of

the projected plant life.

In some reactors, it is planned to move surveillance capsules from

one location to another location during the reactor lifetime. In such cases,

the neutron fluence reached at each position must be taken into account in

determining the total capsule fluence at time of capsule removal.

Figure 4 is a schematic illustration of how the RTNDT is

determined for a capsule with a Lead factor of 3.0 with respect to the

inner surface of the pressure vessel wall, The "effective" lead factor

of the surveillance capsule with respect to the 1/4 to 3/4 thickness is

obviously greater than 3.0, because of the attenuation of the neutrons

passing through the wall. The upper limit line of the band of Charpy data

used must, of course, be adjusted upward if the data from the surveillance

capsule under consideration falls above the existing upper limit line.

CAPSULE AND SPECIMEN PROBLEMS

The analysis of surveillance capsule data can yield results no

better than the specimens and methods of testing. Most capsules removed

from commercial power reactors have been found to perform well the job for

which they were intended, but problems have been experienced with a few

capsules. The following are problems that have occurred in various

surveillance programs in recent years, and should be considered in design

and implementation of surveillance capsule programs for future reactors.

1. Surveillance capsule installation. There have been problemswith surveillance capsules installed in reactors which includecapsules falling off their hangers during reactor operation,and capsules which could not be removed from their hangers.Obviously, major consideration has to be given to construction andplacement of capsules in reactors.

2. Difficult capsule disassembly. Specimens are packed in surveillancecapsules very tightly in order to achieve good heat transfer betweenthe capsule and the reactor coolant. In many cases capsules areslightly bowed or distorted after irradiation, making specimensdifficult to remove. Spacers should be provided between specimensand the capsule wall when there is a possibility specimens can bedamaged during capsule disassembly. Also, detailed surveillancecapsule drawings should be available during capsule disassembly.

3. Loss of capsule integrity. Capsules are normally checked beforeinstallation in reactors to ensure the capsule bodies have no leakswhich would allow water to enter. We recently disassembled the firstcapsule removed from a reactor, and discovered water had enteredcausing very slight surface corrosion on the mechanical property specimens.

170

Although this problem was not serious at the time thefirst capsule was removed, it will be a serious problemfor the other capsules in the same reactor if they alsoleak but are in the reactor five or ten times longer thanthe first one.

4. Inadequate documentation. The inventory of specimenssupposed to be in a capsule recently examined did notcompletely agree with actual specimens in the capsule.

5. Poorly marked specimens. In some surveillance capsuleprograms, the mechanical property specimens are poorly markedmaking specimen identification inside a hot cell very difficult.

6. Improperly machined specimens. The classic example ofspecimens not being machined to expectations occurred anumber of years ago. Upon disassembling the capsule, wediscovered the tensile specimens were in the form ofunmachined cylindrical rods.

7. Thermal monitors. Some thermal monitors are in the form ofwires and are sealed in Pyrex tubes. If the wires have melted,the capsule has presumably been above the thermal monitormelting point. We have seen some monitors that show evidenceof melting at one end, possibly the end next to the region of thetube that was sealed off. In the same capsule there were nearbythermal monitors with a higher melting point that showed noevidence of melting. It is possible, in other words, that somethermal monitors display evidence of melting before they are eveninstalled in a surveillance capsule.

8. Missing specimens. A capsule recently examined supposedly includedtwo unshielded and two cadmium-shielded aluminum 0.15 percent cobaltdosimeter wires. These wires are used to determine the thermalfluence. Neither cadmium shield contained aluminum-cobalt wires,which were apparently omitted during capsule assembly.Therefore, the thermal fluence could not be determined.

171

NEW METHODS FOR DETERMINING RADIATION

EMBRITTLEMENT IN REACTOR VESSEL SURVEILLANCE

Dr. Richard A. Wullaert

Fracture Control Corporation

330 S. Kellogg Ave.

Goleta, California 93017

ABSTRACT

The radiation embrittlement data required by current U.S. standards

are reviewed using recent results from the Maine Yankee reactor surveil-

lance program. Additional data obtained from the Maine Yankee program

using new test and data analysis procedures are presented, including

initiation and propagation energy curves, brittleness transition temp-

erature, dynamic yield strength, microcleavage fracture strength; and

dynamic fracture toughness (elastic and elastic-plastic). The supplemental

surveillance data were obtained by minor modifications to the standard

Charpy V-notch impact test, and the application of notch bend and fracture

mechanics theories.

INTRODUCTION

The functions of a reactor pressure vessel surveillance program

are to continually monitor the neutron embrittlement of the ferritic

materials and to use the data to verify the reactor operating curves.

Current U.S. standards specify that tensile and Charpy V-notch data be

used to monitor the neutron embrittlement. Charpy V-notch data are used

to establish the adjusted reference temperature RTNDT, The adjusted

reference temperature is then used to verify the original operating limit

curves. The recently completed evaluation of a surveillance capsule

from the Maine Yankee reactor illustrates the application of current

U.S. surveillance standards 3) . For a discussion of the standard

surveillance data, the reader is referred to reference 3 (appended).

173

The main purpose of the current paper is to present additional

methods for determining radiation embrittlement from surveillance type

specimens. For the Maine Yankee surveillance program, supplementary

results were obtained by precracking the unirradiated Charpy specimens

and instrumenting all impact tests.

TESTING AND DATA ANALYSIS PROCEDURES

Reliable load-time and energy-time results from the instrumented

impact tests were obtained by meeting the impact velocity, inertial

loading, time to fracture, and frequency response requirements specified

in the procedures developed as part of the EPRI Fracture Toughness

Program4. The "EPRI procedures" have been experimentally verified 5 '

(7,8)and statistically evaluated '. The reader is referred to the above

references for more details on the testing procedures used.

For precracked Charpy tests in which fracture occurred before

yielding, the fracture load was used to calculate dynamic elastic fracture

toughness Kid. For precracked specimens which fractured after yielding,

the energy to maximum load (corrected for system compliance contributions)

was used to calculate elastic-plastic fracture toughness. The equivalent

*energy fracture toughness Kd and the J integral fracture toughness KJd

were calculated assuming that crack initiation occurred at maximum load.

This assumption was necessary because of the difficulty in determining

crack initiation in a dynamic test.

The energy absorbed by the Charpy specimens (both V-notch and pre-

cracked) was normalized by dividing by the fracture area. The energy

to maximum load (initiation energy) and the post-maximum load energy

(propagation energy) were also normalized by the fracture area. The

dynamic yield strength oyd and the microcleavage fracture stress of

were calculated using techniques described in reference 9. For a current

description of the test and data analysis procedures pertinent to the

surveillance data which follows, reference 10 is recommended.

174

RESULTS AND DISCUSSION

Dynamic Fracture Toughness

The dynamic fracture toughness curves for the unirradiated Maine

Yankee surveillance materials are shown in Figures 1-4. The curves are

based on instrumented precracked Charpy tests performed at stress intensity

5 1/2rates K of approximately 3 x 10 ksi-in /2s. Data obtained from trans-

verse specimens is indexed to the reference temperature RTNDT and compared

to the KIR curve. Note in every case that RTNDT equals the nil-ductility

transition temperature NDTT. A horizontal line on each figure separates

the elastic KId data (fracture before general yield) from the elastic-

plastic data (Kd or Kd). The two measurements of elastic-plastic toughness

*are essentially equal (Kd = KJd) when crack initiation is assumed to occur

at maximum load. In all cases, the baseline data for the Maine Yankee

surveillance program fall above the KIR curve. However, since fracture

may initiate prior to maximum load,.the elastic-plastic values shown may

not be conservative. It should be noted that the dynamic fracture tough-

ness data are for information purposes only, and that .as far as the ASME

Code is concerned, the KIR curve is assumed to describe the Maine Yankee

materials (when properly indexed by RTDT).

Energy Partitioning

The normalized initiation energy (EI/A), propagation energy (Ep/A)'

and total, energy (ET/A) for the unirradiated (u) and irradiated (i) Maine

Yankee ,surveillance materials are shown as a function of temperature in

Figures 5-7. Similar information for the standard reference material

(SRM, HSST plate 01) is shown in Figure 8. The radiation-induced decrease

in the Charpy V-notch upper shelf values of EI, Ep, and ET is summarized

in Table 1. An interesting observation is that for a given material,

the percent drop in each energy component (i.e., EI or Ep) is approximately

the same as the percent drop in the total energy. That is, radiation

175

produced the same percentage drop in all three measurements of energy.

However, on an absolute basis, the major influence of radiation on the

Charpy upper shelf energy (ET) is to reduce the post-maximum load energy

E (defined as propagation energy).

Load-Temperature Diagrams

The general yield loads (PGy) and maximum loads (PM) obtained from

instrumented Charpy V-notch tests on the Maine Yankee surveillance materials

are shown as a function of temperature in Figures 9-12. The radiation-

induced shift in the brittleness transition temperature TD is indicated

on each figure and tabulated in Table 2. The ATD values rate the radiation

sensitivity of the materials in the same order as the AT30 and AT50 values

obtained from the standard Charpy energy curves (see Table 3, reference 3).

The ATD values are not much different than the AT30 values.

The dynamic yield strength was calculated from the general yield

load using techniques described in reference 9. From Figures 9-12 it can

be seen that the radiation-induced increase in dynamic yield strength

(Py) is independent of temperature. This is consistent with the theory

that radiation increases the athermal component of the friction stress.

The effect of radiation on the dynamic yield strength of the various mate-

rials is given in Table 2. The Aoyd values were determined at 2000F

because this was the lowest temperature at which all materials exhibited

yielding. Table 2 indicates that the weld experienced the most extensive

radiation hardening, which is consistent with the static tensile results

and the high copper content of the weld (Tables 1 and 2, reference 3).

The microcleavage fracture stress af was calculated from the load-

temperature diagrams using techniques described in reference 9. The

*radiation-induced change in of for the Maine Yankee surveillance, materials

fis given in Table 2. The of values are based on the value of PGy at the

temperature where PF/PGy = 0.8. Knott ) and Server ( 2 ) have proposed

techniques for calculating of from the value of PGy at TD. Table 2

176

indicates that the load at TD is essentially unaffected by radiation, and

* *thus of calculated from PGy values at TD would indicate that of is indepen-

dent of radiation.

SUMMARY

Until recently (1972), the design of nuclear reactor pressure vessels

was based on the transition temperature concept of fracture-safe design.

