ML20207K717
| ML20207K717 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/19/1986 |
| From: | Howard J, Russell J TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20207K654 | List: |
| References | |
| C010603-SQN, C010603-SQN-R03, C10603-SQN, C10603-SQN-R3, NUDOCS 8701090497 | |
| Download: ML20207K717 (22) | |
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- TVA EMPLOYEE CONCERNS REPORT NUMBER: C010603-SQN SPECIAL PROGRAM REPORT TYPE: Sequoyah Nuclear Plant Element REVISION NUMBER: 3 (Final Report)
TITLE: Bolting-Naterial Compatibility PAGE 1 OF 7 REASON FOR REVISION:
Incorporate SRP and TAS conuments. Revision 1 s
To update safety significance on Attachment A. Revision 2 Incorporate line management response to C/A and finalize report Revision 3 3
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DATE N/A MANAGER OF NUCLEAR POWER DATE CONCURRENCE (FINAL REPORT ONLY)
- SRP Secretary's signature denotes SRP concurrences are in files, h
jl 2164T 8701090497 861224 d PDR ADOCK 05000327 l P PDR
TVA EMPLOYEE CONCERNS REPORT NUMBER: C010603-SQN SPECIAL PROGRAM REVISION NUMBER: 3 PAGE 2 0F 7 I. INTRODUCTION This report addresses three concerns, IN-86-183-001, IN-85-021-X04, and IN-85-824-001 which were determined to be generic to Sequoyah Nuclear Plant (SQN) by the Watts Bar Nuclear Plant (WBN) Employee Concern Task Group (ECTG) evaluation. These three concerns raised the issue of material compatibility, more specifically, the use of carbon steel bolts in stainless steel flanged connections including stainless steel valves.
II.
SUMMARY
OF PERCEIVED PROBLEMS Carbon steel bolting material was installed in stainless steel flanged connections including valves. Replacement of carbon steel bolting was started in valves but halted before all bolting replacement was completed.
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III. EVALUATION METHODOLOGY t
l The issue (as stated in section I above) was made generic to SQN as a result of the evaluation made at WBN. The following evaluation process was used at WBN for the Material Compatibility Element Report.
- A. Reviewed Nuclear Safety Review Staff (NSRS) Investigation 1 Report I-85-483-WBN which addressed the use of carbon steel bolts in stainless steel flanged connections. Determined the thoroughness of the report with respect to how adequately the j
concern was addressed as stated by the concerned individual (CI),
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and if sufficient documentation relevant to the concern was reviewed. Also, determined if the report addressed the use of
, carbon steel bolting in stainless steel valves, as well as flanged connections.
B. Reviewed applicable criteria to determine the bolting requirements for stainless steel components to include TVA-WBN Class A, B, and C systems. Also researched records to determine if documentation had been initiated (i.e., nonconformance reports (NCRs)) to record any previous deficient areas with respect to the subject concerns.
C. Interviewed cognizant individuals in responsible organizations at WBN to determine if they agreed with the findings in the aforementioned NSRS report and evaluated their input on the relevance of the report to stainless steel valves. Also, -
interviewed responsible OE and Design Services personnel to determine the requirements / design intent with respect to the subject concern.
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TVA EMPLOYEE CONCERNS REPORT NUMBER: C010603-SQN SPECIAL PROGRAN REVISION NUMBER: 3 PAGE 3 OF 7 IV.
SUMMARY
OF FINDINGS A. A detailed review of NSRS Investigation Report I-85-483-WBN, t revealed that it was sufficiently comprehensive in that it addressed directly the subject concern and provided adequate
' documentation to verify the findings and conclusion. It was also
! determined that the NSRS report could be applied directly to carbon steel bolting in stainless steel valves.
! B. A review of TVA General Construction Specification G-29M revea]6d l
that section 3.1.7.4 addressed stainless steel flanges but did not '
directly apply to stainless steel valves. A review of BLN-NCR-59 revealed that it applied to an ASME Class III-3 stainless drain tank cooler which had carbon steel bolting in closure connections.
l The disposition of the NCR was to "use-as-is" since the vendor had supplied the tank with the carbon steel bolting installed.
1 As stated in the NSRS investigation report, NRC-IE Bulletin 82-02, addressed a corrosion wastage problem with carbon steel bolting in systems where borated water was present. This subject was also
!? addressed in NRC Information Notice 80-27. The correspondence L listed below applies directly to carbon steel bolting in stainless steel flanged piping, valve and pump flange connections:
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-Nemorandum NEB 791030 108
-Nemorandum SWP 791109 047
, -Nemorandum NEB 810521 264
-Nemorandum NEB 820401 276
-Nemorandum B45 860113 268 l
The summary of this correspondence is that carbon steel bolt degradation was identified / documented as a problem in borated water
, syatoms. As a result, corrective action has been initiated to replace on a "no delay to schedule basis" carbon steel bolting in the affected TVA designed flanged connections during routine
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surveillance / repair / maintenance activities. To this end, ECN 3522 l was written on September 14, 1982 to add bolt replacement criteria to mechanical B/Ms. However, in memorandum B45 860113 268 it was noted that considerable confusion had resulted in the original B/M changes in accordance with ECN 3522, so another ECN (6004) was U initiated on January 31, 1986 to clarify the wording and give more specific bolting substitution guidelines. It should be noted that in Class B and C systems, the SA 193 Gr B7 carbon steel bolting was to be replaced with specific stainless steel bolting material. .
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- TVA EMPLOYEE CONCERNS REPORT NUMBER: C010603-SQN SPECIAL PROGRAM REVISION NUMBER: 3 PAGE 4 0F 7 However, if replacement bolting was not immediately available, the SA 193 GR B7 carbon steel bolting was to remain installed. This substitution is also addressed in the aforementioned NSRS report.
A review of the 47BM series of drawings for various piping systems and components revealed the following:
! 1. For Class A systems, in all cases SA 453 Gr 660 stainless i steel material is specified for stainless steel flanged connections. .
- 2. For Class B systems, the note initiated per ECN 6004 is g
included which allowed substitution of SA 193 Gr B7 (carbon steel) for SA 564 Type 630 (stainless) if the latter was not readily available. The stainless bolting was to be installed
' on a "no delay to schedule basis" as a replacement when it became available, f 3. For Class C systems, if SA 193 Gr B7 was installed, it was to be replaced as detailed in item number 2.
l A review of vendor valve drawings (Westinghouse) for Class A
[ systems, check, globe, and gate valves was conducted. Of 18 f separate drawings reviewed at random, no Class A valves were found r to be furnished with carbon steel bolting. Also, a review of
) approximately 20 Class B check, gate, and globe valve drawings was f
conducted with the same result--no carbon steel bolting was furnished for the flanged bonnet-to-body connection.
s C. The cognizant OE individual, who was the author of the memorandum '
referenced above, was interviewed. He indicated that OE's intent L
of the memorandum was to include the valve body-to-bonnet bolted i ,
connection with respect to carbon steel bolting replacement. Their interpretation of the valve body-to-bonnet connection was that it was also a flanged connection. He also provided (informally) documentation with respect to OE's position on carbon steel bolting used in stainless steel flanges and valves.
f The cognizant site Design Services representative was interviewed.
He concurred with the statement by OE that the valve body-to-bonnet connection was, in fact, a flanged connection and therefore, B
was addressed as requiring carbon steel bolting replacement in P accordance with the correspondence in section IV, paragraph B, of this report, f!e also indicated that it was reasonable to assume that the bolt replacement addressed by the subject concern on stainless valves was a result of the carbon steel bolting replacement recommendations per the correspondence referenced in section IV, paragraph B, of this report.
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, - TVA EMPIAYEE CONCERNS REPORT NUMBER: C010603-SQN SPECIAL PROGRAN REVISION NUMBER: 3 PAGE 5 0F 7 l
l D. In preparation of the Subcategory Report " Bolting (Construction),"
which includes the issue addressed in this report, the DNE Supervisor--Nuclear Engineering Branch, Codes, Standards, j Naterials Corrosion, and Coatings--was interviewed and the-subject i concerns were discussed. This individual is a member of the joint Atomic Industrial Forum (AIF) and Metal Properties Counsil (NPC)
Task Group on bolting. He was coeditor of the paper " Improved
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Technology for Critical Bolting Applications " presented at the 1986 Pressure Vessels and Piping Conference and Exhibition and l
i published by ASNE (attachment B). The paper summarizes i
industry-wide bolting failure experiences, the AIF bolting program. -
i the result of the work performed under the AIF program, and the AIF
! program recommendations. The AIF report documents detailed
! research on boric acid wastage of carbon steel bolts.
l The essence of his comuments was that, assuming an adequate f preventative maintenance program, leakage of carbon steel bolted l
connections because of boric acid wastage will be detected prior to l
a significant loss of the pressure boundary. Thus, the concern is i considered to be a reliability issue rather than a safety issue.
l However, no evidence of a systematic review of vendor qualified components on safety related borated water service systems was found for any TVA PWR plant that would ensure that all critical connections were identified. In fact, the DNP procedure that will control the bolt replacement program at WBN specifically excludes vendor qualified components. This exclusion is contrary to the DNE technical position memorandums and the documented scope of the j industry-wide bolting wastage problem.
