ML20058A865
| ML20058A865 | |
| Person / Time | |
|---|---|
| Issue date: | 01/31/1974 |
| From: | Serpan C Advisory Committee on Reactor Safeguards |
| To: | |
| References | |
| ACRS-GENERAL, WASH-1285, NUDOCS 8207220127 | |
| Download: ML20058A865 (88) | |
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WASH-1285 l
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REPORT ON THE
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INTEGRITY OF REACTOR VESSELS FOR LIGHT-WATER POWER REACTORS i
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JANUARY 1974 l
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THE ADVISORY COMMITTEE ON I
REACTOR SAFEGUARDS
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t LEG AL NOTICE This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commisuon. nor any of their employees, nor any of their contractors, subcontrac tors, or their employees, makes any warranty, espress or imphed, or assumes any legal liabihty or responubihty for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe pnvately owned rights.
I For sale by the Superintendent of Documents, U.S. Government Printing Offlee. Washington, D.C.3M02 - Price 81.32
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED S'TATES ATOMIC ENERGY COMMISSION
' WASHINGTON. D.C. 20545 JAN 141974 Honorable Dixy Ice Ray Chaiman U. S. Atenic Energy Ccrmission Washington, D. C.
20545
Subject:
I17fEGRITI OF REACIOR EMR MR IJGTf>@.'IER IGER REACIOPS
Dear Dr. Ray:
In a letter, dated I'.ovceber 24, 1965, to the Atcxtic Energy Ccrrission, l
the Mvisory Cemittee on Ecactor Safcauards exoressed the belief that the proMhility of a mjor reactor vessel failure was very Icxt, but roccimunded that work be done to reduce the probability of failure still further, and also to develoo reans for avlinrating the consecuences of I
such a failure.
Since that letter was written, irprovercnts have been made in reactor vessel technolcgy, and understanding of the bchavior of cressure vcasels has been creatly enhanced try extensive research and develog.rnt by the AEC, by industry, and by others. The Corrittee has reviscx1 the present state of the art and has cut its findings and rcccrrendations in the fom of a recort, " Integrity of Icactor Vcssels for Licht '.?ater Ibder Peactors",
a copy of which is attached. In the process of prenaring the recort, the Ccanittee nut with representatives of nuny organizations,. and also was assistcx1 by its consultants.
The primary purposes of the revicw were to evaluate the adeauacy of current requircments relating to light *ater pcuer reactor vessels and, as far as practical with existing data, to estirate the probability of disruptive failure of a reactor vessel.
'Ihe report incitries a critical revica of currcnt construction and coera-ting practices relating to reactor vessels. In the course of the revis7, the Ctzmtittee identified several areas where ir.rovancnts in current practice are desirable, and the report includes snecific reccrrendations for supolarcntary recuircrents over and above the rcxmircrcnts of the AS'C Boiler and Pressure Vessel Code.
l, llanorable Dixy Ice Ray JAN 141974 l
Detemination of the probability of a reactor vessel failure presents a probicn because the nteber of reactor vessels or vessel-ycars is too sm,11 to pcmit valid statistical inferences, Wercyts the truch greater body of statistics coverina non-ntclear vessels includes maarf vessels of poor or unkncren cinlity, he Ccr:nittee has revicecd statistics of
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defects in conventicnal vessels on the basis of data rublished in the United States, the Unitcd Kingdom, and the Federal Pc[nblic of Ccrmrr/.
Ec OIntittee also rade direct inquirics of U. S. insurance can'nies and vessel ranufacturcrs and, toccther with the Directorate of Pcactor Standards, obtained a largo cuantity of U. S. boilcr information. We Cormtittee concitrics that the pro'moility of disruntive failure of U. S.
boiler pressure vessels (non-nuclear) constructed to carcrcial staMards, and conventionally onerated, is less than 10-5 ocr vessel-year (less than onc disrtr;tive failure per 100,000 vessel-years).
%c remrt proceeds to acrrare toiler vessel practico with reactor vessel practice, inclnling differenccc in design, rcaterials, fabrica-tion, insmetion during fabrication, oncrating conditions, anl inservice insucction. On balonce, the Cccrittee concludes that t'n orcbability of disrtptive failure of a reactor vessel is ruch lcrer than that esti atal for non-nuclear vessels. It is difficult to cruantif the differcmco f
botueen the tv:o classes of vessels, but the Corrittee helieves that there is reasonable assurance thtt tha probability of disruntive failure in norml service, cf a reactor vcssal dccicnal, constructol erd ocorated l
in accordanca rith ZEI: Code Sections III and XI is less than 10 0 por vessel-year (less than one disrtative failure per nillion vessel-years).
.The rercrt recxr.mnds continued surcrt of reactor vessel roccarch and develo.:,tcent, particni,rly in studying the pro crties of heavily irradiated pressure vcscel steal and in devoleping nei tecMicucs for inservice in-spection. We report also rcccmcnds that further stidy of rossible desien approaches to protect against tressure vassal failure cheuld Ec perfomil.
tlc Ccrcittee remrt a glics pri arily to vessels desianed and coerated in accordance with current st.nylards. W e last section of the renort nnhcs a rcccrrendation relating to early vessels that do not rcet current staMards.
%c Ccrrittee will davote continuin7 attention to the ratter of pressure vessel intcarity a:rl will. uke further roccrmndations for continu3d research and develo;mnt as a.rropriate.
Sincerely yours, U. R. Stratton ChairTin l
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WASH-1285 Uc-78 Report on the INTEGRITY OF REACTOR VESSF' S FOR LIGHT-WATER POWER REACTORS by The Advisory Comittee on Reactor Safeguards January 1974
CONSULTANTS The following consultants to ACRS participated in the Committee's review:
W. E. Cooper, Vice President and Technical Director Teledyne Materials Research H. T. Corten, Professor Department of Theoretical and Applied Mechanics University of Illinois A. W. Pense, Professor Department of Metallurgy and Materials Science Lehigh University C. A. Siebert, Professor Emeritus Department of Materials and Metallurgical Engineering University of Michigan A. S. Tetelman, Professor of Engineering Chairman, Materials Department University of California, Los Angeles W. E. Vesely, Staff Scientist JRB Associates J. D. Wilding, Consulting Engineer 39 Van Brackel Road Holmdel, New Jersey R. Maccary, Directorate of Reactor Licensing U. S. Atomic Energy Commission, also assisted the Committee.
Opinions expressed by the consultants and Mr. Maccary were the views of the individuals and not necessarily those of the organizations with which the individuals are associated.
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TABLE OF CONTENTS Page l
SECTION 1 I NT RO DUCT IO N......................................
1 1
1.1 Review of Previous ACRS Activity.............
I 1.2 Scope of Report 2
1.3 Limitations in Scope of Report 3
1.4 S ub c o mm i t t e e M e e t in g s........................
3 SECTION 2 CONSTRUCTION PRACTICES 5
2.1 Materials 5
h 2.2 Design.......................................
8 2.3 Fabrication..................................
14 2.4 Installation.................................
18 2.5 Relevant Work in Other Countries 19 2.6 Adequacy of Nuclear Vessel Construction Practice 19 2.7 Conclusions with Respect to Reactor Vessel Construction Practice......................
21 SECTION 3 POSSIBLE MODES OF VESSEL FAILURE..................
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3.1 S low C ra ck G row th............................
24 3.2 Ra pid C rack Pro paga tion......................
26 3.3 Application of Basic Failure Mechanisms to Reactor Vessels 27 l
3.4 Degradation of Toughness 30 i
3.5 Condi tions o f Ope ra t ion......................
30 3.6 Character and Location of Postulated Vessel Failures in No rma l Ope ra tion...............
31 3.7 Examples of Defects in Reactor Vessels 33 3.8 Accident Conditions that Could Cause Vessel Failure 37 SECTIG.9 4 OPERATIONAL CONSIDERATIONS 42 4.1 Operational Limits on Applied Loads 42 4.2 Environmental Influences on Ductile Response..
43 4.3 Preservice and Inservice Inspection..........
46 4.4 Failure-Limiting Provisions 51 i
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TABLE OF CONTENTS (cont.)
Page SECTION 5 PRESSURE VESSEL FAILURE STATISTICS AND FAILURE PROBABILITIES 54 5.1 Definition of Failure 54 5.2 Types of Disruptive Failure Considered in Assessment of Failure 56 5.3 Bases for Evaluation 57 5.4 United States Pressure Vessel Failure Statistics 58 5.5 Foreign Statistics 63 5.6 Comparison of United States and Foreign Pressure Vessel Failure Statistics 68 5.7 Probability of a Non-Nuclear Vessel j
Disruptive Failure--Committee Appraisal.....
68 5.8 Dif ferences Between Boiler Practice and Reactor Practice 71 5.9 Probability of Disruptive Failure of a Reactor Vessel--Committee Appraisal.........
74 5.10 Probability of Disruptive Failure Beyond the Capability of the Engineered Safety Features.
75 5. 1 1 S u mm a ry.......................................
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SECTION 6 CONCLUSIONS AND RECOMMENDATIONS 78 6.1 Some Limitations in Scope of Report 78 6.2 C u r re nt P ra c t ic e..............................
78 6.3 Probability of Disruptive Failure 79 6.4 Recommended Supplementary Requirements for Reactor Vessels 80 6.5 Reactor Vessels Not Covered by Report 83 6.6 Future Action by ACRS 85 iv
1 I NTRODUCTION This section reviews the long-standing interest of the ACRS in assur-ance of the integrity of reactor vessels for light-water power re-actors, and describes the purpose of this report.
1.1 Review of Previous ACRS Activity The very remote possibility of a reactor vessel failure has, from the beginning, been a matter of concern to the ACRS and others.
On November 23-24, 1965, an ACRS Committec-of-the-whole held a meeting to review the status of pressure vessel technology and factors that might affect reactor vessel integrity. At that time, a good under-standing of brittle fracture of ferritic steels had resulted from the work of the Naval Research Laboratory on nil-ductility temperature and from the work of others on fracture mechanics. The Committee had the benefit of presentations by representatives of the Allis-Chalmers Manufacturing Company, Babcock and Wilcox Company, Bethlehem Steel Company, Combustion Engineering Company, Commonwealth Edison Company, General Atomics, General Electric Company, International Nickel Company, Inc., Lessells and Associates, Lukens Steel Company, Massachusetts Institute of Technology, Oak Ridge National Laboratory, Southwest Re-search Institute, U. S. Naval Research Laboratory, U. S. Steel Company, Westinghouse Electric Corporation, and Yankee Atomic Electric Company.
Subsequent to the meeting, the ACRS released a letter to Dr. Seaborg dated November 24, 1965, suggesting, in part, ".
. that the industry and the AEC give still further attention to methods and details of stress analysis, to the development and implementation of improved methods of inspection during fabrication and vessel service life, and to the improvement of means for evaluating the f actors that may af fect the nil-ductility transitian temperature and the propagation of flaws during vessel life".
While this report is concerned primarily with assurance of vessel integrity, the Committee recognizes the importance of a clearer description of the consequences of varying degrees of vessel failure and of the second recommendation in the letter of November 24, 1965, "that means be developed to ameliorate the conse-quences of a major pressure vessel rupture".
A copy of the letter is appended to Sec. 1.
In 1967, the AEC established its Heavy Section Steel Technology Program (HSST) to explore a wide range of technological problems associated with thick-walled reactor vessels made of SA-533, Grade B, Class 2 steel and similar materials. Important parallel work includes the Pressure Vessel
s Research Committee (PVRC) Industrial Cooperative Program, the EEI-TVA Non-Destructive Test Development Program, and programs conducted by industry and other US agencies. The Committee has continued its interest in these programs and has directed the Reactor Pressure Vessels i
Subcommittee to review the current status of reactor vessel dependability.
1.2 Scope of Report During the past five years, the technology of pressure vessels and pressure vessel steels has been greatly advanced by the HSST program and other programs and investigations. Successive editions of Section III of the ASME Code
- and addition of Section XI have greatly enhanced the quality assurance of reactor vessels, as constructed and in service. The inservice examinations required by Section XI will contribute to greater vessel reliability in service by timely detection of conditions that could impair vessel quality.
The primary purpose of this report is:
1.
To attempt, as far as practical with existing data, to quantify the reliability of reactor vessels.
2.
To evaluate the adequacy of current requirements relating to reactor vessels.
i In support of these objectives, the report addresses the following:
1.
Current nuclear vessel construction practice (Sec. 2).
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Basic feilure mechanisms and their applicability to reactor vessels (Sec. 3).
3.
Operational Considerations (Sec. 4).
4.
Published and unpublished vessel failure statistics, and a comparison of nuclear and non-nuclear practice and condi-tions of service as they relate to vessel failure probability (Sec. 5).
5.
Conclusions and recommendations (Sec. 6).
Section III refers to Section III Nuclear Power Plant Components, Division 1 Metal Components;Section XI refers to Section XI Rules for Inservice Inspection of Nuclear Reactor Coolant Systems.
1.3 Limitations in Scope of Report The discussions and conclusions are limited to reactor vessels of the type and materials currently used in PWRs and BWRs.
The failure probability developed in Sec. 5 is for vessels operated in accordance with Design Specifications, it does not include failures from gross overstress s ;ch as conceivably could result from postulated accidents or sabotage of the reactor system. The probability of vessel failure from such accidents would have to be derived from system analysis to determine the probability of the accident and calculations to determine the resulting stresses in the vessel.
Irradiation effects must be considered in an estimate of failure rate, but irradiation data are still being accumulated. Through the next few years, the Committee believes no large vessel constructed to Section III of the Code will have been in service long enough to have suffered serious embrittlement by irradiation. The Committee's estimate of failure probability may require subsequent reappraisal for heavily irradiated vessels.
1.4 Subcommittee Meetings The Reactor Pressure Vessels Subcommittee held a series of meetings in preparation of this report between January 6,1971, and December 15, 1973.
During several of these meetings, the Subcommittee met with represent-atives of the Division of Reactor Development and Technology *, AEC National Laboratories, U. S. Government and privately-owned research organizations, and nuclear steam supply system vendors for discussions in the following areas:
Area Organization (s) 1.
HSST Program Results and Division of Reactor Development Anticipated Activities and Technology; Oak Ridge National Laboratory--Program Of fice 2.
Fracture Mechanics Westinghouse Electric Corporation Techniques 3.
EEI-TVA Non-Destructive Tennessee Valley Authority Test Development Program 4.
Current Acoustic Emission Battele Memorial Institute, Capabilities and Develop-Pacific Northwest Laboratory; ments Dunegan Research Corporation; Jersey Nuclear Corporation; Southwest Research Institute; Teledyne Materials Research; Westinghouse Electric Corporation
- Now Division of Reactor Research and Development i
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Area O rganiz ation(s) l 5.
Code-Required NDT vs General Electric Company I
Inspections Performed by l
Vessel Fabricators 6.
Fracture Behavior of U. S. Naval Research Laboratory Steels The Committee wishes to thank the many participants in these conferences, f
and also the following individuals for their assistance:
E. H. Phelps, U. S. Steel Company; and representatives of the Directorate of Licensing, U. S. Atomic Energy Commission.
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON 25, D.C.
NOV 24, 1965
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Honorable Glenn T. Seaborg
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Chairman U. S. Atomic Energy Commission Washington, D.C.
Subject:
Dear Dr. Seaborg:
The design of pressurized and boiling water nuclear power plants has undergone many improvements with regard to safety, improvements which markedly reduce the risk of significant radiation exposure to the public in the unlikely event of certain accidents or system failures in such reactors.
There is a facet of current pressurized and boiling water reactor design practice which should be recognized, however. Containment design is generally prediceted on the basis that a sudden, large-scale rupture of the reactor pressure vessel or its closure is in-credible. Reactor designers have supported this view by detailing the extreme care to be taken in design, fabrication, and inspection of a vessel, and by specifying pressurization only at temperatures above the nil ductility transition temperature. They further cite the excellent record for large pressure vessels which comply with the ASME Boiler and Pressure Vessel Code.
The Committee believes, with the industry, that the probability of a sudden major pressure vessel failure leading to breaching the con-tainment is very low.
Nevertheless, it seems desirable and possible to make some provisions in future designs against this very unlikely accident.
1.
To reduce further the already small probability of pressure vessel failure, the Committee suggests that the industry and the AEC give still further attention to methods and details of stress analy-sis, to the development and implementation of improved methods of inspection during fabrication and vessel service life, and to the improvement of means for evaluating the factors that may affect the nil ductility transition temperature and the propagation of flaws during vessel life.
l lionorable Glenn T. Seaborg Nov 24, 1965 i
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The ACRS also recommends that. means be developed to ameliorate the consequences of a major pres iure vessel rupture. Some possible approaches include:
(a)
Design to cope with pre isure buildup in the contain-ment and to assure that no iiternally generated missile can breach the containment.
(b)
Provide adequate core c ooling or flooding which will function reliably in spite c f vessel movement and rupture.
(c)
If breaching the containment cannot be precluded, pro-vide other means of preventing uncontrolled release of large quantities of radioactivity to the atmosphere.
In view of the very small probalility of pressure vessel rupture, the Committee reconfirms its b311ef that no undue hazard to the health and safety of the public exists, but. suggests that the orderly growth of the industry, with concomitant i ncrease in number, size, power level, and proximity of nuclear power reactors to large population centers will in the future make desirab.c, even prudent, incorporating in many reactors the design approaches whose development is recommended above.
Sincerely yours,
/s/
W. D. Manly Chairman
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2 CONSTRUCTION PRACTICES The quality level initially built into a nuclear reactor pressure vessel is an important f actor in the expected service reliability of the vessel.
This quality 1cvel is related directly to the construction practices used in the manufacture of the vessel, and an appraisal of the consequent reliability of the vessel can be developed only if the engineering pro-cesses and construction practices are fully defined. The framework for this definition of construction practice is provided in Section III, Nuclear Power Plant Components, of the ASME Boller and Pressure Vessel Code, which prescribes current requirements for materials, design, fabrication, examination, testing, inspection, and certification of pressure vessels used in nuclear power plants.
Section III of the Code was developed under the sponsorship of The Americen Society of Mechanical Engineers in recognition of the much more etringent safety requirements demanded of nuclear power plant components than were provided by the rules of Section I, Power Boilers, and Section "1II, Pressure Vessels, under which pressure vessels for fossil-fueled plants are constructed.
The Section III Code introduced, for the first time, many improvements and advances in design and construction practices that contribute significantly to enhancement of the quality level of pressure vessels used in nuclear power plants. These improvements and advances have been made in the areas of materials, design, fabrication, examinations, testing, and, in particular, in the newly developed measures for quality assurance.
2.1 Materials Although current Code-approved materials that may be used in the construction of reactor vessels permit a broad selection, reactor vessel manufacturers, in the earliest days of nuclear power plants, limited their se' 'etion to those relatively few low-alloy steels that had demonstrated accaptable ser-vice performance in vessels used in other industries and for which extensive data on physical properties were available. The general philosophy of employ-ing well-characterized materials of uniform properties and consistent behav-ior has been followed in the construction of reactor vessels now in service, as well as those being built today.
2.1.1 Steel Selection. The earlier and somewhat thinner-walled reactor ves-sel shells were constructed of stainless-steel-clad normalized ASME SA-212B plate (now designated SA-515 or SA-516), a carbon-manganese steel with a minimum tensile strength of 70,000 psi, and SA-105, SA-350, and SA-182 forg-ings. More recent heavier-walled reactor vessel shells have been constructed of SA-533* plate, a manganese-molybdenum-nickel steel in the quenched and tempered condition. This plate material, with a minimum tensile strength of
- Throughout the report, SA-533 means SA-533 Grade B Class 1 _ _ _ _ _ - _ _ _ _ _ _
80,000 psi, provides better strength and toughness than the SA-212 plate material. Moreover, the chemistry of this steel provides properties that result in minimal heat-to-heat variations in heat-treated plates, and relatively uniform properties in thick plates. The current widely used f
forging material is SA-508,* which is similar in strengh and toughness I
to SA-533 but has chromium additions and less manganese than SA-533.
These two more recently used plate and forging materials have evolved, generally from materials of leaner alloy content, over a period of about 15 years. The current materials, although stronger than their predeces-sors, have by metallurgical control of their composition attained improved toughness properties at the " upper shelf" or operating temperature range.
This improved toughness permits safe operation of pressure vessels with heavy section thicknesses over the range of operating stress levels nor-mally experienced by reactor vessels.
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2.1.2 Property Changes in Service. The physical properties of nuclear vessel materials are subject to possible change during the life of the structure under the influence of thermal, mechanical, and environmental operating conditions. These effects, such as strain aging, temper embrit-tlement, and irradiation embrittlement, may be strongly influenced by fabrication practices. In general, fabrication practices are controlled to minimize these effects and to assure that properties developed by fab-rication processes cannot be degraded unacceptably by operational condi-tions.
For example, since the upper service temperature limits in nuclear vessel operation are very low (about 6000F) when compared with the post-f abrication annealing temperature ( > 1000 F), the inservice operating tem-peratures have very little influence on strain-aging.
f Delayed or cold cracking (hydrogen cracking) in weldments of SA-533 and SA-508 materials may be avoided by following now well-established welding procedures that experience has shown to be adequate.
(Intergranular cracks in the heat af fected zone of the as-fabricated vessel are dis-cussed in Sec. 3.1.2.)