Size limitations in surveillance. programs require the use of small test

specimens. Thus, the embrittlement of reactor pressure vessels currently

in operation is almost entirely determined by the Charpy V-notch impact

test. The safety and economic aspects of reactor surveillance dictate

that the maximum amount of information be obtained from the Charpy

specimen.

'The current work illustrates that slight modifications to the test

specimen and test equipment allow additional fracture toughness data to

be obtained. The additional data obtained can be used to supplement the

transition temperature concept for fracture-safe design or to calculate

fracture mechanics and metallurgical fracture parameters.

RERERENCES

1. Sheckherd, J.W. and R.A. Wullaert, "Unirradiated Mechanical Propertiesof Maine Yankee Nuclear Pressure Vessel Materials," Effects Tech-nology, Inc. Report CR 75-269, February 1, 1975.

2. Wullaert, R.A. and J.W. Sheckherd, "Evaluation of the First MaineYankee Accelerated Surveillance Capsule," Effects Technology, Inc.Report CR 75-317, August, 1975.

3. Wullaert, R.A., J.W. Sheckherd, and R.W. Smith, "Evaluation of theMaine Yankee Reactor Beltline Materials," presented at the ASTM8th International Symposium on the Effects of Radiation on StructuralMaterials, St. Louis, May, 1976.

4. Ireland, D.R., W.L. Server, and R.A. Wullaert, "Procedures for Testingand Data Analysis," Effects Technology, Inc. Report TR 75-43,October, 1975.

177

5. Server, W.L., R.A. Wullaert, and J.W. Sheckherd, "Verification ofthe EPRI Dynamic Fracture Troughness Testing Procedures," EffectsTechnology, Inc. Report TR 75-42, October 1975.

6. Server, W.L., R.A. Wullaert, and J.W. Sheckherd, "Evaluation ofCurrent Procedures for Dynamic Fracture Toughness Testing," to bepresented at the Tenth National Symposium on Fracture Mechanics,Philadelphia, August, 1976.

7. Oldfield, W., R.A. Wullaert, W.L. Server, and T.R. Wilshaw, "ControlMaterial Rjund Robin Program; Task A - Topical Repor.," Effects Tech-nology, Inc. Report 75-34R, July 1975.

8. Wullaert, R.A., W. Oldfield, W.L. Server, and T,R. Wilshaw, "StatisticalAnalysis of Interlaboratory Dynamic Fracture Toughness Data," to bepresented at the Dynamic Fracture Toughness Conference, London, July 1976.

9. Wullaert, R.A., D.R. Ireland, and A.S. Tetelman, "Radiation Effects on

the Metallurgical Fracture Parameters and Fracture Toughness of PressureVessel Steels," Irradiation Effects of Structural Alloys for NuclearReactor Applications, ASTM STP 484, American Society for Testing andMaterials, 1970, pp 20-41.

10. Server, W.L. and R.A. Wullaert, "Dynamic Three-Point Bend Analysis forNotched and Precracked Samples," Fracture Control Corporation ReportFCC 76-8 (preliminary draft).

11. Knott, J.F., Fundamentals of Fracture Mechanics, Butterworth, London, 1973.

12. Server, W.L., "Dynamic Fracture Toughness Determined from InstrumentedPrecracked Charpy Tests," UCIA-ENG-7267, August 1972.

178

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192

EVALUATION OF THE MAINE YANKEE

REACTOR BELTLINE MATERIALS

R.A. Wullaertt

J.W. Sheckherdt

R.W. Smith

Fracture Control Corporation, Goleta, California 93017

Yankee Atomic Electric Company, Westborough, Massachusetts 01581

ABSTRACT

Tensile and Charpy V-notch specimens of the base metal, heat

affected zone metal and weld metal from the beltline region of the Maine

Yankee pressure vessel (A5331B-] steel) were irradinted in an arccelerated

surveillance capsule. 'I'h specimens were exposed to a Fluence of 1.3 x

109 n/cm2 (>1 Mev) at 5500F. Charpy V-notch specimens of a standard

reference material (SRM) were also irradiated in the surveillance capsule

as a correlation monitor for dosimetry. Irradiation increased the yield

and ultimate strength and decreased the ductility of all of the Maine

Yankee materials. The yield strength increased 50 percent for the weld

metal and 35 percent for the base and heat affected zone materials. Rad-

iation-induced shifts in the Charpy V-notch curves at the 30 ft-lb, 50

ft-lb, and 35 mil levels were measured. The decrease in the Charpy upper

shelf energy was also measured. The largest temperature shift occured at

the 35 mil level for all materials, and this shift was used to determine

the adjusted reference temperature. The increase in reference temperature

ranged from 140 F for the base metal to 345 F for the weld metal. The

weld metal also showed the largest drop in the Charpy upper shelf energy

(44 percent) versus 23 to 31 percent for the other materials.

The critical beltline material for determining the new operating

limit curves for the reactor was the weld metal, with an adjusted refer-

ence temperature of 315°F and a Charpy upper shelf value of 57 ft-lb.

193

The high copper and phosphorus content of the weld (0.36 percent copper,

0.015 percent phosphorus) caused the irradiated Charpy data to fall above

the general trend curve for A533B steel. A trend curve for the weld

metal was constructed using independently generated irradiation data on

the same weld melal.

INTRODUCTION

Pressure-temperature limitations for heatup and cooldown of the

reactor coolant system during operation and test conditions are provided

in the Technical Specifications for each plant. Theese pesstre-tepera-

ture limits are imposed on the reactor coolant pressure boundary to provide

adequate safety margins against nonductile or rapidly propagating failure

of the ferritic pressure vessel materials. Appendices G and H of 10 CFR

50t specify the requirements for the reactor vessel pressure-temperature

limits. These limits are based on the ASME Code, Section III, Appendix G,

which provides a fracture mechanics basis for determining allowable limits.

The.operating limit curves in the plant Technical Specifications are

based on the baseline mechanical properties of the reactor vessel adjusted

by the anticipated embrittle.ent of the beltline region of the vessel due

to neutron exposure. Trend curves of change in fracture toughness versus

fluence are used to predict the embrittlement for the specific type of

steel, residual element content, and operating temperature of the reactor.

The functions of a reactor pressure vessel surveillance program are to con-

tinually monitor the neutron embrittlement of the ferritic materials and to

use the data to verify the original operating limit curves.

The Maine Yankee nuclear reactor is a pressurized water reactor built

by Combustion Engineering. The surveillance program design, selection of

materials, specimen and capsule fabrication, and installation of the cap-

Amendments to AVEC regulallion TiLle 10, Part 50 pithll hed in Federal

Register, July 17, 1973.

194

sules were performed by Combustion Engineering as part of the Maine Yankee

construction contract . The unirradiated and irradiated mechanical proper-

ties of the Maine Yankee surveillance materials have recently been measured2 '3.

The surveillance examination determined that the test specimens were irradi-

ated at 550 °F to a fluence of 1.3 x 1019 m/cm2 (>1 Mev). Details of the

surveillance material characterization, irradiation capsule configuration

and location, capsule disassembly, and neutron dosimetry can be obtained

from these reports and will not be presented here, The purpose of this

paper is to present the radiation-induced changes in mechanical properties

and describe the determination of the adjusted reference temperature RTD T .

MECHANICAL PROPERTY RESULTS

The various materials and mechanical property test specimens con-

tained in the first accelerated capsule are shown in Table 1, along with

the chemical analysis of the materials. All tests were performed in accor-

dance with appropriate ASTM standards and internal procedures 5

Tensile Tests

Three irradiated tensile specimens of the Maine Yankee base metal,

weld metal and heat affected zone (HAZ) material were obtained from the

surveillance capsule. The tensile specimens were standard ASTM type 0.252

inch diameter specimens with a one-inch gage length. One specimen of each

material was tested at room temperature, the reactor operating temperature

(566°F) and the reactor design temperature (650°F). The cross head rate

through 0.2 percent yield strength was 0.005 in/in/min. This cross head

rate was maintained after yielding. An extensometer was used to obtain

load-elongation curves through yielding and a load-time (cross head travel)

curve was obtained past this load. The tensile properties measured were

0.2 percent yield strength, ultimate tensile strength, total elongation

195

and reduction in area. A summary of the irradiated tensile properties is

given in Table 2 for comparison. An irradiation dose of 1.3 x 1019 n/cm

(>1 Mev) at 550 F produced an increase in the yield strength ol approximately

50 percent for the weld 1.nd 35 percent for the bare .ild H1AZ ma.tltril.

It should be noted Lhat irradilated tensile data play no direct role

in determining the adjusted reference temperature RT'rNT. However, the

tensile results provide valuable back up information and are thus included

in the paper.

Charpy Impact Tests

The surveillance capsule contained four Charpy impact compartments.

Each compartment contained twelve Charpy V-notch specimens of a given mat-

erial. All Charpy specimens were the standard size 0.394 in (10 mm). The

Charlpy V-notch tests were performed in accordance with ASTM E23 . All

tests were performed on a 220 ft-lb impact machine at an impact velocity of

16.7 ft/s. Army Materials and Mechanics Research Center (AMMRC) calibration

specimens were tested prior to the irradiated specimens to ensure the cali-

bration of the impact machine. The impact machine was instrumented to obtain

additional test data. The instrumented Charpy results will not be discussed

here, but are presented in Reference 3.

Twelve irradiated Charpy V-notch specimens of the base metal, weld,

HAZ and standard reference material were tested over a range of temperatures

to generate a full Charpy transition curve. Test temperature, dial energy,

fracture appearance and lateral expansion were recorded for each test and

the results for the various materials are plotted as a function of temp-

erature in Figures 1-4. The transition curves for the unirradiated mat-

erials are included in the figures so that the various transition temper-

ature shifts can be calculated. Shown in the figures is the shift in the

energy curve at the 50 ft-lb level (T5 0) and the shift in the lateral expan-

sion curve at the 35 mil level (T35M). Vertical arrows indicate the drop-

196

weight NDT temperature for the unirradiated materials and the reference temp-

erature, RTNDT for the unirradiated and irradiated materials. Also, shown

in the figures is the upper shelf energy value for the unirradiated and ir-

radiated materials. Table 3 summarizes the radiation-induced changes in the

Charpy energy curve for the four materials in terms of the shift at the

30 ft-lb level (T30), the 50 ft-lb level (T5 0 ) and the upper shelf energy.