1 E. The Employee Concern Task Group (ECTG) investigation revealed that f the use of carbon steel bolts on flanged connections in borated r water service was common to SQN, as well as WBN. However, no ECTG evaluation of SQN was performed beyond the examination of documents f ,
pertaining to this issue. None of the SQN documents examined i
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addressed conclusively all the issues raised by these concerns. In f particular, the SQN responses to IEB 82-02 did not indicate that all vendor qualified components in borated water service were reviewed to determine if carbon steel bolts were used in flanged L connections nor was there any evidence to suggest that SQN's l maintenance program would be revised to incorporate the AIF bolting 3
program reconumendations. Consequently, deficiencies may exist at l
SQN similiar to the deficiencies identified at WBN.
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IVL EMPLOYEE CONCERNS REPORT NUMBER: C010603-SQN SPECIAL PROGRAM REVISION NUMBER: 3 PAGE 6 0F 7 CONCLUSION
- 1. There is evidence to support the fact that carbon steel bolting is to be replaced with stainless steel bolting in TVA-designed flanged connections where borated water is present, including bolted pipe to valve connections. It is conceivable that the CI witnessed valve bolt change-out at WBN on TVA-designed flanges in borated water service and it is also conceivable that the change-out was halted before completion because:
- a. All available replacement bolting material had been used--
no more was readily available.
- b. Since the bolting replacement was on "no delay to schedule basis " more critical work may have required completion and therefore caused the bolt change-out to be halted for an indeterminate period of time.
- 2. A partial review of valve drawings and B/Ms revealed no Class A systems where valves were furnished with carbon steel bolting, and in Class B and C systems, the use of carbon steel SA 193 Gr B7 bolting is acceptable and will only be replaced in borated water pressure boundary systems. According to the referenced correspondence in section IV, paragraph B, the SA 193 Gr B7 carbon steel bolting is even considered an acceptable substitute for the replacement stainless bolting until the latter can be installed in the Class B and C systems.
The statement by the CI may well be a true statement with respect to carbon steel bolted stainless steel valves and the CI may also have witnessed a bolt change-out in the subject valves at WBN. However, based on the findings of this
, evaluation, the concern has not identified a condition adverse to quality. Also, this evaluation is in agreement with the findings of the NSRS Investigation Report I-d5-483-WBN.
- 3. The evaluation and investigation of this element at WBN did not reveal a procedure or program providing for the systematic evaluation of vendor qualified bolted closures or flanged connections used on borated water systems. In addition, no evidence was found to indicate that the recommendations to be made by the AIF Bolting Task Force will be reviewed by the appropriate organization (s) and, where applicable, incorporated into the site maintenance program (s). Further evaluation and -
corrective actions will be required to resolve these deficiencies for WBN and SQN (CATD No. 10603-SQN-02).
TVA EMPLOYEE CONCERNS REPORT NUMBER: C010603-SQN SPECIAL PROGRAN REVISION NUMBER: 3 PAGE 7 0F 7
- V. ROOT CAUSE I None
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VI. CORRECTIVE ACTION l
The corrective actions supplied by line management at SQN were:
j Review vendor supplied bolted closures and flanged connections to l ensure that all critical connections are identified and included in the
- J bolting replacement program as applicable. (CATD 10603-SQN-01) i Develop a program for implementation of AIF/MPC bolting program
! recommendations for SQN and WBN. (CATD 10603-SQN-02) i These corrective actions are not a SQN restart requirement.
f VII. GENERIC APPLICABILITY
{ The concerns comprising this issue were determined to be generic to WBN, SQN, and BLN based on the evaluation of the concerns at WBN.
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a FrrsR7 Mr. Cc/0403 -5 94 Rff/9/W h. 3 lll REFERENCE = ECPS120J-ECPS121C
-- ATTAC HMENT A -
TENNESSEE VALLEY AUTHORITY PAGE =
i FREGUENCY = REQUEST OFFICE OF NUCLEAR POWER RUN ftME - 12:37 22 ONP = !$$$ = RHM .
EMPLOYEE CONCERN PROGRAM SYSTEN (ECPS) RUN DAtt
- 11/10/86 LIST OF EMPLOYEE CONCERN INFORMAi!ON CATE00RYs CO CONSTRUCTION-PROCESS SUSCATE00 rye 10600 90LTING
. S OENERIC KEYWORD A N APPL SIC /NSR$ P KEYWORD 8 CONCERN SUB R PLT $83W INVESi!Gai!0N S CONCERN KEYWORD C NUM8ER CAT CAT D L OC FL45 REPORT R DESCRIPTION KEYNORD D e 13 021-204 C0 10400 N WSN NYYY NO H5NP UNIT 01. STAINLESS STEEL VALVES SPECIFICATIONS T30052 REPORT (CHECKr OLOBE AND OATE VALVE $s DIFF ERENT SI2tSs 2*O AND UP) MAVE STUD B PIPING OLTS S NUi$ SHOULD DE SIAINLESS STEE VALVES L. IN 1983 AND 1984, STEAMFillERS S TARTED TO CMANGE OUT THESE CARSON Si EEL STUDS S NUTS, IME CHANGE DUT Ha$
STOPPED IN 1984 SEFORE ALL THE VALV ES HERE COMPLETED. VALVES WliN CARS ON SIEEL STUDS S NUTS ARE NON INSULA TED. LOCA110Ne REACIOR BUILDING. (A CCUMULATOR ROOMS) AND AUX BUIL DING U NIT 1. CI COULD NOT RECALL SYSTEM 0 R VALVE NUMBERS.
IN +45 424-001 CO 10400 N NON NYYY NO UN!? 1. 'ALL OVER* = STAINLESS STEEL DES!CN REVIEW 150071 REPORf VALVES INSTALLED WlfM CARBON STEEL CORRECTIVE ACTION STUDS INAT HAVE $!NCE BEEN COVERED 0 P! PING WER HITN INSULATION. NO ADD 1110NAL VALVES INFORMAi!ON AVAILABLE.
!% 183-001 CO 10400 N NON NYYY I-85-443-NSN NO CARSON STEEL 80LTS ARE INSTALLED IN SPECIFICATIONS T50129 REPORT STAINLESS STEEL FLANOED CONNECTIONS. MONCONFORMANCE AN EXAMPLE OF TN!$ CAN DE FOUND IN PIPING UNIT 2 IN THE AUX SUILDINO ELEVATIO FliTIN05 N T13' NEAR 13414 AND U. 00 ABOUT 1 0' 10 THE NORTH DOWN NALL 10 A ROOM ON THE LEFT. AN EXAMPLE IS ABOUT 3' 0FF THE FLOOR ON SOME 6' PIPE. If EXISTS ALL OVER THE PLANT. CI MAS N 0 ADDIf!ONAL INFORMATION. CONST DE PT CONCERN. e 3 CONCERNS FOR CATEGORY C0 SUSCATE00RY 10400 0
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- ATTACHMENT B MPC-Vol. EG Report No. C010603
~
Improved Technology I
4 For Critical
! Bolting Applications
- r 1 .
1 1, presented at
! THE 1986 PRESSURE VESSELS AND PIPING i CONFERENCE AND EXHIBITION CHICAGO, ILLINOIS .,
JULY 20 - 24,1986 l
l sponsored by THE METAL PROPERTIES COUNCIL IB C.
jointly with
! THE ELECTRIC POWER RESEARCH INSTITUTE i, THE PVP MATERIALS AND FABRICATION COMMITTEE, ASME edited by ,
E. A. MERRICK TENNESSEE VALLEY AUTHORITY
. M.PRAGER I . THE METAL PROPERTIES COUNCIL, INC.
t i ,
li J.