The internal surfaces of components fabricated from these two materials are generally clad with austenitic stainless steel or a nickel-base alloy. A unique condition reported for some SA-508 forgings is the incidence of small fissures underneath the stain-less steci cladding material deposited by certain weld cladding processes (see Sec. 3.7.1). Again, such fissuring is now controlled by using appropriate welding procedures. On the whole, changes in properties intro-duced during operation do not appear to affect the reliability of the vessel to an appreciable degree since the vessel steels used exhibit exceptional ductility and toughness, and cre relatively insensitive to conditions that could induce cracking.
- Throughout the report, SA-508 means SA-508 Class 2 _ _ _ _ _ _ _ _ _ _ _ _ -
The reactor vessel steels may undergo a decrease in toughness as a result of neutron fluence.
Under a specific combination of stress, temperature, and a service-induced flaw, the reduced toughness could decrease the margin available to control fracture behavior. The Code has therefore established specific rules for the fracture toughness properties of such materials and has provided guidance with respect to the margins needed to preclude fracture. The AEC requirements for fracture toughness as established in 10 CFR Pa rt 50, Appendix G, invoke the ASME Section III fracture tough-ness rules and supplement them with further material properties require-ments.
Consideration of additional conservatism with respect to allowable p re ssu re-t empe ra t u re relationships is discussed in Section 3.3.3.
The AEC requirements provide conse rvative margins with respect to the fracture toughness properties required under all reactor operating condi-tions.
These margins include particular consideration of the changes in material f racture toughness caused by neutron irradiation in the high-flux beltline region of reactor vessels. To monitor such changes, 10 CFR Part 50, Appendi x 11, requires that reactor vessel material surveillance specimens be withdrawn periodically f rom capsules in the vessel. The measured changes in toughness derived f rom tests on these specimens allow the operational pressure-temperature limitations during startup and shutdown to be modified so that reactor operation will always be in a range where fracture toughness is adequate. The fracture toughness properties of specimens not only verify the predicted changes that occur under the actual environmental conditions of a pa rticula r reac tor vessel, but, by positioning the specimens in a higher neut ron flux than seen by the vessel walls, also permit prediction of the vessel materials properties in advance of the anticipated changes.
As supporting evidence of the adequacy of the AEC requirements, experimental investigations performed under the Heavy Section Steel Technology Program, administered by the Oak Ridge National Laboratory, demonstrated that thick-walled vessels const ructed of SA-533 and SA-508 materials do exhibit sub-stantial toughness in the operational temperature range for nuclear pres-sure vessels. The data available on welds and heat-affected zones in these I
steels are more limited, but stt ength and toughness of welds made in accord-ance with Code requirements are essentially the same as for plate and forg-ing materials.
2.1.3 Possible Improvements. Recent developments in the metallurgy of reactor vessel steels have produced improvements in their overall properties. Lower-ing the residual alloy element content, particularly phosphorus and copper, in SA-533 and SA-508 materials has reduced their sensitivity to irradiation embrittlement. Although controlled specification limits on the re s id ua l alloy element content of materials used in the high neutron irradiation beltline regions of reactor vessels are not required by Code rules, vessel purchasing such controlled specification steels from mill manu f ac t u re rs a re suppliers. The use of steels and weld metal with low residual alloy con-inbegline regions of reactor vessels expected to incur neutron flu-tent ence of 10 or greater is recommended. If f racture toughness recovery _ _ _ _ _ _ -
from irradiation embrittlement by proposed in-situ annealing of reactor vessels can be demonstrated to be effective, this requirement will be less necessary, although still desirable.
Experience has also shown that the specification of maximum as well as minimum strength levels for steels, such as those used in reactor vessels, is useful in producing optimum toughness-strength combinations in service.
While the maximum strength level is not included in Code specifications for some materials (SA-508, for example), the trend in recent procure-ment practice is toward fixing the upper and lower strength limits on materials used in reactor vessels. Specification of an upper strength limit for all vessel steels should be required.
2.1.4 Recommended Current and Future Practice. The Committee believes that the materials currently used in nuclear pressure vessels provide the strengtl. and toughness necessary to ensure highly reliable operation of a nuclear vessel without either frangible or rapid low-energy ductile fracture during its service life. The currently used steels have substantial tol-erance to slow flaw growth before any form of unstable crack growth could occur.
Future design requirements for larger reactors may possibly call for the use of still heavier walled vessels and higher strength steels. Although several high strength materials such as SA-542 and SA-543 are of potential interest for nuclear vessel application, the AEC has not evaluated the
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acceptability of such materials. The historic practice of employing only J
well-characterized materials of uniform properties and consistent nuclear service behavior is considered by the Committee to be an essential factor 1
in establishing confidence in the reliability of nuclear pressure vessels.
The substitution of new materials for currently used ones, even if these new materials have an acceptable history of prior service in other appli-cations, is recommended only when their behavior in nuclear service is well understood and documented and the ability to control the properties has been fully established by extensive test programs.
2.2 Design Nuclear vessel design practices are based on detailed stress analyses of the integrally connected structural elements comprising the pressure retaining boundary. These consist of the shell structure, penetrations and nozzles, flanged openings and bolted connections, and the attached reactor internals and vessel supports. The Code requirements for vessel design acceptability are defined in Subarticle NB-3311 as follows:
"(a) The design shall be such that the requirements of Sub-articles NB-3100, General Design Rules, and NB-3200, Design by Analysis, are satisfied, and I
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"(b) The rules of this Subarticle are met.
(In cases of conflict between Subarticle NB-3200 and NB-3300, the requirements of NB-3300 ahall govern.)"
Subarticle NB-3100 defines the various design and operating conditions that must be taken into account, the special considerations of environ-mental ef fects such as corrosion and neutron irradiation, special rules for cladding, and consideration of the attachments to the vessel in pro-i ducing constraints on the pressure boundary.
Subarticle NB-3200 provides a complete set of definitions and design cri-teria for " Design by Analysis." The background of these criteria appli-cable to components of the reactor coolant pressure boundary are described in the booklet entitled " Criteria of the ASME Boiler and Pressure Code for Design by Analysis in Section III and VIII, Division 2," published by the ASME in 1969.
Subarticle NB-3300 provides special rules and interpretations applicable only to vessels. These special rules for vessels cover primarily the design of nozzles and welded joints. Specifically invoked is the require-ment of Subarticle NB-3214 that "A detailed stress analysis of all major structural components shall be prepared in sufficient detail to show that each of the stress limitations of NB-3220 and NB-3230 is satisfied when the vessel is subjected to the loadings of NB-3110."
q These Section III design rules differ substantially from those applied to conventional pressure vessels. The design of nuclear vessels requires extensive stress calculations, including analysis for combined mechanical and thermal stresses. Safe design of nuclear vessels does not rely upon empirical design conservatisms, simple code rules, and acceptable design details, such as have evolved for conventional pressure vessels.
The previously discussed Section III rules incorporated, for the first time, an approach to vessel design that includes consideration of the service and transient loadings and environments under which reactors must operate.
Un-like the Section I Code rules for construction of power boilers in fossil-fueled power plants,Section III design rules take into account the effects of metal fatigue from cyclic loadings which may be expected under " normal" and " upset" plant operational conditions. This departure in vessel design practice also includes requirements for thermal utress analyses as a part of the fatigue evaluation.
4 Non-mandatory analytical design methods are presented in Appendix A of Section III, but these techniques are not generally used in the analysis of reactor vessels because of the development of more efficient computer programs. The technology upon which these analytical methods are based 9-l 1
I is described in Chapter 6.1 of USAEC Report ORNL-NSIC-21.* The major change since the publication of this document has been the development and implementation of many more computer programs utilizing finite ele-ment analysis techniques. With these techniques, most of the generalized assumptions and approximations applied with previous stress analys.s tech-l niques are eliminated. For example, shear deformations and local flexibility effects between classical ring and shell elements are automatically included with finite element techniques. In addition, local stress concentration effects may be evaluated directly rather than through the use of empirical correlations. Correlations with experimental data are significantly better with finite-element predictions in the vicinity of the more highly stressed local regions than with solutions obtained by classical analyses. These refined design techniques result in increased confidence in the design adequacy of nuclear pressure vessels.
2.2.1 Loadings. Analytical techniques are applied to evaluate the stresses and deformations that exist within a structure of given geometry when it is subjected to a set of loading conditions described in the Design Speci-fication provided by the vessel owner or his agent. With respect to the reactor vessel, the significant loadings are internal pressure, changes l
in the temperature of the coolant, internal heat generation by radiation absorption, and reactions from internal or external attachments. Attach-ments include reactor internals, piping, vessel supports, and control-rod-f I
drive housings; it is necessary to consider all interaction effects between the reactor vessel and such attachments.
i Design of an LWR vessel for internal pressare, internal heat generation, and pipe reaction is straightforward. Evaluation of the primary stresses l
that result from internal pressure and design mechanical loads involves l
relatively simple analytical techniques. In addition, and most important, the recurring geometry and loading conditions have resulted in the accumula-tion of considerable experience with this type of evaluation, including experimental verification, over the past 15 years. It is extremely unlikely that a reactor pressure vessel would be incapable of sustaining pressure and externally or internally induced mechanical loads because of design deficiency.
More severe thermal transients are generally limited to local regions of a vessel. A vessel nozzle, with or without thermal sleeve, is an example of a region where thermal cycling may result in local fatigue crack propagation rather than violation of the overall pressure-retaining boundary of the en,__t.
A more pronounced thermal stress effect could result from quenching of the interior of the vessel with cool water during initiation of emergency core cooling. This situation has been studied by various nuclear power plant licensees and license applicants, and is being studied at Oak Ridge National Laboratory. Present indications are that this thermal event is not likely
- " Technology of Steel Pressure Vessels for Water-Cooled Nuclear Reactors,"
USAEC Report ORNL-NSIC-21, Oak Ridge National Laboratory, Nuclear Safety Information Center, December 1967. _
i I
to result in rupture of the reactor pressure vessel boundary, because the heat transfer mechanisms are not capabic of imposing thermal strains exceed-l ing the strain capacity of the vessel material.
l f
Both the loss-of-coolant accident and seismic events result in dynamic loadings on the vessel f rom internal and external attachments (Sec. 3. 8).
l The loss-of-coolant accident may cause significant pressure differentials within the vessel during the very short period of the blowdown. Both types of event laad to conditions that are more dif ficult to define than j
other operating conditions, and the stresses induced within the vessel by such loadings may not be determined with the same degree of conserva-j tism.
However, such events will not result in a large number of stress cycles. The effects of these loads upon the potential for failure are discussed in Sec. 4.1 of this report and are found to be adequately con-trolled by current design methods.
Because the loadings are of varying types and structural significance, I
the relative margin of safety for a vessel varies with the specific region of the vessel elements. As discussed elsewhere in this report, measures i
are taken to ensure that major flaws do not exist when a vessel is placed in service and to ensure operating procedures that do not lead to gross i
over-pressurization. In this situation, the major concern of the designer is the elimination of suberitical defects whose growth would nost likely occur by a fatigue mechanism. The re fore, the best relative index of the potential for subcritical crack growth is probably the cumulative fatigue usage factor computed for various portions of the vessel with due con-i sideration of the difficulty of determining this factor in each region.
The computed cumulative fatigue usage factor in pressure-retaining members of light-water-cooled reactor vessels is generally negligible except for (in the approximate order of decreasing magnitude):
1.
main closure bolting, l
2.
thermai sleeve attachment welds in those nozzles subjected to large (greater than 50 F) and repeated thermal transients, 3.
the inner corners of nozzle penetrations that are subject to a number of moderate thermal transients, 4.
the closure head adjacent to the main closure flange, and l
S.
the shell of a boiling-water reactor in the vicinity of the core shroud support plate.
None of these regions experiences significant neutron fluence. The usage factor f or the main closure bolting is usually about twice that of the l
next most highly stres sed region, but is well within the Code permissible design limit of 1.0.
l l l
l 1
l
A special evaluation of the fatigue usage factors is presented in the Stress Report required for each vessel. These factors have exerted significant influence on the inservice inspection program adopted for the reactor vessel as provided by Section XI of the Code.
Because of the application of this fatigue evaluation and inservice examinations, the Committee is of the opinion that a f atigue-induced crack developed during the service lifetime of the reactor vessel will not grow to a size sufficient to propagate rapidly.
2.2.2 Integrally Welded Vessel Attachments. The Code requires that con-sideration be given to the effect of attachments on the integrity of the pressure boundary. In current designs, only integrally welded attach-ments are of significance. Attachments of interest range from very small members for retention of instrumentation to the relatively large boiling-water-reactor shroud support, which is continuous around the entire circum-ference of the vessel. The most common internal attachment is a lug or bracket that locates or supports other internal structures. Similar lugs or brackets or support skirts constitute the most common external attach-ments.
With the possible exception of the shroud support and attachments to unclad removable heads in boiling-water reactors, internal attachments are f ab ri-cated from an austenitic stainless steel or a nickel-base alloy which ex-hibits superior toughness properties over the entire temperature range of operation.
In either case, the attachment material and the weld to the pressure boundary are controlled by Section III rules.
In some cases, the entire attachment is built up of weld metal.
Stresses arise in the pressure boundary adjacent to the attachment as a result of mechanical forces (weight, flow, seismic), strains to which the attachment must conform, dif ferential expansion between austenitic attach-ments and the ferritic vessel steel, and as a result of the differential thermal response to transients between the attachment and the thicker ves-sel pressure boundary. Loads and constraints generated by those portions of the attachment that are not defined as a portion of the reactor pressure vessel are required to be described in the Design Specification and are evaluated in the vescel Stress Report.
Wi th the possible exception of forces that result from flow-induced vibra-tion and those that result from pressure transients developed during a loss-of-coolant accident, the applied forces can be specified clearly and reliably.
Flow-induced vibration effects are further evaluated in connec-tion with AEC Regulatory Guide No. 1.20, " Vibration Measurements on Reactor Internals," during preoperational testing. The forces induced by a loss-of-coolant accident are generally established on a conservative basis and ac-commodated by the design without undesirable compromises.
Because attach-ments are avoided in the region of the reactor core, internal heat genera-tion within the reactor vessel wall seldom has a significant effect on attachment stresses. _ _ _ _ _ _ _ _ _ _ _
The stress analysis in the attachment region may generally be performed by simple strength-of-materials calculations coupled with more exacting solutions (for example, the techniques of Welding Research Council Bulletin 107*) of stresses resulting from external loads acting on a cylinder or sphere. Where such techniques are excessively conservative or where more exact knowledge related to deformations is required, finite element analysis techniques are generally employed.
The most significant stresses resulting from the presence of an attach-located directly at the junction between the attachment and ment are the pressure boundary.
Because they arise from incompatible deforma-tions between the attachment and the boundary, these stresses are sig-nificant mainly with respect to fatigue.
With the possible exception of such major attachments as the boiling-water-reactor shroud support or vessel support skirt, it is very unusual for the attachment to cause structural deformation of the pressure boundary. The majority of attachments, therefore, have insignificant effect with respect to the structural reliability of the pressure boundary. Those structural attachments which do have significant effects are designed and analyzed to accommodate the discontinuity strains within the limits specified by the Code.
The Committee believes integrally welded attachments to reactor vessels do not constitute an area of concern with respect to pressure boundary integrity.
2.2.3 Control Measures.Section III rules impose design control measures on both the owner and manufacturer of the reactor vessel by requiring a certified document identified as a vessel Design Specification. This specification must include (1) the specific functions and complete range of operating conditions expected over the service lifetime of the vessel, (2) the mechanical and thermal loads which the vessel will be expected to withstand during service, (3) the predicted environmental conditions, sc;h as irradiation, to which the ferritic material of the vessel will be ex-posed, (4) the transient loading conditions that may result as a conse-quence of system malfunctions (upset and emergency conditions), (5) the dynamic loadings associated with an earthquake event at the plant site, and (6) the system loadings from a postulated loss-of-coolant accident (faulted condition).
The Committee recognizes that a principal factor in safe vessel design is the establishment of proper design parameters as part of the vessel Design Specification. Such specifications are generally prepared by the nuclear steam system supplier (as the agent for the owner) who, by virtue of his direct knowledge of the system conditions under which the O " Local Stresses in Spherical and Cylindrical Shells Due to External Loadings", Wichman, llopper, Mershon, August 1965..
reactor vessel is expected to operate, is better qualified to specify the required vessel design conditions. In recognition of the importance of these design parameters to safety, the AEC reviews the design condi-tions stated in Safety Analysis Reports in order to confirm that appro-priate loading conditions under which the reactor vessel may be called upon to operate safely are included.
Not only are detailed stress reports required for all portions of the reactor vessel, but Section III also makes mandatory a certification procedure for the Stress Repart prepared by the vessel manufacturer.
A certification procedure applies also to the organizatiot. responsible for the preparation of the Design Specification. This organization reviews the Stress Report in order to certify that design and operating conditions defined by the Design Specification have been properly con-sidered and applied in the stress calculations. The Code places responsi-bility for structural integrity on the manufacturer, but requires that the
(
owner define the appropriate design and operating conditions and then assure himself that the manufacturer has properly interpreted these conditions.
The Committee believes such measures provide adequate assurance that no gross design oversight will be committed whose consequences could potentially lead to a vessel failure.
The Committee further believes that proper implementation of the Section l
III design rules and the review and certification procedures for the ves-sel Design Specification and Stress Report provide adequate design control measures to assure conformance of reactor vessel design to the specifica-tions required by the supplier of the nuclear steam system, floweve r, proper implementation can be assured only if the vessel designer is qualified to perform the required analyses. The Committee believes that the Code rules could be strengthened in this area, as indicated in the conclusions under Sections 2.6 and 2.7.
2.3 Fabrication The non-nuclear vessel failure statistics discussed in Sec. 5 include a few instances where poor fabrication quality was a cause of failure, and some instances where fabrication methods were not suited to the materials selected for the application. To preclude problems of this nature in nuclear ves-sels,Section III of the Code provides fabrication rules that are directed primarily toward controls, nondestructive examinations, and inspections performed during the course of vessel manufacture. Although customary ves-sel fabrication practice is the basic for many of the requirements, the Code philosophy is generally guided by the principle that the needed quality in the completed nuclear vessel is best attained through the imposition of rigorous controls during each stage of each of the major fabrication pro-cesses (forming, welding, heat treating, etc.).
The Code quality assurance requirements provide measures that must be strictly implemented to assure adherence to this Code philosophy.
j,
l l
2.3.1 Fabrication Control Measures. The rigor of the Section III rules is maintained throughout each stage of the fabrication processes through a quality assurance program. Control measures for the timely detection of manufacturing deviations and initiation of corrective actions, as necessary, to remedy fabrication faults or errors are provided by the quality assurance program to assure that the required level of vessel fabrication quality is attained.
Current nuclear vessel practice requires close checks on the material manufacturers and suppliers by personnel of both the vessel manufacturer and the steam system supplier, who independently review material proper-ties and chemistry as reported in mill certifications. Identification markings must be carried by all vessel materials, and these markings must be traceable by documented records of material source and fabrication process history. This identification practice is an important aid in preventing the use of improper or nonconforming materials during vessel fabrication. The effectiveness of the materials controls is monitored separately by the Code Inspector's review and by the owner's audits of the quality control records.
The fabrication work is also monitored by separate quality assurance con-trols in the fabricator's shops. The Code Inspector monitors the fabri-cation work by comparison of inspection records with documented procedures and personally reviews the results of nondestructive examinations. The owner's representative independently audits the fabrication quality con-trol system. The fabrication quality is finally affirmed by comprehensive base-line examination of all welded joints in the fabrication shop or at the nuclear power plant site.
With the number of such check points, the likelihood of a fabrication error in a nuclear vessel is relatively small when compared with that for non-nuclear vessels for which there is no comparabic system of inspection and quality assurance. The Committee recognizes that thorough monitoring alone will not assure the ultimate quality level needed in the manufacture of reactor vessels. Ef fective implementation of these programs requires that each fabrication step be clearly defined by documented procedures based on well understood principles. Up to now, questions regarding the quality of reactor vessels have originated from lack of clarity in Code rules and shop procedures and some omissions of important control actions in the quality assurance programs. The Committee believes that quality assurance implementation deserves the highest level of attention because of its crucial value in attaining reactor vessel reliability. If the atten-tion is adequate, the quality assurance program should result in a vessel fabrication quality level in accordance with the Code intent.
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2.3.2 Nondest ruc t ive Examinations. The measure of soundness and final quality achieved in the completed reactor vessel is established directly by the number and extent of nondestructive examinations conducted and the sensitivity of the examination methods employed to detect fabrica-tion flaws.
Examination practice for nuclear reactor vessels requires an extensive nondestructive examination program which includes visual, radiographic, and ultrasonic examination during all phases of material j
manufacture and fabrication and vessel construction. The examination techniques have been selected to provide the best opportunity for expos-ing significant flaws. Liquid penetrant and magnetic particle examina-tions are used to detect surface fla ws. Radiography techniques are used for full-penetration welds and any other fabricated regions where this technique is capable of exposing flaws. Ultrasonic examination is re-quired of all base metal material, including rolled plate and heavy-
{
section forgings. The ultrasonic technique is also used for weld geome-1 tries not suited for radiography.
f l
Noteworthy is the signi ficant increase in the extent of nondestructive volumetric examination required by Section III for nuclear vessels when I
compared with the examinations required by Section I or Section VIII.