DISCUSSION

Adjusted Reference Temperature

The fracture toughness tests required in surveillance programs are

specified in Appendix H to 10 CFR 50, Section III. The adjusted reference

temperature for the irradiated materials (RTNDT(i)) is established by adding

to the unirradiated reference temperature (RTNDT(u)) the amount of the temp-

erature shift in the Charpy V-notch test curves between the unirradiated

material and the irradiated material, measured at the 50 ft-lb level (AT5 0 )

or the 35 mil lateral expansion level (AT3 5M), whichever temperature shift

is greater. The shifts in the 50 ft-lb and 35 mil Charpy V-notch levels

for the Maine Yankee surveillance materials are.tabulated in Table 4. Note

that for every material AT35M > AT5 0

Thus, the adjusted reference temperature for the irradiated materials

is calculated by

RTNDT (i) = R'NDT () + T3 5 M (1)

Thd unirradiated reference temperatures were reported in Reference 2

and were-established according to Article NB 2300 of the ASME Code, Section

III. Article NB 2300 requires that RNDT be based on test results from

transverse specimens. For all of the unirradiated Maine Yankee surveillance

materials, the transverse Charpy V-notch results exceeded 50 ft-lb and 35

mils at NDTT + 60F. Thus, the reference temperature for the uni.rradiat.ed

materials was controlled by the NDT temperature, and RTNDT(u) = NDTT(u).

197

Only longitudinal Charpy specimens of the standard reference material (SRM)

were available for the unirradiated tests, so the 50 ft-lb and 35 mil

criteria could not be determined. For the SRM, Lt was assumed that

RTNDT(u) = NDTT(u).

The adjusted reference temperatures for the Maine Yankee surveillance

materials were calculated using Equation 1 and the new RTNDT values are

listed in Table 4. The AT35M values for the SRM and base material are

based on shifts in longitudinal Charpy data, whereas the code specifies

that the shift be based on transverse Charpy data. Although orientation

influences the energy absorbed in a Charpy test, there are no published

results that indicate that the radiation-induced shift in Charpy curves

is orientation dependent.

In addition to establishing the radiation-induced shift in the RTNDT,

it is important to determine the influence of irradiation on the Charpy

upper shelf energy. Regulations in 10 CFR 50, Appendix G require that the

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

cannot drop below 50 ft-lb during service (irradiation). Table 3 summarizes

the upper shelf behavior of the Maine Yankee materials due to irradiation.

Note that the upper shelf decreased from 23 to 45 percent, with the weld

metal showing the greatest sensitivity to radiation. For a fluence of

1.3 x 101 n/cm (>1 Mev), none of the materials exhibited a Charpy upper

shelf energy less than 50 ft-lb.

Standard Reference Material

The standard reference material used in the Maine Yankee surveillance

program was A533B Class 1 plate from the Heavy Section Steel Technology

Program (HSST plate 01). Berggren and Stelzman7 have published a radiation

embrittlement trend curve for this material. Figure 5 shows the radiation-

induced shift in the 32 ft-lb Charpy level as a function of irradiation

temperature. The shift has been normalized to 1 x 1019 n/cm2 (>1 Mev).

From Table 3, AT3 0 = 150°F for the Maine Yankee SRM irradiated at 550°F.

198

Figure 5 shows that the data point for the Maine Yankee SRM falls slightly

above the trend band, indicating that the SRM experienced a fluence some-

what greater than 1 x 1019 n/cm (>1 Mev).

The Maine Yankee SRM Charpy specimens were taken from the quarter

thickness of HSST plate 01 and in the longitudinal direction.(]/4T, RW).

If only the 1/4T, RW data points in the trend curve at 550'F are compared

with the SRM data, there is an indication that the fluence obtained by the

SRM was substantially greater than 1 x 1019 n/cm (>1 Mev).

To resolve this question, a trend curve for HSST plate 01 irradiated

at 550°F has been constructed using the 1/4T, RW data discussed above and

8 9recent data by Stelzman and Berggren and Hawthorne , This trend curve is

shown in Figure 6 and is based on radiation-induced shifts in Charpy curves

at both the 30 ft-lb and 50 ft-lb level. The values of AT30 = 150°F and

AT50 = 165°F for the Maine? Yankee SRM have been entered on the trend curve

as horizontal dashed lines. Their intersection with the appropriate trend

curve defines a fluence range for the SRM of 1.65 to 1.85 x 1019 n/cm2

(>1 Mev). Thus, a fluence of 1.7 x 10 n/cm (>1 Mev) was used for the

SRM in evaluating the fluence for the Maine Yankee accelerated surveillance

capsule.

Fluence Estimates

The neutron dosimetry and neutron fluence calculations for the first

Maine Yankee accelerated surveillance capsule are presented in detail in

reference 3 and will only be summarized here. Fluence calculations using

current ASTM practices and the ANISN computer code predicted a fluence of

9 x 1018 n/cm (>1. Mev). Since the standard reference material indicated

a fluence of 1.7 x 10 n/cm (>1 Mev), the initial dosimetry was reviewed

and additional techniques for calculating fluence were tried. The nominal

calculated fluence based onactual core operating history and rodded power

distribution was estimated to be 1.33 x 1019 n/cm2 (>1 ev). This fluence

199

value seems very acceptable when compared to the fluence of 1.27 x 109

n/cm2 (>1 Mev) obtained from the Mn-54 activity and thi '1.7 x 1019 n/c 2

(>1 Mev) fluence estimated from the SRM data. Also, the shifts in the

Chalrpy curves for all a f tlhe i r-l;adll.t ,d M.iline Y.anucc nm;tl ritl:l wer ('i 1Ir)I

nearly In agreement with a fluence of 1.3 x 1019 n/cm (>1 Mev) than for

a fluence of 9 x 101 n/cm (>1 Mev). Thus, the best estimate of the

fluence for the first Maine Yankee gurveJ.llance Capsule is 1..3 x 1019

n/cm2 (>1 Mev).

Trend Curves

Recently Bush compiled a comprehensive data base on radiation

damage in light water reactor pressure vessel steels. Figure 7 shows the

radiation-induced shift in the Charpy curves at the 30 ft-lb level for

A533B and A508-2 steels. 'These steels had varying copper and phosphorus

contents and were irradiated over a wide range of temperatures. The AT30

values from Table 3 for the Maine Yankee surveillance materials have been

entered on this figure at a fluence level of 1.3 x 1019 n/cm 2 (>1 Mev).

The Maine Yankee base metal (P) contained 0.15 percent copper and 0.013

percent phosphorus, whereas the Maine Yankee submerged arc weld contained

0.36 percent copper and 0.015 percent phosphorus. Points representing

the same chemistry and irradiation temperature can be found in the near

vicinity of the Maine Yankee base metal point (X-P). The high copper and

phosphorus content of the Maine Yankee weld metal resulted in a AT3 0

substantially higher than the general data base of points in Figure 7.

The only data point near the Maine Yankee value (X-W) that has the same

irradiation temperature is just to the right of the Maine Yankee point.

This point is for a weld with 0.23 percent copper and 0.011 percent phos-

phorus irradiated at 550 F to a flutnce of 2.5 x 109 n/cm (>1 Mev)9 .

This weld not only showed the same shift (AT30 = 270°F) as the Maine

200

Yankee weld, but also the same percent drop in the Charpy upper shelf

energy (44%, 125 to 70 ft-lb).

Critical Beltline Material

Current federal regulations (10 CFR 50 Appendix H, Section III) specify

that the highest adjusted reference temperature and the lowest upper

shelf energy level of all the irradiated beltline materials shall be used

Lo establish the new operating limit curves. Based on the Maine Yankee

surveillance results, the weld metal is the limiting or critical belt-

line material.

By a rather fortunate coincidence, this same weld metal was studied

as part of a radiation sensitivity study by the Naval Research Laboratory

and Combustion Engineering . The,NRL-CE Charpy curves for the unirradia-

ted and irradiated Maine Yankee weld are shown in Figure 8. Also included

in the figure is the Charpy curve for the same weld obtained from the pres-

ent surveillance program. Note that the unirradiated curves obtained by

both laboratories are very similar (NDTT =-30°F, Charpy upper shelf equals

105 ft-lb versus 107 ft-lb, see Figure 2). Also note that the upper shelf

values after irradiation are essentially identical. It is certainly en-

couraging that different laboratories can have such close agreement on

Charpy curves from completely independent studies.

The shift in the Charpy curves at the 30 ft-lb level (AT30) has been

indicated in Figure 8 for the three fluences involved. These AT30 values

are plotted as a function of fluence in Figure 9, which has been reproduced

from the NRL-CE study . The data point for the weld irradiated in the

first Maine Yankee surveillance capsule is indicated on the figure. The

surveillance capsule fluence of 1.3 x 10 9 n/cm2 (>1 Mev) seems appropriate

when compared to the NRL-CE data. It is obvious that the high copper con-

tent of the Maine Yankee weld causes the radiation-induced embrittlement

to exceed the normal trend for A533B material.

201

CONCLUSIONS

The following conclusions were reached concerning the irradiation

response of the Maine Yankee reactor beltline materials:

1. The base and heat affected zone materials exhibited a radiation-

induced embrittlement similar to the standard reference material and con-

sistent with current embrittlement-fluence trend curves.

2; The enhanced radiation.sensitivity of the weld metal is consistent

with current theories concerning the detrimental effects of high copper

content.

3. Radiation induced shifts at the 35 mil level exceeded those mea-

sured at the 30 and 50 ft-lb levels for all materials. Thus, the 35 mil

shift was used to determine the adjusted reference temperature,

4. The critical beltline material for determining the new operating

limit curves for the Maine Yankee reactor was the weld metal, with an adjus-

ted reference temperature of 315°F and a Charpy upper shelf value of 57

ft-lb.

REFERENCES

1. CENPD-37, "Summary Report on Manufacture of Test Specimens and Assem-

bly of Capsules for Irradiation Surveillance of Maine Yankee Reactor

Vessel Materials," Combustion Engineering (December 30, 1971)

2. Sheckherd, J.W. and R.A. Wullaert, "Unirradiated Mechanical Properties

of Maine Yankee Nuclear Pressure Vessel Materials," Effects Technology,

Inc. Report CR 75-269 (February 1, 1975).