? ,
l THE AMERICAN SOCIETY OF MECH ANICAL ENGINEERS '
! United Engineering Center 345 East 47th Street New York, N.Y.10017
- _ - _ _ _ _ . - _ . . , . _ . . _ _ _ _ . - _ _ _ . _ . _ . . _ _ _ _ _ _ . _ . . _ . . ,_.m.__ _ _ _ _ , _ . _ . . . - . _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ _ . . . . _ _ _ _ _ . _ _
A OVERVIEW OF ISSUES RELATED TO NUCLEAR BOLTING APPLICATIONS E. A. Merrica Tennesm Valley Authority Knonville, Tennessee T. U. Maiston Electric rower Rmerch Instituto Palo Alto. California K. L Metthaney Tenneim varier Avihority
, Knonville. Tenne.
A8STRACT
=
The service history of bolting is good; however, #
during the past several years, the U.S. Nuclear This paper will furnish information on the failure industry has experienced an increase in the number of expertence, an overview of the AIF program, sunusartre reported bolt failures, failure or degradation has some of the results of the work performed, and provide recomendations on the issues, been reported in several generic areas which may impact plant safety or re11abt11ty. The generte bolting applications where failures or degradation have been BOLTING EXPERIENCE IN U.S. NUCLEAR PLANTS expertenced by the industry include pressure boundary manways and flanges, component supports, and There are millions of bolts used in commercial embedmonts, as well as botting used in component nuclear plants. In each unit, two oa three thousand of Internals. An aggressive program to assure the these are used in the primary reactor coolant pressure continued integrity of bolted joints is nearing boundary components, their internals, and supports completion, While the number of reported bolting failures has increased over recent years, there is some evidence INTRODUCTION which indicates that the increase is a function of the increased number of Installed bolts. It appears also This paper updates reference (1), previously that as plant maintenance personnel gain experience presented at the 8th International Conference on during early plant operatton, the incidents of leaking Structural Mechanics in Reactor Technology, in joints and reported failures decrease. The success Brussels, 8elgium, November, 1985. The Atomic history of fasteners is excellent when compared to the Industrial Forum (AIF), in conjunction with the number of failures.
Electric Po*wer Research Institute (EPRI) The Materials Properties Council, Inc. (MPC) the utt11 ties and other The U.S. Nuclear Regulatory Comission (NRC). In industry organizations, responded to the bolting need NUREG-0933, designated the Generic Safety Issue 29 as a by formulating a comprehensive program to address the high priority issue "Solting Degradation or Failure in 1ssues. The goals of this program are to provide Nuclear Power Plants * [2), and indicated its concern:
"There are numerous bolting applications in nuclear definition of the critical issues involved and to power plants ....'
consolidate industry resources in order to supply an appropriate response to the bolt integrity question.
Work includes corrosion and fracture mechantes studles, "The number of bolting related incidents has non-destructive examination (NDE) development, codes increased ... therefore, there is increased concern and standards activities, and maintenance and training regarding the integrity of the primary pressure tasks. Technology exchange has been effective in boundary in operating nuclear power plants and the assuring increased attention to the behavior of bolting reliability of the component support structures in U.S. nuclear plants. Nothing has been discovered to fo11 ewing a LOCA (loss of coolant accident) or l' earthquake."
raise concern regarding bolting integrity, primartly due to the redundant nature of bolting in critical closurejoints. There are four distinct bolting issues grouped by e appitcation and apparent cause as described below:
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- Group I - Oegradation or failure of pressure program and examines both pressure boundary and boundary bolting 'due to general borated water . structural support bolted connections to assess the
. corrosion (wastage or erosion / corrosion). The degradation of the overall bolted connection. The cause of .fallure is attributed to high rates'of consequences of joint degradation in terms of leakage corrosion of low alloy steels in the presence of and leak-before-break margin for pressure boundary borated water. This is primarily a maintenance joints, and support stiffness and faulted load problem. , condition margin for structural support joints, is being exaralned. This approach is an alternative to
- Group II - Degradation or failure of pressure individual fastener integrity assessment. Since one of boundary bolting due to stress corrosion cracking the principal design features of a bolted connection is
($CC). The cause of these failures can be its stru.ctural redundancy, this alternative seems more attributed to an undesirable combination of realistic, provided that acceptance criteria for both stress, environment, and material condition. safety and reliability can be met.
Generally, these types of failures are associated with leaking gaskets or the use of certain The ultimate goal is to use the generic analytical
-lubricants and/or sealants. Failures can te methodologies developed by EPRI for bolted joint eliminated through proper use of tensioning integrity assessment, supplemented by industry techniques, lubricants and sealants. expertence and data - both nuclear and non nuclear - to demonstrate the safety margins in both pressure Group III - Degradation or failure of internals boundary and structural joints, and to recommend bolting due to fatigue and stress corrosion realistic inspection or maintenance programs for cracking. Failures are generally related to uttllties.
materials, heat treatment, forming technique, high steady-state stresses, or cyclic service. CURRENT ACCOMpL!stt1ENTS Fallures can be eliminated by. alternate materials .
selection and design modifications. The integrated bolting program is scheduled for completion in May 19C6. Currently, most of the Group IV - Degradation or failure of supports and identified deliverables are available. The most
, embedmont bolting due to SCC. Failure can be notable are: overall bolting failure and success attributed to a combination of high stress, rates, fracture mechanics (FM) assessment, generalized susceptible material condition, and a wet closure Integrity (leak-before-break) model, bolt-up environment. They can be eliminated by attention procedure guidelines, thread lubricants evaluation, and to pretension and materials. . mechanical maintenance training tapes.
1 Fallure groups I, II, and IV above have been Overall Bolts Failure Probabilities addressed generically under the auspices of a joint AIF/MPC Task Group on Bolting. Pressure boundary An experience base of bolt failures and/or bsiting (Groups I and II) has a greater influence on problems has been generated and analyzed. Of the five
' system integrity and the highest priority in the types of pressure boundary closures studied, industry program. Internals bolting (Group III) is specifically pressuriters, steam generators, reactor being effectively addressed by component vendors and coolant pumps, and valves, steam generator manway owners' groups and is not being considered generically. bolting exhibited the highest total reject rate. The Supports /embedmont bolting (Group IV) presently has a reject rate Gr manway bolting attributed to wastage
! s2condary priority in the industry prograe. The wa .2 x 10 and the rats for cracking was 2.3 x l
remainder of this paper will focus on the pressure 10 b:undary and supports /embedmont generic program.
- Fracture Mechantes Assessment GENERIC INDUSTRY PROGRAM ON PRES $URE B0UNDARY AND SUPPORT 5 BOLTING The initial technological thrust regarding bolted joint integrity was directed toward the evaluation of The question of fastener integrity is very complex individual fasteners using probabilistic fracture and involves many disciplines (e.g., metallurgy, mechanics analysts. As with any FM evaluation, loads, fracture mechanics, mechanical and corroston properties, and crack sizes had to be known. The NDE
, engineering) and activities such as bolt tension requirements for the FM based approach exceeded the control, NOE, design, specifications and standards, state-of-the-art capabilities. It soon became evident manufacturing, and quality assurance / control. Research that a more prudent and simpler approach to bolted activities have focused on understanding, Identifying joint integrity assessment was possible and desirable,
- and implementing solutions to.the issues. The research Nevertheless, the fracture mechanics research has work on bolting is in three key areas - structural greatly improved the state of information regarding l Integrity analysis (including
- nondestructive bolts and other threaded fasteners. Cipolla(3)has.f.
