ML20085C081
| ML20085C081 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 04/30/1991 |
| From: | Beth Brown EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20085C085 | List: |
| References | |
| CON-FIN-D-6022 EGG-MS-9201, NUDOCS 9108230175 | |
| Download: ML20085C081 (62) | |
Text
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f ' e'll TECHNICAL. EVALUATION REPORT w :;>y
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TECHNICAL EVALUATION RFPORT ON THE f!RST Q.,
U* /daho 10 YEAR INTERVAL IN$ERuCE INSPECTION PROGetAM -
) PLAN: COMMONWEALTH EDISON COMPANY, BRAIDWOOD Nat/onal HUCLEAR POWER STA110N, UNITS 1 AND 2,
.~-,. . Engineering
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r TECHNICAL EVALUATION REPORT ON THE FIRST 10 YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
COMMONWEALTH EDISON COMPANY, ,
BRAIDWOOD NUCLEAR POWER STATION, UNITS 1 AND 2.
DOCKET NUMBERS 50 455 AND 50 457 i
B. W. Brown J. D. Mudlin Published April 1991 I
Idaho National Engineering Laboratory EGtG Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for:
U.S. Nuclear Regulatory Commission Washington, D.C. 20555 under DOE Contract No. DE AC07 761001570
- FIN No. D6022 (Project 5) ,
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ABSTRACT
- This report presents the results of the evaluation of the Braldwood Nuclear Power Station, Units 1 and 2. First 10 Year Interval inservice Inspection- ,
(151) Program Plan, through Revision 4, including the requests for relief from the American Society of Hechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements that the Licensee has determined to be l impractical. The Braidwood Nuclear Power Station, Units 1 and 2, First 10 Year Interval ISI Program Plan is evaluated in Section 2 of this report for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of xamination sample, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the Nuclear Regulatory Commission (NRC) review before granting an operating license. The requests for relief from the ASME Code requirements that the Licensee has determined to be impractical- for the first 10 year inspection interval are evaluated in Section 3 of this report.
f This work was funded under:
U.S. Nuclear Regulatory Commission FIN No. 06022, Project 5 Operating Reactor Licensing issues Program, Review of ISI for ASME Code Class 1, 2, and 3 Components 11
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SUMMARY
The Licensee, Commonwealth Edison Comppny, has prepared the Braidwood Nuclear Power Station, Units 1 and 2, First 10 Year Interval inservi:e
- Inspection (ISI) Program Plan, through Revision 4, to meet the requirements of the 1983 Edition, Summer 1983 Addenda of the ASME Code Section XI except th&t Clan ? portions of Erergency Core Cooling, Containment Heat Removal, and Residua? Eeat Removal systems were selected / exempted based upon the 1974 Edition, tvme 1975 Addenda of the ASME Code Section XI as required by 10CFR50.55a(b). The first 10 year interval began July 29, 1988 for Unit I and October 27, 1988 for Unit 2.
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The information in the Braidwood Nuclear Power Station, Units 1 and 2, First l 10 Year Interval 151 trogram Plan, Revision 2, submitted December 15, 1988, was reviewed, including the requests for relief from the ASME Code
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Section XI requirements that the Licensee has determined to be impractical.
I As a result of this review, a request for additional information (RAl) was prepared describing the information and/or clarification required from the Licensee in order to complete the review. The Licensee provideo additional ,
information and Revision 3 of the ISI Program in a submittal dated l
August 15, 1990. Further additional information and Revision 4 of the ISI Program were submitted in a letter dated December 13, 1990.
1 Based on the review of the Braidwood Nuclear Power Station, Units 1 and 2,
- first 10-Year Interval ISI Program Plan, through Revision 4, the Licensee's responses to the NRC's RAI, and the recommendations for granting relief from
! the ISI examination requirements that have been determined to be impractical, it is concluded that the Braidwood Nuclear Power Station, Units 1 and 2, First 10 Year Interval Inservice Inspection Program, through
. Revision 4, with the exception of Request for Relief No. NR 12 (in part), as diccussed in Section 3.0 of this report, is acceptable and in compliance l with10CFR50.55a(g)(4).
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CONTENTS I l
A B S T RA C T . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 SUMHARY ............................................................... iii
- 1. INTRODUCTION ..................................................... 1 1
- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN ...................
- 2.1 Documents Evaluated ............................................ 4 1 2.2 Compliance with Code Requirements .............................. 4 2.2.1 Compliance with Applicable Code Editions . . . . . . . . . . . . . . . . . . . 4 2.2.2 Acceptability of the Examination Sample .................... 5 2.2.3 Exclusion Criteria ......................................... 5 2.2.4 Augmented Examination Commitments .......................... 5 2.3 Cotclusions .................................................... 6
- 3. EVALVATION OF RELIEF REQUESTS ..................................... 7 3.1 Class 1 Components ............................................. 7 3.1.1 Reactor Pressure Vessel .................................... 7 3.1.1.1 Request for Relief NR 9, Rev. 3, Examination Category B A, Items B1.11 and Bl.40, Reactor Pressure Vessel Shell and Head-to Flange Circumferential Welds .................................. 7 9
3.1.2 Presturizer ................................................
3.1.2.1 Raquest for Relief NR-4, Rev. 3 (Part 1 of 2),
Examination Category B-D, Item B3.120, Pressurizer Nozzle Inside Radius Section ............... 9 3.1.3 Heat Exchangers and Steam Generators ....................... 9 3.1.3.1 Request for Relief NR-4, Rev. 3 (Part 2 of 2),
Examination Category B-D, item B3.140, Steam Generator Nozzle Inside Radius Section ................ 9 3.1.3.2 Request for Relief HR-5, Rev. 2, Examination Cate ory B F, Item B5.70, Steam Generator Noz z e t o- El bow Wel d s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 iv
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3.1.4 Piping Pressure Boundary ................................... 12 i
3.1.4.1 Request for Relief NR 2, Rev. 2, Examination Category B J, item B9.11. Reactor Coolant System Component-to Fitting Welds ...................... 12 3.1.4.2 Request for Relief NR 6, Rev. 2, Examination Category B J, Item B9.ll, Reactor Coolant System
- Ci rcumferenti al Piping Welds . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.1.4.3 Request for Relief NR-7, Rev. 2, Examination !
- Category B J, item 89.11 Reactor Coolant System Circumferential P1 ping Welds (Unit 1 only) . . . . . . . . . . . . . 17 3.1.4.4 Request for Relief NR-13 Rev. 3 Examination Category B J, item B9.11, Reactor Coolant System Ci rcumferenti al Piping Welds . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.1.4.S Request for Relief NR 18. Rev. 3. Examination Category B J, item B9.11, Reactor Coolant System Circumferential P1 ping Welds (Unit 2 only) . . . . . . . . . . . . . 19 3.1.4.6 Reauest for Relief NR 19 Rev. 1, Examination C&tegory B J, item B9.31, Reactor Coolant System Branch Pipe Connection Weld ............................ 21 3.1.5 Pump Pressure Boundary ..................................... 21 3.1.S.1 Request for Relief NR-14, Rev. 2. Examination Category B L 2, item B12.20, Class 1 Pump Casings ...... 21 3.1.6 Valve Pressure Boundary .................................... 23 3.1.6.1 Request for Relief NR-3, Rev. 2. Examination Category B H-2. Item B12.50, Class I h1ve Bodies ..... , 23 3.1.7 General (No relief requests) 3.2 Class 2 Components ............................................. 27 3.2.1 Pressure Vessels ........................................... 27 3.2.1.1 Request for Relief NR 8, Rev. 2, Examination Category C A, Items C1.10 and Cl.20, Class 2 Pressure Vessel Shell and Head Circumferential Welds ........................................ ......... 27
. 3.2.1.2 Request for Relief NR-10, Rev. 2, Examination Category C A, Item C1.10, Pressure Retaining Velds in the_ Letdown Heat Exchanger .................... 28 3.2.1.3 Request for Relief NR-ll, Rev. 2, Examination Category C A,-Itern C1.10, Pressure Retaining Welds in the Excess Letdown Heat Exchangee . . . . . . . . . . . . . 30 V
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- l 3.2,1.4 Request for Relief HR-12, Rev. 4. Examination Category C 8, Jtems C2.21 and C2.22 Residual Hest Removal Heat Exchanger Nozzle to Vessel Welds and Nozzle Inside Radius Sections ................ 32 3.2.1.5 Request for Relief NR 17, Rev. 3. Examination Category C-A, items C1.10 and C1.20, Class 2 Pressure Vessel Shell and Head Circumferential Welds .................................................. 34 3.2.2 Piping ..................................................... 34
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3.2.2.1 Request for Relief NR 1, Rev. 2, Examination Category C-F, item C5.31, Class 2 Main Steam System Pipe Branch Connection Welds .................... 34 3.2.2.2 Request for Relief NR 16, Rev. 2 Examination Category C F, Item C5,.21, Class 2 Main Steam and Feedwater System Circunferential Piping Welds (Unit 1 only) .......................................... 36 3.2.3 Pumps ...................................................... 38 3.2.3.1 Request for Relief NR-IS, Rev. 2, Examination Catagory C-C. Item C3.30, Integral Welded Attachments on Class 2 Pumps ........................... 38 3.2.4 Valves (No relief requests) 3.2.5 General (No relief requests) 3.3 Class 3 Components (No relief requests) 3.4 Pressure Tests (No relief requests) 3.5 General ........................................................ 41 3.5.1 Ultrasonic Examination Techniques (No relief requests) 3.5,2 Exempted Components (No relief requests) 3.5.3 Other ...................................................... 41 3.5.3.1 Request for Relief CR-1, Rev. 2, Subarticle IWF 1300, Suppert Examination Boundaries for Nonexempt Component Supports on Insulated L.ines ..................
41 3.5.5.2 Request for Relief CR-2, Rev. 1, Subarticle IWF 2430, Additional Examinations of Component Supports .......... 41 3.5.3.3 Request for Relief SR 1, Rev. 2. Subarticle IWF 1300, Support Examination Boundaries for Nonexempt Safety Related Snubbers on Insulated Components ............... 43 i
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3.5.3.4 Recuest for Relief SR 2, Rev.1. Subarticle IWF.2630, >
Adcitional Examinations of Safety Related Snubbers .....
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- 4. "0NCLUSION ........................................................
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- 5. REFERENCES ........................................................ 53 L e
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. TECHNICAL EVALVATION REPORT ON THE FIRST 10 YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
COMMONWEALTH EDISON COMPANY, !