3. Wullaert, R.A. and J.W. Sheckherd, "Evaluation of the First Maine

Yankee Accelerated Surveillance Capsule," Effects Technology, Inc.

Report CR 75-317 (August, 1975).

4. "Surveillance Tests for Nuclear Reactor Vessels," ASTM E185-73, Annual

Book of ASTM Standards, ASTM, Philadelphia, PA, 1974.

202

5. Sheckherd, J.W. and L.K. Prince, "Quality Assurance Program for Mat-

erials Testing of Unirradiated Baseline Specimens and Irradiated

Surveillance Capsule Specimens," Effects Technology, Inc. report to

Maine Yankee Atomic Power Company, September 1974.

6. "Notched Bar Impact Testing of Metallic Materials," ASTM E23-72,

Annual Book of ASTM Standards, ASTM, Philadelphia, PA, 1974.

7. Berggren, R.G. and W.J. Stelzman, "Radiation Strengthening and Em-

brittlement in Heavy Section Plate and Welds," Nucl. Engng. Design,

17, (1971) 103 - 115.

8. Stelzman, W.J. and R.G. Berggren, "Radiation Strengthening and Em-

brittlement in Heavy Section Steel Plates and Welds," ORNL - 4871

(June 1973).

9. Hawthorne, J.R., "Postirradiation Dynamic Tear and Charpy-V Per-

formance of 12-in. Thick A533B-1 Steel Plates and Weld Metal,"

Nucl. Engng. Design, 17 (1971) 116 - 130.

10. Bush, S.H., "Radiation Damage in Pressure Vessel Steels for Com-

mercial Light Water Reactors (March 1974).

11. Hawthorne, J.R., J.J. Koziol and R.C. Groeschel, "Evaluation of

Commercial Production A533B Plates and Weld Deposits Tailored for

Improved Radiation Embrittlement Resistance," presented at ASTM

Symposium on Effects of Radiation on Structural Materials (June 1974).

203

Table 1. Materials, Specimens and Chemistry for Maine Yankee Capsule Number 1.

A533B-1PLATE

SPECIMEN (D-8406-1)

Charpy V-notch 12(L)+

Tensile (0.252in) 3

TOTAL 15

CHEMISTRY

Si .22

S .013

P .013

Mn 1.27

C .22

Cr .11

Ni .59

Mo .57

B .0004

Cb <.01

V <.001

Co .010

N .006

Cu .15

Al .021

Ti <.01

W .01

As .01

Sn .009

Zr .001

SUBMERGEDARC WELD

(D-8407-1,-3)

12

3

15

.22

.012

.015

1.38

.14

.07

.78

.55

.0002

<.01

.003

.013

.012

.36

.004

<.01

.01

<.01

.001

.002

SRM(HSST-O1)

12(L)

12

.22

.008

.008

1.37

.22

.15

.66

.54

.02

.18

__ _

__ _

HAZ(D-8406-2)

12

3

15

__ _

__ _

__ _

__ _

TOTAL

48

9

57

_ __

SRM - Standard reference material, A533B-1 steel, plate

Heavy Section Steel Technoloty (HSST) program.

+(L) - Longitudinal orientation

01 from the

204

Table 2. Tensile Properties of i:radiated flaine Yankee

Beltline Hlaterials 3550F 1.;3 x 1019 n/1cm 1 (>1 Mev)]

SPECIMENNO.

TEMP,(CF)

11 .

(.2% Y.S.(ks i

UI..T.S, ITOTAl(k si) E1 . (%)AT ERI AL

Base (L) 1UP R,T. 8'1 .6(61. 6)

i... ....... ....

102.2

I(71.1)

103.7

( 37 4. ,

(87.2)

27 .6

(2'9.0)

24 ,..2

:(2 8.5)

RA.(s)

54.8

(71. 3)i.... - .. .

52. 1

(70.4 }

Weld 3J' R. T.

- --- - - -- - - --t -"-* - 1-- ----- -* -;--

HAZ 4JU) R.T, 83.7 10T .7 22.7 55.4

i 6 ) t4.3) g (22..' _1 7.8)._ . _ .-....... .„. : z'--...:... _._ _ _ 1 -- -:..-:, _ . .L?. ..i_'.. _ -L ..

Base (L) 104 566 73.9 5 3 22. 5 45.8

_ _ .. ... _(. !. .......l :8 .6 i ._ 3 ,r8.)

Weld 3J3 566 93.3 10, 8 21 .8 42. 8

... . . . 62_ . -. ... ...... . ). .

HAZ 4KJ 566 78.6 97.8 O.,' 47.6

___s_ (.7Li 834vz (20.2) .. 3+..-.",'-- -" --.- _ _ _ - .-.

* ... _.. .... . ._. _ ........ _.... .

Base (L) 1JM 650 69.2 92.0 25.0 49,9

Weld 3i:J* 650 -- -- 1. l01 -18.2 -44.0

..... (-_4. .... ( 81.6) (?2 5.) 6 .)HJA2 <4K2 650 | 69.8 9|3.0 22.. . 61,

_._ . . _ .. ... _ __. _J._,....i~~ .... ,... .... r1(81_;6^ i~?3_3l,_ L68^_3t( ) - Average value for three tests on unirradiated material.

* ~ Accidental prestraining of specimen prior to testing resultedin unreliable yield strength measurement and questionabletest results,

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216

MATERIALS SURVEILLANCE PROGRAM FOR C-E NSSS REACTOR VESSELS

JOH'N J. KOZIOL, Supervisor, Metallurgical ServicesNuclear Power Systems

C-E Power SystemsCombustion Engineering, Inc.

Windsor, Connecticut

ABSTRACT

Irradiation surveillance programs for light water NSSS

reactor vessels provide the means by which the utility can assess

the extent of neutron-induced changes in the reactor vessel materials.

These programs are conducted to verify, by direct Lmeasurement, the

conservatism in the predicted radiation-induced changes and hence

the operational parameters (i.e., heat-up, cooldown, and pressuriz-

ation rates). In addition, such programs provide assurance that

the scheduled adjustments in the operational parameters are made

with ample margin for safe operation of the plant.

During the past 3 years, several documents have been pro-

mulgated establishing the criteria for determining both the initial

properties of the reactor vessel materials as well as measurement

of changes in tiese initial properties as a result of irradiation.

These documents, ASTM E-185-73, "Recommended Practice for Surveil-

lance Tests for Nuclear Reactor Vessels," and Appendix H to 10 CFR

50, "Reactor Vessel Material Surveillance Program Requirements,"

are complementary to each other. They are the result of a change

in the basic philosophy regarding the design and analysis of

reactor vessels. In effect, the empirical "transition temperature

approach," which was used for design, was replaced by the "analytical

fracture mechanics approach." The implementation of this technique

was described in Welding Research Council Bulletin 1975 and

Appendix G to ASME Code Section III. Further definition of require-

ments appears in Appendix G to 10 CFR 50 published in July 1973.

It is the intvent of this paper to describe (1) a typical materials

surveillance program for the reactor vessel of a Combustion

Engineering NSSS, and (2) how the results of such programs, as

well as experimental programs provide feed-back for improvement of

materials to enhance their radiation resistance and thereby further

improve the safety and reliability of future plants.

217

TYPICAL C-E SURVEILLANCE PROGRAM

The following discussion describes a typical reactor vessel

irradiation surveillance program, provided by C-E for most of the

NSSS already in service or which will be placed in service in the

near future. This program satisfies the requirements of both ASTM

E-185-73 and 10 CFR 50 Appendix H, and, in addition, contains suf-

ficient flexibility, by providing for an adequate number of capsules,

to permit monitoring of vessel.annealing in the unlikely event

that this should become necessary at some point in the life of the

vessel.

Since NSSS nuclear reactor vessels designed by C-E are fabri-

cated primarily from plates, they contain weldments in the belt-

line (irradiated) region. An adequate surveillance program,

therefore, includes test samples containing base metal, deposited

weld metal, and the heat affected zone (HAZ). Sample material is

obtained from each of the six plates that comprise the beltline

region after the vessel shell segments are formed.

MATERIALS SELECTION

The materials selection procedures ensure that surveillance

materials are a representative sample of the materials in the

reactor vessel. The base metal material used in the surveillance

program is selected from that plate, within six comprising the

"beltline" region, which might limit the operation of the reactor

vessel due to irradiation. Consideration is given to (1) initial

reference temperature RTNDT, (2) chemical composition, and (3)

upper shelf energy. Thus, material in the surveillance capsules

is taken directly from the plates actually used in fabricating the

vessel. The section of plate which is used for the base metal test

material and for weldments is adjacent to the material employed

in ASME Code Section III tests and is located at least one plate

thickness from any water-quenched edge. Weld metal is prepared

using the same weld wire/flux as that usedin fabrication of the

vessel beltline.

218

FABRICATION PROCEDURES

Sections of this test material are welded using procedures

and welding materials identical to those employed in the beltline

region of the vessel, then all the test materials (plate and weld-

ments) are heat treated to provide metallurgical conditions equiva-

lent to those within the vessel. This heat treatment is performed

with the vessel or separately under identical conditions. Some

of the test material provides samples for pre-irradiation testing

(drop weight, Charpy impact, tensile) and the balance is included

in the irradiation test capsules (Charpy impact and tensile).

TYPE AND QUANTITY OF SPECIMENS

Samples of base metal and HAZ material are obtained from

the quarter thickness (1/4 T) locations. Weld metals are obtained

from the entire thickness of the weld except for a typical com-

plement of samples shown in Table I. This includes plate material

taken from a standard heat of A533B steel provided by the USNRC

sponsored Heavy Section Steel Technology (HSST) Program. The

number of test samples exceeds the minimum number recommended by

ASTM E-185-73 and is provided to ensure more accurate determination

of both pre-irradiated and irradiated properties. An ample quantity

of unirradiated specimens is provided to permit pre-cracking of

specimens for instrumented Charpy tests along with the standard

Charpy tests.

FLUX MONITORS AND TEMPERATURE MONITORS

The design of an irradiation surveillance program should pro-

vide for measurement of neutron fluence, neutron energy spectra,

and irradiation temperature of the samples. Measurements of fluence

and energy spectra are obtained from analyses of a number of

fission monitors and neutron threshold detectors (Table II) which

are incorporated in each of the surveillance capsules.