- examination) corrosion studles, and maintenance studied the application of fracture mechanics to l
i improvements. fasteners, and has presented a simplified analysts j method for approximating the stress intensification EPRI is completing a " Generic Bolted Joint factor for an elliptical surface crack at the root of a Integrity" project which integrates the industry thread. This approach will be Integrated into design, ;
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a ilfe and failur,e prevention analyses, teak Tightness Generalized Joint Integrity Model The most desirable attribute of a bolted joint is Probably the most significant contribution of the its leak tightness. A recent NRC survey (5) integrated bolting program is the development of the demonstrated that over 90% of all bolted connections in generalized joint integrity model, wherein the closure the primary pressure boundary are leak tight. The key is modeled accurately, incorporating the load sheddirg elements that contribute to leak tightness are:
and redundancy inherent to bolted connections. With adequate joint design, proper cleanliness, proper the evaluation based upon the overall closure. the gasketing, uniform and sufficient preload. It should detalls of Individual fastener degradation are not be noted that all but the first element are controlled required. As a result, the complexity of the by station maintenance. The integrated bolted joint calculations and data burdens are substantially reduced program addresses two aspects; the uniformity of and many of the areas of greatest uncertainty are preload and an evaluation of thread lubricants. The avoided. Either wastage or stress corrosion cracking findings indicate that a joint can be prepared with can be accomodated in the model. Equally impo,rtant is uniform preload in the bolts even with simple torque the fact that the NDE requirements of the generalized wrenches, if particular attention is paid to bolt up, joint integrity model are well within the Recomendations loclude stepped torque values with state +of-the-art capabilities. multiple passes and verification of proper preload by ultrasonic or other method. Studs are preferred over The philosophy behind the model is analogous with bolts for many applications. Leak ttgntness can be assured with proper care.
the leak-before-break philosophy used in fM evaluations of other pressure boundary components. The steps required to achieve the desired result, t.e., A second task evaluates thread lubricants. The demonstrate that a degraded joint (due to wastage, work includes laboratory and field tests and indicates cracking, etc.) has ample margin against catastrophic > that the nickel based lubricants can be substituted for failure when the leakage from the joint reaches levels those using molybdenum disulphides (MoS ) without that have a very high probability of detection, modifying the nut factor and, thus, the 2torque values.
include: knowledge of the degree of load shedding to A recomendation is made to not use these Mo$gbased adjacent fasteners due to fastener degradation, lubricants when dissociation is even suspecteo, f.e.,
knowledge of the joint opening profile accounting for when the joint w111 be exposed to water and high gasket spring back and flange distortion, realistic- temperatures. Reviews by Rungta and Majumdar (2) also calculation of leak rates through the degraded joints, included work in this area. The Materials Properties margin demonstration in load-carrying capabtitty of
- Council has done extensive studies on various thread degraded joint and margin definition. lubricants and their contributton to SCC [5).
This philosophy was initially successfully applied RESOLUTION OF THE GENERIC !$$UES to a steam generator manway cover to show its load carrying capability. Example calculations indicate Industry resolution of the generic issues is that for a typical sixteen bolt joint, about three scheduled for 1986, and is planned to be accomplished bolts must fall before a " detectable" leak (10 gos) is utilizing AIF and EPRI guidelines, and American Society generated. The stresses in the adjacent bolts increase for' Testing and Materials (ASTM) and American Society by less than 30% (well below the yleld strength of the of Mechanical Engineers (ASME) standards activities.
i bolts).
l On September 20, 1984, the Institute of Nuclear l Cipolla and Proctor [4] continued this work and Power Operations (INPO) issued $0ER No. 84-5 entitled, i
have proposed a method to establish leak rate margins " Bolting Degradation or Failures in Nuclear Power Plants." INPO conducted an independent review of the and nondestructive examination limits for bolting materials comonly used in primary pressure boundary issues and arrived at conclusions which reinforced closures. Analysis methods for determining the previous AIF recomendations. The SOER serves to structural behavior and leakage of a bolted closure for highlight and provide a " road map through the issues to various amounts of bolt degradation were refined and ut111ttes." The !NPO $0ER depends heavily on AIF/MPC and EPRI programs, calculations completed for several additional essential components (check valve flanges, reactor coolant pump main flange, and pressurizer manway flange). Initial ASTM comittees, having responsibility over results indicated that leak rates between I gpm and 10 bolting material specifications, have reviewed
' gpm are possible without compromising the closure specifications under their jurisdiction to determine integrity. These analyses ~ should provide sufficient need for modifications based on comercial nuclear j ' basis for recomending rev,lstons to present Code NOE industry experience and input from the Industry requirements. Use of leak-before break criterlon is program. Several speelfications have already been,'
considered an effective method of assuring closure modified. At the November 1985 meeting of the ASTM Integrity while reducing demands on NDE. These Subcomittee F16.02 (on structural bolting), it was calculations clearly demonstrate the degradation recomended that a new subcomittee be formed within g tolerance of bolted connections. ASTM to rationallte ASTM structural bolting QA requirements. This recomendation is under consideration by Comittee F 16, 3
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EpRIisdevelodingtworeferencemanualsto industries expected to benefit from BTC technology address how to identify safety-related joints, includes the fastener, chemical, petroleum, aerospace, selection of appropriate procedures, assembly and nuclear power, automotive and manufacturing industries.
dissembly procedures using various methods, inspection cnd verification of previous procedures to solve problems which have occurred in the past. These CONCLU$!ON manuals will serve as a repository of useful ,
information learned from EPRI analytical and The purpose of the AIF/MPC Task Group was to experimental programs and, when published, they will develop and execute the AIF program for resolution of give the utility industry guidelines for bolted joints. generic bolting issues. $1nce this has been
- Several utt11tles are already using drafts of these accomplished and all work is nearing completion, the guidelines in their efforts to enhance their bolting Task Group was disbanded in November 1985. The AIF program. It is believed that the bolting reference Subcomittee on Materials Requirements will be manuals will satisfy the industry's need for guidance available to handle any residual problems.
In this area.
Plans are to develop two comprehensive documents EPRI is developing an Interpretive p' aper on pressure boundary bolts and structural bolts for use explaining the current ASME Boller and Pressure Vessel by utilities. These would be considered the final Code Rules for Bolting. ANS!/ASME Section !!!, Nuclear products from the program. They will be approved by
- Power Plant Components, has been reviewed and the AIF Subcomittee on Materials Requirements and recomunendations have been developed for clarification distributed to ut111ttes and the NRC for their use and and ampitf tcation of existing requirements. ANS!/ASME information in resolution of the issues.
Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, is considering recomendations The bolting issue is an example of a cooperative to rationalize bolting inspection requirements to focus effort between the ut111 ties, the vendors, and the NRC on "at-risk" applications in service sensitive lines. to resolve a problem with potential safety Results from the EPRI generic joint integrity program significance. Shortcomings in design, material t;lll be used as input to develop empirical rules for specification, procurement, and maintenance were inspection frequency and acceptance requirements. NDE identified. Flxes to alleviate concerns regarding rules will be modified to accomodate recent pressure boundary bolt integrity were formulated and developments in NOE technology. are being implemented. The implementation of the recomendations will result in leproved plant An EPRI/MEAC workshop was held in November 1985 availability and reliability, with reduced usintenance, in Charlotte, NC to enhance technology transfer. This man-rem exposure, and inspection costs.
workshop was aimed primarily at informing maintenance personnel of the issues and alerting them to the REFERENCES practical tools being developed in this program to aid I them in achieving leak free joints. The workshop was a 1. Merrick, E. A. and Marston, T. U., " Industry tremendous success with participants agreeing that Response to the issue of Bolting Degradation or additional workshops would be useful. Failure,' Paper No. 06/1, Presented at SMIRT/8, Brussels, Belgium, August 1985.
The Bolting Technology Council was formed to ;
provide opportunities for threaded fastener and tool 2. NUREG-0933, "A Prioritization of Generic Safety I users to engage in cocperative activities. As stated Issues," Safety Program Evaluation Branch, e la their Bylaws, the purpose of the Council is "to Olvision of Safety Technology, U.S. NRC, December sponsor research; to recomend practices; to act as a 1983, New Generic issue 29, " Bolting Degradation clearinghouse for information; and to provide education or Failure in Nuclear Power Plants."
c:ncerning the art and science of the installation and behavior of mechanical fasteners and their Interaction 3. Cipolla", R. C., " Stress Intensity Factor with the joints they are used in." Approntmations for Semi Elliptical Cracks at the Thread Root of Fasteners, improved Techneteoy for Although a large number of engineering and Critical Bolting Applications ASME Pressure industrial societies have been organized to deal with Vessel and Piping Conference, Chicago,, July 1986, various aspects of f asteners and joints, very little attention is paid to the important job of installing 4. Cipolla, R. C. and Proctor, R. R., "Appilcation of fasteners correctly. It is therefore the Intent of the Leak Before-Break Analy:Is Methods to Primary Council to complement rather than dupilcate the work of System Bolted closures," Improved Technoloar for-cthers. Critical Boltina Applications ASME Pressure Vessel and Piping Conference, Chicago, July 1986.
1he Bolting Technology Council is affiliated with The Materials Properties Counct), Inc. (MPC), formerly 5. U.$, Nuclear Rew1 story Comission, NURIG 1095, The Metal Properties Council. Speelfic areas of " Evaluation of Response to IE Bulletin 82 02,
, interest include achieving, maintaining and post- Published May 1985, assembly changes in preload, behavlo* of the joint 6. The Metal Properties Council, Inc., " Progress under load, and joint failure modes. A partial list of Reports and Background Technical Information on MPC Bolting Task Group Study of Lubricants and
- 4 Sealants,*, December, 1985. .
i APPLICATION OF LEAK 8EFORE BREAK ANALYSIS METHODS TO PRIMARY SYSTEM BOLTED CLOSURES R. C. Cipsele, Principet E ngineer end R. R. Proctor. Engineer Aptoch Engineering Servleen. One.