BRAIDWOOD NUCLEAR POWER STATION, UNITS 1 and 2 DOCKET NUMBERS 50 456 AND 50 457
- 1. INTRODUCTION Throughout the service life of a water cooled nuclear power facility, 10 CFR 50.55a(g)(4) (Reference 1) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice !
examination requirements, set forth in the ASME Ccda Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," (Reference 2) to the extent practical within the limitations of desiga, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinettons of components and systeu pressure tests conducted during the initial 120 month inspection interval sh&1l comply with the requirements in the latest editinn and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the Jkte 12 months prior to the date of issuance of the operating license. -
subject to the lis t p ions and mndification listed therein. The components (including supports) may meet requirements set forth in subscouent editions [
and addenda of this Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein. The Licensee, Commonwealth Edison Company, has prepared the Braidwood Nuclear Pcwer Station, Units 1 and 2, First 10 Year Interval ,
inservice Inspection (ISI) Program Plan, through Revision 4 to meet the requirements of the 1983 Edition, Summer 1983 Addenda of the ASME Code Section XI except that Class 2 portions of Emergency Core Cooling (ECC),
Containment Heat Removal (CHR), and Residual Heat Removal (RHR) systems were selected / exempted based upon the 1974 Edition, Summer 1975 Addenda of the ASME Code Section XI as required by 10 CFR 50.55a(b)(2)(iv). The first 10 year inspection interval commenced with the start of commercial operation on July 29, 1988 for Unit I and October 27, 1988 for Unit 2.
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As required by 10 CFR 50.55a(g;(5), if the licensee determires that certain Code examination requirements are impractical and requests relief from them, the licensee shall submit information and justifications to the Nuclear Regulatory ComrIssion (NRC) to support that determination.
Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's determinations that Code requirements are impractical; alternatively, pursuant to 10 CFR 50.55a(a)(3), the licensee must demonstrate that either (1) the proposed alternatives would provide an acceptable level of quality and safety or that (ii) code compliance would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. The Commission may grant relief and may impose alternative requirements that 'are determined to be authorized by law, will not endanger life or property or the common defense and security, and are otherwise in the pubile interest, giving due consideration to the burden upon the Itcensee that could result if the requirements were imposed on the f acility.
t The information in the Braidwood Nuclear Power Sta ion, Units 1 and 2. First 10 Year Interval 151 Program Plan, through Revision 2, rubmitted
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Decebtr lh,1980 (Referer.ce 3), wu reviewed. This documer.t included the requests for relief from the ASME Code Section XI requirements that the Licensee has determined to be impractical for the first 10 year interval. ,
The review of the ISI Program Plan was performed using the Standard Review Plans of NUREG 0800 (Reference 4), Section 5.2.4, " Reactor coolant Boundary Inservice Inspections and Testing," and Section 6.6, " Inservice Inspection of Class 2 and 3 Components."
In a letter dated May 21, 1990 (Reference 5), the NRC requested additional information that was required in order to complete the review of the 151 Program Plan. The Licensee provided additional information and Revision 3 to the ISI Program in a letter dated August 15, 1990 (Reference 6). Further additional information ar.d Revision 4 to the ISI Program were provided in a
. letter dated December 13, 1990 (Reference 7).
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The Braidwood Nuclear Power Station, Units 1 and 2, First 10-Year Interval l 151 Program Plan, through Revision 4, is evaluated in Section 2 of this l
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report for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the appi! cation of system or component examination exclusion criteria, and (d) compliance with ISI related commitments identified during the NRC review before granting an operating licente, i
.. The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code,Section XI,
.- 1983 Edition, Summer 1983 Addenda. Specific inservice test (IST) programs -
for pumps and valves are being evaluated in other reports.
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- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN This evaluation consisted of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any license conditions pertinent to 151 activities. This section describes the submittals reviewed and the results of the review.
. 2.1 Documents Evaluated Review has been completed on the following information from the Licensee:
(a) Braidwood Nuclear Power Statior,, Units 1 and 2, First 10 Year Interval 151 Program Plan, Revision 2, submitted December 15, 1988; (b) Letter,datedAugust 15, 1990, additional information and Revision 3 pages to the ISI Program; and (c) Letter, dated December 13, 1990, additional infornation and Revision 4 pages to the 151 Program.
2.2 Comn11ance with Code Reauirements 2.2.1 [pmoliance with Aenlicable Code Editions The Inservice inspection Program Plan shall be based on the Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b).
Based on Operating License dates of July 2, 1987 for Unit I and May 20,19P8 for Unit 2, the Code applicable to the first interval ISI program is the 1983 Edition, Summer 1983 Addenda. As stated in l .
Section 1 of this report, the Licensee prepared the Braidwood Nuclear Power Station, Units 1 and 2. First 10 Year Interval 151
. Program Plan, through Revision 4, to meet the requirements of the 1983 Edition,-Summer 1983 Addenda of the ASME Ccde Section XI except that Class 2 portions of Emergency Core Cooling (ECC),
l Containment Heat Removal (CHR), and Residual Heat Removal (RHR) 4 e
systems were selected / exempted based upon the 1974 Edition, summe 1975 Addenda of the ASME Code Section XI as required by 10 CFR 50.55a(b)(2)(iv).
2.2.2 Acceptability of tb_q_{ygmination Sample inservice volumetric, surface, and visual examinations shall be performed on ASME Code Class 1, 2, and 3 components and their
. supports using sampling schedules described in Section XI cf thet ASME Code and 10 CFR 50.55a(b). With the Licensee's comdtreents in the December 13, 1990 submittal, the selection of weids ano wld sampic size have been implemented in accordance with the Code and appear to be correct.
2.2.3 Exclurion Critnia The criteria used to exclude components from examination shall be consistent with Paragraphs IWB-1220, IWC-1220, IWC-1230, IWD-1220, and 10 CFR 50.55a(b). The exclusion criteria have been applied by the '.':'nsee in accordance with the Code as discussed in Section 2.4, " Exempt Components," of the Braidwood, Units 1 and 2, 151 Program Plan, and appear to be correct.
2.2.4 Aucmented Examination Commitment 1 The Licensee has stated, in the 151 Program Plan, that the fallowing augmented examinations are being implemented during the first 10 year inspection interval:
(a) Augmented examinations will be performed on Class 2 and 3 high energy piping systems outside containment rhere breaks have not been postulated. Volumetric examination of circumferential welds shall be performed on 6-inch and larger diameter piping in the main steam (MS) and main feedwater (FW) systems from the
. centainment wall to the MS and FW torsional restraints downstream of the outboard containment isolation valves (within the valve room).
(b) Augmented examinations per Regulatory Guide 1.14, " Reactor Coolant Pump flywheel Integrity," (Reference 8) will be performed.
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l (c) Augmented examinations of the turbine rotors and turbine disks will be performed.
(d) Augmented volumetric examinations of the shell to transition ;
cone weld #6 of the loop 1 steam generator ari the lower to upper intermediate circumferential shell weld #8C of the pressurizer will be performed (Unit 1 only). ;
(e) Augmented volumetric examinations of the loop 1 elbow to isolation valve weld #2RC-01-04 will be performed (Unit 2 only).
. (f) Augmented volumetric examinations will be performed on Class 2 '
ECC, RHR, and CHR systems (chemical and volume cor, trol, residual heat removal, safety injection, and containment spray) that are not currently subject to volumetric examination as required by the Code. The inspections will include 7.5%
sampling of the total population of circumferential piping welds (greater that 4 inches nominal pipe size) that contair stagnant borated water. Nominal pipe wall thickness and pressure / temperature exemptions do not apply.
(g) Eddy current inspection of the steam generator U tubes will comply with NRC Regulatory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes,*
(Reference 9) and the Braidwood Technical Specification 3/4.4.5, " Steam Generators."
2.3 Conclusions Based on review of the documents listed above, it is concluded that the Braidwood Nuclear Power Station, Units 1 and 2, first 10-Year Interval Inservice Inspection Program Plan, through Revision 4, is acceptable andincompliancewith10CFR50.55a(g)(4).
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- 3. EVALUATION OF REllEF REQUESTS The requests for relief from the ASME Code requirements that the Licensee has determined to be impractical for the first 10-year inspection interval l are evaluated in the following sections. l 3,1 Class 1 C.groenents 3.1.1 & actor Pressure Vet,te,1 3.1.1.1 bauest for Relief NR-9. Rev. 3. Examination ___Cateoory B- A.
Items 81.11 and 91.40. Reactor Pressure Vessel Shell and_
Head-to-Flance Circumferential Welds Code Reauirement: Section XI, Table IWB 2500 1 Examination ,
Category B A, item B1.11 requires a 100f. volumetric examination of the Reactor Pressure Vessel (RPV) circumferential shell welds as defined by figure IWB 2500-1. Item Bl.40 requires both 100% volumetric and surface examinations of the RPV head to flange welds as defined by figure IWB 2500-5.
Licensee's Code Relief Reauest: Relief is requested from examining 100f. of the rode requ Sed volume on RPV circumferential shell welds IRV 02 002 (Unit 1) and 2RV-02 002 (Unit 2) and RPV head to flange welds 1RV 03+001 (Unit 1) and 2RV-03 001 (Unit 2).
Licensee's Proggigd Alternative Examination: None. The RPV is examined remotely using the immersion technique, and no alternative examination is proposed, However, each of the subject walds will be examined to the fullest extent practical usin3 ultrasonic examination techniques. In addition, a leak test at r.ominal operating pressure will be performed during each refueling outage as required by the Code. The Code-required surface examination will also be performed.
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Lietnsee's Basis for Reouestinn Relief: Lower shell cours9 to Dutchman welds IRV 02 002 and 2RV 02-002 have six core support guide lugs welded to the interior surface of the reactor pressure vessel approximately 3.5 inches above the weld. These lugs restrict the automated inspection tool from inspecting the required volume from the shell course side in the areas of these lugs. All of the wold and heat affected zone can receive 100% coverage from at least one direction,
- however, the required base metal cannot be fully inspected in the areas of the core support guide lugs.
Closure head flange to Dutchman forging welds 1RV-03 001 and 2RV 03 001 are located just above the tapered portion of the flange that physically obstructs the ultrasonic transducer from examining the requirtd scan krea, part of the three larger lifting lugs also f all in the required scan area. The Code required surface examinations will be performed on the accessible areas.
Evaluation: The sketches attached to the relief request show that the Code-required volumetric examination of the subjut welds is obstructed by core support guide lugs, lifting lugs, and the tapered portion of the flange. The reactor vessel design, therefore, makes the volumetric examination of the subject welds impractical to perform to the extent required by the Code. In order to examine the weids in accordance with the requirement, the reactor vessel would require extensive design modifications. Imposition of this requirement on Commonwealth Edison Company would cause a burden that would not be compensated significantly by an increase in safety above that provided by the limited examination.
- Commonwealth Edison Company has stated that the volumetric
, examination will be performed to the maximum extent practical and that the Code required surface examination will be performed. Based on the design of the reactor pressure vessel, 8
an acceptable percentage of the Code required volume will be examined. Thus, the limited Section XI volumetric examination, along with the Code required surface exam).ution and leak test, will provide adequate assurance that unallowable inservice :
flaws have not developed in the subject reactor vessel welds or that they will be detected and removed or repaired prior to the
. return of the RPV to service.
.. Qnclusions: It is concluded that the volumetric examination ,-
of the subject RPV welds is impractical to perform at Braidwood, Units 1 and 2, to the extent required by Section XI of the ASME Code and that public health and safety will not be endangered by allowing the limited examination to be performed in lieu-of the Code requirement. Therefore, it is recommended that relief be granted as requested.