The 5.3 year Co-60 formed by Cu-63 (n, a) reaction and

the 28 year Sr-90 formed by fast fission of U-238 provide a means

of automatically integrating the fast neutron flux. The Co-60

provides integrated fast neutron flux for the first 10 to 15 years

219

TABLE T

TYPE ANTD QCLU.\NTITY OF SPECISENS

Type of Specimen

Pre-irradiation

DW

C

Tensile

Orientation*

L

T

L

TrL

T

BaseMetal

16

50

30

18

1.8

112

.lantity of Specimens

WeldMetal HAZ SRM**

16

30

18

64

16

30

15

Totals

16

32

45

90

36

55

255

18

64 15Sub- totalIrradiation

cv L

T

L

T

Tensile

48

72

18

138

250

18

90

154

72

18

90

154

24 72

216

36

18

342

597

Sub-total

Total

24

39

*L = Longitudinal; T = Transverse

**SRM = Standard Reference Material

TABLE II

FLUX MONITOR MATERIALS

Material

Uranium

Sulfur

Iron

Nickel

Copper

Titanium

Cobalt

Reactor

U 2 8(nf) Sr 90

S 3 2 (n,p)p 3 2

Fe 5 4 (n,p)Mn 5 4

Ni 8 (n,p)Co5 8

Cu 6 3 (n,C)Co 6 0

Ti46(n,p)Sc 4 6

CO5 9 (n, 'r)Co60

Threshold Energy (MEV)

0.7

2.9

4.0

5.0

7.0

8.

Thermal

Half-Life

28 Years

14.3 Days

314 Days

71 Days

5.3 Years

84 Days

5.3 Years

220

of plant operation while the Sr-90 provides integrated flux for

the entire life of the plant owing to its 28-year half life. All

other reactions become saturated quickly and are only useful in

determining the fast neutron spectrum over the desired energy range.

A good estimate of the maximum temperature of the irradiated

samples is obtained by post-irradiation examination of temperature

monitors, (Table III) which contain materials with melting points

within the operating temperature range of the vessel. The

standard reference material from the HSST Program, referred to

above and irradiated with some bf the Cv samples, provides a good

cross-check on the dosimetry by means of changes in impact properties.

These data provide a basis for correlating the results of this

program with other surveillance programs as well as with data from

experimental irradiations.

TABLE III

COMPOSITION AND MELTING POINTS

OF CANDIDATE MATERIALS FOR TEMPERATURE MONITORS

Composition (WT%) Melting Temperature (F)

80.0 Au, 20.0 Sn 536

90.0 Pb, 5.0 Sn, 5.0 Ag 558

97.5 Pb, 2.5 Ag 580

97.5 Pb, 0.75 Sn, 1.75 Ag 590

DESCRIPTION OF CAPSULE ASSEMBLIES

The test specimens are placed within corrosion-resistant

capsule assemblies (1) to prevent corrosion of the carbon steel

test specimens by the primary coolant during irradiation, (2) to

physically locate the test specimens in selected locations within

the reactor, (3) to provide a means by which the irradiation con-

ditions (fluence, flux spectrum, temperature) can be determined,

and (4) to facilitate the removal of a desired quantity of test

specimens from the reactor when a specified fluence has been

attained.

221

Lock Assembly

Wedge Coupling AssemblyTensile -Monitor.

Compartment

ITensile -MonitorCompartment -

Charpy Impact Compartments

Charpy Impact Compartments

Tensile -MonCompartmer

Fig. 1: Surveillance capsule assembly

222

Capsule Assembly

A typical capsule assembly, (Fig. 1) consists of a series

of seven specimen compartments, connected by wedge couplings, and

a lock assembly. The wedge couplings also serve as end caps for

the specimen compartments and position the compartments within thecapsule holders which are attached to the reactor vessel. The

lock assemblies fix the locations of the capsules within the holders

by exerting axial forces on the wedge coupling assemblies whichcause these assemblies to exert horizontal forces against the sides

of the holders. The lock assemblies also serve as a point of

attachment for the tooling used to remove the capsules from thereactor.

Each capsule assembly is made up of four Charpy impact testspecimen (Charpy impact) compartments and three tensile test speci-

men -- flux/temperature monitor (tensile-monitor) compartments.

Each capsule compartment is assigned a unique identification so

that a complete record of test specimen location within each com-partment can be maintained. The capsule identification incorporates

a four-symbol alphanumeric code that identifies the reactor vessel,

the capsule assembly, the relative position of a compartment within

a capsule assembly, and the type of test material contained within

each compartment.

Charpy Impact Compartments -- Each Charpy impact compartment

(Fig. 2) contains 12 impact test specimens. This quantity of

specimens provides an adequate number of data points for establish-

ing a Charpy impact energy transition curve for a given irradiated

material. Comparison of the unirradiated and irradiated Charpy

impact energy transition curves permits determination of the RTNDTchanges due to irradiation for the various materials.

The specimens are arranged vertically in four 1 x 3 arrays

and are oriented with the notch toward the core. The temperature

differential between the specimens and the reactor coolant is

minimized by using spacers between the specimens and the compart-

ment and by sealing the entire assembly in an atmosphere of helium.

Tensile -- Monitor Compartments -- Each tensile-monitor com-

partment (Fig. 3) contains three tensile test specimens, a set

of flux spectrum monitors, and a set of temperature monitors for

223

Wedge Coupling - End Cap

Charpy Impact Specimens

Spacersr: :

-Rectangular Tubing

.Wedge Coupling - End Cap

Fig. 2: Charpy impact compartment assembly

224

Wedge Coupling - End Cap- ,'

Flux Monitor Housing-

Stainless Steel Tubing-Threshold Detector /

Flux Spectrum Monitor

Flux Spectrum MonitorCadmium Shielded

-'Stainless Steel Tubing'Cadmium S-hield\Threshold Detector

.ingTer peratu re

TemperatureHousing Alloy

Tensile' Specimen

Split Spacer

Tensile Specimen Housing

Rectangular Tubing

*Wedge Coupling - End Cap

Fig. 3: Tensile monitor compartment assembly

225

estimating the maximum temperature to which the specimens have been

exposed. The entire tensile-monitor compartment is sealed within

an atmosphere of helium.

REMOVAL SCHEDULE

Surveillance capsules must be placed at locations where

they will receive an exposure equal to but not greater than three

times the exposure of the reactor vessel. Capsule positions for

a typical C-E surveillance program are shown in Fig. 4. The

capsule holders are welded to the cladding of the vessel, thereby

accurately establishing the location of the samples with respect

to the vessel. A typical removal schedule for the six capsules

included in the program are presented in Table IV. The target

TABLE IV

TYPICAL CAPSULE ASSEMBLY REMOVAL SCHEDULE

AzimuthalCapsule Location Removal Target 2

No. (degrees) Time (yrs) Fluence (n/cm )

1 97 7 6.0 x 101 8

2 104 19 1.6 x 1019

3 248 30 2.5 x 1019

4 263 Standby

5 277 Standby

6 83 Standby

fluence levels are determined at the azimuthal locations at the time

intervals indicated in the withdrawal schedule in 10 CFR 50 Appendix

H, Section II C. 3. b. The fluence values in Table IV are accurate

within +10 percent, -40 percent. The uncertainty is composed of

errors in the calculational method and errors in the combined radial

and axial power distribution.

Withdrawal schedules may be modified to coincide with those

refueling outages or plant shutdowns most closely approaching the

withdrawal schedule.

Three standby capsules are available for any contingency that

might necessitate additional surveillance tests, e.g., to assess

226

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228

the effects of possible annealing and subsequent reirradiation

of the reactor vessel. The type and quantity of samples in each

of the six surveillance capsules is presented in Table V.

TESTS

Following removal, standard Charpy-V impact, and tensile

tests are conducted. In addition, instrumented Charpy-V tests are

conducted, and where sufficient samples are available, pre-

cracked Charpy-V samples are tested.

Neutron dosimeter wires are analyzed to determine total

fluence. Pre-irradiation and post-irradiation data are compared

to determine the extent of change. These data are, in turn, com-

pared with earlier predictions of property change as a function

of neutron exposure. Modifications to operating parameters are made,

when necessary.

APPLICATION OF RESULTS

As previously mentioned, the results of these programs are

used to verify the predictions of RTNDT shift versus fluence.

Design curves, such as that shown in Fig. 5 are used for this pur-

pose,' These are.based on extensive experimental irradiations con-

ducted for over a decade. Analysis of the experiments led to

studies of residual element influence on radiation behavior through

limitations of residual elements. It has been found that signifi-

cant improvement in properties is achievable (improved materials

curve in Fig. 5) by limiting the content of Cu, P, S, and V. By

limiting these elements to a greater degree (controlled material

curve in Fig. 5), additional improvements are achievable,

Note the various changes in RTNDT increase achievable through

composition control. This leads to:

1) Smaller beltline region

2) Fewer capsule removals and post-irradiation tests

3) Less adjustment in operating parameters

4) Elimination of annealing considerations.

2 i . , -

229

L 400 1312U-

= 300 -c ~~~'"300'U c Improved

i .^-' .^^ ^Ccn;'o;!d- -C Oo£ 2.C0 - ^^ >^ ^^^-Mtctrial

C2 0

1920Ix1018 xO19 x1020

Fluence, n/cm2 1 Mev

Fig. 5: Comparison of C-E design curves for controlled, improved

and uncontrolled residual beltline material (550 F

irradiation)

SUMMARY

Surveillance programs containing Charpy-V specimens can be

used to adequately monitor changes in RTNDT (reference temperature)

and fracture toughness of reactor vessel materials. Use of in-

strumented Charpy-V impact testing techniques further extends the

measurement capabilities and introduces quantitative values of

fracture toughness. New criteria for selection, fabrication,

irradiation, and testing further ensure adequacy of toughness of

vessel materials under all operating conditions. Composition

control reduces the amount of property change and thereby further

contributes to the margin of safety.

230

REPORT ABOUT ACOUSTIC EMISSION ANALYSIS ON THE

REACTOR PRESSURE VESSEL OF ETE FIRST AUSTRIAN

NUCIEAR POWER PLANT

Kurt K. WISCOIN

Vienna

ABSTRACT

The acoustic emission analysis performed in conjunction with the

hydraulic test of the reactor pressure vessel of the Tullnerfeld nuclear

power plant is briefly described.