Palo Alto, California A851RACT basis for integrating appropriate altigating measures, A strategy is proposed that will establish leak
, such as preload control, nondestructive examination rate margins and nondestructive examination llotts for (POE), and leak detection capabilities. In order to botting materials consonly used in primary pressure assure the Integrity of the primary pressure boundary.
boundary closures. In the application of leak before. A Bolted Joint Integrity Program has been break analysis methods to closures, an analogy is sponsored by the Electric Power Research Institute drawn between a welded joint and a bolted joint with (EpRI) with the main objective of obtaining a better regard to structural redundancy, load shedding understanding of the behavior of bolted closures.
behavfor, and early warning detection created by the Primary emphasis is placed on the safety acceptance of presence of a leak. Analysis methods for determining the degraded bolted closure, but it is espected that the structural behavior and leakage of a bolted improvements in closure reliability will occur as closure for various amounts of bolt degradatfon are well. The purpose of this paper is to present a leak.
pres ented. Calculations have been completed for two before break strategy for resolving bolted closure stears generator / pressurizer manway cover designs, two integrity issues as the continuation of past work (1) and to show how this approach could be implemented check valve flanges, and a reactor coolant pump main flange. Results indicate that leak rates in excess of through the ASHE Boller and Pressurt Yessel Code (6).
1 GPH (0.042 kg/s) and as high as 10 GPt1 (0.42 kg/s) are possible without compromising the closure LEAK.BtFORE.
- ~ BREAK EVALUATIONS FOR CLOSURES integrity significantly.
The leak.before-break criterton was originally ,
. INTRODUCTION proposed in the late 1960s as a means of estimating I the necessary toughness of pressure vessel steels so
! Recent service emperience with primary pressure that a surface crack could grow through the wall.
causing leakage nf vessel contents to detectable boundary bofting .in pressurised water reactors (PWR)
Indicates carbon steel fasteners can become degraded levels before fracturing. As a result, this philo.
as the result of prolonged contact with primary sophy has been effectively used in the assessment of integrity issues for welded pressure vessels and i
coolant water at elevated temperatures (),,1, 3 . The piping components fabricated from ductile materials.
closuresthathaveesperiencedboltingdegradai)on t include primary side manway covers of steam If a component exhibits a leakage failure mode prior and pressurtaers, coolant pump main flanges,and generators to the point where the actual integrity becomes some questionable, then the demands on NDE methods other primary valve flanges. Of the closures listed, the than leak detection can be reduced. Hence, the steam generator manway covers have been the most troublesome (4).
objective of any leak before break analysis is to show Ipdividual fasteners have been observed to suffer that leakage will always precede failure by a suitably safe margin.
fromgeneralcorrosion(wastage)atthethankor threaded sections or from stress corrotton tracking The basic sf allarttles between a bolted closure
, ($CC) at the thread root. Although degradatto.1 of and a welded joint with respect to material selection Individual fasteners has aised some questions with design requirements, control of fabrication processes,,
.i regard to closure integri y, operating experience also and preservice inspection suggest that an assessment suggests that only a small neber of closures have plan for closures could make use of a leak before.
break philosophy in much the same fashion as with actually degraded while in service. By focussing on welded pipes or vessels. $1nce one of the principal
[ '
these ' service sensitive" closures, a generic plan for design features of a bolted connection is its addressing the integrity of a joint could be deve. structural redundancy it seems plausible that a e
loped. Such a plan would also provide a rational bolted closure, even w,ith some degraded or failed 5
l i
L_ -
f asteners, could meet acceptance criteria consistent The parameters that govern bolt degradation and with current industry practice provided that emple ultinately the integrity of a closure would naturally
' safety margins and closure reliability could be include the material condition, the closure loads, demonstrated. As an alternative to current emphasis and the environment benng contained. Because the $CC within the A5ME Code on Individual fastener integrity. susceptibility of low alloy steels increases with an assessment strategy is proposed that will establish increasing strength, those parameters that ef fect the acceptance of a closure provided that the variability in strength are the most important:
following conditions are met: specifically, material spectitcation, heat treatment, ,
and nominal strength level. The stress-related -
e Leak-before-break of the closure is assured variables include preloading method, preload level, under the design basis conditions for the plant anticipated service loadings, and the joint stiffness and load redistribution characteristics of the e The safety consequences of closure leakage are closure. Given suitable numerical methods, the acceptable closure displacements and bolt stresses can be computed for a wide range of degraded bolt conditions.
e The margin against break at the point when the Finally, the environment variables include tempera-leakage becomes detectable exceeds an . ture. humidity, and the presence of corrosive agents.
acceptance level These environmental effects are used to estimate the range of possible initial degraded bolting conditions A proposed assessment strategy for bolted clo- prior to the application of service loadings. Based sures that exploits the leak-before-break philosophy on these postulated conditions, requirements for ,
is depicted in Figure 1 The suitability of this alternative NDE measures can be proposed once the strategy to closure evaluations will depend on resulting leak rates and available safety margins are available margins as dictated by the conditions estabitshed for a given closure design.
required for closure failure, the amount of external leakage from the closure, and the availability of leak SERVICE SENSITIVE CLOSURES detection instrumentation. Clearly, the character- '
istics characteristics of joint behavior in terms of The focusing of inspection and maintenance load redistributton and gasket unloading followed by lctivities on service sensitive closures will allow flange separation must be quantified for a valid and for more effective resolution of equipment leakage accurate detemination of safety margins. Load problems. During the investigation of primary changes within the joint are due to postulated bolt pressure boundary bolting problems, the AIF/MPC Task degradation (wastage or cracking due to corroston) Group on Bolting and EPRI developed a Bolt Failures that will cause the degraded region of the closure to Data Base with a specif!c objective of identifying unload at the expense of neighboring regions which now troublesome closures. The failure data were compiled I must carry a greater portion of the pressure loadings. ' primartly from uttitty responses to IE Bulletin 82-D2
- and Licensee Event Reports up to September 1984. It was the intention of the AIF/MPC Task Group that this Cia.<* field infomation, along with historical data on plant
'"** specific closure performance from preservice and hydrotesting, will help define the service sensitive closures.
o The Bolt Failure Data Base was used to estimate rejection rates for fasteners used in five closures t ir toi ** @ . steam generator manways, pressurizer aanways,
,mierte, valves, reactor coolant pump (RCP) seals, and RCP i
rui. ,s Wa*fa
- , .ri tes flanges. A surnary of bolt rejection rates for all l reported causes including boric acid corroston, i mechanical damage, cracking, etc., is given in
" Table 1. The rejection rates were computed on two
- bases: first on the total number of bolts at risk and tua . again on the total service years for the bolts at
'd$'g** . risk. On either basis, the ranking of closure type is the same with the steam generator manways exhibiting l' the highest frequency of fastener replacements. The o RCP main flange, pressurizer manway, and valves 9reater than six inches (15 cm) in diameter were also o o troublesome but all exhibited rejection rates less than half that for the steam generator manway.
cin.re The cause for rejection of generator manway studs
/,'M', rativre was principally due to boric acid corroston as shown
- of the causes for rejection. ,,*
J '
f Although one can argue on the overall conclusions-
! that can be reached from these limited data, the
"g',','**,' information does help to focus the types of components
=coas requiring utility attention for improved maintenance practice as well as identifying candidate closure designs for evaluation by leak-before-break analysis Figure 1 - Closure Integrity Assessment Strategy. methods.
8
Table 1 the gasket. The finite element mesh for the 16-stud REJECbONRATESFORBOLTINGINPRIMARY cover is shown in perspective view in Figure 2. The PRES 5URE BOUNDARY CL0iURES (Att CAUSES) studs were modelled by beam elements which were connected directly to the solid elements. Orthogonal 1 %
rigid links were connected to the beam element end Total Bolts Total Bolt nodes to induce stud bending when cover and flange Closure Type At Risk Service Years surf aces do not remain perpendicular to the stud during loading. To simplify the analysis, the gasket Steam Generator Manways was modelled as an elastic foundation represented by 5.81% 3.98% discrete untantal elastic springs. The elastic Reactor Coolant Pump 2.85% 2.48% loading and unloading behavior of a spiral wound Main Flange asbestos filled gasket was inferred from experimental Pressurizer Manways cyclic stress deflection curves (8,) and used to 2.28% 1.20% define the spring element stiffnesses. The Valves (>6 Inches 2.10%* --- deflections from the tests were matched to the actual (15.2 cm) Diameter) manway by relating the gasket properties through Reactor Coolant Pump rattos involving stress area and gasket thickness.