3.1.2 Pressurizer 3.1.2.1 Eta gst for Re\ief NR 4. Rev. L (Part 1 of 21. ExLmination
{ateoory-B D. Itea B3.120, Pressurizer Nozzle inside Radius 1E1110
((QII: Request for Relief NR 4 was withdrawn by the Licensee in the August 15, 1990 response to the NRC request for additional information. Commonwealth Edison is evaluating the feasibility of performing these examinations.
3.1.3 h it Exchanaers and Steam Gfnerators 3.1.3.1 Raouest for Relief NR-4 Rw. 3 (Pa t_2 of 21. Examina_ tion Cateoory B D- Item BM4A Steam Generator f{ozzie inside Ragigi, Section flDJ1: Request for Relief NR-4 was withdrawn by the Licensee in l
the August 15, 1990 response to the NRC request for additional L
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a i information. In that submittal, the Licensee stated that "for l l the interim, Braidwood Station has removed Relief Request NR 4 from the ISI Program. Commonwealth Edison will continue to evaluate the feasibility of performing these examinations. If at some later date we feel relie' is justified, we shall submit a revised relief request to the NRC."
- 3.1.3.2 Eggggst for Egligf NR-5. Rev. 2. Examination Cateagry B-F. Item B5.70. Steam Generator Nozzle to Elbow Welds (2ds_Reauirement: Section XI, Table IWB-2500 1, Examination Category B-F, Item B5.70 requires both 100% volumetric and surface examinations of steam 9enerator nozzle-to safe end butt welds as defined by figure IWB-2500 8. In addition, Appendix !!!, Supplement 7, requires that ultrasonic examination sensitivity be establir.hed using 1.0. notches with a depth of 10% wall thickness.
Licensee's Code Rgliff._Egqqgit: Relief is requested from examining 100% of the Code-required volume of the following steau generator nozzle-to safe end welds:
Unit 1 __
Unit 2 _
line Number Weld Number Line Number Weld Number 1RC01AA-29 1RC-01 8 2RC01AA-29 2RC-01 8 1RC02AA-31 1RC 01-9 2RC02AA 31 2RC-01 9 1RC01AB-29 1RC-02-19 2RC01AB 29 2RC-02-19 13C02AB-31 1RC-02-23 2RC02AB 31 2RC 02-23 1RC01AC-29 1RC-03 8 2RC01AC 29 2RC-03 8 1RC02AC 31 1RC-03-9 2RC02AC 31 2RC-03 9 1RC01AD 29 1RC-04 9 2RC01AD 29 2RC 04 9 1RC02AD 31 1RC-04-10 2RC02AD 31 2RC-04-10 Licensee's Proposed Alternativ.g_Ex. amination: None. The Code required volumetric examination will be completed to the maximum extent practical (an ultrasonic examination-from the carbon steel side and a best effort ultrasonic examination, i
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based on state of the art techniques, from the cast stainless side). The Code-required surface examinattors and leakage and hydrostatic tests will be performed.
Licensce's Basis for Requntina Relief: The welds listed above are cast austenitic stainless steel (SA 351-CF8A)-to cast
. carbon steel (SA 216 GR WCC) with austenitic ste.inless steel cladding. The technique used to examine the cast stainless
- side of these welds will detect large flaws (257. or greater through the w311), therefore, this sensitivity is less than that required by the Code, in addition, welds IRC-02-23 and 2RC-02 23 had limited circumferential scans near the weld toe area. This limitation is due to the transducer system's inability to mainttin coupling while scanning over the weld to base metal transition.
Evaluation: The Licensee's submittal has been reviewed and it was determined that examination of the Code-required volume is limited due to the high attenuating properties of the cast austenitic stainless steel. Commonwealth Eoison has made a reasonable effort to develop, within the state-of the-art, effective ultrasonic equipment and examination procedures for the examination of these cast stainless steel welds. The Licenses has also committed to continue evaluating advanced techniques and, as they become available, to incorporate them in the ISI program plan for the first 10 year interval. In order to examine the welds in accordance with the requirements, the nozzles, and thus the steam generators, would have to be redesigned, fabricated and installed. Imposition of this requirement on Commonwealth Edison Company would cause a burden
. that would not be compensated significantly by an increase in safety above that provided by the limited examination.
Tne Licensee has stated that the volumetric examination will be performed to the maximum extent practical and that the Code required surface examinations will be performed. The 11
_ _ _ _ . m_______
limited Section XI volumetric examination, along with the Code required surface examinations and leakage and hydrostatic tests, will provide adequate assurance that unallowable inservice firws have not developed in the steam generator nozzle to safe end welds or that they will be detected and removed or repaired prior to the return of the steam generators to service.
o
Conclusions:
It is concluded that the volumetric' examination of the subject steam generator nozzle to safe end welds is impractical to perform at Braidwood, Units 1 and 2, to the extent 'equired by Section XI of the ASME Code and that public health and safety will not be endangered by allowing the limited examinations to be performed in lieu of the Code requirement. Therefore, it is recommended that relief be granted as requested.
3.1.4 Pioina Pressure Boundary 3.1.4.1 Reauest for Relief NR-2. Rev. 2. Exemination Cateaory B J. Item B9.11. Reactor CoollDt System Component-to-Fittina Welds Code Reauirement: Section XI. Table IWB-2500-1. Examination Category B J, item 89.11 requires both 100% surface and volumetric examinations of circumferential welds in Class 1 piping systems.as defined by figure IWB 2500 8. In addition, Appendix !!!, Supplement 7, requires that ultrasonic examination sensitivity be established using 1.0. notches with a depth of 10% wall thickness.
Licensee's Code Relief Recues.t: Relief is requested from examining 100% of the Code-required volume of the following reactor coolant system fitting-to component welds:
12
Unit 1 Unit 2 line Number Feld Number Line Numbe.t Weld Number IRC01AA-29" 1RC-01-4 2RC01AA 29" 2RC-01-4 1RC02AA 31" 1RC 01 17 2RC02AA.31" 2RC-01 17 2RC01AD 29" 2RC 04 !
i 2RC02AD-31" 2RC-04 18 licensee's Proposed Alternative Examination: None. The Code required volumetric examination will be completed to the maximum extent practical (a best effort ultrasonic examination of the cast stainless welds based on state of the art techniques). The Code-required surface examinations and system leakage and hydrostatic tests will be performed.
Licensee's Basis for Recuestina Relief: The welds listed above are cast stainless steel elbows to either cast pumps or cast valves. The optimized ultrasonic technique used for the welds in statically cast stainless steel will detect large flaws (25%
or greater through the wall), therefore, this sensitivity is less than that required by the Code.
In addition, due to the unwieldy characteristics of the contoured wedge search units and variations in the machined surfaces, the welds listed above experiento axial and circumferential scanning limitations.
Evaluation: The Licensee's submittal has been reviewed and it was determined that examination of the Code-required volume is limited due to the high attenuating properties encountered in the cast austenitic stainless steel. The Licensee has made a reasonable effort to develop, within the state-of-the art, effective ultrasonic equipment and examination procedures for the examination of these cast stainless steel welds. The Licensee has also committed to continue evaluating advanced i
techniques and, as they become available, to incorporate them in the ISI program plan for the first 10 year interval. In order to examine the welds in accordance with the requirements, l
l
\
13
the components would have to be redesigned, fabricated, and
~
installed. Itaposition of the requirement of Commonwealth Edison Company would cause A burden that would not be compensatedbyanincrejseinsafetyabovethatprovidedbythe limited examination.
The 1.icensee has stated that the volumetric examinatloa will be performed to the maximum extent practical and that the Code required surface examination will be performed. The limited Section XI volumetric examination, along with the Code required surface examination and leakage and hydrostatic tests, will provide adequate assurance that unallowable inservice flaws have not developed in the subject reactor coolant system welds or that they will be dettcted and removed or repaired prior to the return of this system to service.
Conclusions:
It is concluded that the volumetric examinatinn of the subject fitting-tu-component welds is impractical to perform at Braidwood, Units 1 and 2, to the extent required by Section XI of the ASME Code and that public health and safety will not be endangered by allowing the limited e,tamination to be performed in lieu of the Code requirement. Therefore, it is recommended that relief be granted as requested.
3.1.4.2 Reauest for Relief NR-6. Rev. 2. Examination Cateaorv 0-J. Item B9.ll. Reactor Coolant System Circumferential Pioina Veldi
~
.Cade Reautrement: Section XI, Table IWB 2500-1, Examination Category B J, item B9.11 requires both 100%-surface and -
volumetric examinations of circumferential welds in Class 1 4
piping systems as defined by Figure IWB 2500 8. In addition, .
Appendix !!!, Supplement 7, requires that ultrasonic m:f.. stion sensitivity be established using 1.0. notches with s W,1 of 10% wall thickness.
14
, ..- .- . .,,r.__ . , -
Littape's Code Relig.f_Bigu111: Relief is requested from exam'ning 100", of the Code required volume of the following reactor coclant system welds:
Cast Stainless Steel (SA 351 CF8A) Cibow to Stainless Steel (SA 376 GR. 304N) Pipe Line Number Weld Number.(11.,
IRC02AA-31" (Unit 1) 1RC 01 10, 16 1RC03AA 27.5" (Unit 1) 1RC 01-30 2RC02AA 31" (Unit 2) 2RC 01-10, 16 2RC03AA 27.5" (Unit 2) 2RC-01-30 2RC02AD 31" (Unit 2l 2RC 04 17 Cast Stainless Steel (SA 351-CF8A) Pump to Stainless Steel (SA 376 GR. 304N) Pipe Ling _N.jamber k l.d Numb.t.t IRC03AA 27.5" (Unit 1) 1RC-01-1B*
2RC03AA 27.5" (Unit 2) 2RC 01-18*
Cast Stainless Steel (SA-351 CFSM) Valve to Stainless Steel (SA-3?E GR. 304N) Pipe line Number Weld Number IRC03AA 27.5" (Unit 1) !RC-01 22 Licensee's Pro 90std Alternative Examination: None. The Code-requ.*eo volumetric examination will be completed to the maximum extent practical (an ultrasonic examination from the non cast side and a best effort ultrasonic examination based on state-of the-art techniques from the cast stainless side). The Code-required surface examination and system leakage and hydrostatic te;ts will be performed.
Licensee's.jasis for Reqv,qstina Relief: The welds listed above are cast stainless steel to wrought stainless pipe. The optimized ultrasonic technique used for the statically cast stainless steel welds will detect large flaws (25% or greater through the wall), therefore, this sensitivity is less than 15
that required by Code. In addition, due to the uneteldy
~
characteristics of the contoured wedge search units and variations in the machined surfaces, those welds listed with an asterisk experienced axial and circumferential scanning limitations. ,
, haluation: The Licensee's submittal has been reviewed and it was determined that examination of the Code-required volume is
. limited due to the high attenuating properties of cast austenitic stainless steel. The Licensee has nde a reasonable effort to develop, within the state of theott, offective ultrasonic equipment and examination procedures for the extamination of these cast stainless steel welds. The Licensee has also committed to continue evaluating advanced techniques and to incorporate them in the ISI prr, gram plan for the first 10-year interval as they become (,vai'aable. In order to examine the welds in accordance with the rrquirement, the system piping would have to be redesigned, fabricated, and installed.
Imposition of the requireraent on Comonwealth Edison Company would cause a burden that would not be compe'1 sated significantly by an increase in safaty above that provided by the limited examination.