The first Austrian nuclear power station using a boiling

water reactor of the KWU-type with 760 MW output is approaching

completion and will be approximately going into service in the

next year. For safety reasons the Federal Ministry for Build-

ings and Technology responsible for the security of the pressurized

components in this nuclear plant has imposed an acoustic emission

analysis of the reactor pressure vessel during the first hydraulic

test.

This new testing method enables to inspect the reactor pres-

sure vessel including all piping till to the first shutoff valves

against cracks and other flaws, because it can not be thoroughly

excluded, that using the usual non-destructive testing methods

for material and welding, defects have been overlooked. On the

other hand, in all safety philosophies it is assumed, that in all

cases of accidents the integrity of the reactor pressure vessel

is guaranteed.

The hydraulic test on the reactor pressure vessel of the

Gemeinschaftskernkraftwerk Tullnerfeld in Zwentendorf was started

in December 1975, but had to be interrupted at a pressure of 44 bars

due to a leakage of an instrument penetration in the vessel. In

January 1976 the hydraulic test was resumed and could be successfully

completed at 115 bars.

231

The acoustic emission analysis performed in conjunction with

this hydraulic test was carried out by Exxon Nuclear Company, Inc.,

Richland, USA. At the beginning of the acoustic emission analysis

some problems arised due to little noises of leaking fittings, dis-

turbing the emitted signals. After eliminating this leakages the

acoustic emission test could be finished without further trouble.

According to the final report of Exxon, the reactor pressure

vessel is virtual flawless, apart from some little emitting sources

(Grade 1) insignificant to structural integrity. Besides of this

for the safety of the reactor pressure vessel important statementst

the possibility of finding little leakages on fittings etc may be

a further advantage of the new testing method. The Austrian autho-

rity is therefore of the opinion, that for all further nuclear po-

wer plants in Austria with reactor pressure vessels this method

should be applied. It is considered, that in the case of a success-

ful development of the acoustic emission analysis for application

in nuclear service, this Lmethod could be imposed for the inspecting

of the reactor pressure vessel during service. This method could

then replace a good deal of the existing NDT-methods.

232

In service insiection of the VVR-S reactor

F.Jonak, L.Kaisler

Nuclear Research Institute, Fez

ABSTRACT

ISI results of the VVRS reactor in Rez, CSSR are described.

Details are given in connection to equipment, procedures used

as well as results obtained. Recommendations are prepared for

future inspections.

A research reactor of an initial thermal power of 2 IW'

is in operation in the Nuclear Research Institute since the

year 1557 . In 1974 a gradual reconstruction of this reactor

was started to increase its power to 10 'iW, which involves

changes in the reactor core, fuel, cooling circuits, pro-

tection and safety systems, etc.

In connection with the reconstruction new principles

were also applied to secure an operational safety of the

nuclear reactor and primary cooling circuit. New measures

also include the introduction of a system of in-service

inspection as a part of the quality assurance program. In

the first stage the state of the reactor vessel was checked.

The inspected vessel

The reactor vessel is shown in Fig. 1. It is made

from aluminium alloy SAV-1. The thicknesses of individual

parts vary from 12 to 20 mm. A cylindrical outer shell and

an inner shell excentrically positioned containing the re-

actor core are welded to the common disk-shaped bottom with

inlet and outlet fillers of the cooling circuit. Horizontal

233

ex rr.f..nt. chane'."s an a t.'.r.a.. "·,.olun,. &.are: .o.'.uxI'teii c o','

cantr-.i.cally ., ro'.u.d, lthe re a:., tor .Co..

The reaictor' is also.. e Im.ped;. wit several expari.^'' " L'

ver tical cnannrei s

Presrva,.=ion of the ine!:pe. ta .on

The 'iahe to prepare the inspection was very snort -

2 months on-l.., It .vas. necessary therefore?, ^tIo e .the n&&.a

wh:ich c.oulo be quicklyiY provliaed .,and ,heare vwas no ti.i to

elbaor 4 tde in detai.ls a method to lear-n i.t o'.d to t.r.'.i..fl ir-'

sorrnnelr The esserntial aethod. used. was visual ilspection .,i.

closed-loop television.. eelds ca internal 've.ss:e par ti-

were a:lso. stu.die, by radorbaphy for orien:taT.4o.. It wa;.

not oossible to oti r to operate cwn e quip,.ent of' ,^t' ,-,

loop television and to master its operation. Therefo're i.et-i

a.se .abLis for common uses we'ie applied.

ThIe preparation itself was con.centrated on *the d-.es.

and .:roduc:tion of a convenient maniapulator, ilaLi..n.ation,

and determination of a method for a durab.e reacord with

a possibility of a repeated reproduction of tle indJ.catlcri

point.

ian ioulator

A simple manipulator was used with maanual con.troi, which

is shown schematically in Fig. 2. In. addition to a verti-

cal displacement of the mast with a head holding the tle--

vision camera and headlamps or an X-ray tube tne mLsa.pul;.a-

tor also performs three independent motions:

- rotation concentrically, with the axis of the outer

shell of the vessel, in which motion the possibility of

anuaal and motor di.splacements is used by .aontin 'the1 -L-

i.ipulator on a toothed ring for the rotary reactor vesea,.

head,

234

- displacement on rails within the toothed vessel

head ring, and

- rotation of the basic manipulator frame with the

mast on a special turn ring.

The latter motion is necessary owing to the excentric

position of the inner vessel shell with respect to the

outer vessel shell. Except the rotation by means of the

toothed ring of the reactor cover all other motions are only

manual. The vertical displacement of the mast is carried out by

means of ropes and a windlass with a ratchet and a pawl.

For visual inspection a head with four headlamps and

a television camera ( Fig. 3) is mounted on the flange of

the mast. The headlamps are of the common automotive halogen

fog design, 12 V, 50 W, provided with a ground foil for

beam dispersion. The headlamps are switched on independently

from the operator a panel.

The motion of the head with the television camera in

the reactor vessel was controlled with a second television

loop the camera of which was positioned on the upper mani-

pulator part ( Fig. 4 ).

During radiography tests an X-ray tube is mounted on

the flange of the mast of the manipulator. The film casette

is attached to a sheet segment on the lower end of a bar

which is suspended by means of an adjustable cross beam in

a convenient groove in an extension of the lower horizontal

beams of the manipulator ( Fig. 2 ) and it is balanced witn

mild tension by means of a rope.

C.osed-loop television assembly

Different closed-loop televisions were used.

At the first stage or the visual inspection television

cameras, line ampLifiers, and monitors or Hungarian produc-

235

tion were used, in combination witn a monitor adapted

from a common purtable television set or Czechoslovak pro-

duction ( Minitesla ). At this stage pictures were taken

photographically from the monitor since no videorecorder

was available.

Later a Japanese television assembly was successfully

provided including a videorecorder SONY and the whole con-

trol was repeated to obtain a videorecord. At the same time

the control procedure was recorded with a report camera

and the SONY videorecorder.

During the general inspection two monitors were situa-

ted directly on the reactor platform for information of the

manipulator operators. The main monitors were situated at

a greater distance in a partially darkened area.

At a repeated inspection involving videorecording the

monitors were positioned op the reactor platform only.

Electric supply

The television cameras during the first phase of the

inspection and the headlamps to illuminate the inspected

space were supplied from 12 V accumulators. This alterna-

tive was convenient since it excluded a possibility of an

injury by electric current during the work with the manipu-

.lator.

Other equipment was supplied from 220 V, 50 lz mains.

Photographic recording

The surface indications:were recorded with a photoca-

mera NIKKON F on a stand at a distance of approximately 80 cm

from the monitor CRT. The cine-film ORWO NP 20 was used at

diaphragm 4, exposition 1/30 sec.

236

Doses of ionizing radiation

During the inspection tt reactor was shut down for appro-

ximately one month and fuel was removed. Since the instru-

mentation for operational control was not watertignt water

was discharged down to the level of horizontal channels,

i.e. to the middle plane of the reactor core.

Before manipulator installation dosimetrical measure-

ments were carried out and following values were measured:

- at the upper edge of the reactor core: 5 R/hr

- at the level of the reactor platform: above the middle

of the reactor core - 180 mR/hr ; at the vessel edge - 10

to 40 mR/hr.

The delay on the reactor platform was determined to

3 hr/day at working for several successive days. Since the

work required a longer' period the manipulator personnel

was changed.

The inspection of the vessel including the installation

and dismantling of the testing equipment lasted from Novem-

ber 24 to December 12, 1975, totally 15 working days. Du-

ring this time the daily dose was exceeded by one worker

but the weekly admissible dose was not exceeded.

Scope of the inspection

The scope of the visual inspection was limited by tech-

nical possibilities, which were given by the configuration

of the vessel parts and by the size of the head with the

television camera and the headlamps. They enabled an inspec-

tion up to the level of the upper part of the reactor core

and the thermal column. The inner vessel.could be inspected

on its whole inner-surface, the outer vessel only within

the inner pazt of approximately 2/3 of the excentric space

237

between the two vessels. The outer surface of the inner

vessel could be inspected to the same extent.

The radiography test was carried out only for orienta-

tion on a part of the welds of the inner vessel.

Indication system for the inspected places

Initially, a coordinate system was considered which

would agree with manipulator motions and which would be

indicated 'simultaneously with the inspected surface on the

monitor ,r.d hence also on the videorecord or photographic

record. It appeared, however, as unreal to be done during

such a short period for preparing the inspection. As a

substitution individual welds were denoted by letters which

were attached on the CRT tube during photographing (Fig.5).

In case of a contact of two welds the letters denoting

both welds were attached ( Fig. 6 ). During the inspection

the procedure of work was noted.in the diary together with

the direction of the inspection of individual welds.

By videorecording labelling welds on the record was

not possible. The procedure of the inspection was indicated

again in the diary by using the same labelling of welds.

The direction of the inspection can be also determined

by the displacement observed on. the CRT tube.

The same labelling of welds combined with common me-

thods of radiogram labelling was used in testing by irra-

diation.

Procedure of work in the visual inspection

Before the start of the visual inspection the selected

principles of the television loop and illumination were

tested at first in laboratory. Especially following condi-

tions were iesied with the provisionally installed head

238

with the camera and the headlamps ( Fig. 3 ) on a sheet

from alluminium alloy:

- convenient illumination method ( the necessity of

using a ground foil on the headlamps was found ),

- the effect of illumination intensity on the contrast

and discernibility of the indications,

- the effect of the direction of illumination on the

discernibility,

- the difference between the indication distinction

by television transmission and by mere eye,

- techniques of photographing the CRT picture.