0.82% 0.85 Because of the massiveness of the vessel flanges, the Seal Flange flange surface was assumed to be rigid. Stud preload was established in the model by imposing a
- (Note:
Estimated BoltsFor All Primary at-risk Valves is the total By 5tatistical population Analysis of bolts inser-(4) differential temperature between the studs and the vice for the given closure during the reporting period) cover. A nominal preload of approximately 30 kst (207 HPa) and internal pressure of 2235 pst (15.4 NPa) were used in the study. Stud degradation was
, Table 2 modelled by changing the area of individual bolt STEAM GENERATOR MANWAY elements to simulate partial wastage or by removing bolt elements to model complete fastener fallure.
STUD REJECTIONS BY CAUSE Cover separation was predicted in the 20-stud' model when approximately two studs were assumed to Bolts % Of j beve failed; whereas, in the 16-stud manway, the Cause For Rejection Rejected Total cover first lif ted away from the gasket when one stud Boric acid corrosion 116 37.1 was assumed to have failed. When increased amour.ts Galled / mechanical damage / 65 21.3 of degradation were permitted, including multiple thread damage / removal damage stud failures, a redistribution in both gasket and Pitting / removal damage 65* 21.0 stud loads was observed. The change in gasket load Stress corrosion cracking 32 10.3 in a 20-stud manway from the "preload only" case Linear indications through to various degrees of stud failures under 16 5.5 internal pressure of 2235 psig (15.4 HFa) is shown in Cracks 'S 1.6 Figure 3. The unifom gasket load becomes nonlinear Corrosion / erosion / steam cut 4 1.2 Corrosion / mechanical damage 3 1.0 as the stud 3 degrade and eccentric pressure loading Other 1.0 causes gasket compression to shift. The anqular
_4 position at 2ero gasket load indicates the extent of cover separation.
TOTAL 310 100.0 Stud load redistribution was most significant for
- 61 at one facility for one event the five studs nearest to the degraded region.
Figure 4 illustrates the load shedding and redistri-bution characteristics of the 20-stud manway for a ANALYSIS OF ' PICAl. CLOSURES Primary Manwa . Cover Although there are more than 300 primary manway covers in use in steam generators and pressurizers of United States plants, the basic design is very similar .N.a I in all appilcations. flost covers are typically 27-inch (69 cm) diameter circular plates covering a
- t
' 16-inch (40.6 cm) opening. The cover is 5.75 inches (14.6 cm) thick and held to the vessel by 16 studs.
The 16 studs are fabricated from AISI 4340 steel l according to either ASTM A193-87 or A320-L43 spect-fications and are 1.875 inches (4.76 cm) in diameter.
A 20-stud manway cover of similar geometry is also used by one PWR vendor. The 20 studs are smaller in sire, typically 1.3 inches (3.3 cm) in diameter and 7.'j"*"
I f abricated from similar materials.
s,-wr eim A three-dimensional finite element model was davstcpad to study the deformation behavior of both cover designs as a function of stud preload and J.
.u different degrees of stud degradation. A general s..wi .i
( """
purpose finite element computer program called ANSYS Q) was used to solve for cover displacements as a j function of circumferential position, and the o
conditions under which the cover would separate from Figure 2 - Sixteen-Stud itanway Cecir Model.
, 7
t i i e i i i i i i e i
,, _ to.si.e ==., c. e _
,,g, to
- i. - _
is - _
e,ei..e ne e,en.<e
'i
,n . tail si.e iaisiin ,
/ -
r i io - _
t - _
j: One it.d reae.ee s -
""d' .
e - ,,,,,,,,,, se... ii.43 .
r - .
I f f I f I f f I I
,0 30 60 9 gio 110 640 l
aa,.ier e.isti.0.. e toe,reeil
! Figure 3 - Gasket Load Redistribution for Different Stud Failure Conditions.
, (NOTE: 1 KIP = 4.45 KN) range of conditions including a worst case of seven The finite element model representing 180' adjacent or contiguous studs completely failed (1005 segment of the pump casing, flange, and cover is shown degraded). It is observed that the two studs nearest u to the failed region receives the greatest increase in '
to-sted maar Cow, load, while the second and third nearest neighbors 38 ~
receive a smaller fraction. The load in the fourth ,,,,,,, ,,,i,,,
and fifth closest stud decreases with the unloading es
- 22n est 05.4 wel f.e -
caused by the reduced stiffness of the cover / flange =**
joint. The applied pressure loading performs a 9 greater amount of work in deforming the more flexible t.s -
(degraded) portions of the closure, while slightly less work is done on the greater stiffnesses of the f.e -
s undergraded portions of the closure.
A similar trend in load change is observed in the 2.2 -
- 16-stud design except that the load increase in the nearest stud is greater due to the fewer number of .o -
k studs and greater angular distance between the faste-y ners. Here, load redistribution was most significant i.a -
l for the three nearest studs to the degraded region, as !
- si.e aeuest i.
shown in Figure 5. Only the first two studs share an
~
increased load whereas the third nearest is observed j ,,,,,,
to unload. Percentage-wise, the stud stress in, creases : ,,, .
=areis faster in the 16-stud manway for a given amount of
- closure damage, but larger amounts of leakage would [ ,,,
inere .ren also be expected.
Reactor Coolant Pump Main Flange The main flange and cover of a Type E RCP was o.s -
\r rin .e2 evaluated in the same manner as the manway closure. " * ** ""' 7 The pump cover is composed of an insert plate and 88 -
bolting ring with a bolt circle diameter of approxi-mately 58 inches (147 cm). The insert and ring is E'
' held to the pump casing by 16 studs 4.75 inches (12 i cm) in diameter and approximately 36 inches (91 cm) long. The opening of the pump casing is 48 inches 02 -
E (122 cm) in diameter and the outside diameter of the ' ' ' ' ' '
ring is approximately 80 inches (203 cm). The studs eo, , ,'
. are fabricated from AIS! 4340 steel. Because the % , ,,c ,d ,,,,,,,n,, g ,
mating flange on the casing is of comparable size to i
- the cover, the pump casing was also modelled so that Figure 4 - Load Redistribution in Five flearest l the compilance of the mating flange is well repre- Studs to Degraded Region in a sented. 20. Stud Hanway.
8
t.* ' ' ' '
studs (bcams) and the co. .*/ body to give an approxi-I* -
S t ** *
- mate stretch of 25 mils (0.64 m) translating to a oO O stud stress of about 25 ksi (172 MPa). Internal pump
, . . . . . , pressure was assumed to be 2250 psi (15.5 MPa).
t8 -
a . S ., . on , . a. iireis - The urloading of he flange as studs were removed
- m. ,,, ta. eegree ti.at was similar to that outerved for the manway cover to - i, air. . e ei...
, except that gasket unineding was more uniform with ei a tras est, its.a **t little or no incressa in gasket cornpressive load.
7, . ,,,,,,,
a..,*,
- Ring / insert plate it expected to separate when only one stud is assumen *.o have failed. The increase in g* '
stud stress for various degrees of stud degradation is
- shodn in Figure 7. 1he four closest studs to the
- 2. 8' ~ degraded region are cbserved to carry increased 3
~
amounts of load above their original level of approxi-sn.as aesreit mately 35 kst (240 I N ). As with the manway cover,
- 12 -
/ -
the two studs adjace,it to the degraded region receive t
A the largest increase in load. The load ratios are I.0 - - - - - - - . - greater than the manways because the pressure load is about nine times greater for the pump cover.
There me. rest 3.2
.6 - g g ; , ,
. u - .c w , .a,.
" ~
~
u -
" in.";"*iits'"::1 -
0.t -
2.6 -
t t ' ' I O
e I i 3 4 5 ., t.4 -
n . .e c.ati . reises st. . 5t.e reiseemesrei.t resi . t.