The nondestructive examinations prescribed by Section III are compared with those prescribed by Section I rules in Table 2-1.
As a prime exam-ple, reactor vessel shell and head materials are subject to ultrasonic 1
examination covering 1007 of the metal volume. This examination is not required of materials used in power boilers (Section I) of fossil-fueled power plants or in conventional unfired pressure vessels (Section VIII),
j even though such vessels are subject to similar or more severe service conditions than nuclear reactor versels. In addition, the Code requires that weld joints for nuclear reactor vessel shells and connected reactor coolant piping nozzles be volumetrically examined by radiography, and supplemented by surface examination techniques. This requirement places a desirable limitation on the type of weld joint designs that may be applied.
A point of concern relative to examination techniques is that previously inspected areas may be altered during subsequent fabrication processes, leading to the possibility that flaws may develop af ter the examinations have been completed.
For this reason Section III requires surface examina-tion of weld joints following the hydrostatic pressure test of the vessel.
Section XI of the Code further requires that, after vessel construction is completed, the vessel welds (including portions of base metal adjoining the welds) must be subjected to a comprehensive volumetric examination (e.g., ultrasonic) to map the presence of flaws, if any, as a base line for comparison with future inservice examinations.
The acceptance standards for each nondestructive examination method specified by Section III have evolved from experience over the past 30 years. They are based on the attainment of high fabrication quality l
l Table 2-1 COMPARISON OF NONDESTUCTIVE EXAMINATION REQUIRIMEhTS REACTOR VESSEL VS. BOILER DRUM ASME Section III ASME Section 1 Reactor Vessel Boiler Drum Materials Exam.*
Extent Exam.*
Extent Plates UT 1007. Vol.
Non-mandatory Substantially less than Section III Forgings, and UT, and 100% Vol.
Non-mandatory Substantially Bolting MT or PT 100% Surf.
less than Section III f
Fabrication Weld Grooves MT or PT 100% Surf.
Visual 100% Surf.
Shell and Head RT, and 1007. Vol.
RT 1007. Vol.
Weld Joints MT or PT 100% Surf.
Nozzle Welds RT, and 100% Vol.
RT 100% vol.
Penetration Welds MT or PT 1007. Surf.
Visual 1007. S u r f.
! and each
! progressive weld layer After Hydrotest All Welds MT or Pr 100% Surf.
Visual 100% Surf.
Preservice (Section XI)
Shell, Head, and Nozzle weld joints UT 100% Vol.
Ncne Nozzle Inner UT 100% Vol.
None Radius Section Vessel Supports-welds UT 1007. Vol.
None
- RT - Radiography Exam. - 1007. of volume of metal examined.
UT - Ultrasonic Exam.
100% of volume of metal examined.
MT - Magnetic Particle - 100% of surface area of material or weld.
PT - Liquid Penetrant - 100% of surface area of material or weld.
through excellent workmanship using proven fabrication practices. The acceptance standards have been defined in terms of the sensitivity limits o f available examination techniques. Ilowever, mino r indica-tions of a random nature, though detectable by radiography, may not require correction; but a non-random series of such indications sug-gesting a crack would not be acceptable without investigation and possibly correction.
In most cases, the Code requires corrective action for flaws that can have structural significance, even if they are relatively minor. Service experience with non-nuclear vessels and boilers has demonstrated the general adequacy of these standards.
Although it is clear that the intent of the Code is to be extremely con-servative with respect to the acceptability of detectable flaws, the current Code requirements cannot assure that all flaws will be detected, since the available examination tcchniques cannot detect some flaw geom-etries.
Improvements in flaw detection and interpretation methods by increasing the sensitivity of radiography, by increasing the precision of ultrasonic techniques, or by bringing acoustic emission methods into practical use, would be valuable in providing further assurance of ves-sel reliabitity through reduction in the probability that undetected flaws will exist.
Pending resolution of uncertainties arising from flaw detection limita-l tions, reliability is best assured by selecting materials whose toughness characteristics under the conditions of service make them tolerant of flaw sizes that are well within the detection capability of the examina-tion methods, thus giving maximum assurance that critical flaw growth potential can be recognized and avoided. The Committee believes that the operating limitations imposed for normal operation, including design transient loading, shutdown, and structural testing should provide suf-ficient margin of safety against fracture to compensate for limitations of flaw detection.
2.4 Installation Installation errors or improper handling of components have been suggested as possible contributors to reduction in the vessel quality level.
Ves-sels have been subjected to mechanical shock, environmental contamination, and improper installation. The experience with such faulty practices has shown that if they are known to have occurred, their damage ef fects can be assessed by reinspections and re-examinations if the vessel has not yet been placed in service. The types of defects that could be caused by these installation faults are fundamentally the same as those of concern during fabrication. ___-_
The known installation errors that might affect the vessel are abuse of the vessel in handling, deficiencies in the vessel supports and in the intermediate structural attachments between the foundation and the vessel, misapplication of vibrational restraints provided to resist seismic loads, and improper placement of piping restraints that could result in unanticipated pipe reaction loads on the vessel. A number of such faults have already occurred in nuclear power plant installations, mainly because of lack of attention to design and installation details; in each case, a review has led to the conclusion that these faults did not increase the probability of disruptive failure
- of the vessel. By proper checking of structural movement during non-nuclear hot operation and system functional tests, and with careful attention to attachment details, the Committee believes it should be possible to reduce to negligible proportions the probability that installation faults of this type will contribute to gross vessel failures.
)
2.5 Relevant Work in Other Countries Emphasis has been rl'ied on design and fabrication practices covered by relevant U.S. Codes such as ASME Section III; however, work in the construction and operation of nuclear and non-nuclear pressure vessels in other countries of ten is pertinent. The Committee and its con-sultants have a substantial familiarity with such work through the Pressure Vessel Research Committee Code activities at the International level, International Conferences, and participation through CREST on the CEC-NEA Experts Group on Mechanical and Material Problems Relating to Safety Aspects of Steel Components in Nuclear Plants.
2.6 Adequacy of Nuclear Vessel Construction Practice The foregoing considerations provide strong evidence to support the belief that the quality cf nuclear vessels conforming to the rules of Section III is superior to that of the already high quality vessels in non-nuclear service built under the rules of Sections I or VIII. The reasons can be summarized as follow:
1.
For nuclear service, the properties of materials are defined more strictly and much more attention is paid to making certain that the proper environmental conditions are considered. The strength and toughness of welds and heat-af fected zones have been shown to be in the same general range as for plate and forging materials.
- Refer to Sec. 5.1 for definition.
2.
Design practices are much more highly refined and emphasis is placed on careful analysis of design details in preference to reliance only on the nominal membrane stresses and gross factor of safety. Com-parison between requirements of Section III and those of Section I and Section VIII, Division 1 shows that the non-nuclear codes gain all of their conservatism by increased wall thickness requirements, evaluated solely for membrane loads and with a factor of four between design and ultimate strength values. In contrast, although Section III requires a nominal factor of only three, this Section of the Code provides rules for treating design features affected by fatigue loads and other secondary loading conditions which are known to be the major contributors to failure of vessels.
3.
Section III specifies that the vessel function, design conditions, and environmental loads be carefully defined in the Owner's Design Specification so that designers can account for all conditions that j
the vessel may experience over its service lifetime. Non-nuclear codes do not have such a requirement and the majority of non-nuclesr vessel failures reported were associated with conditions not intended by design.
4.
Fabrication and installation methods are much more carefully con-trolled under Section III than for most non-nuclear service appli-cations, because of the extensive quality assurance requirements of Section III.
The Committee believes that the superior quality of nuclear vessels does provide considerable assurance that the necessary vessel reliability will be attained. At the same time, because of the cautionary considera-tions discussed in the following paragraphs, the Committee recommends against over-dependence on Code rules as the only defense against nuclear reactor vessel failures.
J The effective exercise of the responsibilities assigned by Section III to the various parties involved in the manufacture of a reactor vessel are most important in providing assurance to the owner of vessel adequacy and integrity. The important elements in this chain of responsibilities are (a) the ASME survey of the vessel manufacturer, which establishes the qualifications required before the manufacturer receives an ASME certifi-cate of authorization to build nuclear vessels, (b) the preparation by the vessel manufacturer of a Quality Assurance Manual in accordance with Section III QA requirements and the implementation of the QA program before vessel fabrication commences, (c) the contracting of Code inspection services by the vessel manufacturer from a qualified inspection agency which provides the Code Authorized Inspector, whose duties during vessel fabrication are defined in Section III, and (d) the qualification requirements and tests of Code Authorized Inspectors, who are held responsible for certifying con-struction of the completed reactor vessel in accordance with Section III.
In order to assure that the individual responsibilities assigned to the various parties are effectively carried out, an independent means of audit is considered essential. Since 10 CFR Part 50.55a, " Codes and Standards," adoptsSection III as part of the Code of Federal Regulations, and 10 CFR Part 50, Appendix B, Quality Assurance Require-ments, establishes the need for a clearly specified arrangement by which the owner can demonstrate that each responsible party involved in the manufacture of the reactor vessel is performing the Code designated duties, the Committee believes that contractual provisions should permit audit of each responsible party by both the AEC and the owner.
Both fabrication and examination practices have undergone an evolution-ary process. Some fabrication problems have been experienced, including problems with new fabrication methods such as electroslag welding and other problems referenced in Sec. 3.7; these problems have been solved by changes in procedures and, in some cases, by design changes. The technology of fracture mechanics, while providing an im-portant new tool for analysis of flaw propagation, requires elaborate non-destructive examinations and specially trained personnel for mean-ingful interpretation of the characterization and significance of flaws.
2.7 Conclusions with Respect to Reactor Vessel Construction Practice The Committee believes th 2t continuing participation of the AEC Regulatory Staf f on Code writing groups and liaison by ACRS does provide substantial opportunity for contributing to assurance that vessel requirements related to public safety are properly covered by Code rules. The interests and needs of the AEC in exercising its regulatory responsibility have usually been considered by the Code groups and in most cases rules have been written in response to these needs. In addition, where necessery, the AEC has issued supplementary requirements to the Code as a part of its regula-tory program to assure that appropriate care is taken in nuclear reactor vessel practice. Nevertheless, the Committee sees the nced for enhanced vigilance to assure that maximum advantage is taken of the available measures for defense against vessel failure. It is incumbent on the nuclear induc. rial community as a part of its responsibility for public safety to make certain that the following measures are properly implemented.
'. More definitive material toughness requirements for emergency and faulted conditions should be stated in the Code rules, taking ac-courit of the variable relationship between vessel temperature, level of stress, and fracture toughness. Fracture toughness requirements should be governed by this relationship for transient conditions as well as by normal design requirements../
i.
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L.
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2.
Where owners ' or manufacturers ' quality requirements ordinarily exceed minimum Code requirements, consideration should be given to upgrading the Code to conform to practice.
f 3.
Explicit requirements for limiting residual elements, such as cc aper and phosphorus, that are believed to increase irradiation embr.ttlement, should be set forth for materials subject to neutron 18 fluentes exceeding 10 nyt.
4.
The present Code rules require that the owner,or an agent seting i
l for the owner, prepare a Design Specification; that the manufacturer, f
or an agent acting for the manufacturer, prepare a Stress Report; l
and that the Stress Report be reviewed by the organization prepar-l ing the Design Specification to establish that the two documents are consisteat. It is considered essential that these requirements be interpreted as involving two independent organizations. If the two independent organizations are part of a single corporation, the owner, or a qualified agent not associated with the single corporation, should satisfy himself that the review is properly done. If all three steps are performed by a single organization, a separate confirmatory design review should be required.
5.
Code rules or AEC regulations should require that first-of-a-kind vessel designs be subject to confirmatory design review by the owner or his agent.
By a confirmatory design review, it is meant that the review-ing organization undertakes whatever steps it considers necessary to certify conformance of the vessel design to Design Specification and Code Stress Report requirements. The review should be conducted by an organization responsible to the owner, and should not be one of those crganizations responsible for the Design Specification, the Stress Report, or the reviews of the Stress Report relative to the Design Specification. The confirmatory design review, the Design Specification, the Stress Report, and the review of the Etress Report for Conformance with Design Specification should be made under the direction of an engineer or engineers whose technical competence for such responsibility can be demonstrated by education and experience.
6.
The nondestructive examination techniques presently specified in the Code should be reassessed to improve the correlation between flaw detection, defect characterization, and required quality level.
Specific attention is needed to improve and clarify the acceptance s tanda rd s for nondestructive examinations. The relation between pe rmis s ib le flaw size and character and their influence upon the strength, ductility, and toughness of the materials should be con-sidered in defining acceptance standards.
New nondestructive examina-tion techniques, such as acoustic emission and acoustic holography, should be investigated for possible application to reactor vessels when these techniques are proved practical. - _ _ _ _ _ _ _
7.
The individual responsibilities carried cut by (a) the ASME survey teams that evaluate the vessel manufacturers' facilities to qualify reactor vessel fabricators, (b) the manufacturer's quality assurance organization in developing the QA Manual, (c) the in.ipection agency, which provides the Code authorized inspector, (d) the organization which controls the system of qualifying Code authorized inspectors, and (e) the Code authorized inspector's performance should be subject to audit by the owner and the AEC in order to assure acceptability in meeting the AEC regulations 10 CFR Part 50, secticn 50.55a and Appendix B.
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3 POSSIBLE MODES OF VESSEL FAILURE This report is primarily concerned with " disruptive" failures (as defined in Sec. 5.1) whose consequences cannot be limited by current engineered safety features, i.e., failures that could cause:
1.
blowdown at a rate exceeding the design capability of the emergency core cooling systems or the containment, or 2.
loss of coolable core geometry, or 3.
generation of missiles not considered in current designs.
Structural failure usually involves the slow growth of a crack to a criti-cal size at which failure occurs by rapid crack propagation.* Almost invariably such crack growth is the extension of either a crack which initiates in a highly stressed region or a crack-like defect which remains af ter fabrication. In the latter case, the defect may have been too small to be detected during examination or may have been over-looked due to errors in examination.
In either case, high stresses are required for subsequent crack extension at temperatures at which the material is ductile.
In considering the types of vessel structural failure possible in testing or in service, it is convenient to divide the failure processes into two stages:
1.
slow crack growth under cyclic or constant load.
2.
final failure by rapid crack propagation.
3.1 Slow Crack Growth Slow crack growth may occur by fatigue or corrosion fatigue, by hydrogen cracking, or by stress corrosion cracking.
3.1.1 Fatigue and Corrosion Fatigue. Fatigue damage and crack growth may be caused by cyclic stresses f rom pressure, mechanical, or thermal loading, including transients. Very small flaws that are too small to be detected by ASME Code examination procedures will not propagate to disruptive failure under conditions encountered in normal reactor ves-sel service. Larger flaws and cracks that are not detected, through error or poor inspection procedures, might grow sufficiently to cause either leakage or localized rapid fracture, especially in the vicinity of a nozzle.
- Gross overload may cause failure by deformation, with or without ultimate rupture. _
During cyclic loading a crack may extend a small amount (10~7 to 10-4 in.)
during each cycle. The largest rate of crack propagation at a given 4 K*
appears to occur for a corrosion fatigue condition at a relatively slow cyclic loading rate. While this condition is still undec investigation, it appears that only large cracks would experience more than minute amounts of crack extension in normal reactor vessel service.
3.1.2 Ilydrogen Cracking. Low yicid-strength nuclear vessel steels (SA-533 and SA-508) are not highly susceptible to hydrogen-induced crack formation. Ilowever, Kussmaul and co-workers in the Federal Republic of Germany have reported cases of hydrogen cracking in the heat affected zone of weldments of 21 NiMcCr 37 steel, which is equivalent to SA-508**.
It is well recognized, both in the FRG and in the US, that some combi-nation of severe constraint, insufficiently dried electrodes or other welding materials, insufficient time and duration of preheat or postheat, etc., may lead to hydrogen cracking. Proper welding techniques will minimize the risk of cracking. Proper inspection techniques should detect significant cracks prior to placing the vessel in service, but small or incipient cracks might not be detected.
In the FRG, an extensive investigation is proposed to explore the effect of all the important parameters on hydrogen cracking and embrittlement of 21 NiMcCr 37.
The Committee recommends that, even though such cracks have not been reported in the US, a serious investigation be made into the sensitivity of SA-533 and SA-508 welds to hydrogen damage as a function of variations in the pertinent parameters.
In a vessel initially free from hydrogen-induced defects, there appears to be no mechanism by which large amounts of hydrogen could enter and be retained in the steel and cause subsequent embrittlement or crack formation in service.
3.1.3 Corrosion and Stress-Corrosion Cracking. Localized pitting and crevice corrosion could occur in base metal exposed by localized clad-ding failure. These phenomena do not lead to cracking, and the rate of chemical attack is expected to be so slow that large flaws should not develop. Control of the reactor coolant composition and pli should elim-inate any possibility of caustic embrittlement of either austenitic or ferritic steels.
O Range of K (stress intensity factor, Ksi) in.)
- See Reference 7, Sec. 5.
aensitized austenitic materials are subject to intergranular attack and/or stress-corrosion cracking. This effect would be insignificant in clad-ding, but could be serious in wrought austenitic materials used for safe-ends, small diameter nozzles, and reactor internals. The use of sensitized materials should be avoided; if sensitized materials are used, careful inservice inspection is necessary. With current reactor pressure vessel design, cracking of such material would not result in a disruptive f ailure of the vessel.
3.2 Rapid Crack Propagarirn Rapid crack propagation can occur in structural metals by two prccesses, cleavage or ductile tearing rupture. Figure 3-1 illustrates the effect of thickness and temperature on the fracture tcughness of SA-533 steel, as measured in a Dynamic Tear (DT) test. The NDT* temperature as measured in the standard Drop Weight Test corresponds to the toe of the UI curve, from which point the toughness increases rapidly with increasing temperature. Cleavage is the predominant mode of fracture at tempe ra tu res below NUT. At higher temperatures, increasing microscopic ductility causes the fracture made to change to the ductile tear mode.
For thick sections (thickness B;>3 in.) with temperatures in the vicinity of NUT + 200 F, the fracture mode is ductile tear and the upper shelf is reached.
In the transition temperature region, from NDT to approximately I
NDT + 200 F, fracture occurs by a mixture of cleavage and ductile tear.
The fraction of ductile tear increases as the temperature is raised or as the wall thickness decreases. This leads to a transition in vessel toughness between the NDT and approximately NDT + 200 P.
Based on this well documented behavior. two approaches may be used to characterize and control the toughness of the reactor vessel steel.
3.2.1 Transition Temperature Range Approach. The first approach, some-times referred to as the transition temperature range approach is based on the fact that, for nuclear vessel steels of normal quality, fracture by the ductile tear mode requires extremely high energy for crack propa-gation -- so high that gross overload and bulging of the vessel would probably occur before a crack could propagate. Assurance that no rapid crack propagation will occur can be achieved by pressurizing a nuclear vessel of a given thickness only at a temperature above the transition range from cleavage to predominantly ductile fracture, and by confirm-ing that the " upper shelf" toughness of the material is satisfactory.
To provide assurance that any one vessel or part of a vessel will always be operated in a safe temperature range, the maximum permissible transi-tion temperature range of the vessel steel is specified prior to service and determined by surveillance specimens throughout its life. Operating condidons that preclude full pressurization of the vessel below the upper shelf can thus be determined and maintained.
- Nil Ductility Transition.
I l
i 3.2.2 Fracture Mechanics Approach. A second approach to fracture con-trol of a nuclear vessel relies on a fracture mechanics characterization j
of the material and its stress environment. Using this characterization, j
the stress in any portion of the vessel, in conjunction with any exist-ing flaw, can be compared with the stressed-flaw tolerance of the mater-ial, a material parameter such as KIc.*
Using this parameter,,the stress in the vessel can be limited such that, in the presence of an assumed j
flaw size so large as to ensure detection, no rapid crack propagation can occur. Above NDT, the fracture toughness of the materials used in nuclear reactor vessels increases greatly, as seen in Fig. 3.2.
Thus, the crack tolerance of the material at the normal operating temperatures is f
high. Under this system of f racture control, rapid fracture prevention is assured by control of stresses and flaw sizes. For nuclear vessel materials of normal shelf fracture toughness, very large cracks would be required to cause the onset of rapid crack propagation at operating pressure. In regions of high local stresses, such as nozzle corners, ductile tearing could commence at smaller cracks or lower pressure but, as the tear extended into the region of lower nominal stress of the ves-sel wall, rapid fracture would again require very large cracks.
3.3 Application of Basic Failure Mechanisms to Reactor Vessels i
It is necessary to consider, separately, the temperature and pressure conditions during power operation, and during heatup and cooldown.
3.3.1 Power Operation. In the range of temperatures in which reactor vessels normally operate, either the transition temperature range ap-proach or the fracture mechanics approach will lead to the conclusion that the probability of brittle fracture is extremely remote and that, l
in the absence of gross overpressurization, ductile (shear) failure can occur only in the presence of a long deep crack. For rapid shear frac-ture, the crack would have to be of such length and depth that the local stress in the remaining ligament approached the ultimate strength of the material.