The photo6raphic record of the television indication

on a sheet in laboratory conditions is shown in Fig. 7.

If a convenient contrast is adjusted the distinction is

significantly better than by mere eye. The cross in Fig. 7,

which was indicated on the observed sheet with pencil as

well as other indications ( scratches and corrosion spots

on the sheet ) were discernible only difficultly with mere

eye and. the discernibility depended on the direction of ob-

servation. Neither illumination direction nor cross po-

sition influenced significantly the discernibility by te-

levision.

None of the television cameras used was equipped with

the telecontrol of the objective. Therefore, it was necessa-

ry to carry out focusing and diaphragm adjusting in the

position of the camera above the reactor platform. After

focusing the manipulator had to be displaced so that the

distance from the abserved surface was not changed, i.e.

vertically and concentrically with the observed surface.

Since the inner and outer vessels are connected with

239

armouring ridges in the upper part the inspection could be

carried out only by small sections and the focusing had

to be adjusted with respect to the possibility of a passage

of the head with the camera and the headlamps into the

lower vessel part.

During the vessel inspection the manipulator was con-

trolled by an operator from the reactor platform.'He

followed the hints of the worker observing the surface

state in the monitor, recorded in the diary and labelled

the welds on the CET. This operator also gave instructions

for photographing.

During the inspection the illumination of the inspect-

ed surface was changed . It enabled to determine whether

unevenness appearing on the surface or situated below

the surface is indicated.

Procedure of work duringradiography

'The irradiation was carried out with the X-ray appa-

ratus Super Liliput. The control panel was situated du-

ring the irradiation on a walkway to the reactor plat-

form at a sufficient distance from the vessel. The X-ray

apparatus was rotary mounted on the flange of the manipu-

lator mast so that the apparatus axis was in a vertical

position a cable connection being on the lower part. The

lower part of the apparatus with the cable loop was

wrapped in a polyethylene foil to avoid occasional oil

drops into the reactor vessel but also to avoid apparatus

contamination at a contact with the reactor core.

The first operation step was adjusting the distance

of the apparatus from the irradiated weld by means of a

measuring rod as well as the indication of the vertical

240

position on the mast. Then all manipulator displacements,

except vertical one, were blocked. After pulling the appa-

ratus above the platform, level the measuring rod was removed,

the apparatus was lowered to the denoted level and heating

was switched on. Similarly the depth was adjusted on the bar

for the film. After pulling out the X-ray film casette with

labelling was mounted on the suspension bar. Then the sus-

pension bar with the casette was suspended as quick as

possible on the manipulator and balanced by pulling the

tape out of the upper bar end. Immediately after the with-

drawal of the worker performing this operation the exposi-

tion was switched on. After its end the rod with the casette

was removed immediately from the reactor. Since the radio-

graphy was carried out only on the vertical weld it was

possible to adjust subsequent positions of the X-ray appa-

ratus and the casette suspension bar by corresponding film

size.

By this method the delay of the film in the vessel

was reduced.

Results of the visual inspection

Following main conclusions were drawn froum the results

obtained by visual inspection from the viewpoint of the

indications found:

1. lignlricant corrosion indications were found in some

places on tne vessel surface, especially on welds ( Fig. 5 )

tne surface of whlcn was not machined.

2. Arter cleaning corrosion products ( by ecening

with 2 % dau 3 , rinsing with distilled water and cleaning

mechanically by tampons ) corrosion craters were found the

depth of which was evaluated up to 5 mm maximum in a single

case.

241

3.Since the deeper corrosion craters occur only in the

region of weld reinforcement and no tendency to their joining

was revealed on the inspected places it .may be assumed

that there is no immediate danger of failure since the vessel

load is small.

4. It has been confirmed by the inspection that the in-

spection was useful and repeated inspection must be prepared

after 2 to 3 years of continued operation with respecting

following principles:

- the inspected region should be extended to the

places which were not inspected up now,

- corrosion products should be removed by etching and

by mechanical cleaning before the.inspection as a part

of its preparing,

- during the inspection the welds in the places. of

significant corrosion craters should be machined gradually

by telemechanical. treatment at simultaneous measurement

of feed and at observing the television to determine the

depth of the corrosion attack,

5. During the preparations of the next in service in-

spection the discernibility of the indications should be

solved systematically , as well as their evaluation, tech-

nical conditions of the inspection from the manipulator

viewpoint, recording the inspected place, illumination and

television loop1 The state of the preparation, equipment

function, and personnel qualification should be checked in

simulated conditions on a model of the inspected object.

Generally, the application of a remote visual .inspection

by a television loop in testing radioactive objects offers

following essential advantages of the method :

242

a) Possibility of observing a remote place at simulta-

neous control of picture quality with a possibility of aa-

justing brightness and contrast in a wide range, convenient

illumination and taking representative photographic pictu-

res or videorecords.

b) Possibility of working in a strongly radioactive

environment without influencing the indication discernibi-

lity.

c) Possibility of .an immediate and additional evalua-

tion as well as the possibility to compare the changes at

repeated inspections on the same place.

d) Relatively good discernibility of the surface state,

in some cases significantly better than with mere eye. By

using convenient optics to attain satisfactory primary

magnification a good discernibility can be attained even

for the indications of crack type in the direction of the

CRT lines.

The Figs. 8,9, and 10 show the discernibility of the

surface state on an example of a standard with mean geo-

metrical roughness 1.6 /u and on cracks on the surface.of

test welds.

Generally valid recommendations for the development

.of methods, manipulators,and instrumentation for remote

visual inspection can be formulated by the following

way:

A. A manipulator for the inspection with a suffici-

ently rigid design with continuous displacements in all

working positions.

B. Minimum sizes of the camera and the illuminating

lights with variable illumination direction or also inten-

sity.243

C. Remote focusing and adjustin, of the camera objective

in a wide range and a possibility of working with the trans-

mission of the primary picture by a tilting mirror at ad-

justable angles from 0 to 45°.

D. Displacement of the television head Independently

on the position of the manipulator itself.

.* E. Watertightness of the television camera and the illu-

minating lights.

. . Determination of the position of the observed ob-

ject in dependence on the coordinate system which agrees

with the manipulator disp;acemnts. These coordinates should

be indicated on the monitor, photopicture, and videorecord.

G.. Carry out systematic work to evaluate the discerni-

bility of the indication by black-white television tech-

nique and to determine the maximum discernibiiiy of the

crack-type defects.

H. Check and in affirmative case apply the fill num-

ber Of 625 lines instead of the usual 312 lines of the

closed-loop television if it leads to a corresponding in-

crease of the discernibility of the crack-type defects.

I. Check the usefulness of the application of coloured

television.

,K. lalorra-e en "ltias of typical iridicaions" for usie

checked system having the optimum discernibility for per-

sonnel training. to carry out the evaluation at a remote

visual inspection by television. The atlas should be ela-

borated in three parallel variants: for direct photography,

of the indications, photography on the monitor, and for vi-

deorecords.

244

Experiences from the X-ray.radiography

The carried out radiograms showed a good discernibili-

ty of the indications in irradiated welds and a negligible

effect of the emitted radiation of the inspected object on

the film. It is convenient to prepare the inspection by

X-ray radiography in a larger extent in the given case of

the research reactor. An X-ray apparatus of the least size

possible with a short focal length must be secured for this

purpose.

In a general sense the usefulness of the radiograpnicwas confirmed

objects. According to further analyses some trends in the

development of the inspection of radioactive objects by

radiography can be formulated:

A. Determination of. the radiation spectra of the used

nuclear facilities, especially those of primary circuits

of nuclear reactors. Determinations of the conditions and

limitations for radiographic tests with X-ray tubes and with

other sources. Selection and confirmation of the convenabi-

lity of using films which are less sensitive to the spec-

trum components of the radiation emitted by tested objects

and also the convenability of using converters.

B. Confirmation of the possibility to use X-ray tele-

vision technique in cases of an acces from both sides to

the inspected object. Development of the instrumentation

and determination of the method and conditions of individual

application.

Conclusion

On basis of the experiences obtained by the inspection

of the reactor vessel of the VVR-S reactor and by the study

of the experiences in other laboratories following gehe-

245

ral conclusions can be recommended from the viewpoint of

non-destructive defectoscopy:

1. the methoa of visLal inspection by closed-loop te-

levision can be conveniently comprised into the methods or

non-destructive vesting as a special method and in practi-

cal application following principles should be met:

a) the workers applying the method of visual inspec-

tion should be trained for this method and they should

prove corresponding knov'leCec necessary for its applica-

tion and for the evaluation of indications by an examina-

tion;

b) the method is applied with the instruments and

equipment which-has been tested for the given purpose from

the viewpoint of a sufficient discernibility of indications;

c) a set of typical irniications should be available

for the evaluation of indications ;

d) the application of a renote' visual inspection is

convenient especially at a repeated application to compare

the cnanges during the operatioeal exploitation of the

investigated objects

e) providing a rec.ord of observed surfaces with a possi-

bility of recording the placers f any indication is a

condition of a puaposeful application.

2. The method of the radiographic study of objects must

be considered as a special modification of a general ra-

diography method and following principles should be met

at its application:

a) personnel qualification for general radiography

should be proved including the training for special appli-

cations in radioactive cbjects ;

246

b) application of the method should be confirmed for

a given case theoretically and also experimentally ,if

possible, at least in simulated conditions.

It can be assumed that necessary conditions for a con-

venient application of both methods and sufficient assump-

tions of the reliability of the conclusions of the inspec-

tion for a safe operation of nuclear facilities will be

given at keeping the principles mentioned. In this way

assumptions will be also formed for a gradual introduction

of both-methods into the.common exploitation from the re-

search and special spheres.

Fig.l Vertikal section of the reactor WR-S: 1-separator, 2-fuel section,3-expeller with an air cavity, 4-supporting grid, 5-loop channel, 6-channelwith a control rod, 7-control system channels, 8-confuser, 9-beryllium re-

flektor block, 10-beryllium side expeller,ll-supporting platform of the con-trol system, 12-inner vessel.