2.s -
Figure 5 - Load Redistribution in the Three Nearest Studs Due to Stud Degradation 28 -
sne,a aemit -
in a 16-Stud Hanway. ,
in Figure 6. The model is comprised of 1200 solid elements with the studs being represented by beam t, 1.6 -
elements and attached to the solid body in the same .
m,e .,eit manner as the canway model. Two pressure retaining .4 -
gaskets are used in the actual assembly of this pump; J however, a single line of gasket (spring) elements of equivalent area and location is employed in the model j t.2 -
'; 8 eta aemit to simpilfy the codel geometry. The studs are ., -
preloaded by a differential temperature between the ,
0.8 -
"# ~
a, drt elete)
! 0. 4 -
0.1 -
g, f I I f f 0 . I 2 3 8 s 6 no r .e coati, reites st .
Figure 7 - Load Redistribution in Reactor
, Coolant Ptsnp Main Flange.
-Check Valves Two bolted flange check valves, one small sis-Dl" inch (15 cm) swing check and another larger ten-inch (25 cm) check, were analyzed in similar manner as the
- previous closures. The valve flanges complied with ANSI B16.5 steel pipe flange design. 1500 lb class.
Check valves were selected for evaluation becaus/ they
' / exhibited the most flange leakage problems as docu.
mented in the Bolts Data Base.
The six-inch (15 cm) valve has a 14.25 inch (36.2 cm) diameter cover with a neck diameter of 7.825
. Figure 6 - Reactor Coolant Pump Model. inches (19.9 cm). Twelve 1.25 inch (3.2 cm) diameter 9
+
i st~ds hold the cover to the body with a specified The unloading of the flange due to stud degrada-proload torque of 500 f t-lbs (680 J). The ten-inch tion was similar to the manway cover analysis except (25 cm) valve has a 19.875-inch (50.5 cm) bonnet that the ten-Irch (25 cm) valve was less uniform, ccvering a ll-Inch (28 cm) diameter opening. Statten probably because of the nonsymetric valve flange body studs,1.625 inches (4.1 cm) in diameter, are used in geometry. The redistributton of the original 31 kst this design. The stud matertal is the same for both (255 HPa) bolt stress is shown in Figures 10 an 11.
valves, specifically ASTM A193 87. Cover separation is expected to c. cur at some point A three-dimensional finite element model of each after two contiguous studs have f ailed although the .
valve is shown in Figures 8 and 9. Both models are specific analysis to show conditions for bonnet lift-180' symetric representations centaining approxi- off was not performed. The load redistribution in the cately 700 elements in each. Because of the impor- ten-inch (25 cm) valve was relatively uneven between tance of flange stiff ness on stud load, the valve two and four contiguous stud failures; however, bodies were also modelled. The basic modelling of the because of the greater density of studs, significa".
studs and gasket follow that of the previous analyses. load is carried by the two studs nearest the degraded A uniform preload of approximately 35 kst (240 HPa) region of the closure as compared to the smaller t*as applied to the studs and the internal pressure was valve.
2250 psi (15.5 HPa).
' I i ! I g 3.0 gt .lach Chett val.e n, t.s -
N l'UNM.INf r.6 -
. n.
l.4 -
t2 -
" ~
e..r.it si.e to t.slee region
,' 5
. i.6 -
, . ; n.. -
a si.e .ii== . . j
% .e.it
. s.t -
/ sec e -
Figure 8 - Six-Inch (15 cm) Check Valve flodel.
- l .0 7 N,
... . ..../
0.6 - ' -
0.4 - -
34aae t --- 0.2 - -
0.0
,g 7 ,, [
a 1 2 3 e s 6
//I/
- see et ce.tig rati e st.as
[ .
i !U
/// lO l y"', Figure 10 - Load Redistribution in Three Nearest f Studs to Degraded Region in a 6-Inch f
x (15 cm) Check Valve.
s x
,k i LEAK RATE PREDICTIONS m y ,
j l / I I # Htxfel Description
, i g f / f Selection of an appropriate codel for predicting fp p- ~ j . , ,. p% , , .
gy, p
g flow through a slit will depend on the fluid conditions and geometric characteristics of the crack (Figure 12). In this case, the sitt is represented by, -
d' the gap between the unloaded portion of the gasket and the previously mating flange surface. The ratio of flow path length to characteristic dimension (i.e.,
hydraulic diameter) defined as L/D. is used to specify the degree of thermal nonequilibrium of the escaping fluid. A leak rate model following this approach Figure 9 - Ten-Inch (25 cm) Check Valve Model. based on Henry's homogeneous nonequilibrium crit,1 cal 10
_. m . . . . - _ _ __ _ __ .
S
. 3.3 8 3 1 4 5 6 6 flow fills t slet L so -
chottag I
,,..ig.c.,,,,,,,,
_ l- ,
su %wsW,ww\wwwumwosw. wwws.
4 u - 3'"**'*a* d*** ' S'1 /
ei 2 50 ein'its".s wai - ~.
,
- i _ ,* .
., ts - ..
- 4. ' . .
" ~ gawww mwwwwm mwg%wwwwwwwwwg %
~
saci.e1l,f;*,"'l ..a.i.
t.: -
_ lievid Jet ., alsture l
t.o -
~
si.e .,.it i.
(a) Two-Phase Flow Through a Long Narrow u -
'* "'d t a _ $11t.
L4 -
e
. l.4 -
uf atti 4 sn a ....,
c , '-
- .' /g tatt plane 1.o g /
- g,g faird nearest ,
,e Al (area at LiO = lt) a.. - ..._... _ - , ..,.. .,.
o.e - _
, [
- a. . .
- /
,Otrectionofflow I f f f f f f
o I s .
m 3.e c o,. reites st.**
, n (b) Critical Flow Model For Leakage Through a $11t.
Figure 11 - Load Redistribution in a Ten-inch Figure 12 - Critical Flow Model For Leakage (25 cm) Check Valve. Through a Slit.
flow model (9) has been developed by Collier (10 and where s' is the momentum density equal to citature subsequently modified by Abdo11ahlan and( ChexaT-)H). density (p for the homogeneous flow assumption.
The general features of the discharge of initially Equation (f) ) can be integrated along the flow path to subcooled or saturated 11guld through a slit is shown evaluate the overall pressure drop across the slit as in Figure 12. In the region 0 < L/0 < 3, a liquid jet the sum of individual drop components to give:
surroundee by a vapor annulus is fomed. For lengths between L/0 a 3 and L/D = 12. the liquid jet breaks up AP =
into droplets at the surface and small bubbles are total APe + APse+ APas+APf (3) entrained within the jet. It is assumed that no mass ,
or heat transfer takes place between entrance and where AP ls the entrance pressure loss, AP and AP
- L/0 = 12 and also the friction pressure drop in this are the Icceleration pressure drops due to Nufd phaf$
region is negligible, change and area change, respectively, and AP is the f
The flow is assumed to be isenthalpic and homo- frict ss s geneous, and pil nonequilibrium effects are introduced e abcVe expressions requires an through a single parameter which is a function of iterative process for a given set of stagnation equilibrium quality and flow path length to di.ameter conditions and slit geometry. The detafis of the ratio. L/D. The one-dimensional mixture mass and numerical procedure are given elsewhere (11 12 . For momentum conservation equations are used to evaluate situations where the flow is not choked, tKe, Tea)k rate pressure drop componerts. The continuity et;uation is calculated from single phase relations with f riction included:
4 dG g+gg=.0 G dA (1) G = 2g (P, - gP b c v I4I o
where G is the mass flux, A is the slit opening area, d rec n of flow coordinate. The p ,,9ri and hW tatio W ac M W i % P specific volume at stagnatio#, W vand P is m
the back pressure. Calculatedleakratesbythe$bove 2 2 methods have agreed well with experimental studieg .
~
dP 1 d GA fC (10).
H " {1.I M ('Tr) t .
- 26 ;- (2)
Computed Closure Leakages The leak rate for each closure was calculated by PICEP (_12,) which was modified to acconinodate the 11 2
t l -
3- -
= - - - _ - . . _ _ _ , -- - , - _ - - , .-- ,
. . .- .. - - - =_= - - _ . - . - -
, v.
' .4
" espected slot openings for the bolted flange relatively high margins at the 1 GPM (0.042 kg/s) leak connections as detennined from the finite element rate.