3.3.2 IIcatup and Cooldown. Restrictions are placed on pressurization during heatup and cooldown. Ilowever, in the lower temperature range, e.g.,
ambient to 200 F (higher for irradiated material), fracture con-trol procedures differ between the two approaches employed to prevent rapid crack propagation. Under the transition-temperature approach, the control procedure would be to prohibit any substantial ** pressuriza-tion below NDT + d
, where
- increases with thickness. Under the fracture mechanics approach, the permissible stress in the vessel would
/
- Fracture toughness, Ksi \\ in!. Other parameters that measure fracture l
toughness may include kid, EIa, (COD) c, JIc
- Pressurization above about 25 percent of operating pressure has not been permitted. -
he set, quantitatively, by the fracture toughness of the material at the particular temperature and assumed flaw sizes. Thus, based on an assumed flaw size, the operatin;; stresses may be set over the entire temperature range of vessel operation.
Because the fracture toughness of the material decreases rapidly at temperatures below the upper shelf, and flaw sizes are estimated with a substantial safety allowance, the stress permitted i
decreases sharply to low values as ambient temperatures are approached.
The similarity of the applications of the two approaches is illustrated in Fig. 3-3.
The transition temperature approach was used in 10 CFR l
Part 50, Appendix G " Fracture Toughness Requirements", published for comment in the Federal Register in 1971. The fracture mechanics approach is adopted in the revised Appendix G to 10 CFR 50, " Fracture Toughness Requirements," proposed by the AEC Division of Regulatory Standards in 1973.
In the particular example shown in Fig. 3-3, the fracture mechan-ics approach (Appendix G, 1973) is somewhat more conservative (lower allowable coolant pressure) at temperatures in the vicinity of the NDT and in the upper part of the transition range, whereas the transition temperature approach (Appendix G. 1971) is somewhat more conservative s
in the range between these two extremes.
Advantages of the transition temperature approach are its simplicity In use and its assumed independence of a postulated or measured defect size.
An advantage of the fracture mechanics approach is that it per-
)
mits an assessment of the effect of all detected or probable flaws, at any temperature, and thus provides assurance that operating stresses and flaw sizes will at all times be limited to values well within the tolerance of the material. Some engineers are concerned, however, that flaws larger than T/4 (Appendix G) could remain undetected and that the use of unjustified, less conservative assumptions with respect to flaw sizes in f racture mechanics calculations will eventually lead to less conservative estimates of allowable reactor coolant pressures.
The two approaches have much in common.
Both provide for decreased operating pressure at low temperatures, and also provide for an increase in stress with increasing temperature. In comparison with the normal operating conditions under nuclear heat, both indicate that vessel heatup can currently be carried out in a conservative manner.
Detailed analysis of current data suggests that both approaches, as now applied, can be refined. As stated earlier, NDT + 200 F provides a more conservative definition of the lowest temperature of the upper shelf than provided by Section III of the Code, independent of thickness.
Therefore, the transition temperature approach might be further refined by restricting pressures during heatup to the minimum required for pump shaft seals and prevention of cavitation, until temperatures close to NDT + 200"F are achieved in thick walled vessels. Such a condition is shown for a spectiic condition of wall thickness and irradiation by Curve A in Fig. 3-3.
Similar restrictions would apply during cooldown. __
In the f racture mechanics approach, a better definition of the KIR* curve can be obtained by statistical treatment of toughness determinations of larger number of heats of material. As the phenomenon of crack arrest a
is c larified, further adjustments in the KIR curve may be required. In addition, the safety factor may be changed in the stress calculations.
Such a condition is illustrated for reduction in the factor of safety from 2 to 1.5 by Curve B in Fig. 3-3.
All of these adjustments could lead to cnanges in reactor coolant pressures as indicated by the cross-hatched area in Fig. 3-3.
By either approach, at temperatures in the vicinity of the upper shelf, a crack may be initiated in a small embrittled region and propagate rapidly until it encounters high toughness material.
If the embrittled region is not large, the crack should be arrested in surrounding ductile metal.
This condition may also occur in those reactors in which actuation of emergency cooling systems cause local chilling to below the upper shelf.
Measurements show that both dynamic fracture toughness and apparent arrest fracture toughness increase rapidly in the temperature range between NDT and NDT + 120 F.
Above NDT + 200 F, the dynamic fracture toughness exceeds the static fracture toughness. It is, therefore, believed that a crack initiated as postulated would be arrested before it could result in escape of fluid at a rate greater than can be handled by protection systems. This conclusion should be reviewed for heavily irradiated vessels when more irradiation data become available.
The above considerations have led the Committee, in Sec. 3.3.3, to recommend conservatism in application of proposed Appendix G.
3.3.3 Application o f 10 CFR Part 50, Appendix G.
It may be noted that both Appendix G and the 1971 proposal, shown in Fig. 3-3, are considerably more conservative than the Porse criteria formerly believed to be suffi-ciently conservative. Ilowever, considerations discussed in Sec. 3.3.1 and Sec. 3.3.2 suggest that all reasonable conservatism should be used in applying Appendix G.
Operating procedures and reactor safety features give a very high degree of assurance against vessel failure by accidental gross overpressure be-yond design pressure; but, if such overpressure should occur at tempera-tures below NDT + 200 F, the fracture could be partially brittle.
1 Until a vessel has become signfficantly irradiated, there should be no need to stress a vessel during heatup or cooldown to the limit permitted by Appendix C.
Fig. 3-3 shows a wide margin between the Appendix G limit line and the limit line for a P'4R using the circulating pumps for heatup. For BWRs, current practice is likely to be conservative with respect to Appendix G.
- Fracture toughness at Reference Temperature.. _ _ _ _ _
The Committee, the re fore, recommends that Technical Specifications give operating heatup and cooldo in pressure-temperature curves that can be shown to be as conservative w.th respect to Appendix G as practical.
It is believed that this course kill give time to develop a more exten-sive knowledge of material proper :Les, and better flaw detection capa-bility without imposing an unreas(nable burden on operation.
3.3.4 The Leak-15e fo re-Fireak C rite r i on.
It has been postulated that a
crack propagating progressively ty stress cycling will penetrate the pressure boundarv and reveal its presence by leakage before it becomes large enough to be unstable. The Committee therefore believes that reliable and sensitive systems for detecting and locating leaks are of great importance in assuring vessel integrity. However, the Committee believes that it has not been proven that crack will penetrate the a
l wall before it reaches critical size, and therefore rejects the " leak-l before-break criterion" as an infallible warning of impending disruptive I
failure.
1.4 Degradation of Toughness Several processes could conceivably cause the toughness to be lowered during service, thus causing a decrease in KIc, an increase of transition 1
t empe rature, and/or a decrease in toughness at the upper shelf. These processes are t empe r emb r i t t lement, strain aging embrittlement, and irradiation embrittlement. The low alloy SA-533 and SA-508 materials used in nuclear pressure vessels are relatively insensitive to the first two of these embrittlement phenomena.
Also, they are given a post-fabrication anneal at 1,000 F, which will make them still less suscep-tible to both temper and strain aging embrittlement at peak operating temperatures below 600"F.
The toughness of both SA-533 and SA-508 steels may be dec reased by irradiation, particularly at the belt line where the fluence is highest. Even for irradiation sensitive heats, the combina-tion of low working stresses in this region,plus careful surveillance p ro g rams to monitor the change of toughness, make it extremely unlikely that rapid crack propagation could develop in this regig. Ilowever, for J
a vessel that may be exposed to a neutron fluence of 10 or more
( > 1 Mev), the owner should assure himself that the system permits a practical procedure for annealing the vessel in service in case this should become necessary.
3.5 Conditions of Operation In considering modes and probabilities of failure (as discussed in Sec. 5), it is necessary to distinguish between vessels operated in accordance with design specifications and vessels subjected to acci-dent conditions that represent gross departure from design specifica-tions.
Vessels designed, constructed, and operated in conformance with _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
AEC and Code requi rements and f ree from flaws capable of growing to critical size should not be subject to failure. Ilowever, the possi-bili ty of error in design or ccnstruction, of material deficiency, or of the presence of large undetected cracks cannot be precluded.
The proposed failure rate discussed in Subsection 5.9 is not applicable to events that cause gross overload. Overload conditions are discussed in Sec. 3.8.
3.6 Character and Location of Postulated Vessel Failures in Normal Operation In this section, it is postulated that reactor vessel failures occur.
The expected characteristics of such failures are then discussed for various locations.
3.6.1 Failure in the Cylindrical Shell.
Failure in the cylindrical wall remote from constraints is highly improbable. As discussed in Sec. 3.2.2 and Sec. 3.3.1, a longitudinal crack would have to be very long and deep to cause disruptive failure by loagitudinal splitting of the vessel. Such a crack could be postulated in a longitudinal shell weld or heat af fected zone, but it is unlikely that so large a flaw would escape detection during manufacturing, preoperational testing, and inservice inspection, or that it could grow to such size during operation without leakage. Circumferential f ailure remote from con-l straints is still less probable, because the stress normal to the postulated break is less.
If a failure in the shell were to occur, assuming no severe embrittlement of the cylindrical portion, it would be most likely to occur at an integrally welded circumferential attachment, such as a flange, support skirt, or integral internal ring or shroud.
3.6.2 Failure at Nozzles. Some nozzles are subject to relatively seve e cyclic stress conditions ir. thermal sleeve attachment welds and at the inner corner of the nozzle penetrations (Sec. 2.2.1).
If a crack in one of these regions should become large without devel-oping a leak or without being otherwise detected, a critical combina-tion of stress and crack geometry could cause rapid crack propagation longitudinally into the vessel wall, or along the nozzle, or both.
A crack running longitudinally into the vessel wall should propagate only a limited distance, where the relatively low stress would permit crack arrest. The crack might penetrate the wall, but it would not be expected to open an area for coolant discharge greater than provided for in the ECCS. Dynamic ef fects of such breaks require specific analy-sis on components outside the vessel. It is believed that the consequen-ces of a nozzle split would probably be similar...
3.6.3 Failure in Vessel Head. The shell of the vessel head, at the junction with the head flange, is subject to relatively severe fatigue conditions, but cumulative 100 percent inspection of this region every ten years makes gross failure at this point improbable.
3.6.4 Safe-End Failure. The probability of failure of a safe-end or of a bimetallic weld between a vessel nozzle and a pipe may be higher than others discussed here because of cyclic stresses resulting from the dif ference in the coefficients of thermal expansion of ferritic steel and austenttic steel, and because of possible cyclic loadings transmitted by improperly supported piping. The Committee believes that the present practice of designing the biological shield (vessel cavity walls in PWRs) and other structures to withstand the effects (pressure differences, jet forces, and the possible generation of misslics) of failures of a safe-end should be extended to include failures of the thin-walled part of the nozzle adjacent to a safe-end.
A single safe-end separation could not lead to consequences greater than for a double-ended pipe break, provided that pressure and jet forces can be sustained by the surrounding structure; moreover, a crack in the relatively thin wall at the safe-end is likely to propagate through the wall and leak before severance occurs. The Committee believes that simultanecus multiple safe-end failures are very improbable in normal operation.
3.6.5 Bolt Failures. Failure of a suf ficient number of bolts in a ductile mode would cause leakage before total failure of the joint.
Howeve, some common mode failures such as fatigue cracking and stress corrosion cracking could cause bolts to break with little elongation.
Section XI of the Code requires one hundred percent examination of the bolting during each ten-year inspection period, but does not require l
that studbolts be removed from the lower flange for this examination.
I Some operators have routinely removed studbolts for refueling and others
]
have not, but the problem of protecting the studbolts and threads of the lower flange from corrosion by boric acid has led all man'facturers of PWRs to recommend removal of studbolts at each refueling. The Committee recommends that Section XI examination be interpreted as requiring surface examination of both threaded ends of studbolts for all LWRs.
A Regulatory Guide addressing this subject is in preparation by the Division of Reactor Standards.
Subject to adoption of the above recommended examination policy, the Committee believes that failure of the bolted joint is not a signifi-cant hazard. <
3.6.6 Other Failure Locations. Other conceivable failure modes include a ligament crack between control-rod nozzles propagating from nozzle to nozzle across the head, a circumferential nozzle crack propagating into the vessel (including a crack originating in a nozzle serving as a ves-sel support), bell-mouthing (virtually impossiale in a vessel with heavy flanges), and blow-out of a large nozzle or a piece of vessel wall in an otherwise unimpaired vessel (not reasonably conceivable). These failure modes are considered to have a negligibic probability of occurrence.
Failure of a control-rod nozzle in a manner leading to control rod election could lead to an uncontrolled accident.
It is the opinion of the Committee that nozzle-to-head joints are designed in a manner that makes the probability of nozzle ejection very small in comparison with the other low probability failures comprising the overall f ailure rate.
3.7 Examples of Defects in Reactor Vessels There have been several instances where errors in design, f ab rica t ion,
or construction have caused defects to form within the reactor pressure boundary. In some instances operating conditions have contributed to the generation or growth of defects. The following is a representative, but not all-inclusive, listing of incidents that have led to defects of varying significance in reactor vessels.
3.7.1 Cladding Problems. Early reactors such as Yankee Rowe and the Experimental Boiling Water Reactor (EBWR) used wrought stainless steel spot-welded to the vessel wall as a cladding. Cracking was observed in the EBWR cladding, predominantly in the steam region, after several years of service. This cracking was attributed to stress-rupture; however, there is some evidence that stress-corrosion or corrosion-fatigue con-
[
tributed to the cracking. At Yankee Rowe no cracking was observed in the reactor vessel cladding, but the similar cladding in the pressurizer cracked in the steam zone. The cracks at EBWR and Yankee did not propa-gate into the vessel wall.
At Yankee Rowe, a surveillance capsule containing tensile and impact specimens broke loose from its supports and perforated the cladding of the bottom head by wear.
Calculations indicated that the rates of hydrogen generation from such sources as galvanic coupling, ferritic steel corrosion, and radiolysis were much too small to have a signifi-cant sffect on the exposed vessel material. The steels used in the pres-sure vessel were of low yield strength and relatively immune to hydrogen embrittlement. _ _ _ - _ _
The Elk River Reactor (ERR), Monticello, and Japanese JPDR BWR reactor vessels all used a weld overlay cladding of stainless steel similar to Type 308.
Cracking was observed in the ERR cladding prior to startup, and enhanced c racking was cbserved after one year of operacion. S imila r behavior was observed at JPDR and Monticello. The intergranular cracking at ERR was believed tc be due to excessive mixing of base metal with weld metal forming a nartensitic alloy. The Monticello and JPDR cracks were believed to be due to low levels of delta ferrite. The cracks did not propagate into the base metal.
The Slc pressure vessel, a prot;cype for submarine reactors, suffered cracking of the austenitic cladding due to stress corrosion. A Viton gasket (a fluoro-organic compound), used as a head seal during hot standby tests, decomposed, releasing fluoride ions that caused cracking in the cladding; the cracks did not penetrate into the vessel wall.
The Sequoyah pressure vessel was observcd to contain cladding cracks after the acceptance hydro tests.
Stress corrosion in conjunction with the low delta ferrite content of the weld overlay cladding was considered to be responsible. The cracking was superficial.
Underclad cracking has occurred in several pressure vessels, including Zien and Surry.
These c racks, which were quite small, resulted when high heat input strip or multi-wire weld overlay processes were used to deposit the austenitic stainless steel cladding. Analytic studies o f the significance of these cracks indicated that there should be very limited growth in c rack size throughout the projected life of the vessels. There is therefore no rcason for concern about disruptive failure initiating from these cracks.
l l
l The Ccmmittee believes, based on experimental and analytic data, that s
the rate of propagation of small cracks under weld overlay is acceptably low.
New precedures eliminate the underclad cracking problem.
3.7.2 Bolts and Studs.
There have been several instances of cracking or failure of bolts and studs in nuclear service. Generally, failures have been due to excess ive ha rdness in the bolting materials which caused the critical flaw size to be small, or to a combination of stress corro-sien and tmbrittlement.
The SM-1 reactor, a small power plant developed for the U.S. Army, suf-fered failure of two Type 410 stainless steel main-flan;e bolts during operation.
Failures were attributed to a combination of excessive hard-in this ma rtensitic stainless steel and stress corrosion cracking.
ness The two f acto rs combined to form a critical size flaw that led to rapid failure of the twa bolts. This material is not used in commercial nuclear reactor vessels. _ _ _ _ _
The Palisades reactor suf fered corrosion attack of the main-flange studs due to leakage of the primary coolant and concentration by evaporation of the borate and lithium hydroxide.
3.7.3 Reactor Nozzles and Sa fe-Ends.
A complete circumferential failure in a safe-end corresponds to a double-ended pipe break; however, there is a very remote possibility that a crack might propagate from the safe end into the nozzle.
Partial safe-end failures have occurred at Elk River, lACBWR, and Nine Mile Point due to a combination of furnace sensitized stainless steel, high stresses, which in one instance were very high due to an error in piping design, and an oxygen-rich environment. The most severe failure was at Nine Mile Point where inadequate attention was given to stresses induced by system thermal expansion. One nozzle exhibited a through-wall circumferential crack which extended over
'O-180.
Fabrication cracks have been detected in Fermi-2, llatch-1, LaSalle-1, and Shoreham reactor vessel nozzles. The cracks occurred in the welds between the nozzle and the vessel shell. These cracks occurred near the center line of the vessel wall and were caused by a combination of setup and welding procedures. All occurred in the same fabrication facility. The cracks were not detected by the radiographic inspection required in ASME Section III, nor by a cursory non-code ultrasonic inspection. For the llatch-1 vessel the cracks were detected af ter "N" Code stamping and during the ultrasonic baseline examination required by ASME Section XI.
These cracked welds were repaired and positive taken to modify Section XI to review the significance of measures such weld defects. The cracks in the other vessels were detected by ins pec t ion in the shop. Welding procedures have been modified to climinate this problem.
l Defects have been observed in nozzle welds of one PWR, D. C. Cook, and on the shell side of a nozzle weld heat-af fected zone interface in another PWR, Trojan. In the first instance (Cook) incorrect position-ing of the source during radiography resulted in the weld not being examined in several nozzles. Subsequent ultrasonic examination af ter the hydro test detected slag, porosity and two cases of nonfusion.
These de fects were removed and RT procedures modified.
Planar defects were detected on the shell side of the weld heat-af fected zone during UT examination. These defects, which were removed, were attributed to inadequate metal removal of the arc gcuged region on the shell cut-out for the nozzle. The residual layer exposed to the high t emperatures of are gouging apparently degraded the properties or resulted in a plane of weakness.
Procedures have been modified to ensure suf ficient metal to prevent recurrence of this problem.
3.7.4 Reactor Vessel Attachments and Internals. The Oyster Creek and Tarapur BWR reactor vessels suffered cracking in the stub tubes and in welded-on internals. Since the internals were installed prior to post-weld heat treatment o f the vessels, the Type 304 stainless steel was l
I severely sensitized during the subsequent heat treatment. In the case of the stub tubes the design was such that very high residual stresses existed near the stub-tube-to-vessel weld. The lack of environmental controls, while at the site for Oyster Creek and during shipment for the Tarapur vessels, f avored stress corrosion cracking, which was the mechanism considered responsible for the failures. Corrective meas-ures included modification of the stub tube design and substitution of Ni-Cr-Fe Alloy 600 for Type 304 stainless steel.
These stub tubes are located in the lower head, but loss of water fol-lowing failure of a stub :ube weld would be easily compensated for by the ECCS.
3.7.5 Ma jo r Core Internals. Examples of operational failures of poten-tlat concern to the reactor vessel boundary are the thermal shield failures in SENA, SEl.NI, and Yankee Rowe.
The thermal shield design is unique to these vessels, where the throat of the vessel is narrow relative to the vessel diameter. As a result, the thermal shield con-sis ted o f segment s inserted separately and then bolted together. The variable hydraulic loads occurring during operation caused flexing and ultimate failure of the bolting and straps. The SENA and SELNI plants also incurred failure of a large fraction of the bolts connecting i
the core barrel courses in each plant. It is noted that reactor designs have been changed so that core barrels are now of one piece construction and thermal shields, where employed, are generally not segmented.
The econee reactor vessel suf fered f ailure of several core internal components during prestartup flow testing due to hydraulically in-duced vibrations, and major design changes were made to remedy this problem; the pressure boundary was not jeopardized.
l The inadvertent introduction of chlorides into the primary system has caused stress corrosion cracking of austenitic stainless steel internals and small cracks in the stainless steel cladding in a few reactors, but this condition does not affect the integrity of the pressure boundary which is of ferritic steel.
3.7.6 Operating Incidents. Several operating incidents of potential concern to reactor vessels have occurred.
In the case of the SL-1, a small experimental reactor, a major reactivity incident occurred, because of an improper action while shut down, which melted the core, generating one-hundred to two-hundred megawatt-seconds of energy, and causing bulging of the vessel and vessel jump.
PWR, developed a leak in the upper head region.
1he Swiss NOK reactor, a As in the Palisades reactor (Sec. 3.7.2), water containing sodium tetra-borate and lithium hydroxide leaked, and the solutes were concentrated by evaporation on the outside of the upper head. Corrosion occurred over a substantial portion of the head, with penetration as deep as 2-3 cm.,
3.7.7 Conclusion. Even though these events have had no reactor vessel safety consequences, their occurrence emphasizes the need for a strong quality assurance proFram in design, fabrication, construction, and operation.