Fig. 2 Scheme of the manipulator design

navijAk- winch

ozuben' segment - toothed segment of the reactor cover

pojezdovA draiha - travelling path

stolek pfiEn6ho pojeadu - stage of the transversal travel

todna - turnring

nosnA konstrukce - supporting frame

voditka trubky - mast guides

zavesnA trubka - suspension mast with a flange

plosina reaktoru - reactor platform

gachta reaktoru - reactor shaft

nosnik rtg.film'l - beam for X-ray films

plist nAdoby - vessel shell

vnitini nAdoba reaktoru - inner reactor vessel

247

Pig.3 Head with a television camera and headlamps at tests in the laboratory.

Fig. 4 Manipulator in working position above the reactor.

Fig.5 Corrosion products on the welds of the outer vessel.

Fig.6 Corrosion craters on the surface of a cleaned weld.

Fig. 7 Indication on a sheet in laboratory conditions.

Fig.8 Standard surface with mean geometrical roughness 1.6 jum

Fig.9 Television indication on cracks on the test weld ( ferritic steel )

in laboratory.

cFig.10 Diretly photographed indication on the same cracks as in Fig.9 on

the test weld in laboratory.

248

'2320

11

H'~·1

gora. -1 Li -1

tB

O.

0

12

;

//I

!t 8 7Fig. 1

249

1

-. p

ZAVESNA TRUBKA

. A '°S PRIRUBO

REAKTORU

d

I0

Pig. 2

250

Fg. 9

Fig. 10

254

RECOMMENDATIONS OF SPECIALISTS MEETING AT PLZEN, CSSR, ON

SURVEILLANCE OF REACTOR PRESSURE VESSELS FOR IRRADIATION DAMAGE

After a comprehensive review and comparison of surveillance

programmes, status and results, from several nations it is

recommended:

1. That it is important to utilize surveillance for irradia-

tion and other service degradation of reactor pressure

vessel material as one part of the broader consideration

of reactor vessel reliability which also includes, the

properties of reactor vessel steels, the means for de-

tecting flaws, other in-service inspection procedures,

design and operational considerations, etc.

2; The supoort of achieving greater reactor pressure vessel

reliability through the pursuit of the following tasks

which are listed by order of priority in two categories:

a. Tasks related directly to surveillance:

1/ encourage improvement of neutron dbsimetry for des-

cribing neutron exposure in surveillance programs for

selecting life time peak fluences on the vessel wall;

2/ assure criteria for new surveillance programs which

enforce selection of capsule location at peak influence

location and selection of specimens of vessel component

projected to become the weakest in terms of vessel

reliability;

3/ promote improvements in techniques for in-reactor cap-

sule design and application for surveillance;

4/ recommend collection and organization of all surveillance

and research data in terms of composition and response to

neutron irradiation;

255 ,

5/ promote the advancement of criteria for reducing overcon-

servatisms which may be attributed to currelit methods of

surveillance in a fracture analysis;

6/ surveillance data can be augmented significantly by in-

clusion of a reference steel which may provide a useful

guide on the condition that the reference steel has been

investigated exhastively, subjected to well defined ir-

radiation conditions, is more sensitive to irradiation

than the steel used for the pressure vessel and that the

scatter in properties measured is within reasonable limits.

Such use of a reference steel may be especially helpful

when the neutron spectra in a reactor are not well

established;

7/ in support of surveillance programs it is recommended

that an adequate supply of archive material representing

the vessel be obtained and stored for the possibility of

future need;

8/ encourage development of improved surveillance specimens

based on quantitative fracture mechanics and by correla-

tion or other means advance our understanding of Charpy

V-notch toughness data or data from notched pieces of

similar size since these will continue as a primary re-

ference for many years.

b. Tasks supportin_ g the improvement of the resistance of

the vessel materials to irradiation induced chances:

1/-. promote the advancement of limits on residual element

content and microstructure which may tend to reduce

fracture toughness of reactor vessel steels and also

encourage research to better understand the phenomena

involved;

2/ promote understanding of the size effect in fracture

mechanics application to pressure vesselsT especially

for steels of low upper shelf toughness as well as for

vessel steel behavior at service temperatures;

256

3/ promote fluller understaarLding of the implications of

the gradient in toughness through a reactor vessel

wall;

4/ encourage development of data on stainless steels used

for construction of internal structural components with

the goal of determining potential failure modes which

might require surveillance in the future.

257

List of Particia.nts

Austria

1. Mr. K. Wischin

2. Mr. Rudolf Steiner

3. Mr. J. Zeman

Bundesminioteriumn flir Bautenund TechnjikStubenring 1A-1010 Vienna

OsterreichischeStudiengesellschaft fUrAtomenergieLenaugasse 10A-1080 Vienna

Technischer UberwachungsvereinVienna

4. Mr. H. Theiretzbacher

5. Mr. T. Mager

6. Mr. Philippe van Asbroeck

7. Mr. Ginther Rottenberg

ManagerMaterials EngineeringLa Soci6t6 Westinghouse73 rue de StalleB-1180 Bruxelles

CEN-SCK; 200 Boeretang2400 Mol; Belgium

,

C.S.S.R.

8. Mr. Radislav Filip SkodaPlzen

National Corporation

9.- Mr. Pavel Mrkous

10. Mr. Milan Brumovsky

11. Mr. Jirl Prepechal

12. Mr. Stanislav Stepanek it -

It -

_ 11 -

r »

Ir"

13.-l

Mr. Antonfn Urban

14. Mr. Josef Sulc

15. Mr. Karel Mazanec Corresp. Member of Academy of ScienceTechnical University Ostrava

259

16. Mr. Stanislav Havel DirectorNuclear Research InstituteRez near Prague

Nuclear Research InstituteRez near Prague

17. Mr. Jir£ Cervdsek

18. Mr. Miroslav Vacek

19. Mr. Ladislav Kaisler -_ -

20. Mr. Jan Korycdnek

21. Mr. Josef Homola

22. Mrs. Jirina Davidova

3enmark

23. Mr. Arved Nielsen

Czech Technical UniversityFaculty of Nuclear Physicsand EngineeringBrehovS 7, Prague 1

Nuclear Power Plant A-1Jaslovsk6 Bohunice

Atomic Energy CommissionSlezsks 9Prague 2

The Atomic lEergy Commission'sResearch Establishment RisfDK-4000 Roskilde

Fed. Rep. of Germany

24. Mr. Klaus Peter INW-TUV, Hagen, Buscheystrasse 30

Finland

25. Mr, Jarl Forsten Technical ResearchFinlandWLnnrotinkatu 37SF-00180 Helsinki

Centre of

18

WFance

26. Mr. P. Petrequin

27. Mr. B. Barrachin

Departement de Technologie duCentre d'Etudes Nucleaires deSaclay, B.P. No. 291190 Gif-sur-Yvette

Departement du Surete NuoleaireService d'Etudes Technioues deSureteCommissariat , l'Energie AtomiqueCentre d'Etudes Nuclaires de SaclayB.P. No. 291190 Gif-sur-Yvette

260

India

28. Mr. K. S. Sivaramakrishnan Radiometallurgy Section, MetallurgyDivisionBhabha Atomic Research CentreTrombay, Bombay 00 085Technical Services SuperintendantTaraput Atomic Power StationDist. Thana, Maharashtra 401504

Italy

29. Gius6ppe Crocenzi

30. Vittorio Vaccari

CNE2, National Committee forNuclear Energy, Roma

A.N.C.C./ Associazione Nazionaleper il Controllo della Combustione/Sorvizio Impianti NucleairiVia Depretis b&o Roma

Japan

31. Mr. S. Miyazono Senior Research EngineerChief of Mechanical Strength andStructure LaboratoryJapan Atomic Energy Research InstituteTokai Research EstablishmentTokai-mura, Naka-gan, Ibaraki-ken

Netherlands

32. Mr. B. Korff

33. Mr. Louis Bernard Dufour

34. Mr. Eduard Lybrink

Service for Steam-EngineeringThe Hague

N.V. Kema Utrechtseweg 310Arnhein, Holland

Reactor Centrum NederlandWesterduinweg 3Petten, Holland

Spain

35. Mr* Roberto RodriguezSolano

Junta de Energia NuclearCiudad UniversitariaMadrid-3

261

Sweden

36. Mr. G. Oestberg Lund Institute of TechnologyS-220 07 Lund 7, Sweden

Switzerland

37. Mr. Ernst Sandona

38. Mr. G. Prantl

Chief, Quality AssuranceKernkraftwerk BeznanCII-5312 Dottingen

Institute feddral de recherchesen matirre de reacteursCH-5303 Wirenlingen

United Kingdom

39. Mr. A. Cowan

40. Mr. Darleston

Risley Engineering and MaterialsLaboratoryU.K. Atomic Energy AuthorityRisley, Warrington, Lanes

CEGB

United States of America

41. Mr. L. E. Steele

42. Mr.'Thomas Keenan

43. Mr. Charles Z. Serpan

44. 'r. John Koziol

45. Mr. Arthur L. Lowe

46. Mr. R. A. Wullaert

.Thermostructural Materials BranchEngineering Materials DivisionNaval Research LaboratoryWashington D.C. 20375

Yankee Atomic Electric Co.20, Turnpike RoadWestboro, Ma. 01587

Metallurgy and Materials BranchDivision of Reactor Safety ResearchUS Nuclear Regulatory CommissionMail Stop G-158Washington D.C. 20555

Combustion Engineering, Inc.1000 Prospect Hill RoadWindsor, Ct. 06095

Babcock and Wilcox Co.P.O. Box 1260Lynchburg, Va. 24505

Fracture Control Corp.330 South Kellog AvenueGoleta, Cal. 93017

262

47. Mr. James Perrin Battele Memorial505 King AvenueColumbus, Ohio

Institute

43201

OECD/EA

48. Mr. N. de Boer OECD/NEA38, Boulevard Suchet75016 ParisFrance

CEC.^_

49. Mr. H. A. Maurer Direction gendrale des affairesindustrielles et technologiquesCommission of European Communities200 rue de la Lai1040 BruxellesBelgium

IAEA

50. Mr. I. S. Zheludev

51. Mr. I. K. Terentiev

Deputy Director GeneralInternational Atomic Energy AgencyKErntner Ring 11P.O. Box 590A-1011 ViennaAustria

Division of Nuclear Power and ReactorsInternational Atomic Energy Agency

263

reactor pressure vessel surveillance

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