The Licensee has stated that the volumetric examination will be performed to the maximum extent practical and that the Code required surface examination will be performed. The limited Section XI volumetric examination, along with the Code-required surface examination and leakage and hydrostatic tests, will provide adequate assurance that unallowable inservice flaws have not developed in the subject piping welds or that they will be detected and removed or repaired prior to the return of the piping to service.
Gnclusions: It is conc 19ded that the volumetric examination of the subject piping welds is impractical to perform at Braidwood, Units 1 and 2, to the extent required by Section XI 16
.m
1 of the ASME Code and that public health and safety will not be .
endangered by allowing the limited examination to be performed in lieu of the Code requirement. Therefore, it is recommended
-that relief be granted as requested.
3.1.4.3 Reauest fer Relief NR-7 Revt 2. Examination Cateaory B-J. Item __ .
B9.11. Reacine Coolant System Circumferential Ploina Weldt, !
(Unit 1 only)
Code Reautrement,: Section XI, Table IWB 25001, Examination Category B J. Item B9.11 requires both 100% surface and volumetric examinations of circumferential welds in Class 1 piping systems as defined by Figure IWB 2500 8.
Licensee's Code Relief Recuest: Relief is requested from examining 100% of the Code-required volume of the following Unit I reactor coolant system welds: ;
)
Percent Not Line Number Weld Number Examinable Beason for Limited Exam.
1RY0lC-4" 1RC-16-2 16% Elbow inner radius, ,
reducer geometry IRY01B 6" 1RC 16 5 16% Elbow inner radius ,
IRYO3AA 6" 1RC 32-1 16% Elbow inner radius, nozzle geometry IRYO3AB-6" 1RC-32-7 16% Elbow inner radius, nozzle geometry '
IRYO3AC-6" 1RC-32-13 16% Elbow inner radius, nozzle geometry IRYO2A-6" 1RC-35 1 16% Elbow inner radius, nozzle geometry Licensee's Prooosed Alternative Et tmination: None. The Code required volumetric examination will be completed to the maximum extent practical. The Code-required surface ,
. examination will be performed.
Licensee's Besis for Reauestina Relief: The welds listte above have interfering conditions on each side of the weld. These i
17
, . - , . - .. , . - -- . . - . .e - . . - - _ _ _ .
. : interferences can cause: poor coupling of the transducer, limited movement of the transducer, redirecting of the sound beam and, in same cases, complete restriction of a portion of a particular scan, These conditions limit the axial scans so as to leave the listed percent of weld length uninspected. The estimates of weld lengths unable to be examined are extremely-
. conservative and are actually a percent of weld length for which there is less confidence that the entire required weld
- volume will be examined. Based on the required surface and leakage tests, the structural integrity of these welds is assured.
Eyaluatiqn: Review of the Licensee's submittals show that the nozzle or fitting geometry is such that the Code required volume of the subject welds cannot be fully examined.
Therefore.- the volemetric examination of the subject piping welds is impractical to perform to the extent required by the Code, in order to examine the welds in accordance with the requireirent, the subject piping would have to be redesigned, fabricated, and installed. Imposition of this requirement on Commonwealth Edison Company would cause a burden that would not
~
be ccmpensated significantly by an increase in safety above that provided by-the limited examination.
The Licensee has stated that the volumetric examination will be performed to the maximum extent practical: and that the =
Code required surface examination will be performed, it is noted that_a'significant' percentage (84%) of the Code-required volume can and will be namined. Therefore, the limited volumetric examination,'along with the Codt required surface
. examination, will provide adequate assurance that unallowable inservice flaws have not developed in the _ subject piping welds
- or that they will be detected.and removed or repaired prior.to
' the' return of the subject piping'to service.
Conclusions:
It is concluded that the volumetric examination of the subject piping welds is impractical to perform at 18
l Braidwood, Unit 1, to the extent required by Section XI of the ASME Code and that public health and safety will not be endangered by allowing the limited examination to be performed i in lieu of the Code requirement. Therefore, it is recommended that relief be granted as requested.
3.1.4.4 Recuest for Relief NR 13. Rev. 3. Examination Cateaory B J.
Item 89.11. Reactor Coplag.t System Circumferential Pioina Weldi NQll: Request for Relief NR-13 was withdrawn by the Licensee in the August 15, 1990 response to the NRC request for additional information. In that submittal, the Licensee stated that "A 'best effort' volumetric examination will be performed in accordance with the ISI schedule. If the code required volume is not able to be examined, a relief request will be submitted to the NRL at that time.'
3.1.4.5- Recuest for Relief NR-18. Rev. 3. Examination Cateaory B-J.
Item B9.11. Reactor Coolant Systrm Circumferential Pioina Welds _
(Unit 2 only)
Code Reauirement: Section XI, Table IWB 2500-1. Examination Category B-J. Item B9.11 requires both 1007. surface and volumetric examinations of circumferential welds in Class 1 piping systems as defined by Figure IWB-2500-8.
Licensee's Code Relief Reouest: Relief is requested from performing the Code-required volumetric examination of reactor coolant system pipins weld 251-09 17 on line 2RCZ9AC-10" and weld 251-13 28 on line 2RC29AD 10".
Licensee's Prooosed Alternative Examination: None. The Code-required surface examination of the accessible regions (weld 2SI-09 17) and system leakage and hydrostatic tests will be performed.
19
licensee's Basis for Recuestinn Relief: The Licensee states that weld 251-13-28 is encased in a permanent whip restraint, making it inaccessible to both surface and volumetric examinations. Weld 251-09-17 is adjacent to a permanent whip restraint, making it accessible for surface exarainattun but inaccessible for volumetric examination since it is a valve-to pipe weld on the upstream side. ,
Evaluation: As shown in the sketches attached to the relief request, weld number 251-13-28 is inaccessible for both Code-required surface and volumetric examinations because it is encased in a permanent whip restraint, and weld 251 09 17 is accessible for the Code required surface examination but is betwean a' permanent whip restraint and a valve with no access for the Code-required volumetric examination, in order to examine the welds in accordance with the requirements, extensive' design modifications would be required, imposition of the requirements on Ccmmonwealth Edison Company would cause a burden that would not be compensated significantly by an increase in safety above that provided by the limited examination.
The Licensee has stated that the surface examination of weld 2SI-09-17 will be performed to the maxLmum extent practical.
Other similar welds will receive the Code required volumetric and surface examinations; thus, the continued inservice strcctural integrity will be verified by sampling.
[onclusions: It is concluded that the Code rcquired examinations of the subject welds are impractical to perform at Braidwood Unit 2, and that public health and safety will not be endangered by allowing the limited examination to be oerformed in lieu of the Code requirement. Therefore, it is recommended that relief be granted as requested.
20
\
I 3.1.4.6 Reouast for Relief NR-19. Rgr. 1. Examination Cateaorv B-J.
11.em B0.31. Reactor Coolant System Branch Pine Connection Weld NDII: Request ~ for Relief NR-19 war withdrawn by the Licensee in the August 15, 1990 response to the NRC request for
_ additional information. In that-submittal, the Licensee _ stated
" th<t "A 'best effort' volumetric examination will be performed in accordance with the 151 schedula. If the cod' required volume is not able_ to be examined,. a relief reoccet will be submitted to the NRC at that time."
3.1.5 Euro Pressure Boundar.y 3.1.5.1 Reauest for Relief NR-14. Rev. 2. Examination Cateaorv m B-L-2.
Item B12.20. C', ass 1 Pumo Casinas Code Reauirement: Section XI, Table IWB 2500-1, Examination Category B-L-2, Item B12.20 requires a VT-3 visual examination of the internal surfaces of at least one pump in each group of pumps performing similar fonctions in the system. The
, examiaation may be performed at the end of the 10-year interval.
Licensee's Code Relief,P.anggst: Relief is requested from performing the Cede tu' 1 VT-3 visual oxamination of the internal pressure be-ndtr surfaces of one (one for each unit) of the following reactor coolant pump casings:
Unit 1 Unit 2 1RCCIPA 2RC0lPA 1RC0lPB 2RC0lPB IRC01PC 2RC0lPC 1RC01PD 2RC0lPD Licensee's Proonsed Alternative Examination: None. The VT-3 examination will be performed on one of these pumps if disassembly is required for maintenance purposes but will r.ot 21
'Y
- exceed more than one inspection per interval. -In addition, a VT-2 visual examination will be performed after each refueling outage.
Licensee's Basis for Egmstino Relief: The Code requirement, to disassemble at least one pump from each group of pumps performing similar functions, results in high radiation -
exposures and does not produce a proportionally higher potential for identification of service-induced flaws or degradation. The industrial performance of these pumps has proven their excellent ability to resist service degradation or fl awing. The Licensee also reports that the inappropriate balance of antential flaw detection and the large impact on expenditurr,s of manpower t?hout-substantially increasing component reliability is considered impractical. Coupled with this is the highly negative impact on the Station's ALARA program.
- valuation: The examination requirement for internal surfaces af pumps necessitates complete disassembly of the pump. The sisassembly of the reactor coolant pumps for the sole puroou of visual ext ,ination of the casing internal surfaces is a major effort ano requires many manhours from skilled maintenance and inspection personnel. In order to examine the internal surfaces of a reactor recirculation pump in accordance with the requirements, complete disassembly of the pump would be reacired which, in addition to the possibility of damage to the pump, would result in personnel receiving excessive radiation exposare. Therefore, the Code requirement is impractical. The visual examination is performed to determine if unanticipated severe degradation of the casing is occurring due to phenomena such as erosion, corrosion, or cracking.
However, previous experience during examination of similar pumps at other plants has not shown any significant degradation of pump casings. Imposition of the requirements on Commonwealth Edison Company would cause a burden that would not 22 m
- be compensated significantly by an increase in safety above that provided by the proposed examination.
Commonwealth Edison company has stated that the Code required visual examination will be perfarmed on the internal pressure boundary surface of one reactor coolant pump if disassembly is required for maintenance purposes.
Later editions and addenda of the ASME Code (1938 Addenda) have eliminated disassembly of pumps for the sole purpose of performing examinations of the internal surfaces and state that the internal surface visual examination requirement is only applicable to pumps that are disassembled for reasons such as mairitenance, repair, or volumetric examination. Therefore, the concept of visual examination of the internal surfaces of the pump casing, if the pump is disassembled for maintenance, is acceptable, Since no major problems have been reported in the industry with regard to pump casings, the Licensee's proposal will provide adequate assurance of the continued inservice structural integrity.-
Conclusions:
It is concluded that the disassembly of a pump for the sole purpose of inspection is impractical to perform at Braidwood, Units 1 and 2, and that public health and safety will not be endangered by allowing the proposed examination to be performed in lieu of the Code requirement. Therefore, it-is recommended that relief. be granted provided that, if the pumps have not been disassembled, this fact should be reported by the Licensee in the ISI Summary Report at the end of the interval.