3.8 Accident Conditions that Could Cause Vessel Failure It is not within the scope of this report to consider the causes and ultimate consequences of all accidents that conceivably could lead to loss of vessel integrity. This subsection gives examples of a few possible causes of accidents not covered by the failure rate discussed i
in Sec. 5 and gives a general indication of possible vessel response to such accidents.
3.8.1 Reactivity Insertion. A possible, although highly unlikely, cause of conditions that may jeopardize vessel integrity is an uncontrolled reactivity insertion such as might result from ATWS (anticipated transient without scram) without provision for accommodation; loss of control arising from operator error, sabotage, fire or missiles; sudden change in core configuration or composition, including mechanical failure of core components, control-rod drop-out or ejection, rapid change in moderator density or composition, etc.
The SL-1 incident described in Sec. 3.7.6 is an example of uncontrolled reactivity insertion.
The immediate consequence of a large reactivity insertion is an increase in pressure, an effect that may be autocatalytic if the reactor has a strong positive pressure coefficient of reactivity, as in a BWR. A fast pressure transient may also cause disarrangement of the core, which could either mitigate or aggravate the reactivity transient. Meltdown is a possible sequel.
The probability of inserting sufficient reactivity to provide the energy necessary to cause vessel jump, including pipe shearing, is believed to be extremely low in large LWRs.
3.8.2 Response to Over-Pressurization. Although over-pressurization may result from a large reactivity insertion, as described in Sec. 3.8.1, it could also result from inadequate pressure-relieving capacity for conditions not covered by design specifications, e.g., failure of shut-down mechanisms coupled with loss of heat sink. Depending on the pres-sure, the vessel would be expected to exhibit minor distortion; general bulging; a limited ductile split; or, in the unlikely event of gross overpressurization, complete disruption..
\\
l 3.8.3 Departure from Heating and cooling Cycles. Failure to follow specified heating and cooling procedures, or inadvertent rapid cooling by depressurization or cold water injection could lead to a prohibited temperature-pressure condition. Such events increase the potential for crack growth in varying degree. However, considerable departure from heating and cooling procedures can be tolerated because of conser-vatism in specifications. A fracture from local rapid chilling should be limited to the affected region. General rapid chilling would be ac-companied by depressurization, which would reduce the potential for failure.
The Committee believes that these considerations may have greater significance in vessels that have been subjected to high irradiation and recommends continued consideration of this problem.
3.8.4 Response to Seismic Forces. Postulated vessel failures caused by an earthquake could result from failure to scram, failure of inter-nals or core components, failure of the skirt or nozzles supporting the vessel, failure of substructures transmitting the load to founda-tions, or multiple nozzle failures. The design takes into account earthquakes up to the Safe-Shutdown Earthquake.
The possibility of multiple failures of external piping unrelated to vessel movement or failure should be considered separately and is outside the context of this report.
Current designs are believed to provide adequate assurance against vessel failure or vessel movement for carthquakes up to the S-fe Shutdown Earthquake.
3.8.5 Response to Missiles. Active systems required for safe shutdown
)
of the reactor are more vulnerable than the reactor vessel to externally and internally generated missiles. The Committee believes that missiles for which such systems are protected could not directly cause a vessel failure.
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4 OPERATIONAL CONSIDERATIONS The service reliability of a vessel is strongly dependent on opera-tional loading and environmental conditions, and on the effectiveness of inspection and surveillance measures to ensure the initial and con-tieuing quality o f the vessel.
4.1 Operational Limits on Applied Loads The operational limits for which reactor vessels are designed include expected conditionq o f normal operat ion, anticipated deviations from no rmal conditions, and low probability emergency and faulted conditions.
It is possible to conceive even more severe conditions than those covered by design, but the probability of accidents that lead to such conditions is extremely low.
The conditions associated with such highly improbable events a re not, and should not be, included in the vessel l
Design Specification. The Committee 1.
believes that the current procedure of excluding condi-tions for extremely low-probability events (i.e., beyond faulted conditions) f rom the Design Specification is proper and desirable; to design for such events could well lead to overall reduction in safety by forcing designers to such measures as the use of extreme wall thickness or I
higher strength steels having lower toughness, 1
2.
believes, however, that such events should be taken into i
account by other means, 1
3.
suggests in Sec. 4.1.4, how they can be taken into account if an overall evaluation of reliability d'_tates the need.
4.1.1 Operating Conditions Covered by the Design Specification and Code Desian Stress Limits.
Five operating conditions are recognized by the Code:
1.
No rma l condi tions (defined in NB-3113.1 of Section III of the Code) cover all intentional modes of operation (including power operation, startup, and shutdown) except testing.
2.
Upset conditions (de fined in NB-3113.2) include unintentional de-viations f rom normal conditions that must reasonably be expected to occur with moderate frequency during the life of the plant.
3.
Emergency conditions (defined in NB-3113.3) are infrequent incidents that must be provided for in the design, most of which have a low probability of occurrence during the life of the plant.. - _
4.
Faulted conditions (defined in NB-3113.4) are extremely low-probability postulated accidents that may impair the integrity of the system, and whose consequences involve consideration of the health and safety of the public.
5.
Testing conditions (defined in NB-3114) include ten pressure tests at not less than 25 percent over design pressure, and such other tests as may be required by the Design Specification and by the inservice inspection code requirements.
Associated with each of these operating categories is a set of allowable stresses which are related to the failure modes pertinent to the specific operating category. The allowable stresses vary from category to category.
For example, fatigue evaluation is required for normal and upset conditions, but is not required for emergency and faulted conditions.
For loadings imposed in the lower temperature ranges of heating and cooling, material fracture toughness properties are specified to protect against crack propagation.
4.1.2 Abnormal Conditions not Covered by Design Specification. Severe overpressure cannot occur in fossil-fueled boilers, because the safety valves are capable of discharging at the full rated boiler capacity, and because the combustion rate is limited by fan ratings. In contrast, reactors are generally not provided with safety valve capacity suf ficient to remove from the primary system, on a continuous basis, the steam equivalent to the nominal recctor power rating. Mc reover, postulated low probability nuclear transients may cause a large increase in power above nominal rating (Sec. 3.8.1). The basis for setting the relief capacity is defined in the Overpressure Protection Report, which is subject to review by the AEC. The Committee believes that further studies should be performed to determine the degree of primary system safety valve capacity most desirable from the point of view of safety.
Conditions of severe overpressure which have no counterpart in fossil-fueled boilers, are not factored into the Design Specification.
Section 5.2.2. suggests a procedure for consideration of those extreme incidents that appear reasonably conceivable. Some consideration might be given to designing systems with added protection against particular accidents 11 analysis should show that significant reduction in risk can be achieved.
4.2 Environmental Influences on Ductile Response Several environmental factors may af fect the notch toughness of reactor vessel steels. The most obvious is neutron fluence; however, coolant chemistry may play a role, depending on the type of steel, fabrication history, and control of operating parameters.
4.2.1 Neutron Fluence.
Pressure vessel steels SA-533 ggd SA-508 are relatively9 insensitive to neutron fluences below 5-8x10 nvt ( >> 1 M ev).
1-5x10 nvt there may be a shift in the transition temperature of as At much as 200-300 F.
This is of ten accompanied by a decrease in upper shelf energy; however, most upper shelf values are well above 50 f t-lb in the worst case of Charpy C specimens taken in the weak direction of y
the plate. The shifts in transition temperature cited at the higher fluences are limited to the belt line region at the inner surface, and the stresses are well understood in this region of simple geometry. The transition temperature shift decreases rapidly as a function of depth through the wall f rom inside to outside because of the fluence drop through the wall.
It may be noted that, if circumstances should lead to long term reactor operation at lower than usual temperature, a greater than normal shif t of transition temperature and decrease in upper shelf energy might be expected.
l The principal effect of high fluence on KIc is to shift the K curve Ic function of temperature to higher temperatures as noted in Fig. 4-1.
as a Basically, this shift is comparable to that of the C and DT curves due y
19 to irradiatien. Even at high fluences,,-,5x10 nyt ( ;> 1 Mev), the KIc values are quite large, ranching 150 KSI)' In" (see Fig. 4-1) well below the temperature at which pressurization increases markedly in a PWR.
If very high loading rates in the pressure vessel are assumed, then kid values would be more appropriate than K c.
Conceivably, either rapid pressuriza-I tion during an ATWS event or water slugging might produce such loading Generally, K d values equal to KIc will occur at a temperature rates.
I approximately 600F higher in the transition temperature range. While there are reliable kid values for unirradiated SA-508 and SA-533 steels, the data are very limited for.rradiated material. Therefore, it would I
be desirable to obtain kid values for irradiated specimens. Major interest is in values near the heel of the curve where K Id is I w but beginning to increase rapidly with temperature. Subsized specimens, such pre-cracked Charpy V-Notch specimens, should yield valid results at as the lower temperatures. The significant factor is to establish whether the shif t of kid with respect to K f r irradiated steels is equal to, Ic greater than, or less than the shif t of kid with respect to KIc for untrradiated steel.
The steel in a BWR vessel, as currently designed, will show very limited shift in NDT ( eC 100 F) because of the low neutron fluence.
In the belt line region of PWRs, since neutron damage may shift the NDT markedly by the end of life in vessels made of material sensitive to radiation, surveillance programs, as required by the AEC, are necessary until sufficient data are available on which to base any decision on possible relaxation of general requirements. -
As stated in Sec. 2.1.3, in some of the newer pressure vessels the level of residual elements, particularly phosphorus and copper, has been controlled in the plates or forgings in the core region. SA-533 or SA-508 steels with low P and Cu undergo very limgted NDT shift or upper shelf depression at fluences as high as 5x10 nyt.
4.2.2 Vessel Temperature Control. A problem inherent in reactor opera-tion is the possibility of ovororessurization while near room temperature, at which critical flaw sizes are smaller than at operating temperature.
Appropriate administrative procedures should prevent this. It is important that such procedures be reviewed carefully by the AEC in the Technical Specifications for every plant.
Heating or cooling the primary system is controlled by administrative procedures. Rigorous control is required since unduly rapid and numerous heatup or cooldown transients may result in the development and/or propa-gation of fatigue cracks. The Code provides a basis for determining the permissible transient conditions, including numoer of cycles and tempera-ture changes, for those events categorized as Normal or Upset Conditions.
The procedure is conservative, so that a margin of safety exists.
The Code rules do not provide for consideration of transient conditions other than those specified, nor do they explicitly provide fracture toughness requirements for Emergency or Faulted Conditions. While the large reservoir of water in the vessel will tend to minimize rapid changes there is no assurance that cold water injection cannot drop in temperature, the wall temperature rapidly. Such events must be considered in the system design, are subject to review as part of the licensing procedure, and are restricted by the Technical Specifications. If an unexpected event should occur, the AEC review procedure requires evaluation of the past and future consequences of the event. At this stage, the Code procedures again provide a basis for evaluation.
4.2.3 Coolant Chemistry. The coolant chemistry of BWRs is approximately neutral pH and about one ppm of oxygen. Stainless steel sensitized under certain conditions is subject to stress corrosion cracking in such an environment. In particular, sensitized safe ends or piping may be attacked.
In the case of the reactor vessel, the delta ferrite in the stainless steel overlay cladding should minimize stress corrosion cracking. The ferritic steels are not sensitive to stress corrosion in BWR environments. While hydrogen can be formed by radiolytic decomposition, the amount entering the vessel wall should be small and it would dif fuse through rapidly. The SA-533 and SA-508 steels have suf ficiently low yield points that hydrogen embrittlement should not be a problem. Little or no loss in toughness occurs at 1-2 ppm of hydrogen. l
A PWR operates with a hydrogen overpressure, and the pH is usually con-trolled by lithium hydroxide. Unless abnormally high concentrations of the lithium hyd roxide occur, the probabi.lity of caustic stress corrosion of the stainless steel is believed to be low.
Caustic embrittlement of the ferritic steel is a remote possibility, provided a suitable concentration mechanism exists.
It is difficult to postulate a concentration mechanism inside the reactor; however, leakage and evaporation could result in concentration on the outer surface as reported in Sec. 3.7.2 and Sec. 3.7.6.
4.3 Preservice and Inservice Inspection AEC regulations (10 CFR 50.55a " Codes and Standards for Nuclear Power Plants") currently require compliance with Section III and Section XI of the ASME Code.
Examinations of Section III vessels employ ultrasonic techniques for volumetric examinations, and liquid penetrant or magnetic particle techniques for surface examinations of the pressure-retaining welds, including the base material for at least one wall thickness beyond the ed o f :. k C.
M L L,, completion of fabrication, testing, exami-I nation, and certi fication in accordance with the construction rules of Section III, and as part of the requirements of Section XI, reactor vessels are required to be further examined nondest ructively prior to, and as a condition for, placement into nuclear power plant service.
These examinations are totally unlike those generally applied to the ins pect ion o f fossil-fueled power plant boiler drums. Inservice in-spection programs for steam boiler d rums, which have been formalized by the National Board Inspection Code, have evolved as a result of l
s pec i fic requ i remen t s imposed by state jurisdictional authorities.
Visual examinations and hydrostatic tests are, in general, the extent of inspection performed in such cases.
4.3.1 Preservice Inspection. The Section XI rules require ultrasonic mapping of all areas of the reactor vessel that a re subject to periodic inservice examination. The ultrasonic examination is extended to in-clude essentially 100 percent of the pressure-retaining welds. The Committee believes that preservice examinations serve two useful purposes.
First, these examinations provide record of any indications, a
such as extremely small flaws, within the limits pe itted by the allow-able indication standa rds of Section XI, under which rules the preservice examinations of the reactor vessel must be conducted. The ultrasonic mapping of small flaws provides a basis for future comparison with the results of subsequent periodic examinations, in the event the flaws grow in size during service.
Evaluation and analysis of the influence on the structural integrity of any detected chanr 2 in size of flaws provide, in turn, a means to assess the safety of the vessel for continued service. -
l The-second purpose served by the preservice examinations is a final confirmation of the structural quality of the reactor vessel before its placement into initial service. These examinations are required despite the fact that all welds and base materials were examined 100 percent volumetrically during the course of vessel fabrication and that the examinations successfully met the acceptance standards for the inspection techniques specified by the construction rules of Section III. The Committee believes such preservice examinations not only provide increased assurance that the vessel meets the intended quality for initial placement into service but also reduce substantially the risk of any significant fabrication flaws escaping detection.
The rules of Section XI of the Code include provisions for establishing requirements pertaining specifically to records and reports of the pre-service examination as well as inservice examinations. Such reports are required to be filed with the enforcement authority having jurisdiction.
4.3.2 Preservice Hydrostatic Pressure Testing. With respect to pre-service testing of reactor vessels,Section III requires a hydrostatic pressure test following completion of fabrication. Although such a test will ger.erally subject the vessel to an overpressure 25 percent greater than the design pressure, it is pertinent to note that, except for very localized areas of geometric discontinuity (e.g., nozzles, flange-to-shell joint, flange-to-spherical head junction), the imposed test stress is significantly below the yield strength of the vessel material -- approximately two-thirds of the specified minimum yield strength at test temperature. For the major portion of the vessel, the hydrostatic test stress is well within the clastic range (i.e., no permanent deformation) of the materials of construction.
Hydrostatic tests are useful to confirm vessel integrity by exposing excessive deformations, faulty materials, or leaks. The initial pre-service hydrostatic test is valuable as a means of attaining " shake-down" in localized stress regions where peak stresses exceed the yield strength. " Shakedown" by localized plastic deformation permits subse-quent loading response to be elastic. The preservice hydrostatic tests also provide an opportunity for exposing the potential for flaw propaga-tion by use of the newly developed acoustic emission techniques.
Frequent postservice tests may not be of essential value for Section III vessels where quality controls over design and construction have been rigorously applied. The overstress required for the hydrostatic test is included as a design loading in the stress analysis but each test still uses some of the fatigue strength of the vessel and thus causes some strength degradation. Furthe rmo re, the hydrostatic test does expose the vessel to crack propagation conditions if steps are not taken to assure a test temperature at which the fracture toughness pro-perties of the vessel are adequate. Thus, positive and negative fac to rs influence hydrostatic test desirability. -
The Committee believes the initial preservice hydrostatic test to have considerable value in verifying vessel strength capability. The Committee also believes a second preservice hydrostatic test on the vessel should be perforced af ter installation, and prior to the baseline mapping required by Section XI.
These test procedures include extensive nondestructive examination not normally performed on boiler d rums and thus enhance the probability that significant flaws are eliminated prior to in-service use.
This flaw elimination should increase the reliability of nuclear vessels beyond the already satisf actory performance of boiler drums, subj ect to the qualification that care is taken to test the vessels at temperatures at which adequate fracture toughness is assured.
4.3.3 Inservice Inspection and Flaw Growth Detection. Compliance with the AEC regulation on Codes and Standards for Nuclear Power Plants im-poses the requirements of inservice inspections in accordance with the rules of Section XI of the ASME Code.Section XI identifies the prin-cipal vessel areas and the extent of these areas to be periodically examined over the entire service lifetime of the reactor vessel. The j
principal areas of the vessel selected for examination are (1) those
}
that are more highly stressed (e.g., nozzle-shell junction), (2) those components for which a representative examination sampling provides an assessment of the overall structural condition of the component materials (e.g., shell weld joints), and (3) those portions of the vessel where environmental effects and irradiation could influence the properties of the vessel materials (e.g., reactor beltline region).
These inservice examinations represent the first application of non-destructive examination techniques to monitor the structural conditions l
of vessels periodically during their service lifetime, and will prove an f
effective means of monitoring the integrity of reactor vessels. Applica-l tion of Section XI of the Code has resulted in many changes in reactor l
vessel design in order to facilitate inspections, as well as modifications of plant layouts to provide more convenient access for the conduct of the I
inspections. As an example, an attempt is made to relocate weld seams to avoid intersection with control-rod nozzle openings. This rearrange-ment permits unobstructed access for examination of these important pressure-retaining welds in the vessel head.
The required inspection frequency of reactor vessel components pre-scribed by Section XI has stimulated industry to develop specialized inservice inspection equipment. Such equipment is expected to provide utilities with means to perform the required examinations by remotely operable instruments and to minimize the time required to conduct the inspections. _ - _ _ _ _ _ _ _ _.
The Committee believes that experience and further development will show that inservice ultrasonic examination of welds is not a very time-consuming or difficult operation. If this is demonstrated, consideration should therefore be given to requiring more extensive vessel inservice ultrasonic examination than presently required in Section XI, including re-examination of highly irradiated plate. It may be noted that licensing authorities in the Federal Republic of Germany, and perhaps in other countries, already require more extensive e:< amination.
4.3.4 Evaluation of Flaws Detected by Inservice Inspection. The application of nondestructive testing techniques for inservice inspection has introduced a new basis for reassurance. Defects detected by the examinations can now be evaluated on an individual basis in order to determine the structural adequacy of the reactor vessel for continued service. Recent developments in the field of fracture mechanics have introduced a practical engineering technique which permits analysis of the influence of defects on the f racture potential of a vessel component and have provided a basis for develop-of a more practical approach for nondestructive examination ment acceptanec standards as specified in the Summer 1972, Addenda to Section XI.
In order to gain confidence in the application of fracture mechanics in the evaluation of inservice inspection results, additional infor-mation will be required with respect to material behavior in the presence of defects. The fracture toughness properties of base metal, welds, and heat affected zones of the vessel components must be availabic to the fracture mechanics analyst. The characterization of the defect must be precisely or conservatively defined, in terms of size, extent, orientation, and location with respect to the stress field in the component, to permit a meaningful evaluation. In addition, knowledge of the rate of growth of flaws from a suberitical size to within an acceptable margin of the critical size becomes of paramount importance in determining the useful and safe operating life of a vessel with known or postulated defects.
To satisfy these requirements, adequate information concerning material properties and the geometry and location of defects is a prerequisite for any fracture mechanics analysis. Much progress has been made recently in obtaining fracture toughness properties of materials used in reactor vessels. Improvements in defining examination techniques which lead to reproducible results are needed, however, to assure optimum defect characterization and location in the reactor vessel.
Current efforts underway on the part of industry in the areas of acoustic emission, acoustic holography, and acoustic spectroscopy are _ _ _ _ _ _ _ _ _ -
expected eventually to yield these required improvements. These techniques should provide a basic tool for determining more precisely flaw location and growth characteristics in reactor vessels and, as such, should make an important additional contribution to vessel reliability.
The Committee believes that the conduct of inservice inspections in accordance with the rules of Section XI constitutes a major con-tribution toward increasing the reliability of reactor vessels in service because such inspections are expected to yield meaningful information concerning the structural integrity of the vessel throughout the service lifetime.
4.3.5 Inservice Hydrostatic Testing. Inservice hydrostatic testing of the reactor vessel, as required by Section XI, is conducted in con-junction with the system hydrostatic test, since in the majority of current system designs the vessel cannot be isolated from the reactor coolant system. These system tests are required to be performed at or near the end of each 10-year service period. The test pressure is established at a level such that the imposed test pressure strains on the vessel correspond closely to those which the vessel experiences during normal operation.
Although not c requirement of Section XI, the reactor vessel is sub-jected to a leaktightness hydrostatic test prior to each plant startup.