I results. The subcooled fluid conditions for a
- pressure of 2235 psi (15.4 HPa) and 2250 psi (15.5 IMPACT OF CLOSURE INTEGRITY ASSESSMENTS ON -
MPa) of a temperature of 600'F.(316'C) were assumed. HONDESTRUCTIVE EI411NAT10N I Leak rate estimates for all the closures analyzed i previously are presented in Figure 13. The pump main . With reference to the requirements under flange showed the greatest capacity for producin9 Section XI of the ASME Code, two areas where closure large leak rates owing to the ?arge diameter of the jntegrity assessments would affect NOE are the extent .
sealing surface and smaller number of studs per arc of examinations (IWB-2000) and the flaw acceptance length. The manway covers and valve bonnets exhibit standards (!WB-3000). The extent of examination for similar leak rates and trends. Being a smaller pressure retaining bolting is divided into two.
closure, the six-inch (15 cm) valve is predicted to categories as dictated by bolt size. Category B-G-1 produce smaller leak rates at lesser levels of stud covers principally volumetric examination of bolting degradation; however, significant leakages are whose diameter is greater than two inches (5.1 cm).
possible once degradation his extended to a larger Category B-G-2 is for bolting two inches (5.1 cm) in percentage of the bolting. diameter or less with visual surface examination specified only. These NDE requirements were developed ior f rom conventional bolted joint fabrication r : 8 I i i applications; however, nuclear power plant field experience presented earlier suggests that the
- g,%
volumetric / visual examination cutoff at two inches ,
) ca.o nin (5.1 cm) may require reassessment. If these field
' ~
06 5td'l data provide a statistically representative measure of
~ '
primary pressure boundary closure performance, service N sce e rie,
- sensitive closures could be identified and 38 ~ ' '"'"dd appropriately ranked and NDE requirements established based on known closure performance and on likely roow , -
failure modes. The NDE requirements developed from
- Asuch an approach would not necessarily be the same as
- ./
Q. ~
those in the present 1983 edition of the Code. It would be expected that any alternative approach would
- / / -
103_ emphasize volumetric examination with supplemental
/ / : 5 visual /volianetric EE for those situations when S / / : a leakage from the closure has occurred during service.
l
- I -- i / Category B-G-1 acceptance standards for nonaxial 2
i :/
3'
- flaws are 0.250 inch (0.64 cm) and one inch (2.5 cm)
' / for axially oriented flaws. Closure assessment based j ,
/
1 on leak-before-break will provide a relationship
, j between leak rate and closure integrity as measured in 3 terms of bolt degradation. By selecting a minimum
. 3 j
/[ / $ N '" "'" _
to- s
~d required safety margin, which may vary for different service loading levels, the results from a closure 3
i, . _ assessaent would give the basis for establishing MDE
- requirements. The logic of integrating a leak-before.
i f
[ l
~
breal; philosophy into a determination of requirements and criteria for NDE is shown in Figure 14. From an
/
4 l
- established set of safety eargins, a range of degraded f #3 conditions would be postulated that maintains a j .
f f
~"
~
constant level of closure safety. Leak rates are computed for the range of postulated conditions and the minimen leak rate used to establish detecta-ll
- 38-8, l j { bility limits. Likewise, the type and extent of m , et catipen raties siws . degradation used in the analysts for leak rates provides the basis for selecting NDE requirements and levels of acceptance. Clearly, the example analyses Figure 13 - Leak pate Predictions For Different presented herein provide sufficient bases for Primary System Closures. ,
initf ating Code revisions.
For the closures analyzed, a leak rate of one St#94ARY Am CONCLUS!0MS gallon per minute or 1 GPl1 (0.042 kg/s) is achieved when approximately one to three studs have failed. Closure integrity assessments will provide a The available margins at 1 GPM (0.042 kg/s) are shown rational basis for reconenending revisions to present in Table 3 where the safety factor is based on load Code E E requirements. Satisfying a leak-before-break required to fall the stud nearest the degraded region criterion is an effective strategy for assuring by net section tensile overload. In this determina-closure integrity while at the same time reducing tion, the direct (o and bending (og ) stresses were demands on MDE. Preliminary analysis of various .a '
conservatively adde 3)and compared with specified primary pressure boundary closures (steam generator minimum strength properties. The six-inch (15 cm) and pressurizer manways, RCP main flanges, and check valve exhibits the smallest margin for the conditten valves) suggests that integrity can be assured by
- where 28% of the studs are gone, but because of the monitoring closurie leakage in excess of operational
, smaller pressure load, a safety factor of 2.2 Still limits. Large leak rates are predicted when a few exists. The pump and manway covers all exhibit fasteners are assumed to have failed. Adequate 12
_ . - ~ . _ _ _ _. _ _. _. -- -_ . ~ . - . . _ __ .,, _ - _ _ ~ _ _ . _, , ~ ,
( .
, Table 3
. ASSESSMENT OF MARGINS AT ONE GPM (0.042 kg/s) LEAKAGE Percentage Computed Assumed failed Studs
- Factor Of Bolting For One GPM Safety At One GPM Closure Type Material teak teakage 16 Stud Manway Cover SA320 L43 15.9% 3.2 20 Stud Manway Cover AS40.B24 14.51 3.0 RC Pump Flange Alg3 87 7.8% 3.3 6 Inch (15 cm) Check Valve A193 87 27.51 2.2 10 Inch (25 cm) Check Valve A193 87 17.8% 2.6 safety margins can be dernonstrated provided that . 3. Hall, J. F., "A Survey of the Literature on Low-closure damage is local and that bolting materials are Allow Steel Fastener Corrosion in PWR Power sufficiently ductile as to tolerate heavy damage Plants," Electric Power Research Institute, induced by corrosion. Topical Report NP-3784 (December 1984). *
- 4. Capener, E. L., and R. C. Cipolla,
- Evaluation of 8"a"* 5d"'
Bolting Service Emperiences in Primary Pressure
""'" '" ~ " Boundary Closures," Aptech Engineering Services.
inc., Report AES 8111290-3, EPRI RP2055-5 (December 1984). .
- 5. Nickell, R. E., R. C. Cipolla, and E. A. Merrick, *
,,u.i.a s.,,,4,4
-(**"" "The Use of Leak-Before-Break Criteria and Assessment of liargins in Addressing Closure i Integrity issues, Paper D6/3, SMIRT-8 4 Conference, Brussels, Belgium (August 1985). .
,,g,,,,, 6. American Society of Mechanical Engineers, Boller nun .
and Pressure Vessel Code,Section II, " Rules For ,
the Inservice Inspection of Nuclear Power Plant Components," 1983 Edition.
- 7. DeSalvo, G. J., and J. A. Swanson. *ANSYS - ,
i Engineering Analysis Systems User's Manual," e s, anau ans soanaau im, Revision 4.1, Swanson Analysis Systems, Houston,
- e..s.u.n uua a .u=" (d "'"
PA (itarch 1,1983).
- 8. Barergut, A., "Short Term Creep and Relaxation Figure 14 - Flowchart Showing the Determination .. Behavior of Gaskets," Welding Research Council, of NDE Ferformance Requirements and Bulletin 294 (1985).
Acceptance Criterion.
- 9. Henry, R. E., "The Two-Phase Critical Discharge l ACKNOWLEDGEMENTS of Initially Saturated or Subcooled Liquid,"
l Nuclear Science and Engineerine, Vol. 41 (1970).-
l This work was performed under the sponshorship of i
the Electric Power Research Institute, Palo Alto, 10. Collier, R. P.', et al., "Two-Phase Flow Through I l
California (USA), Research Project 2055-5. The Intergranular Stress Corrosion Cracks and authors also wish to acknowledge the assistance Resu.1 ting Acoustic Emission " Electric Power provided by the AIF/MPC Task Group on Bolting with Research Insitute, Report (in pubitcation).
Special thanks extended to Mr. Kenneth Moore of the
- Sabcock and Wticox Company, tir. Edgar Landerman of 11. Abdollahten. D., and 8. Chexal, " Calculation of i Westtaghouse Electric Corporation, and tir. Walter Bak Leak Rates Through Cracks in Pipes and Tubes,"
of Coeusttor. Engineering Inc.
Electric Power Research Institute, Report NP-3395 (December 1983).
l
- 12. Norris, D., A. Okamoto, 8. Chexal, and T.
l
- 1. Merrick, E. A., and T. U. Marston, " Background Griesbach. "PICEP: Pipe Crack Evaluation and Industry Response to the Issue of Bolting Program " Electric Power Research Insitute.
Degradation or Failure in U.S. Cornercial Nuclear Special Report NP-3596-SR (August 1984).
- Power Plants " Paper 06/1, SHIRT-8 Conference, ?
l Brussels, Belgium (August 1985).
- 2. Anderson. W. , and P. Sterner, " Evaluation of l Responses to IE Bulletin 82 02," NUREG-1095 (May 1985).
13
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