3.1.6 Valve pressure Boundary 3.1.6.1 Reauest for Relief flR-3, Rev. 2. Examination Cateaorv B-M-2, Item B12.50. Class 1 Valve Bodies l
Code Recuirement: Section XI, Table IWB-2500-i, Examination Category B-M 2, Item B12.50 requires a VT-3 visual examination l
23
9: -of the internal surfaces of v'alve bodies exceeding 4 inches: ,
nominal pipe size. Examinations shall be conducted on at=least
~
. one _ valve'in each- group of valves that are of the _same design,
- such-es globe,Lgate, or check valve, and manufacturing method and_that perform ~similar functions in the system. The.
examinations may be performed at.the end of the 10-year
' interval..
Licensee's Code Relief Reauest: Relief is requested from-performing the Code-required VT-3 visual examination on the internal pressure boundary surfaces of one valve in each of the following groups (12 valvesitotal for each unit):
Reactor Coolant System-(3 reouired for each unii 2 m.
Unit 1 .
Unit 2' '
1RC8001-A, B, C, or D 2RC8001-A, B, C,-or D lRC8002-A, B, C, or D 2RC8002-A, B, C, or D 4 _lRC8003-A, B, C, or D 2RC8003 A, B, C,_or 0 Piessurizer il reouired for each unit) ,
Unit 1- Unit 2 lRY8010 A,-8, or C 2RY8010-A,.B, or C Residual Heat Removal (2 reouired for each uniti-Unit 1 Unit 2 1RH8701-A, er 3 2RH8701-A,.or-B--
1RH8702-A, or B 2RH8702-A, or B-Safety In.iection f6 reouired for each unit)
Unit 1 .__
Unit 2 [
~
ISIB808-A, 8, C, or D 2S18808-A, B,1C, or_D.
ISI8818-A, B, C, or D 2 SIB 818-A, B, C, or D ISI8841-A, or B- -2SI8841-A, or B ISI8948-A, B, C, or D- _2SI8948-A,-B, C, or D ISI8949-A, B, C, or D 2518949-A, B, C, lor D ISI8956-A, B, C, or D 2SI8956-A, B, C, or.D ,
-- ticensee's Proposed-Alternative Examination: None. The VT-3d examination-will be perfo".med on one valve from each design group in each' system performing a similar-function when t
24
- , , - , - , _ . - . _ . . . u . . _ _ . . . _ _ _ _ _ _ _ _ _
- disassembly is required far maintenance purposes but not to exceed once per inspection interval per design group. In addition, the visual examination during the leakage test will be performed after each refueling outage, licensee's Basis for Reouestina Relief: The Code requirement to disassemble one valve from each group of valves performing similar functions for the purpose of visual examination results in high radiation exposures and does not produce a proportionally higher potential for identification of service-induced flaws or degradation. The industrial performance of these valves has proven their excellent ability to resist service degradation or flawing. The Licensee also reports that the inappropriate balance of potential flaw detection and the large impact on expenditures of manpower without substantially increasing component rel: ability is considered impractical. Coupled with this is the highly negative impact on tne Station's ALARA program.
Evaluation: The examination requirement for internal surfaces of valve bodies necessitates complete disassembly of the valve.
The disassembly of the subject valves for the sole purpose of visual examination of the valve body internal surfaces is a major _ effort and requires many manhours from skilled maintenance and inspection personnel. In order to examine the internal surfaces of a valve body in accordance with the requirements, complete disassembly of the valve would be required which, in addition to the possibility of damage to the valve, would result in personnel receiving excessive radiation exposure. Therefore, the Codo requirement is-impractical. The visual examination is performed to determine if unanticipated severe degradation of the valve body is occurring due to phenomena such as erosion, corrosion, or cracking. However, previous experience during examination of similar valves at other plants has not shown any significant degradation of valve bodies. Imposition of the requirements on Commonwealth Edison l
25
Company would cause a burden that would not be compensated significantly by an increase in safety above that provided by the proposed examination.
Commonwealth Edison Company has stated that the Code-required visual examination will be performed on the internal pressure boundary surface of one valve in each of the groups of valves if maintenance activities require disassembly of the valve such that access for conducting the examination is provided.
Later editions and addenda of the ASME Code (1988 Addenda) have eliminated disassembly of valves for the sole purpose of performing examinations of the internal surfaces and state that the internal surface visual examination requirement is only applicable to valves that are disassembled for reasons such as maintenance, repair, or volumetric examination.
Therefore, the concept of visual examination of- the internal surfaces of the valve body, if the valve is disassembled for maintenance, is acceptable. Since no major problems have been reported in the industry with regard to valve bodies, the Licensee's proposal will provide adequate assurance of the continued inservice structural integrity.
Conclusions:
It is concluded that the disassembly of a valve for the sole purpose of inspection is impractical to perform at Braidwood, Units 1 and 2, and that public health and safety will not be endangered by allowing the proposed examination to be performed in lieu of the Code requirement. Therefore, it is recommended that relief be granted provided that, if the valves have not been disassembled, this fact should be reported by the Licensee in the ISI Summary Report at the end of the interval.
3.1.7 General (No relief requests) 26
-- 3.2 (Jass ? Components 3.2.1 Eressure Vessels ,
3.2.1.1 Recuest for Relief NR-8. Rev. 2. Examination Cateoory C A.
Items C1.10 and C1.20. Class ? Pressure Vessel Shell and Head Circumferential Welds UDII: Relief Request NR-8 was deleted from Revision 2 of the Braidwood, Units 1 and 2, First 10-Year Interval ISI Program Pl an. The information in Relief Request NR-8 was incorporated into Note 12 of Section 2.3, " Notes," of the ISI Program Plan.
Note 12 states that Braidwood will incorporate Code Case N-435-1, which allows a surtate examination for welds in vessels with nominal wall thickness of 1/5-inch or less, in lieu of volumetric examination. ASME Code Case N-435-1,
" Alternative Examination Requirements for Vessels with Wall Thickness 2 in, or less,Section XI, Division 1," is referenced in NRC Regulatory Guide 1.147, " Inservice Inspection Code Case Acceptability, ASME Section XI Division 1," Revision 7, as an NRC approved Code Case and, therefore, may be used. The affected welds are as follows:
UNIT 1 Comoonent Weld Number Material Thjckness ICVOSA ILRHXC 2 0.188" ICV 02F ISWRF-Cl 0.165" 1CV02F ISWRF-C2 0.165" 1CV03F 1RCFC-Cl 0.165" ICV 03F 1RCFC C2 0.165" UNIT 2
~
Comoonent Weld Number Material Thickness 2CV05A 2LRHXC-2 0.188" 2CV02F 2SWRF-C1 0.165" 2CV02F 2SWRF-C2 0.165" 2CV03F 2RCFC-Cl 0.165" 2CV03F 2RCFC-C2 0.165" 27
y t L3.2.1.2 Recuest for-Relief NR-10. Rev. 2. Examination Cateoory C-A. i item C1'.10. Pressure Retainino Welds in the letdown Heat '
Exchanoer-t
- Code Reauirement: Section XI. Table IWC-25001, Examination Category C-A,' Item C1.10 requires a.100% volumetric examination of the shell circumferential welds on Class 2 vessels as defined by Figure IWC 2500-1. Examinations may be limited to one vessel-in cases of multiple vessels of similar design, size, and. service, and shall be conducted each inspection
_ _ interval .
'Sub'section IWA, paragraph IWA-2232 requires use of Article 5 of-Section V for vessels less than 2 inches nominal wall thickness. Article 5 of Section V requires examination by: ,
(a) Straight beam for any defects that might affect interpretation of angle beam results;- ,
(b) Angle. beam scanning for reflectors oriented parallel to
.the_ weld,.from two directions where possible; o_
.(c) Angle beam scanning for reflectors oriented transverse to the weld. ,
licensee's Code Relief Reauest: EReliefisrequestedfrom
~
examining 100% of- the Code-required volume of weld number 1HLHXC-01;of Unit I letdown heat exchanger ICV 04AA and of weld 1 number 2HLHXC-01 of Unit 2 letdown-heat exchanger 2CVO_4AA.
Licensee's- Proposed Alternative Ex3mination: A liquid
_ penetrant surface examination-is-proposed in lieu of ultrasonic
. examination by the straight beam method and angle beam examination for reflectors transverse to the welds. The complete ultrasonic examination for! reflectors parallel to the-weld seam will- be performed. In the event the flange bolting is disassembled for maintenance, the straight beam and angle beam examination for-reflectors transverse to the weld will be performed.
28
4 I
Licensee's Basis for Reouestina Relief: The Licensee reports that the subject welds are obstructed by the fiange belting, thus preventing volumetric examination for reflectors transverse to the weld and from straight beam examination.
Evaluation: As shown in the sketch attached to the relief request, the flange bolting limits the Code-required volumetric examination of the subject welds. Therefore, the required volume cannot be examined to the extent required by the Code without disassembling the flange bolting. Imposition of the requirement on Commonwealth Edison Company would cause a burden that would not be compensated significantly by an increase in safety above that provided by the proposed alternative.
The Licensee has stated that a supplemental surface examination will be performed and that the volumetric examination will be performed to the maximum extent practical. The Licensee has committed to perform the remainder of the Code-required volumetric examination in the event the flange bolting is disassembled for maintenance. The limited volumetric examination and the proposed surface examination will provide adequate assurance that unallowable inservice -flaws have not developed in the letdown heat exchanger welds or that they will be detected and removed or repaired prior to the return of the letdown heat exchangers to service.
Conclusions:
It is concluded that=the volumetric examination of the subject wel? - is impractical to perform at Braidwood, Units 1 and 2. to .e extent required by Section XI of the ASME Code and that public health and safety will not be endangered by allowing the prooosed-alternative examination to be l performed in lieu of the Code requirement. Therefore, it is recommended that relief be granted as requested.
l 29
.- 3.2.1.3 Recuest for Relief NR-11. Rev. 2. Examination Cateaory C ,L_
Item C1.10. Pressure Retainino Welsis in thpjxcess Letdiwn Heat Exchanaer Code Recuirement: Section XI, Table IWC 2500-1, fxamination Category C-A, item C1.10 requires a 100% volumetric examinstion of the shell circumferential welds on Class 2 vessels as defined by Figure IWC-2500-1. Examinations may be limited to one vec el in cases of multiple vessels of similar desip, size, and service, and shall be conducted each inspection interval.
Subsection IWA, paragraph IWA-2232 requires use of Article S of
~
Section V for vessels less than 2 inches nominal waU
. thickness. Article 5 of Section V requires examnation by:
(a) Straight t,eam for any defects that might affect interpretation of angle bean results; (b) Angle beam scanning for reflectors oriented parallel to the weld, from two directions where possible; a
(c) Angle beam scanning for reflectors oriented transverse to
, the weld.
Licensee's Code Relief Reauest: Relief is requested from examining h 0% of the Code-required volume of weld number IELHXC-03 of Unit I excess letdown heat exchanger ICV 01AB and weld number 2CLHXC-03 of Unit 2 excess letdown heat exchanger 2CV01AA.
Licensee's Proposeo Alternative Examination: A liquid
~
penetrant surface examination is proposed to supplement the limited ultrasonic examination for reflectors oriented parallel to the weld. The complete ultrasonic examination for reflectors transverse to the weld seam will be performed. The surface examination is to be performed in conjunction with the ultrasonic examinations.