Such tests are generally conducted at a pressure level not in excess of the nominal operating pressure. Both the inservice hydrostatic pressure tests (Section XI) and the leaktightness pressure tests are, in conse-quence, conducted at pressures below the preservice hydrostatic test pressure which the reactor vessel sustained initially.
4.3.6 Periodic Proof Testing Above Operating Pressure.
Periodic proof testing is required for boilers and many non-nuclear pressure vessels, although there is no uniformity in selection of test pressure and frequency.
Present understanding of the mechanism of fracture raises questions, pro and con, concerning the desirability of periodic proof testing at substantial overpressure and concerning appropriate pressures and temperatures if such tests are conducted.
Periodic overpressure testing is not required for reactor vessels, but the Committee believes elimination of periodic tests should not be regarded as a firmly established practice. Continuing research may show that periodic overpressurization is, in fact, beneficial, parti-cularly if such tests can be monitored by improved acoustic emission techniques.
But the Committee does not endorse periodic hydrostatic tests without meaningful accompanying nondestructive examination because the cumulative fatigue damage which must be accepted may not be compensated by improvement in the vessel reliability confidence level. -_.
The Committee believes that the option to perform periodic overpressure tests should be provided and that consideration should be given to the problems of designing nuclear steam supply systems so that proof testing of the reactor vessel to at least design pressure will be practical. The Committee also believes that post service as well as preservice repairs should be verified by hydrostatic test.
4.4 Failure-Limitine Provisions As stated in Sec. 1.1, the scope of this report does not include con-sideration of the consequences of a vessel failure or of means to ameliorate the consequences. Ilowever, a brief description is given here of some studies of systems intended to limit the consequences of a vessel rupture.
4.4.1 Study in the Federal Republic of Germany. An approach now being studied in the FRG for PWRs is the addition of a secondary non-pressure-tight prestressed concrete (or possibly steel) shell around the primary reactor vessel. The secondary shell limits the movement of the primary vessel wall in the event of a fracture. The blowdown rate through the openings in the secondary shell is less than in the postulated double-ended pipe break (DEPB) for which emergency cooling is provided. In the event of a failure in the primary vessel, this secondary shell is intended to prevent a gross breach from exceeding the DEPB and to absorb the missile energy of flying debris involved in a postulated gross rupture of the primary vessel. A criterion of the system is that core geometry will be maintained in the event of a vessel rupture. It has been suggested that the space between the vessels be filled with a suitable packing, possibly steel balls.
Secondary restraint for the head is provided by a holddown device sur-rounding the vessel flange and attached to the head. The device engages with the vessel flange if the vessel head bolts fail and the vessel head lifts off the vessel. This approach would tend to minimize the consequences of a gross rupture beyond the ECCS capability, but it, too, involves some arbitrary failure assumptions. It may not be well suited to BWR applications because the bottom control rod nozzle configuration requires a large number of penetrations, and the large diameter of the reactor vessel and its head flange impose major structural complications and greater cost.
4.4.2 Double-Walled Vessel. Another approach for providing structural redundancy in reactor vessels which has been studied in the US uses an internal vessel to supplement the capability of the primary reactor vessel. The approach consists essentially of a conventional primary vessel into which a secondary vessel with a separate head is inserted.
The inner vessel is keyed to the primary vessel structure for dimensional control and is capable of holding primary coolant in the event of a l
primary (outer) vessel rupture up to a 5-ft-diameter opening, a l
3-ft-wide split for the full length of a vertical seam, a top head I
rupture, or a couplete circumferential scam rupture. Primary and inner vessel heads both use the same holddown flange and bolts.
The outer vessel would be the primary pressure boundary, and its stress design would be conventional. The inner vessel would not be Icaktight and would be subjected to significant pressure only in the event of a rupture of the outer vessel; it would be designed with relatively thin walls, and correspondingly high membrane stresses would be developed in the event of a failure of the outer vessel.
One advantage of this concept lies in the ability to limit the blow-1 j
d own forces resulting from a primary vessel rupture since the inner vessel would serve as a flow restrictor to limit the blowdown rate.
Thus, missile forces would be controllable in a way that vould not j eopardize the containment. In addition, the inner vessel would serve as a backup container for the shutdown core coolant. Flange bolt failure backup is provided by a removable holddown ring keyed j
into the vessel support structure.
This concept has attractive reliability features. Its valce is signifi-l cant only if the failure modes under consideration initiate at the specified design pressure conditions. For accident conditions involving a failure pressure substantially (by a f actor of about 3 or larger) above the design level, there is probably no important gain in re li ab i li t y. The complexity of this concept compared to the FRG concept is difficult to judge. The reliability value with respect to the FRG concept would depend on the failure initiation mechanism for the primary coolant system.
Neither the US nor the FRG design has been carried to a stage where all of the engineering details are fully developed, but considerable increase in engineering complexity for the vessel arrangement is obvious. Inservice inspection capability is claimed for both the US and FRG systems.
4.4.3 Confinement by the Biological Shield. The reactor vessel of a PWR is enclosed in a concrete well, which also serves as a biological shield surrounding the vessel. In three PWR plants, the biological shield has been suf ficiently reinforced to contain the vessel in the event of a longitudinal split or other rupture that could cause lateral movement of the vessel.
The PLOCAP (Pos t Loss-o f-Coolant Accident Protection) system is a pro-posed extensicn of the simple confining system. In conventional designs, the supporting shield wall has lateral openings for personnel access, instrument leads, etc.
In PLOCAP, the cavity around and below the reactor vessel is reduced in volume and is made watertight to above the core level, and flooding tanks and other flooding systems are provided. In the event of a vessel rupture, the flooding system, in. _ _ _ _ _ _ _ _
conjunction with the ECCS, fills the cavity and the vessel to above the core Icvel. One PWR plant (two reactors) has been constructed with portions of PLOCAP systems installed and with provision for completing the installations if necessary.
A determination has been made by the Directorate of Licensing, and concurred in by ACRS, that the degree of irradiation damage to the vessel over the next several years will not be sufficient to require a system such as PLOCAP. This determination may be revised as more data become available or, alternatively, it is envisioned that vessels may be annealed in place if necessary.
4.4.4 Committee Recommendation. The information available to the Committee is not sufficient to permit a conclusion regarding the systems described in Sec. 4.4.1 and 4.4.2.
The Committee believes further study of possible design approaches to protect against pressure vessel f ailure should be performed. !
Figure 4-1
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l l
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50 100 150 200 T E MPE R ATUR E, F
PRE AND POST IRRADIATION TEMPERATURE DEPENDENCE OF THE FRACTURE TOUGHNESS OF A533 B CLASS 1 STEEL (HSST PLATE 02)
- Ref T. R. Mager, " Irradiation of Thick Section Compact Tension Specimens," Paper 22 at HSST Annual Information Meeting, April 25-26,19 72 3332-24 Figure on Page 2.
5 PRESSURE VESSEL FAILURE STATISTICS AND FAILURE PROBABILITIES Operating experience with nuclear reactor pressure vessels is inade-quate, both with respect to the number of vessel-years of service and the number of known defects, to permit a direct statistical determina-tion of the probability of failure. For this reason, an alternate approach has been taken, involving the following steps:
1.
Consideration of operational and f ailure data for non-nuclear types of pressurized components, such as boiler drums and unfired pressure vessels (Sec. 5.3 and Sec. 5.4).
2.
Classification of these non-nuclear vessel failures into three categories dopted for this report, as discussed in Sec. 5.1.
3.
Comparison of the design, construction, inspection, and operating procedures used for boiler drums and unfired pressure vessels with those used for nuclear reactor vessels, and estimation of the ef fects of any dif ferences on the relative probability of failure of the two types of vessels (Sec. 5.8).
4.
Utilization of the information from the above steps, to appraise the probability of disruptive reactor vessel failure (Sec. 5.9).
5.1 Definition of Failure The words " accident" and " failure" have been used in the various reports available to cover a wide range of incidents requiring pressure vessels or boilers to be taken out of service for repairs. In some cases,
" failures" have meant failures of boiler tubes rather than the vessel shell. In other cases, the term " catastrophic failure" has been used for failures that are " catastrophic", not in the sense of consequences to the public, but in the more limited sense of a need to shut down for major repairs to the boiler. To establish a definitive basis for the analysis of vessel performance statistics, " vessel failure" is defined herein as a condition in which a crack, leak, or other defect has de-veloped in the pressure retaining components of the vessel, requiring repair or replacement of the vessel. Two major categories of vessel failure are considered, disruptive and nondtsruptive; nondisruptive vessel failures are further subdivided into "potentially disruptive" and "nonc rit ica l".
These are defined as:
l.
" Disruptive vessel failure": A breaching of the vessel by failure of the shell, head, nozzles, or bolting, accompanied by rapid release of a large volume of the contained pressurized fluid.
2.
"Nondisruptive failures";
"Potentially disruptive vessel failure": A condition a.
of crack growth rate or flaw size that is corrected, and which, if it had not been corrected, could have reached a critical size and led to " disruptive" vessel failure.
b.
" Noncritical vessel f ailure": A local degradation of the pressure vessel boundary that is limited to localized cracking, with or without minor leakage. Such a crack would not reach critical size and Icad to disruptive vessel failure.
The various descriptive terms used to categorize the failures discussed in Sec. 5.5 have been roughly correlcted with the above classifications.
These correlations have been based in part on quantitative data presented by the various investigators, and in part on subjective judgment. The correlations are shown in Table 5-1.
TABLE 5-1 CORRELATION OF PRESSURE VESSEL FAILURE TERMINOLOGY FROM VARIOUS SOURCES Source Pressure Vessel Failure Terminology Nondisruptive This Report Noncritical Potentially Disruptive Disruptive I
l I
i Phillips(gd I
Wa rwick
<-Potent ially Dangerous 1;:
Catastrophic-->
I I
s t
(
Failures I
Kellermann(0) l l
Serious Failures l
i 1
Failures I
l6- - - - - Ca t a s t ro ph ic Slopian - and g
Mieze I
l Kellermann I
6) and Scipel Failures Keller n,
et a1 C a t a s t ro ph ic -->
I
5.2 Types of Disruptive F2.ilure Considered in Assessment of Failure Probabilig Section Ill of the Code provides for a safe vessel for the conditions specified in the Design Speci fication, but does not quantify reliability.
The failure p robab i li t i e s given in this Section represent Committee judgment, and the conditions to which this rate applies need to be defined.
5.2.1 Failures Covered hv Probabilities.
Normal, upset, and test conditions will occur with predictable frequency and duration over the life of the plant, and are f actorel into the fatigue analysis. These are the conditions that have a clear analogy in fossil-fueled power plants, and the probabilities given apply primarily to these conditions.
Emergency and faulted conditions are associated with low probability I
events that may or may not occur during the life of the plant, and that m j or may not have an analogy in fossil-fueled plants. Cyclic
.epetition of these events is not postulated, hence a fatigue analysis l
is not required. Ilowe ve r, the events are factored into the stress analysis, and the Cor.nittee believes that, with proper conservatism,
)
these low frequency events should not contribute significantly to the i
probability of failure in Sec. 5.9.
5.2.2 Failures Not Covered by Probabilities. Section 3.8 described some improbable but conceivable accidents that could stress the reactor vessel beyond the limits associated with emergency or faulted conditions. As indicated in Sec. 4.1, the Committee does not recom-meni that these accidents be covered by the vessel Design Specification, j
and the failure probabilities given in Sec. 5 do not include failures under accident conditions.
Accident conditions must be considered separately as part of an overall accident analysis, and systems must be so designed that the resulting probability of vessel failure, taking into account the consequences, is acceptably low.
For example, the probability of vessel failure per year from a part3 cular incident may be written:
P=fxP i xP 2 where P = the probability of failure per vessel year as a result of the particular incident, f = estimated f requency of the initiating event, events per year Pt = probability that the initiating event will lead, through a chain o f circumstances, to a specified condition of overload P2 = Probability of vessel failure per specified condition of overload. -_
For a particular incident, P may be a summation for several alter-native chains of circumstances.
The accident analysis should lead to evaluation of f and P, and the 1
peak pressure to which the vessel will be subjected. P e n then be 2
estimated for the peak pressure.
5.3 Bases for Evaluation l
Data regarding "f ailures" of non-nuclear pressure vessels and boiler drums are available f rom the UK and FRG. Those data sets of sta-tistically significant size have been analyzed to obtain failure probability ranges. Comparisons have been made among the data sets to determine and reconcile statistically significant dif ferences that may exist.
The following criteria have been applied in the evaluation of the data:
1.
There is a considerable body of statistics pertaining to f ailure during the initial hydrostatic tests performed as a part of the manuf acturing process. Such f ailr es resulted f rom pressurization at the low temperature of the hydrostatic test and occurred within the testing program for qualifying the vessels for service; they did not cccur in vessels completely qualified for service and would not have occurred at operating temperature in service.
Where identified, they have been excluded from calculation of failure probability.
2.
Some types of failures are unique to non-nuclear units (for example, ruptured tubes). In such instances, appro-priate modifications to the statistics have been made.
Such modifications are noted in the pertinent sections.
The following policy has been followed in deriving failure rates and probabilities:
1.
Failure rates quoted in the literature are given as stated.
2.
Other failure rates are calculated simply by dividing the number of failures by the number of vessel-years.
3.
Where po s s ible in 1. and 2.,
the 99 percent confidence upper bound f ailure probability also is given.*
- In a probability range bounded by two numgers, thg upper bound is the larger npmber, e.g.,
in the range 10 to 10 the upper bound is 10 ____-_
A 99 percent confidence upper bound was selected for comparison of the data since it represented a suitably conservative basis. Conversion to other values such as 95 or 99.9 percent confidence upper bound, or to upper and lower confidence bounds can be made, knowing the number of f ailures and the vessel-years, and through use of tables of Chi-square values. Where a f ailure probability is based on zero f ailures, the 95 percent and 99.9 percent confidence upper bounds are, respectively, 0.65 and 1.5 times the upper bound for 99 percent confidence.
5.4 United States pressure Vessel Failure Statistics l
TFe pressure vesr,els considered most comparable to nuclear pressure vessels, based on materials, nominal design and construction practices, and operating conditions, are fossil-fueled boiler drums used by elec-tric utilities for the generation of electricity. These drums are ex-posed to water and steam at temperatures and pressures equal to or greater than those in LWRs. Operational histories on a large number of such drums and pressure vessels have been obtained as the result of a request by the USAEC to the Edison Electric Institute (l). In addition, data were obtained through the Edison Electric Institute--Tennessee Valley Authority funded program.
outhwest Research Institute,(2d) and f rom inquiries to utilities, boiler and vessel manufacturers,(2a) insurance com anies,(2b,2c) and the American Boiler Manufacturers I
Association.( ")
5.4.1 General Inquiries. Inquiries of utility operators, boiler and pressure vessel manufacturers, insurance companies, and code groups, confirm that no welded boiler drum or Section VIII pressure vessel has ever suffered a d "' gig)failureincentralstationservice ru in the United States.
Periodic inspection and timely repairs of developing defects are probably important factors in this good performance.
5.4.2 Insurance Companies. One casualty insurance company reported 150 cases of " accidents" for boilers operated at or above 600 psi in about 3,0Q0 boiler-years of operation during the years 1968, 1969, and 1970.(2b) These " accidents" were all nondisruptive failures in water-tube boilers and most were not pertinent to reactor vessels.
However, personnel of the same reporting organization recalled four or five cases occurring over approximately the past twenty years in which the failures found, while nondisruptive, may have been of the potentially disruptive type. These included one crack in a longi-tudinal welded seam, one crack in a plate, and two or three cases of c racking in manway openings, all found by routine inservice inspection.
Another casualty insurance company reported that, of approximately 100,000 pressure vessels insured, their regards show 13 " failures",
all nondtsruptive, in the last four years.(2c) Only four of the failures were in vessels two inches thick or more, and two of these vessels were of multilayer construction not of current relevance to nuclear reactors.
5.4.3 Edison Electric Institute--Tennessee Valley Authority Data.
A portion of the EEI-TVA orogram on nondestructive examination at the Southwest.Research Institute was the analysis of operati f rom a relatively limited number of electric utilities. gg data The data reported were in the form of operational years of boiler drums and descriptions of specific details concerning defects detected during the inspections. These boiler drums had been fabricated to the requirements of Section I of the ASME Code.
The EEI-TVA atudy involved participation by 55 utilities and included a total of 1033 vessels, 468 operating above 1400 psi and 565 operating between 900 and 1400 psi. The vessel of longest service was com-missioned in 1926. These vessels represent approximately 10,000 vessel-years of operation. Six utilities reported instances of subcritical flaw growth in ten of these 1033 vessels and no diaruptive f ailures (Tabic 5-2).
All of the cracks were associated with small diameter pipe penetrations and most (possibly all) of the failures appear to have been caused by thermal fatigue. All of the vessels were repaired and are currently reported to be in service. The failures are judged to be noncritical and their relevance to reactor vessels is questionable. The 99% confi--4 dence upper bound for noncritical failure probabijity would be 4.6x10 if none of the failures were relevant and 2.0x10-if all were relevant.
There were no disruptive failures, f rom which, because of the small number of vessel-years, it is only possible to conclude that the 4.6x10'{ity of disruptive f ailure (99% confidence) does not exceed probabl.
TABLE 5-2 F'AILtfRES OR DEFECTS REPORTED IN EEI-TVA STUDIES Service Li fe at Time of Detection Com ponen t (s)
(Ye a rs)
Comments on Failure or Defects Lowe r (Mud) D rums (4) 25 Cracks around 1.5 inch I.D.
blowdown nozzles.
Steam Drum ( 1) 2 Circumferential crack 11 inches long in inner weld of feedwater nozzle the rma l s leeve exterior to drum.
Lower Drums (2)
Not known Internal cracks where crossover nc.zzles were welded into d rum.
Steam Drum (1) 25 Thermal crack at feedwater nozzle.
Steam Drum (1)
Not known Crack in welded rein fo rcemen t on feedwater nozzle.
Steam Drum (1) 8 Crack in fillet weld between feedwater inlet nozzle and drum; crack length 7 57, o f weld.
10 Twa cracks observed in welds around piping entering drum.
One 5" the other 2" long.
S.4.4 Edison Electric Institute Boiler Drum and Pressure Vessel Data.
The USAEC requested the EEI to provide relevant data on operational l
histories o f boiler d rums and pressure vessels in fossil-fueled power plants.
}
Several constraints were fixed in the request to obtain data reasonably representative of conditions in nuclear pressure vessels:
1.
Baller drums and pressure vessels should have been fabricated to 1956 or later versions of Section I or Section VIII of the Code.
2.
Wall thickness should exceed 1.5 inches.
3.
Pressures and temperatures should equal or exceed those t ypica l o f LWRs.
Information relating to about 2.2xlb vessel service-years
- is avail-able to date on about 5000 vessels.
- Assumes one service year is 7500 hours0.0868 days <br />2.083 hours <br />0.0124 weeks <br />0.00285 months <br />, or approximately 85 percent plant operating factor. _ _ _ _ _ _.
one noncritical failure was reported, yielding a failure rate of 4.6x10-5 and a 99 percent confidence upper bound of 3.0x10-4 per vessel-year.
There were no potentially disruptive or disruptive failures reported.
For zero failures, the 99 percent confidence upper bound would be 2.1x10~4 for the relatively small number of 22,000 vessel service year;. The 99 percent confidence upper bound values are comparable to those obtained frem the EEI-TVA data (Sec. 5. 6. 3).
5.4.5 American Boiler Manuf acturers Association Data *. The American Boller Manuf acturers Association data pertaining to the number o f ASME Section 1 and Section VIII pressure vessels for fossil-fired water-tube boilers installed in central station electric utility plants cover over 33,000 vessels for the period 1962-1971. Most of these vessels were ASME Section I steam drums and hea ers.
The ASME Section VIII vessels d
were principally, feedwater heaters. The ABMA data were limited to vessels satisfying the same constraints specified in Section 5.3.4 for the Edison Electric Institute data.
In order to obtain a larger and thus more significant vessel population, these data for 1962-1971 have been extrapolated back to estimate the t otal number of comparable pressure vessels manuf actured for central-station boilers and feed water heaters since 1939. This extrapolation was based on statistics reported by the Committee on Power Generation of the Association of Edison Illuminating Companies, relating both to the number of boilers installed during the period 1939-1961 and to their aggregate capacity (Ib/hr steam).
In making this extrapolation, it has been assumed conservatively that the number of vessels (mainly headers) per boiler was l
proportional to the power steaming capacity of the boiler. The average l
years of service (approximately 4-1/ 3 years) sustained by the vessels in operat ion between 1962 and 1971 was determined f rom the Edison Electric Institute data.
For the earlier decades, this period was increased ac-cordingly. The results are shown in the following table which is believed to provide a conservative estimate of the number of vessels an1 the total number of vessel-years of service for the last three decades:
- A report on the ABMA data is in preparation by R. R. Maccary, Assistant Director for Engineering, L..__
1 1
Period of Number of Average Service Manufacture Vessels Years in Service Vessel-Years 1962-1971 33,000 4-1/3 142,000 1952-1961 27,000 14-1/3 387,000 h
1942-1951 8,000 24-1/3 194,000 l
l 68,000 723,000 The statistics discussed in Sec. 5.4.1 through 5.4.4 indicate that there has been no disruptive failure reported in ASME Section I boiler drums and headers or ASME Section VIII pressure vessels, such as are used in feedwater heaters in central station power plants, while the vessels were in service and operating under conditions controlled by fossil-fueled power plant practice.