}
30
t
' <. Licensee's. Basis for Reauestina Relief: --The Licensee repor.ts that each ofLthe-subject welds is obstructed for approximately 70% of its length by four branch connections from one side and!
by a flange on the other.
Evaluation: As shown in the sketches attached t'o the relief-request, the volumetric examination of these welds is: limited due_ to a flange on one side and four branch' connection obstructions on the other side. Therefore, the volumetric
~
examination of the subject welds is ' impractical to perform to the extent required by the Code. In order.to examine the weids in accordance with the requirement, the excess letdown heat exchangers would have to be redesigned,_ fabricated,_and
- installed. _ Imposition of the requirement on Commonwealth-Edison _ Company would'cause'a burden that would-not be compensated.significantly by an increase in safety-above that provided by the proposed alternative.
- The Licensee has-stated that the volumetric examination will be.
- performed to the maximum extent practical and that a supplemental- surface examination will be performed. -Based on the vessel design, an acceptable percentage of the
- Code-required volume can'and will be examined. -Thus,'the limited-Section XI-volumetric examination and the proposed supplemental surface examination will provide reasonable assurance of the continued l inservice structural integrity. g
Conclusions:
- It -is concluded that = the' volumetric examination -
of the subject welds is impractical to perform at Braidwood,
- Units-1 and 2, to the _ extent required by Section XI:of the: ASME-
_ Code _and that.public health'and safety will:-not be endangered by allowing _the proposed alternative _ examination to be-performed'in lieu of the Code requirement. Therefore, it-is-
~
recommended that relief be granted as requested.
l 31
.i :
. 3.2.1.4 ' Reauest for Relief: NP-12. Rev. 3. - Examination Cateoory C-B.
-Items C2,21~and C2.22. Residual Heat' Removal 1 Heat Exchanaer f
~
Nozzle-to-Vessel Welds and Nozzle-Inside Radius Section_s 3
J Code Reautrement: 'SectionXI,_TableilWC-2500-1, Examination:
- Category C-8, Item C2.21 requires both-100% volumetric and
' - ' =
surface examinations of Class-2 nozzle-to-shell welds and item C2.22 requires a 100% volumetric examination of Class 2 vessel !
4 nozzle-inside radius' sections'as defineo by Figure. ,
IWC-2500-4(a)'or (b), for nozzles without reinforcing plate in vessels greater than 1/2 inch' nominal-thickness. Examinations ;
shall'be conducted on nozzles at terminal ends:of pipirg runs -
selected for examination under Examination Category C-F oach inspection interval.
Licensee's Codo Relief Reauest: Relief is requested from
-performing the Code-required volumetric examination'of the following residual heat -removal- heat exchanger nozzle-to-vessel-welds and nozzle'inside radius sections:
[.g_moonent Number - Item Numbers 1RH02AA (Unit 1) RHXN-01 RHXN-02:
1 2RH02AA (Unit 2) RHXN-01
~
RHXN-02 :
Licensee's Proposed Alternative-Examination: zThe subject welds 1 will receive th+ required Section XIL surface examinations. A VT-1 visual examination of the nozzle-inner radit will be performed either directly'or_ remotely to the extent practical when disassembly- is'_ required for maintenance. In addition, the
. Code-required VT-2 visual examination will be -performed each .
inspection period on all pressure-retaining _ components. A best
- effort ultrasonic examination will be performed on the RHR' heat exchanger nozzle-to-shell welds on a frequency _ consistent-with I ASME Section XI.
32
, - _ . _ . _ -_ _ . _ _ . _ - . _ __ ~ _ . -
- Licensee's Basis for Reouestino Relief: The subject nozzles contain inherent geometric constraints that limit the ability to perform meaningful ultrasonic examinations.
The residual heat removal heat exchanger is approximately 7/8 inch nominal wall thickness with 14 inch diameter nozzles approximately 3/8 inch nominal wall thickness. The configuration is best characterized as a fillet welded nozzle
~
using an internal reinforcement pad and, therefore, is not analogous to a full penetration butt welded nozzle as shown in Figure IWC-2500 4. In addition, the inner radius of the reinforcement pad would be representative of the nozzle inner radius required for inspection. The inherent geometric constraints of the nozzle design prevent the performance of the required ultrasonic examination of the nozzle inner radius. A nozzle-to-shell ultrasonic examination may not achieve full coverage of the weld area.
Evaluation: The sketch attached to the Licensee's relief request shows that the configuration is such that volumetric examination of the nozzle inner radius sections is impractical to perform. In order to examine the inner radius in accordance with the requirements, the nozzles, and thus the RHR heat exchangers, would have to be redesigned, fabricated, and installed. Imposition of the requirement on Commonwealth Edison Company would cause burden that would not be compensated significantly by an increase in safety. However, with regard to the nozzle-to-vessel welds, the sketch attached to the Licensee's relief request shows that the configuration is such that the required volumetric exaniination of the nozzle-to-vessel welds can be performed from the 0.0. nozzle side. Therefore, the Code-required volumetric examination of the RHR heat exchanger nozzle-to-vessel welds should be performed. Also, since these nozzle to-vessel welds correspond 33
~2 tofthereinforcingplate:weldsinFigureIWC-25004(c). surface .
examination of the 0.0. of the' nozzle-to vessel welds shohld also be performed._ ,
_[onclusions:1 It is concluded that the Code required volumetric examinations of the subject RHR heat exchanger nozzle'inside radius sections are impractical _to perform at Braidwood, L
Units 1 and 2,.due1to the nozzle design and that public health ,
and safety will not be endangered by allowing the proposed alternative examination to be performed in lieu.of the Code ,
requirements. It is also concluded that_the required volumetric examination of thu nozzle to vessel welds can be performed from the 0.D. nozzle side. Therefore, it is
. recommended that relief be granted for the nozzle inside radius sections and that relief be ' denied for the nozzle to-vessel
. welds.
3.2.1.5 Reauest for Relief NR-17. Rev. 3. Examination Cateaory C-A.
Items C1.10 and C1.20. Class 2 Pressure Vessel Shell and Hgid_ -
Circumferential Welds
, HQII: Request for Relief NR-17 was withdrawn by the Licensee
" in the August 15, 1990 response to the NRC requast for additional information. In that response,_the Licensee submitted Note 13, allowing for exemption of ASME Class 2 vessels from the ~ISI Program' based. on cumulative inlet < and cumulative outlet pipe cross-sectional areas not exceeding four inches-in diameter.
-3.2.2 Pioino 3.2.2.1 Reauest for Relief NR-1. Rev. 2. Examination Cateaory C-F. Item
[3 31. Class 2 Main Steam System Pine Branch Connection Welds Code _Beauirement: Section XI, Table IWF-2500-1, Examination Category C-F, Item C5.31 requires a 100% surface examination of 34
__ ., . ..,_ . , . _ . . . , , . . ,m., m,... _ , , _ . _ , . . . . . . . _ . , . , _ . ...m. _---. - .._J.. .
. the Class 2 circumferential welds in pipe branch connections greater than 4 inches nominal branch pipe size as defined by figures IWC-2500-9 to -13, inclusive. ,
tjsinige's Code Relief Reauest: Relief is requested from performing the Code-required surface examination of the following Citis 2 main steam pipe branch connection circumferential 9 elds:
line Norber_..__. Weld Numbers IMS07AA 28" (Unit 1) IMS-04 25. 26, -27. -28, -29 IMS07AB-28" (Unit 1) *1MS 06-43, -44, -45, -46, -47 IMS07AC-28" (Unit 1) *1MS 08-25, -26, -27. 29 IMS07AD-28" (Unit 1) *1MS-02-37, 38, -39, -40, -41 2MS07AA-28" (Unit 2) 2MS-04-34, -37, -40 43, -46 2MS07AB 28" (Unit 2) *2MS-06-34, .37, -40, -43, -46 2MS07AC-28" (Unit 2) *2MS 08-31, -34, -37, -40, -43 2MS07AD-28" (Unit 2) *2MS-02 36, -38 -42, -44, -47
- These welds are not selected for inspectita as required by the Code. However, they are examined within the scope of auomented high energy lines as described in Note 5 of the ISI Program Plan.
Licensee's Prooosed Alternative Examination: None. A magnetic particle surface examination and visual examination (leak test) will be performed on the saddle plate fillet welds in lieu of the required surface examinations for those pressure retaining welds selected for examination, Licensee's Basis for Recuestina Relief: The subject welds are inaccessible to surface examinations due to saddle plates over the pressure retaining welds.
Evaluation: The Code-required. surface examination of the subject branch connection welds is impractical to perform as the welds are located beneath saddle plates. In order to examine the welds in accordance with the requirement, the connections would have to be redesigned, fabricated, and 35
installed. Imposition of the regt. rement on Commonwealth Edison Company would cause ; burden that would not be compensated significantly by an increase in safety above that provided by the proposed alternative examination.
The surface and visual examinations of the saddle plate fillet
- welds meet the intent of the Code and, therefore, provide reasonable assurance of the continued inservice structural integrity.
[pnclusions: It is concluded that the Code-required surface examination of the subject welds is impractical to perform at Braidwood, Units 1 and 2, and that public health and safety will not be endangered by allowing the proposed alternative examination to be performed in lieu of the Code requirement.
Therefore, it is recommended that relief be granted as requested.
3.2.2.2 Reauest for Rolief NR-16. Rev. 2.-Examination Cateaory C-F.
Item C5.21. Class 2 Main Steam and Feedwater Systyp_.
Circumferential Pioina Welds (Unit 1 only)
Code Reauirement: Section XI, Table IWC-2500-1. Examination Category C-F, Item C5.21. requires both 100% volumetric and surface examinations of Class 2 circumferential piping welds-greater than 1/2 inch nominal wall- thickness as defined by Figure IWC-2500-1.
Licensec's Code Reljlf Recuelt.: Relief is requested from examining 100% of the Code-required volume of the following ,
main steam and feedwater system circumferential piping welas:
(.
36 l
4
-- Percent Not Reason for Limited Linejhtmbn Examination ifWS6AD 16" Weld Number Examinable 1FW-01-01 16%
. Branch connection, valve geometry IFW86AA-16" 1rW 02-01 16% Branch connection, valve geometry
-1FWO3DC-16" 1FW-04-29 25% Elbow inner radius, valve geometry IMS01AD 32" 1MS 01-01 8% Nozzle, gamma plug IMS01BD-30.25" 1MS-02-07 16% Tee, weldolets
. IMS01BA-30.25" 1MS 04 07 26% Weldolots, tee geometry Ligensee's Proo,qsed Alternative Examination: None. The subject welds will receive the required Section XI surface examination in addition to a "best effort" ultrasonic examin& tion, licensee's Basis for Reauestina Relief: The sebject welds have been selected for examination during the first inspection interval. They have interfering conditions or each side of the weld. These interferences can cause poor coupling of the transducer, limited movement of the transducer, redirecting of the sound beam, and, in some cases, complete restriction of a portion of a particular scan. These conditions limit the axial scans so as to leave the listed percent of the weld volume uninspected.