For the total number of vessel-years of operation extrapolated f rom the ABMA data (723,000), and for no disruptive failures among this vesselgopulation, the upper bound probability of a disruptive failure is 6.4x10~
at 99 percent confidence level.
The data referenced in Sec. 5.3.3 for EEI-and Sec. 5.3.5 for ABMA are derived from the same population of vessels in central-station power plants in the United States. Although the EEI-TVA data do not necessarily fall entirely within the same constraints of Sec. 5.3.4, as do the EEI and ABMA data, the EEI-TVA and EEI data are basically subsets of the larger extrapolated ABMA data set.
Therefore, the most significant value is the 6.4x10-6 at 99 percent confidence upper bound for disruptive failure as determined on the basis of the extrapolation of the AINA vessel population and its associated vessel service-years.
5.4.6 United States Navy Experience. The United States Navy reports 123 reactors in operation with an accumulated total of 1075 years of reactor operation without having experienced a reactor accident. It must be reccgnized that this operating experience relates directly only to vessels of the size used in naval reactors and designed, manufactured, inspected, and operated in accordance with Navy requirements.
5.4.7 United States Commercial Reactor Experience. Commercial LWRs in the US have an accumulcted total of about 125 reactor years of operation.
Specific instances of defects found in the vessel or the vessel cladding, all nondisruptive, are discussed in Sec. 3.7.
The number of reactor years is too small to permit relevant statistical inferences of failure proba-bilities. _ _ _ _ _ _ _ _
5.5 Foreign Statistics Statistics of defects found during periodic inspection of boilers and pressure vessels have been compiled in the UK and FRG and failure rates and probabilities have been derived for those defects that are pertinent to pressure vessels. This section brings into sharp focus the confusion resulting from imprecise definition and inconsistent une of terms such as
" failure statistics" and " catastrophic failure". Many of the " failures" cited, because of their nature and locations, we re probably generated during the fabrication stage and remained virtually unchanged until derected during periodit inspection; they did not directly interfere with functional use.
aThe significance of such defects and the possibility of their growth even to detectable but noncritical size is difficult to assess.
Better terminology describing these data might be " defect statistics" rather than " failure statistics".
5.5.1 St at i stical Tr ends in pressure Vessel Failures in the United Kingdom.
A review is made in this section of analyses from the UK which deal with boiler explosions and defects in pressure vessels and piping which have a certain relevance to nuclear systems.
Compliance with pressure vessel codes is not a statutory requirement in t he UK ; applicat ion of codes is the responsibility of the user. This di f fers markedly f rom practice in most of the US where use of the ASME Code (Sections I, III, VIII, etc.) is a statutory requirement.
The principal UK study was performed by Phillips and Warwick(3) whose purpose was to relate failures in non-nuclear systems to probable failure rates in auclear systems.
In this study, the statistics analyzed covered 12,700 pressure vessels and associated systems and 100,300 vessel-years.
A total of 132 service failures in vessels and piping were discussed. An additional 10 f ailures (including three " catastrophic") occurred during hydrotests prior to heing placed in service, usually due to high NDT, and are excluded f rom consideration in the present evaluation.
The failures in service are classified in Table 5-3.
The terms "potentially dangerous" and "catast rophic", used by Phillips and Warwick, are judged to relate to the failure categories defined in this report, as shown in Table 5-1.
On the basis of the data in Table 5-3, it is possible to evaluate failure statistics for f ailures during service in terms o f vesse l-yea rs.
The results are presented in Table 5-4. _ _ _ _ _ _ _ _ _ _
TABLE 5-3 l
DISTRIBUJ_ ION OF 112 SERVICE FAILURES FROM 12,700 VESSELS _(SYSTFNS)
DURING 100,300 VESSEL-YEARS OF SERVICE (PilILLIPS AND WARWICK)(N Total Failures Relevant to Nuclear "Potentially "Potentially Dangerous"
" Catastrophic" Dancerous"
" Catastrophic" Leak No Leak Vessel Fatigue 1
3 0
2 0
l Corrosion 0
12 0
8 0
1 Preexisting Defect 1
3 0
4 s
No: Determined 1
9 0
0 0
Miscellaneous 0
0 0
0 0
M a l o pe ra t i on 0
3 4
0 0
3 30 4
14 0
Piping, Etc.
Fatigue 17 24 2
14 1
Corrosion 0
14 0
11 0
Preexisting De fec't 1
7 1
8 Not Determined 14 11 0
15 0
Miscellaneous 1
1 0
0 0
Ma lope rat ion 1
0 0
0 0
Creep 1
0 0
0 0
35 57 3
48 1
Total 38 87 7
62 1 _
TABLE 5-4 FAILURE RATES AND 99 PERCENT CONFIDENCE UPPER BOUND FAILURE PROBABILITIES DERIVED FROM STATISTICS IN TABLE 5-3 FOR 100,300 VESSEL-YEARS OF OPERATION Failure Rate 997. Confidence Upper Bound Condition "Potentially "Potentially Dangerous
" Catastrophic" Dangerous "C a t a s t ro ph i c"
-5 For All Data 1.3x10" 7.0x10 1.6x10-3 1.6x10-4 Relevant to Nuclear Syst ems--Per Vessel Year.
Vessels & Piping 6.2x10 1.0x10-5 8.3x10-4 6.6x10-5
-4 Piping, Etc.
4.8x10~
1.0x10' 6.7x10-4 6.6x10-5
-4 Vessels Alone 1.4x10-0 2.5x10 4.6x10-5 5.5.2 Federal Republic of Germany Pressure Vessel Failure Statistics. The Institut fur Reactorsicherheit de r Technischen Uberwachungs-Vereine (Institute for Reactor Safety of the Technical Inspection Associations) has, on a continuing basis, accumulated the largest body of pressure vessel data rapidlyincreasig) number available in the world. The statistics represent a of vessels and vessel-years of operation.
n 1952 there were 50,000 vessels in the d ry a set; in 1965, 240,000(6 ; in 1966, 300,000(4) ; and 1972,470,000\\g)
(including over 7000 boiler drums). The corresponding in number of vessel-years starting f rom 1952 were 1.7 million in 1965, l
2.0 million in 1966, and 4.3 million in 1972(10).
These statistics have provided the basis for several papers giving studies of failure probabilities; the findings of some of the papers are described below.
In so large a vessel population, little is published regarding design, fabrication. and operating practice for many of the vessels. M o re-over, because of the analysis techniques employed, it is not always possible to obtain failure rates for the various types of failure.
In many cases, only lumped statistics are given.
Kellermann(4) discusses documented and undocumented histories of 300,000 vessels covering the period 1958-1965. Ilis evaluation of the experience during the period is summarized in Table 5-5.
TA BLE 5-5 FAILURE EXPERIENCE IN FRG FOR PRESSITRE VESSELS (1958-1965)
(KELLERMANN DATAU+1
" Failures" Failure Rate per Vessel Year Total 2.1x10-0 Material Fabrication Defects 1.35x10-4 I
Found During Inspection 5.56x10-5 l
Found During Pressure Test 1.88x10-6
" Serious Failures" Total 9x10-5 1
Material Fabrication Defects 3.8x10-5 Slopianka and MiezeU) made a detailed study of failure trends in vessels inspected by TUV in the period 1952 to 1968. A Weibull distribution was used in establishing pressure vessel reliability and failure rate for each 10,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> operating period from startup through 100,000 operating hours.
(There were 354 " failures" in this statistical sample of vessels manufac-tured during the period 1952-1966). The number of failures per year decreased over this period, pres fromabout4x10-gmablybecauseofim oving vessel quality and inspection, to less than 2x10-Kellermann and Seipel(6) analyzed 547 " failures" in conventional vessels during the period 1950-1965, at the end of which there were 240,000 vessels, and separated failures that might be of nuclear vessel significance. Their analysis, given in Table 5-6, indicates that, for the 49 " failures" ghey consider pertinent to nucicar vessels, the " failure" rate is 1.3x10-per vessel-year. The authors also give a failure rate, based on their obser-vations, of 3x10-6 per vessel-year for " severe vessel damage."
In determining failure probability, the Commitee tyas used the total vessel-years derived from the Slopianka and Mieze report (5). The authors do not specifically state how many of the 49 failures that might be of nuclear vessel significance were disruptive.
If there had been no disru confidenceupperboundwouldbe0.2x10gtive failures, the 99 percent per vessel year; if the other extreme assumption were made that all 49 failures were disruptive, the upper bound would be 4.0x10-5 per vessel-year.
al(7) cited a special study similar to the ABMA and EEI(I)
K e l l e rma nn, et statistics where 7000 high pressure steam drums of water tube boilers were reviewed. These drums were considered comparable to reactor pressure vessels insofar as material selection, f abrication, type of loading and 6
operational conditious are concerned. The 7000 drums have exceeded 500x10 hours over a mean service time of ten years, comparable to 67,000 years at 7500 hours0.0868 days <br />2.083 hours <br />0.0124 weeks <br />0.00285 months <br /> per year. The authors reported no " catastrophic" failures in any of these drums; however, there were a few cases of drums taken out of service due to leaks during operation. The 99 confidence upper bound for " catastrophic" failures would be 6.9x10-gercent per vessel-year.
l In the same study the authors state that, in more than 470,000 pressure vessels being inspected, about 90 "f ailures" per year are reported, the largest proportion being from plants that do not employ qualified personnel, particularly for maintenance. The corresponding failure rate is 1.9x10-4 for failures of all kinds.
TABLE 5-6 FAILURES IN PRESSURE VESSELS
- KELLERMANN AND SEIPEL DATA (6)
Of Nuclear Vessel Causes of Failures Total Reported Significance Design 101
,13 Sa fety Devices Design 33 2
Sa fety Devices Operation 76 4
Material 67 7
Manufacturing 55 6
Aging 11 3
Stress Corrosion 16 4
overload 52 1
Corrosion Inside 63 2
Corrosion outside 17 1
Vibrations 56
_7 TOTAL 547 49**
6 Per 10 Vessel-Year 150 13
-5 Per Vessel-Year 1.5x10'0 1.3x10
- Covers all incidents whether noncritical, potentially disruptive, or disruptive.
- The total is actually 50, but was reported as 49.
The latter has been used since the difference does not significantly affect conclusions.
5.6 Compcrison of Patted States and Foreign
]
Pressure Vessel Failure Statistics The available failure rates and 99 percent confidence upper bounds for all data analyzed are presented in Table 5-7.
The most significant value is derived from the ABMA data where the 99 percent confidence upper bound value is 6.4x 6, quite comparable to the Kellermann and Scipel failure rateof3x10g0-5.7 Probability of a Non-Nuclear Vessel Disruptive Fa ilure--Commi ttee Appraisal From the non-nuclear vessel data available f rom the US, UK and FhG, it is desired to establish the probability of disruptive failure (by modes pertinent to reactor vessels) of non-nuclear vessels.
i 5.7.1 Determination from Very Large Data Sets.
A summary of 99 percent confidence upper bound values for probability of disruptive failure of non-nuclear vessels derived from each of the sources of data reviewed is sh g in Table 5-8.
No results from the Kellermann(4) or Slopianka and Mieze sources are included because insufficient information is available in l
these cases to enable determination of an associated upper bound value.
The data sets for the various sources listed in Table 5-8 are arranged in order of increasing size of data base; that is, in order of increasing j
number of vessel-years of operation. The upper bound values shown are taken directly from Table 5-7.
In each of the first five cases the number l
of disruptive failures is zero.
For zero fa ilures, the statistical 99 l
percent confidence upper bound is determined by the number of vessel-years covered in the set; i.e.,
it equals 4.6 divided by the number of vessel-years. Accordingly, the upper bound value is seen to decrease in the table from 46x10-5 for the EEI-TVA data (10,000 vessel-ars) to 0.64x10-5 for the ABMA data (723,000 vessel years). The most significant and useful of these values is th e A BMA numb e r, that derived from the largest data base.
It is important to recognize that, even if th failure probability were much lower than the 0.64x10-g true disruptive ABMA value, it would be impossible to establish that fact with a data base limited to 723,000 vessel-years.
For example, if the next 723,000 vessel-years of operation accrued still involved no disruptive failure, the upper bound failure probability which would then have been demonstrated is significantly lower; viz, 0.32x10-5 per vessel-year. In fact, even if one disruptive failure were to occur in the additional period, the upper bound value then substantiated (0.46x10-5) would be lower than the value demonstrated by the data availabic to date with no failures. -_-
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TABLE 5-8
SUMMARY
OF 997. CONFIDENCE UPPER BOUND VALUM FOR PROBABILITY OF DISRUPTIVE FAILURE OF NON-NUCLEAR VESSELS l
Number of
- 99. Confidence Upper Data Source Disruptive Vessel-Years Bound, Failures Per Failures of Operation Vessel-Year EEI-TVA (Sec. 5.4. 3) 0 10,000 46x10-5
-5 EEI (Sec. 5.4.4) 0 22,000 21x10 Kellermann, et al
-5 (Sec. 5.5.2) 0 67,000 6.9x10
-5 UK (Sec. 5. 5.1) 0 100,300 4.6x10 ABMA (Sec. 5. 4. 5) 0 723,000 0.64x10-5 Kellermann and Seipel (Sec. 5.5.2) 1,700,000 0.27x10-5 to
-5 4.0 x10
l l
The data sets for the EEI and EEI-TVA sourves do not contribute addi-tionally to determination of an appropriate upper bound value for dis-ruptive failure probability. This is because, as indicated earlier, both constitute subsets of the ABMA data. In addition, the very small number of vessel-years of operation reflected in each prevents their associated upper bound values from being of significant usefulness.
Similarly, the relatively few vessel-years of operation involved in the UK data set make the UK upper bound value of little usefulness in com-parison with the AP2iA number.
In consideration of all the data available, the Committee concludes that an appropriate 99 percent confidence upper bound for probability of d is ru pt ive failure (pertinent to reactor vessels) of non-nuclear vessels is less than 1x10-5 per vessel-year.
5.7.2 Determination bv other Methods. Although very few disruptive failures have been reported in welded boiler drums or comparable pressure vessels, the statistical studies covered by Sections 5.3 and 5.4 report many potentially disruptive failures. If a ratio could be established between categories of failure, it would be possible to extrapolate the statistically significant nondisruptive failure rates to obtain a disruptive failure rate, even with relatively small data sets. Iloweve r, the following problems are encountered in this type of analysis.
1.
Determination of the ratio of potentially disruptive failure rate to disruptive f ailure rate is a function of the initial vessel quality and the ability of the inservice inspection methods to identify flaws of a disruptive potential for timely repair. This ratio is a matter for speculation until a large body of statistically relevant experience is available even though experience with other types of inspection-sensitive failure rates shows substantial improvement in failure rate by improved inspection frequency and methods.
2.
Where no disruptive failures have occurred in a data set, it is necessary to categorize the nondisruptive failures by varying degrees of seriousness in order to extrapolate to disruptive failures.
3.
The differences in terminology discussed in Sec. 5.1 make it difficult to coordinate the data to the terminology used in this report.
For these reasons, the Committee has based its appraisal of failure probability on the procedure described in Sec. 5.7.1.
5.8 Differences Between Boiler Practice and Reactor Practice The appraisal of reactor vessel reliability is based primarily on boiler statistics because no substantial body of statistical data is available for other components in comparable service.
It is recognized that there are impo rt ant dif ferences between boiler and reactor practice, and this sectian reviews conditions that might be construed as either favorable or unfavorable to the reliability of reactor vessels as compared to that of boilers.
S.R.1 Design Di f ferences. In comparison with boiler drums, reactor vessels are larger in dianeter and the walls are thicker. Reactor vessels have a full-diameter flanged head, and in some cases constraining internal or external attachments are welded to the vessel shell. Ilowever, detailed stress analysis and other AEC and Section III requirements lead to essentially uniform confidence in predicting the performance of all parts
(
of a reactor vessel. Moreover, vessels of comparable complexity and wall thickness have given reliable service in other industries even when not designed to the exacting standards required for nuclear reactor vessels.
The Committee believes that the design requirements of Section III of the Code more than compensate for design complexity, with a net result favorable to a Section III reactor vessel in comparison with a Section I boiler vessel.
5.8.2 Fabrication and Inspection Dif ferences. The large size and special design features of reactor vessels have led to fabricating and welding precedures differing in many respects from those used in the manufacture o f boile r d rums, including, in some cases, field fabrication.
Fu r th e rmo re,
some o f the specialized skills unique to the design and fabrication of a complete l' oiler are not required for a reactor vessel. These circumstances have permitted additional manufacturers, both domestic and foreign, to supply complete reactor vessels or to subcontract for parts of vessels.
The qualifications of suppliers are monitored by the ASME, the reactor designer, the AEC, and in some cases by the purchaser of the nuclear steam supply system. In all cases, the requirements of the Code, and other quality assurance requirements imposed by the AEC, the designer, and the vessel manufacturer extend to suppliers of all parts of the vessel pressure boundary.
It is recognized that some mmnufacturers have encountered initial dif fi-culties with new manufacturing and welding procedures. However, the problems have been identified and procedures have been developed to solve them.
The Committee believes that, with respect to fabrication and inspection requirements,Section III should give greater assurance of a sound reactor vessel than can be expected for a Section I boiler drum. The Committee sees no reason to expect significant differences in quality among reactor vessels produced by different authorized manufacturers if Section III requirements are met.
Confidence in the quality of reactor vessels is further enhanced by the preoperational inspection procedures required by Section XI, which, in some vessels, have revealed flaws that escaped detection by the usual shop inspection procedures used to satisfy Section III requirements.
Identification and correction of such flaws reduces the potential for failure in service.
5.8.3 Material Di f ferences. Reactor pressure vessels are fabricated of SA-533 plate and SA-508 forgings. These are manganese-molybdenum steels with additions of nickel or nickel plus chromium. Basically these alloys are the same as SA-302B-modified, used extensively in naval nuclear vessels. These alloys have been used for more than 15 years with excellent results. The scatter band for properties is quite narrow and the SA-533 or SA-508 steels are relatively insensitive to variations in heat treatment or fabrication practice. Uniformly high material performance can therefore be expected.
Although austenitic materials are rarely used in boiler drums, some usq is made of austenitic stainless steels in reactor systems for vessel-wall cladding, in-vessel structures, and nozzle safe-ends. Stress corrosion cracking has occurred in sensitized austenitic materials used in reactor internals and nozzle safe-ends, but revised specifications now minimize the likelihood of this problem occurring in newer reactors.
The consequences of mechanical failure at such locations must be con-sidered in an overall review of the safety of the system, but d is ru pt ive reactor vessel failure is not a possible direct consequence.
The Committee believes that; the uniform properties and high toughness of reactor vessel steels at least compensate for the greater thickness of a renetor vessel in comparison with a boiler drum made of a good conven-tional grade of steel. In comparison with the broad range of materials used for the boiler drums included in the failure data, the Committee oclieves that a Section III reactor vessel l as a decided reliability ad-vantage.
5.8.4 Operating Dif ferences. Operating conditions that require comparison are water chemistry, temperature cycling, radiation, and inspectability.
f 71-f 1
Water Chemistry. In fossil-fueled boilers, there must be good control
/
of water chemistry to protect against corrosion and scale formation.
The same freedom of choice in the selection of additives to control water chemistry is not feasible in nuclear reactors. In BWRs high purity water is used, but radiolysis causes a relatively high oxygen concentration; in PWRs, boric acid is used for nuclear control, with other additives for corrosion control. However, the water temperature is lower in reactor vessels than in modern steam boilers and all portions of the vessel surface in contact with water are clad with austenitic stainless steel.
It is believed that the problem of corrosion of the pressure boundary of reactor vessels is less severe than in boiler drums and of little significance as a possible cause of " disruptive failures".
Thermal Cycling. The temperature transients in normal operation are I
generally less severe in a reactor vessel than in a boiler drum. Further-
)
more, since the number and severity of thermal cycles are specified and 1
allowed for in the design of a reactor vessel, the probability of thermal
{
fatigue cracks should be much lower in a reactor vessel than in boiler drums.
Radition Damage. As stated in 3.3.3 and Sec. 3.4, enough is known about the effects of neutron fluence on SA-533 steel to give assurance that no serious radiation damage has occurred to date in Section III nuclear reactor vessels made of SA-533 or steels of similar composition.
Although SA-533 steel normally shows little susceptibility to thermally induced strain aging, it is not impossible that a synergistic effect on transition temperature and fracture toughness could result from some combination of temperature, stress, stress cycling, and irradiation.
Further study of the effects of radiation is expected to provide suffi-cient data before any potential problem arises. To assure timely detec-tion of long range effects of neutron radiation on the material, each vessel is monitored in accordance with the requirements of Appendix H of 10 CFR Part 50.
Neutron attenuation in the vessel wall leaves most of the outer portion of the wall unaf fected. The possibility of serious adverse effects is further minimized by the low cyclic stress duty and the simple geometry of the vessel in the core belt region where radiation effects are greatest. The recent practice of selecting low phosphorus and low copper steel further reduces concern for this region.
In view of the above, while reactor vessels are subject to this addi-tional potential degradation of properties compared to boiler drums, the i
practical difference in reliability is believed to be small but should l
be the subject of future review.