Evaluation: The volumetric examination of the subject circumferential piping welds'is impractical to perform to the extent required by the Code due to the fitting configurations (listed above) adjacent to the weld. In order to examine the welds in accordance with the requirements, these
< components / fittings would have to be redesigned, fabricated, and installed. Imposition of the requirements on Commonwealth Edison Company would cause a burden that would not be compensated significantly by an increase in safety above that provided by the limited examination.
l 37
_ ]
4 4- :The Licensee-has stated Ithat the Code-required surface -
examination will be performed and that the volumetric :
. examinations will be performed to the maximum extent-practica1'., A significant percentage- (74-to 92%) of the Code-required volume can and will be' examined. Thus, the~
1
- limited volumetric examination and Code-required surface-examination-of the subject welds will provide reasonable assurance of;the continued inservice structural integrity.
Conclusions:
L It is concluded that the volumetric examination -
of Une subject piping welds is impractical to perform at Braidwood; Unit 1, to the extent required by Section XI.of the ASME Code and that public health and safety will not be .
Li endangered by-allowing the limited examination to be performed in-lieu of the Code requirement. . Therefore, it is recommended that relief be granted as requested.
3.2.3 Pumps
-3.2.3.1' Reauest for Relief NR-15. Rev. 2. Examination Cateaorv C-C.
~
Item C3.30. Intearal Welded Attachments on Class 2 Pumni :
. Code Recuirement: Section XI, Table-IWC-2500-1, Examination .
Category C-C, Item C3.30 requires a:100%. surface examination of
-integrally ~ welded attachments.on Class 2 pumps -as defined by -
Figure !WC-2500-5. Examinations shall be conducted on components required to be. examined under Examination Categories C-F'and C-G each inspection interval t
Licensee's Code Relief Recuest: Relief is requested from-examining 100% of the Code-required surface of the following Class 2 pump integrally welded attachments:
i l'
I 38
1 i
-. Component Number Weld Numbers CVC Centrifugal Charging CVP-01, CVP-02, CVP-03, CVP 04 Pump ICV 0lPA (Unit 1)
CVC Centrifugal Charging CVP-01, CVP-02, CVP-03, CVP 04 Pump 2CV0lPA (Unit 2)
Residual Heat Removal Pump RHP-01, RHP-02, RHP-03 IRH0lPA (Unit 1)
Residual Heat Removal Pump RHP-01, RHP-02, RHP-03
. 2RH0lPA (Unit 2)
Licensee's Proposed Alternat.ive Examinatip_q: A VT-1 visual examination will be performed on the inaccessible side of the integrally welded attachments when the surface examination is performed on the remaining portion of the attachments. In addition, the Code-required system leakage test will be performed each inspection period.
Licensee's Basis for Reauestino Relief: The subject welds connect the support lugs to the pump casings. These integrally welded attachments cannot be examined on one of the required sides due to the location of structural supports. The welds are located between the pump casing and the structural supports for the pumps. These welds can only be examined in the areas that allow sufficient clearance.
The 1.icensee reports that these welds are not full penetration welds and thus the proposed VT-1 visual examination, in conjunction with the surface examination, will identify cueloping defects.
Evaluation: The sketches attached to the Licensee's relief request show the location of the subject welds and that portions of the welds are inaccessible for surface examination. Therefore, the surface examination of the subject welds is impractical to perform to the extent required by the Code. In order to examine the welds in accordance with the requirement, the pumps and supports would require extensive 39
- - - . . -. - . . ~ . -- .- .-- . _. - - ,- .
- design modifications. Imposition of the requirement on m
~
!< Commonwealth Edison. Company _ would cause a burden that would not be compensated significantly by an increase-in safety above- .
that-provided:by the' proposed examination, l
ji ~ - The Licensee has stated that the surface examination will be -
- performed to' the maximum extent practical and that a VT-1 visual examination will be performed on the areas inaccessible for surface examination in conjunction with the-surface
[
- examination of the accessible areas. -- As shown .in the sketches ;
attached to.the: relief request, a significant percentage of the-Code-required surface examination will' be performed. Thus, the -
]
- limited. surface examination, in conjunction with-the proposed
- VT-1 visual-. examination,- will provide reasonable assurance of
' the continued inservice structural integrity.
[ fanclusions: _ = It is concluded that- the surface examination of -
the/ subject integrally welded-ettachments is impractical to
! perform at:Braidwood, Units 1 knd 2, to the ' extent _ required.by I Section-XI of- the ASME Code and that public health-and safety J will not.be endangered by allowing the proposed alternativo L.
- examination to
- te performed in lieu of the Code' requirement. .
I- Therefore,'it is-recommended that relief bc granted as
! requested. ,
i' 3.2;4 " Valves (No relief requests)
!- 3.2.5. General - (No relief. requests) s - 3.3- Class 3-Comoonents (No relief requests)
^
? 3,4- Pressure Tests- (No relief requests)
I i-r
- - 40 t
t y--,.n ,, -
e ., , ,m..
. 3.5 General 3.5.1 Ultrasonic Examination Technioues (No relief requests) 3.5.2 Exemoted comoonents (Noreliefrequests) 3.5.3 Other 3.5.3.1 Reauest for Relief CR-I _Rev. 2. Subarticle IWF-1300. Support t
Examination Boundari's for Honexemot Comoonent Supoorts on Jnsulated lines .
1 8Q11: Request for Relief CR-1 was withdrawn by the Licensee in the August 15, 1990 response to the NRC request for additienal information.
3.5.3.2 Recuest for Relief CR-2. Rev 1. Subarticig_IWF-2430.
Additio2al Examinations of Comoonent Suocorts Code Reouirement: Section XI, Subarticle IWF-2430 states, in part, "When the results of examinations require corrective measures in accordance with the provisions of IWF 3000, the component supports immediately adjacent to those requiring l- corrective action shall be examined."
c.icensee's Code Relief Requggi: Relief is requested from_ the Code-required examination of the component supports immediately-l adjacent to those requiring correct!<e action if the adjacent components are not of'the same type or susceptible to the same
[
!~ type of indication or mode of failure.
~
Licenseg's,Erecosed Alternative Examination: A visual examination will be performed on component supports required to be examined by Section XI. When corrective actions are required as a result of these examinations not satisfying the i
L 41 l
4-
-. requirements of IWF-3400, the adjacent supports will be examined if they are generically susceptible to the same type of indication or mode of failure, regardless of support type.
Licensee's Basit_for Reouestinn Religf: The additional exarninations required by failure to satisfy the acceptance standards of IWF-3400 are not limited by Section XI to the type of Indication identified, the type of support being examined, or the types of adjacent supports. This would require that a documented visual examination be performed which, in some cases, would provide no meaningful results. This can cause
. undue hardships and increase ALARA concerns without significantly affecting overall plant safety.
Evaluation: It has been determined that it is irrpractical to examine an adjacent r.omponent support when the indication identified is liraited to a particular type of component support and the adjacent component supports are not of that type. The Licensee has stated that if the indication requiring corrective measurcs can be generically applied to the adjacent component support (s), the component support (s) will be examined. Since additional examinations are required and performed in accordance with IWF-2430 and the component :upports are re-examined per the requirements of IWF-2420 during the next inspection period, this provides a high degree af system reliability and safety while providing ALARA benefits to plant employees. The Licensee has cemonstraed that the proposed alternative will provide an acceptable level of quality and safety and that compliance with the Code requirement woulo result in hardship or unusual difficulties without a compensating increase in safety. .
~
Conclusions:
It is concluded that the proposed alternative is acceptable and will not endanger public health and safety.
Therefore, pursuant to 10 CFR 50.55a(a)(3), it is recommended that relief be granted as requested.
42
. 3.5.3.3 Recuest for Rolitf SR-L _ Rev. 2. Subarticle IWF-1300. Suonort Ey_ amination Boundaries for Nonexemot Safety Related Snubbers on Insulated Components Code Recuirement: Section XI, Table IWF-2500-1 requires VT-3 and VT-4 visual examinations of nonintegral component supports as defined by Subarticles IWF-1300 and IWF-2510 and Figure IWF-1300-1. Component support examination boundaries are defined by Subarticle IWF-1300 and Figure IWF-1300-1, which identify the boundary of a nonintegral support to the pressure-retaining component to be the contact surface between the component and the support.
Licensee's Code Relief Recuest: Relief is requested from removing the insulation from all nonexempt safety-related snubbers on Code Class 1 and 2 insulated lines for the sole purpose of performing a visual examination on the portion of the nonintegral attachment within the insulation.
The following table shows the number of snubbers on insulated lines vs. the total number of snubbers in the population. The numbers will vary with time as a result of snubber reduction and other plant raodifications.
Snubbers Attached to insulated Ptoe 21 stem Unit 1 Unit 2 AB 2 1 AF 1 0 CS 0 9 CV 82 91 00 ~0 10 FW 4 42 HS 12 12 RC 93 80 ;
RH 24 28 RY 31 26
. SD 15 21 Total enubbers on 264 320 insulated lines Total snubber 369 384 population 43 I
1
Licensee's Proposed Alternative Examination: The Licensee states that, in lieu of removing insulation on snubber pipe clamps, the following alternative examination methods will be ,
I employed on all' snubbers that are accessible for direct examination: !
(a) A hands-on inspection of the pipe clamp will be performed to verify that the clamp is tight. ,
(b) Clamp alignment with the load pin axis will be observed to verify alignment is within design tolerances.
(c) The load pin / stud will be inspected to verify its integrity. This will ensure that parts are in place and that the pin is tight. If the load pin is obscured by insulation, the insulation will be removed or modified to allow for this inspection.
(d) Insulation will be checked for evidence of damage due to slipped or loose clamps.
(e) If boric acid contamination or corrosion is observed, insulation will be removed to inspect the pipe clamp.
Ucensee's Bas _is_for Reauestinc Relief: The Licensee states that the visual examination of nonintegral pipe attachments is limited by the insulation installed on the piping. It would impose a great deal of hardship in terms of manpower, time, and radiation exposure to remove insulation to vin ally inspect all snubber pipe clamps, particularly if there are alternative methods that provide an equivalent means of determining pipe clamp integrity.
The majority of snubbers are located inside containment in high radiation areas. Removing insulation on all snubber pipe clamps would require one Health Physics Technician to survey l
44 ,
l
- the insulation prior to its removtl and then a two-man insulator. crew to remove the insulation. This would add three people to the customary two-man inspection crew, which would more than double the man-rem exposures for performing the-surveillance.
" The Ceco SPPM VT 3/4 procedure allows remote inspections to be performed on snubbers that are out of reach for direct inspection. Scaf' folds or man baskets would have to be used to remove insulation on temote snubbers. Extra scaffolds built for snubber pipe clamp inrulation removal would increase congestion in containment and increase the amount of material being handled and surveyed during the outage. It would also product: additional dry active waste (DAW) and result in more scaffold material acquiring fixed contamination during the outage. It would pose an additional burden on the examination in terms of manpower, time, safety, and radiation' exposure. It also defeats the purpose of the remote examination methods allowed in the SPPM.