Inservice Inspection. Some cracks classified as nondisruptive failures may grow and, if not detected and repaired, may eventually cause disrup-
^
tive failure. Periodic visual inspection and timely repair has unques-tionably contributed much to the good safety record of boiler drums.
and llolmes(8) in a study of probabilities of failure of advanced Cave gas reactors, cited a factor of about 100 between failure probability for no inspection versus " full" inspection, and a factor of 10 versus
" partial" inspection. O'Neil and Jordan (9 ) cited a similar relation-
~
ship, namely, a reduction by a factor of at least 100. The Committee believes that experience with non-nuclear vessels and the studies cited emphasize the importance of inservice inspection and strict adherence to the requirements of Section XI.
The presence of internal structures and radioactivity prevent direct visual inspection of major portions of nuclear reactor vessels, and optical devices, while valuable, do not permit equivalent remote in-spection. Moreover, the cladding would obscure any small cracks that might develop in the base metal at the interface. However, visual inspection, either directly or with optical tools, or volumetric inspec-tion (UT) are required for the critical areas where high stresses cause concern. Sensitive leak detection systems are also installed to monitor leakage from the primary pressure boundary, including the pressure vessel.
The Committee believes that the inspections required by Section XI of the Code and the Section III design requirements compensate for the limited access for direct visual inspection. Nevertheless, the Committee believes that reactor designers should provide improved access, commensurate with other safety considerations, for inspectability of the interior and exterior vessel surfaces in order to enhance flaw detection capability. Moreover, it is expected that acoustic methods of inspection will become available in the future that could provide comprehensive in-spection for the entire vessel.
5.8.5 S umma ry.
On balance, the Committee believes the reliability in service of a section III reactor vessel is greater than that of a Section I boiler drum.
5.9 Probability of Disruptive Failure of a Reactor Vessel
='
--Committee Appraisal It is the opinion of the Committee that the disruptive failure probability of nuclear reactor vessels is significantly lower than that of the non-nuclear vessels evaluated in the preceding sections. This is based on the following:
r
_. m l------
1.
The non-nucicar vessels covered in the statistical data evaluated consisted of Section I boiler drums and other types of pressure vessels. As indicated in Sec. 5.8, the Committee has concluded that the failure rate of Section III reactor vessels is lower than that of modern Section I boiler drums. This conclusion is based on differences in design, fabrication and inspection, materials, and operating conditions between Section III reactor vessels and Section I boiler drums.
Of particular importance a ; the preoperational and inservice inspection measures required for reactor "essels by Section XI of the Code.
2.
The non-nuclear vessel statistics include many vessels other than boiler drums; but the failure rate for such vessels would not be expected to be lower than for Section I boiler drums.
In fact, the large number of vessels and long historical period covered by the data make it likely that the statistics include l
many vessels of lower quality and lesser inspection requirements than the Section I boiler drums. The f ailure rate for Section I boiler drums would therefore be expected to be lower than for the total population of non-nuclear vessels evaluated.
While the Committee has not attempted to assign separate factors as result of the above considerations, it believes that some of the a
factors may be substantial, and concludes that the disruptive failure rate of Section III reactor vessels is at least one order of magnitude less than that of the non-nuclear vessels evaluated. Accordingly the Committee concludes that there is reasonable assurance that the disruptive failure rate of reactor vessels designed, constructed and operaged in accordance with Code Sections III and XI is less than 1x10~ per vessel-year *.
The Committee recognizes that there have been other studies of vessel reliability, based largely on the UK and FRG statistics, and that some authors have concluded that the risk of failure is considerably greater than concluded by the Committee.
5.10 Probability of Disruptive Failure Beyond the Capability of the Engineered Safety Fe a tu re s Disruptive vessel failures for which the probabilities have been esti-mated in Sec. 5.9 can occur in a variety of ways, not all of which ex-ceed the capability of the Engineered Safety Features to protect the public. The probability of a vessel failure under design conditions in modes that would lead to loss of core cooling capability or loss of containment is therefore less than the probability of a disruptive f ailure as defined herein.
- Failures at or immediately adjacent to a safe-end are assumed equiv-alent to double-ended pipe failures and are not considered in this failure rate (Sec. 3.6.4) 5.11 Summary As discussed in Sec. 5.1 and elsewhere, the nomenclature used in the vari.us referenced reports is misleading, many of the reported failures are not in the vessel or boiler drum pressure boundary, and f ailures are frequently reported as catastrophic with no implication of disruptive failure.
The foreign vessel statistics, particularly the FRG data, span a long l
service period and do not reflect the improvements in material l
processes, fabrication practices, and design code rules which occurred during the latter part of the periods covered. Moreover, from the large l
I number of vessels covered it is evident that most of the boilers were small and that the studies probably included many heating boilers, for which the average Icvel of quality would be expected to be lower than for central station boilers. It is therefore probabic that the statistics tend to overestimate the f ailu re rate for large fossil-fueled boilers.
Many of the reported failures were in regions, such as tubes, that have no counterpart in a reactor vessel. These failures, where adequately identified, have been eliminated from consideration as not pertinent to reactor vessels.
Many of the reported failures are fatigue failures.Section III of the Code sets limits to peak stresses, and to cyclic stresses as a function of the number of cycles. The Committee concludes that most of these failures would have been avoided if these non-nuclear vessels had been constructed to Section III of the Code.
The Committee concludes that there is reasonable essurance that:
(1) the disruptive failure probability of non-nuclear vessels in central-statign service by modes pertinent to reactor vessels is less than 1x10-per vessel-year; (2) the disruptive f ailure probability of reactor vessels designed, construgted and operated to Sections III and XI of the Code is less than 1x10-per vessel-year; and, (3) the disruptive failure probability of such reactor vessels, beyond the capability of engineered safety features is still lower.. _ _ _ _ _ _
REFERENCES 1.
Letter, E. G. Case, Director, DRS, to Electric Research Council, November 26, 1971 2a. Private Communications 2b. Private Communications 2c. Private Communications 2d. A. G. Pickett and E. R. Reinhart, " Review of Recent Utility Experience with Power Reactor Coolant Pressure Boundary Inspection Regarding Ser-I vice Condition, Defect Detection Capability, Defect Size, and Defect Orientation", Southwest Research Institute Report to EEI-TVA 2e. American Boiler Manufacturers Association letter, William B. Marx
{
to R. R. Maccary, February 14, 1973, Pressure Vessels Built: 1/1/62 to 12/31/71 3.
AllSB(S) R162 - C. A. G. Phillips and R. G. Warwick "A Survey of Defects in Pressure Vessels Built to 1[igh Standards of Construction and Its Relevance to Nuclear Primary Circuit Envelopes" 4.
O. Kellermann, "Present Views on Recurring Inspection of Reactor Pressure Vessels in the Federal Republic of Germany", IAEA No. 81 5.
IRS-134 (1968) - C. Slopianka and G. Mieze, " Failure Rates of Pressure Vessels, Part I:
Evaluation of VdTUV Statistics" 6.
O. Kellermann, H. G. Scipel, " Analysis of the Improvement in Safety Obtained by a Containment and by Other Safety Devices for Water-Cooled Reactors", IAEA SM-89/8 T.P. 89 7.
O. Kellermann, E. Kraegeloh, K. Kussmaul, and D. Sturm, " Considerations About the Reliability of Nuclear Pressure Vessels - Status and Research Planning", Paper 1-2, Pressure Vessel Technology, Part I, Design and Analysis, 2nd International Conference For Pressure Vessel Technology, San Antonio, Texas, Oct. 1-4, 8.
Cave and R. E. Holmes, " United Kingdom, Suitability of the Advanced Cas-Cooled Reactor for Urban Siting" (1967) 9.
R. O'Neil and G. M. Jordan, " Safety and Reliability Requirements for Periodic Inspection of Pressure Vessels in the Nuclear Industry",
Conference on Periodic Inspection of Prescure Vessels, London, May 9-11, 1973
- 10. Letter, O. Kellermann, Director, IRS, Federal Republic of Germany, to R. F. Fraley, Jan. 2, 1974,
6 CONCLUS TONS AND RECOM>ENDATIONS The report reviews current reactor vessel construction practices, possible modes of failure, and operating practices pertinent to vessel integrity. It also reviews f ailure statistics of non-nuclear pressure vessels and gives an assessment of the disruptive failure probability of reactor vessels. Important conclusions and recommendations are recapitulated in this section.
6.1 Some Limitations in Scope of Report Two important limitations have been placed by the Committee on the scope of the report:
6.1.1 Accident Cond i t i ons.
The disruptive failure probability determined for reactor vessels is for vessels designed, constructed and operated in accordance with Code Sections III and XI.
For reasons stated in Sec. 4.1, the probability does not include any contribution attributable to failures which conceivably could result from overstressing of the vessel by system accidents not contemplated in design.
6.1.2 Rad ia tion Damage.
The ef fect of irradiation is a matter requiring continuing nttention.
The Committee believes that, during the next few years, no large reactor vessel constructed to Section III of the Code will have been suf ficiently irradiated for this matter to become a problem.
The ef fects of irradiation will require careful review when larger fluences have been accrued and more data have become available. The material surveil-lance programs and research programs now being carried out should provide such data (Sec. 1.3, Sec. 3.3, Sec. 3.4, Sec. 5.8.4).
6.2 Current Practice The technology of reactor vessels has been advanced greatly in recent years by marked improvement in steel quality, design methods, inspection techniques, and quality assurance programs. Some of these improvements are summarized here as background for recommendations recapitulated in Sec. 6.4.
6.2.1 Materials. The properties of steels used for the reactor vessel pressure boundary, SA-533 and SA-508, have been intensively studied for thick sections in the llSST programs. These steels provide a goed balance of strength and toughness properties, with good metallurgical stability against changes occuring in service, including resistance to irradiation embrittlement (Sec. 2.1).
6.2.2 Design. Design rules under Section III of the Code differ substan-tially from those applied to conventional pressure vessels. The design requirements for nuclear vessels include extensive analytical stress calculations as well as fatigue analyses for combined mechanical and thermal stresses. Assurance of safe design of nuclear vessels does not rely pri-marily upon empirical design conservatisms, simple code rules, and established _
design details as does that of non-nuclear code vessels (Sec. 2.2).
On the other hand, higher design stress levels are permitted, thereby allowing thinner sections for the same application.
Fatigue analyses and inservice examinations give reasonable assurance that a fatigue induced crack developing during the service lifetime will not grow to a size suf ficient to propagate rapidly, and that it will be detected and monitored if growth occurs (Sec. 2.2.1).
Section III rules impose design control measures on both the Owner and Manufacturer of the reactor vessel by requiring a Vessel Design Specificatioq i
prepared by the Owner or his agent, a Stress Report prepared by the Manu-facturer, and certification of these documents (Sec. 2.2.3).
6.2.3 Fabrication.
Section III of the Code provides fabrication rules that are directed primarily toward controls, nondestructive examinations,
{
and inspections performed during each stage of each of the major fabrication processes (forming, welding, heat treating, etc.) (Sec. 2.3).
The nondestructive examination requirements of Section I and Section III are compared in Tabic 2-1.
6.2.4 Preservice and Inservice Inspection.Section XI of the Code requires an enhanced program of preservice inspection, over and above the inspection pnagrams required for fossil-fueled s team boilers, including ultrasonic mapping of all areas subject to periodic volumetric examination over the service lifetime, i.e.,
essentially 100 percent of the pressure-retaining welds. The preservice examinations also serve as a final confirmation of the structural quality of the vessel before it is placed in service (Sec. 4.3.1).
Section XI specifies periodic inservice examination requirements, including volumetric inspcction of representative portions of pressure retaining welds.
These examinations, in conjunction with the preservice mapping are intended to monitor extension during service of flaws in areas examined (Sec. 4.3.4).
6.3 Probability of Disruptive Failure A disruptive vessel failure is defined for purposes of this report as a breaching of the vessel by failure of the shell, heed, nozzles, or bolting, accompanied by rapid release of a large volume of tae contained pressurized fluid (Sec. 5.1).
As assessment of the disruptive failure probabili;y of nuclear reactor vessels has been made, based on:
(1) Consideration of the available failure data for non-nuclear types of vessels, such as boiler drums and unfired pressure vessels, and determination of the disruptive failure probability for such vessels; and, _
(2) Comparison of the design, fabrication, materials, operating conditions, and inspection practices used for non-nuclear vessels with those used for reactor vessels, and estimation of the ef fect of the dif ferences on the relative probability of disruptive failure of the two classes of vessels.
6.3.1 Probability of Disruptive Failure of Non-Nuclear Vessels. The Com-mittee has reviewed available statistics of boiler drums and other non-nuclear pressure vessels and concludes that the 99 percent confidence upper failure (pertinent to reactor vessels) bound for the probability of disrugtive of such vessels is less than 1x10~ per vessel-year (Sec. 5.7).
6.3.2 Probability of Disruptive Failure of a Reactor Vessel. The Committee believes that the disruptive failure probability of reactor vessels designed, constructed, and operated in accordance with Code Sections III and XI is at least one order of magnitude lower than that of the non-nuclear vessels evaluated. Accordingly, the Committee concludes that there is reasonabic assurance that the disruptive failure probability of such reactor vessels is less than 1x10- per vessel-year (Sec. 5.9).
6.3.3 Probability of Disruptive Failure Beyond the Capability of the Engineered Safety Features. As defined, disruptive failures of reactor vessels include failures of various magnitudes and descriptions, not all of which would exceed the capability of the engineered safety features.
Accordingly, the probability of reactor vessel disruptive failure beyond the capability of Engineered Safety Features is lower than the probability for disruptive failures of all kinds addressed in Sec. 6.3.2 (Sec. 5.10).
6.4 Recommended Supplementary Requirements for Reactor Vessels Current practice gives a high degree of assurance against nuclear vessel failure. However, the Committee believes that reactor vessels, because of their greater importance to safety, should be considered in a class above other Class 1 vessels.
The following recommendations are made.
6.4.1 Materials.
1.
Explicit requirements for limits on residual elements, such as copper and phosphorous, which are believed to increase irradiation embrittlement, should be set forth for materials subject to neutron 18 nyt (Sec. 2.1.3, Sec. 2.7, Item 3).
fluences exceeding 10 2.
Maximum strength 1cvels should be included in specifications of all materials, including SA-508 (Sec. 2.1.3).
3.
New and high strength materials permitted by the Code such as SA-542 and SA-543 should not be used until their predicted behavior in nuclear service is well understood and documented and the ability to control their properties has been fully established by extensive test programs (Sec. 2.1.4).._. _
4 The Committee recommends that an investigation be made in the US into sensitivity of SA-533 and SA-508 welds to hydrogen damage as a result of deviation fram optimum conditions (Sec. 3.1.2).
6.4.2 Design.
1.
More definitive material toughness requirements for emergency and faulted conditions should be stated in the Code rules, taking account of the variabic relationship between vessel temperature, Icvel of stress, and fracture toughness (Sec. 2.7, Item 1).
2 First-of-a-kind vessel designs should be subject to confirmatory design review by the owner or his agent (Sec. 2.7, Item 5).
3.
Present Code requirements for the Vessel Design Specification, Vessel Stress Report, and Certification of Stress Report for Compatibility
{
with Design Specification should be interpreted as involving at least two independent organizations in the preparation or review of documents (Sec. 2.7, Item 4).
4.
For a vessel that may be exposed to a neutron fluence of 1018 or more ( > 1 Mev), the owner should assure himself that the system design permits a practical procedure for annealing the vessel in service in case this should become necessary (Sec. 3.4).
5.
The Committee recommends that the biological shield (reactor cavity walls in PWRs) and other structures be designed to withstand the effects of a failure of the thin-walled part of the nozzle adjacent to the sa fe-end (Sec. 3.6.4).
I 6.4.3 Nondes truc tive Examination.
The potential exists for further im-provement in vessel defect characterization by the application of newly developed techniques, such as acoustic emission, acoustic holography, and l
acoustic spectroscopy.
These techniques should provide a basic tool for determining, more precisely, flaw locations and growth characteristics in reactor vessels and may make an important additional contribution to vessel reliability. Attention has recently turned to supplementing cur-rent practice by acoustic emission techniques capable of showing the location of flaws by acoustic response during structural loading. These methods, when developed to the stage where they can quantify defect location and significance, should be applied to the manufacturing and preoperational testing phases, and possibly to the operating phase (Sec. 2.7, Item 6 and Sec. 4.3.4). _____ __
6.4.4 Inservice Inspection and Surveillance. Periodic visual inspection and timely repair of cracks has contributed much to the good safety record of foscil-fueled boilers. Although most of the cracks in boiler drums have resulted from thermal cycling and peak stresses that would not be permitted by Section III, it is important that good inservice inspecta-bility be maintained for reactor vessels.Section XI of the Code requires visual examination of some critical areas, and 100 percent volumetric examination of other representative areas. Sensitive leak detection systems are also installed to monitor leakage from the primary pressure boundary, including the vessel.
The Committee of fers the following additional observations and recom-mendations with respect to inservice inspection and surveillance.
1.
The Committee believes that reactor designers should give greater attention to providing accessibility for inservice inspection.
This applies to volumetric inspection and to visual inspection by optical means (of both interior and exterior surfaces), especially at regions of severe constraint (Sec. 5.8.4 ).
2.
The Committee believes that experience and further development will show that inservice ultrasonic examination of welds is not a very time consuming or difficult operation.
If this is demonstrated, consideration should be given to requiring uore extensive vessel inservice ultrasonic examination than presently required by Sec-tion XI (Sec. 4.3.3).
3.
The Committee believes that, although the " leak-before-break" cri-terion cannot be relied on.a potential failure might be averted by advance warning of leakage through a crack of suberitical size.
More sensitive systems for detection and location of possibic reactor vessel leaks should, therefore, be provided (Sec. 3.3.4).
4.
Periodic overpressure testing is not required for reactor vessels; however, the Committee believes elimination of periodic tests should not be regardeo as a firmly established practice and that further consideration should be given to this matter (Sec. 4.3.6).
5.
The Committee recommends that periodic bolting examination required by Section XI be interpreted as requiring surface examination of both threaded ends of main closure studbolts (Sec. 3.6.5).
6.4.5 Operation. The Committee recommends that Technical Specifica-tions specify heatup and cooldown pressure-temperature limits that can be shown to be as conservative as practical with respect to 10 CFR Part 50, Appendix G (Sec. 3.3.3). -
6.4.6 Research and Develonnent. The adequacy of the Research and Development (R&D) pertaining to reactor pressure vessels has not been examined f ully by the Committee as part of this review. The informa-tion provided f rom the llSST, EEI-TVA, and other programs has been valu-able in assessing vessel reliability; however, the data have not yet been analyzed completely. The status of the R&D ef fort requires fur-ther study to determine what work should be extended, modified, or initiated to support the reliability requirements of nuclear reactor vessels.
The necessary work may include further investigation of vessel failure modes, material properties, and inservice acoustic emission techniques.
Suf ficient work should be carried on to assure the continuing availability of expert current knowledge of reactor vessel technology to the AEC.
6.4.7
- General, 1.
In arder to increase the assurance that QA programs are adequate, and thereby to minimize the probability that defects will remain undetected, the Coumittee recommends that the individual re-sponsibilities of the principal organizations that establish QA procedures and controls be subject to audit by the owner and by the AEC (Sec. 2.7 Item 7).
2.
Where owner's or manuf acturer's quality ccquirements ordinarily exceed minimum Code requirements, consideration should be given to upgrading the Code to conform to practice (Sec. 2.7, Item 2).
3.
The Comnittee believes further study of possible design changes to protect against vessel failure should be performed (Sec. 4.4.4).
6.5 Reactor Vessels Not Covered by Report This report applies primarily to vessels constructed of SA-533 and SA-508 steels, designed and constructed to Section III of the Code, and operated in accordance with Section XI.
The report further stiuplates that future consideration should be given to the effect of increasing irradiation.
The Committee recognizes that some older vessels are constructed of other steels, that they were designed to Sections I and/or VIII of the Code, and that only limited conformance to Section XI is practi-cal.
i Moreover, the belt zones of some of these vessels have been signifi-cantly irradiated. The provisions of Appendix G to 10 CFR 50 should give reasonable assurance against failure, but it may become increas-ingly dif ficult to apply the conservatism recommended by the Committee (Sec. 3.3.2) in applying Appendix G. __
The Committee recognizes that these older vessels are under continuing surveillance by AEC, but recommends that a documented review be made of their present status and of the Commission's rules governing their operation.
6.6 Puture Action by ACRS As additional.information is accumulated, the Committee plans to review matters relating to vessel integrity.
6.6.1 Review of Current Programs. The llSST Program is, with the excep-tion of irradiation investigations, approaching conclusion.
The Com-mittee, which has input to the program through individual memberships of ACRS members and consultants, has closely followed this comprehen-sive program, and concurrent work under other programs. The Committee plans to review the status and projected program of the llSST and other programs and to make recommendations for future work.
6.6.2 General. The Comittee plans to review periodically the matter of pressure vessel integrity, including the adequacy of Sections III and XI of the Code, and AEC supplementary requirements. The review will take into account manufacturing and operating experience, results of continuing research and development in the US and abroad, data from the irradiation surveillance programs required by Appendix 11 of 10 CFR Part 50, and newly developed techniques and methods. _
.__