This relief request is intended for nonintegral attachments on insulated lines. The visual inspections of snubbers are performed using the CECO SPPM VT-3/4 procedure. The inspections are performed on all safety related snubbers every 18 months. Under this procedure, support indications to be observed and documented include the following:
- cracks, pitting
- erosion, corrosion, wear
- loose, missing, damaged parts
- contamination, debris
- weld degradation
- slipped _ clamps ,
- arc strikes, weld spatter, paint
- clearances, settings
- condition of spherical bearings The proposed alternative examination methods listed enhance tne SPPM VT-3/4 inspection procedure. The hands-on check combined 45
~. - . - ~ . -. . - .
m .
li( r v with;the'_VT-3/4 procedure will ensure that snubber pipe-clamps are installed and secure 'on snubbers that can be reached for-
! direct examination. If there are iny indications 'of- .
degradation -the insulation will a removed to allow for a toti clamp inspection. .
i
- For the snubbers that have pipe clamps completely buried in )
insulation, the insulation will be removed for a-complete 1 inspection. -Based onia review of the piping line list and ,
snubber data bases,-this condition occurs primarily on PSA 1/4 ,
~
snubbers' attached to 1 inch or less piping covered with 2 inch ,
thick or greater insulation. Insulation removal to inspect these-pipe clamps will be accomplished several ways. When ;
these' snubbers are removed for functional testing,_the
-insulation will have to be removed to unpin the snubber. The visual inspection of the pipe clamp will be performed at this time. Visual examinations on pipe clamps 'will be coordinated with NDE' inspections during which insulation removed to access welds also exposes the pipe clamp. Snubber examinations are-documented,. so it can be verified that all snubber pipe clamps completely. buried in insulation receive a visual inspection within-the ten-year interval.
-On' snubbers that are inaccessib1'e for direct examination, a remote examination will be performed per the SPPM VT-3/4 procedure. This would include snubbers in heat exchanger pits 3r on overhead runs of-piping that are out of reach. The number of snubbers in this category represents only about 5% of the snubbers in Units 1 and:2. None of these snubbers haven pipe clamps completely buriert in insulation. .They are not in
'high traffic creas where, typically, the most cicmp slippage and_ other daruge is experienced. Boric acid spray from valves or flanges _is also less likely on the majority of these
' snubbers. Because they are in relatively safe areas, an indirect inspection verifying that no outward indications are present will demonstrate that the pipe clamps are secure. When 46 !
i
., - . ~ v,, %_ . -
these snubbers are functionally tested, scaffolds will be built to access them. A hands-on examination of the pipe clamp will be performed at that time. If any of the above conditions are present, a scaffold will be built to more thoroughly investigate-the indication.
Previous inspection experience has not shown an increasing trend in regard to loose pipe clamps. A review of earlier data indicates that loose pipe clamps are rare. The alternative methods proposed in this relief request, combined with the commitment to remove insulation on those pipe clamps completely obscured by insulation, will provide a complete examination of the entire snubber population in Units 1 and 2. This approach meets the requirements of an alternctive inspection and will provide a high degree of confidence that the snubber pipe clamps are in place and secure. By 10miting the number of people required to perform the surveillance, the proposed methods will minimize man-rem exposures. The need for additional scaffolding is also elimineted, which will lower the amount of contaminated material produced during the outege, and reduce the traffic and congestion associated with moving scaffolding in and out of containment. This will promote the efficient and cost effective execution of the visual surveillances.
EvaluatLgn: The visual examination of the subject nonintegral attachments of nonexempt safety related snubbers is impractical to perform to the extent required by the Code because they are covered or partially covered by insulation. Removal cf the insulation for the sole purpose of inspe; tion would result i's personnel receiving excessive radiation exposure. Imposition of the requirements on Commonwealth Edison Company would cause a burden that would not be compensated significantly by an increase in safety above that provided by the alternative.
47 a ^
l
.- Damage to the nonintegral attachments beneath the insulation should be evident by misalignment between the snubber and its nonintegral attachment, crushed insulation, dented insulation, boric acid on the exterior of the insulation, nonintegral i attachment discrepancies on adjacent supports, or.other abnormal conditions. Also, the scope and frequency of snubber
' examinations performed in accordance with the Technical Specifications exceed the Code required scope and frequency.
The Lice.osee's proposed alternative examination will provide reasonable assurance of the continued inservice structural integrity.
Conclusions:
It is concluded that compliance with the Code requirement is impractical at Braidwood, Units 1 and 2, and that public health and safety will not be endangered by allowing the proposed alternative examination to be performed in lieu of the Code requirement. Therefore, it is recommended that relief be granted as requested. .
3.5.3.4 Reauest for Relief SR.2. Rev. 1. Subarticle IWF-2430.
Additional Examinatienj of Safety Reh led !aubbers Lode Reouirement: .iection.XI, Subarticle IWF-2430 states in part, "When the results of. examinations require corrective measures in accordance with the provisions of IWF-3000, the component supports immediately adjacent to those requiring corrective action shall be examined."
LJcensee's Code Relief Reauest: Relief is requested from tha
- Code-rec,uired examination of the component supports immediately adjacent to those requiring corrective action if the adjacent component supports are not susceptible to the same type of indication or mode of failure, regardless of support type.
- 48
. Licensee's Proposed Alternatiyf_fJLmiration: A visual examination is performed on all safety related snubbers. as required by the Braidwood Station Technical Specificat ons.
When corrective actions are required as a result of these examinations not satisfying the requirements of IWF 3400, the adjacent component supports will be examined if they are
' susceptible to thu same type of indication or mode of failure regardless of support type.
Licensee's Bas,is fer Reouestino Ibligi: The cdditional examinations required by failure to satisfy the acceptance standards of IWF 3400 are not ilmited by Section XI to the type of indication identified, the type of component support being examined, or the types of adjacent component supports. This ,
would require that a documented visual examination be oerformed that, in some cases, would provide no meaningful results. The Licensee states that this can cause undue hardships and increase ALARA concerns without significantly affecting overall plant safety.
Evaluation: It has been determined that it is impractical to examine an adjacent component support whM th2 indication identified is limited to a particular type of compencnt support ,
and the adjacent component supports tre not of that type. The Licensee has stated that, if the indication requiring corrective measures can be generically applied to the adjacent component support (s), the component support (s) will be examined. Since the M pection frequency is increased based on the number of failures identified, this approach to performing additional examinations, along with the additional examinations
' required by.lWF-2430 for nonexempt component support (s),
provides a high degree of system reliability and sefety while providing ALARA benefits to plant employees. The Licensee has demonstrated that the proposed alternative will provide an acceptable level of quality and safety and that compliance with 49
the Code requirement would result tri hardship or unusuti difficulties without a compenssting increase in safety.
,Conclusigns: It is concluded that the proposed alternative is acceptable and will not endanger public health and safety.
Therefore, pursuant to 10 CFR 50.55a(a)(3), it is reccmmended
' that relief be granted as requested.
/
t
[
t 4
50 i
. 4. CONCLUSION Puesuant to 10 CFR 50.55a(g)(6) or, alternatively, 10 CFR 50.55a(a)(3), it hw been determined that certain Section XI required inservice examinations cannot be performed to the extent required by the Code, in all cases except RequestforReliefNo.NR-12(inpart),theLicenseehasdemonstratedthat a spe<:ific Section XI requirements are impractical or that alternative extainations should be perfortred. For Request for Rolief No. NR 12 (in p et), it is concluded that the Licensee has not provided information to support the determination that the Code requirement is impractical for the RhD heat txchanger nozzle to vessel weld; requiring th Licensee to comply with the Code requirement would not result in hardshi, Requests for Relief <
NR.4, NR 8, NR-13. WR 17. kR 19, and CR 1 were withdrawn by the Licensee as a result of the NRC request for additional information.
This technical evaluation has not identified any practical method by which the Licensee can meet all the soecific inservice inspection requirements of Section XI of the ASME Code for the Braidwood Nuclear Power Station, Units 1 and 2, facilities. Compliance with all the exact Section XI required inspections would necessitate redesign of a significant number of plant systems, sufficient replacement components to be obtained, installation of the new components, and a baseline examination of these components. Even after the redesign efforts, complete compliance with the Section XI examination requirements probably could not be achieved. Therefore, it is concluded that the public interest is not served by imposing certain provisions of Section XI of the ASME Code that have been determined to be impractical. Pursuant to 10 CFR 50.55a(g)(6), relief is allowed from these requirements that are impractical to implement, or alternatively, pursuant i to 10 W R 50.55a(a)(3), alternatives to the Code-required examinations may be granted previded that either (1) the proposed alternat!ve provides an
- acceptable level of quality and cafety or that (ii) Code compliance would result in hardship or unusual difficulty without a compensating increase in safety. Relief may be granted only if granting the relief will not endanger life or property or the common defense and security and is otherwise in the l public interest giving due consideratinn to the burden upon the licensee l that could result if the requirements were imposed on the facility.
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i , The development of new or isnproved examination techniques should continue to be monitored. As improvements in these areas are achieved, the Licensee should incorporate these techniques in the 151 program plan examination requirements.
Based on the review of the Braidwood tiuclear Power Station, Units 1 and 2
- First 10 Year Interval liiservice inspection Program Plan, through Revision 4, the Licensee's responses to the liRC'; request for additional information, and the recommendations for granting relief from the 151 examiration requirements that have been determined to be impractica'i, it is concluded that the Braidwood liuclear Power $tation, Units 1 and 2, First 10 Year Interval inservice inspection Program Plan, through Revision 4, with the exception of Request for Rollef fio, f4R 12 (in part), is acceptable and in compliance with 10 CFR 50.55a(g)(4).
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, 5. REFERENCES
- l. Code of federal Regulations, Volume 10, Part 50,
- 2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1:
1974 Edition through Summer 1975 Addenda i 1983 Edition through Summer 1983 Addenda
- 3. Braidwood Nuclear Power Station, Units 1 and 2, first 10 Yect Interval Inservice inspection Program Plan, Revision 2, submitted December 15, 1988.
- 4. NUREG 0800 Standard Review Plans Section 5.2.4,
- Reactor Coolant and Section 6.6, "Inservico Doundary inservice Inspection and Testing,"ly inspection of Class 2 and 3 Components," Ju 1981.
- 5. Letter, dated May 21, 1990, S. P. Sands (NRC) to T. J. Kovach (Commonwealth Edison Company (Ceco)), request for additional information required to complete review of the Braidwood Nuclear Power Station, Units 1 and 2, First 10 Year Interval inservice inspection Program Plan, Revision 2.
- 6. Letter, dated August 15, 1990, S. C. Hunsader (Ceco) to T. E. Murley (NRC), additional information and Revision 3 pages of the Braidwood Nuclear Power Station, Units 1 and 2, inservice Inspection Program Plan.
- 7. Letter, dated December 13, 1990, A. R. Checca (Ceco) to T. E. Murley (NRC), additional information and Revision 4 pages of the Braidwood Nuclear Power Station, Units 1 and 2, Inservice Inspection Program Plan.
- 8. NRC Regulatory Guide 1.14. " Reactor Coolant Pump fl, wheel Integrity,"
Revision 1, dated August 1975.
- 9. N;C Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," Revision 1, dated July 1975.
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- s. ra 66 aso sam 6 EGG MS 9201 Technical Evaluation Report on the first 10 Year Interval inservice Inspection Program Plan: a cats aiao" ws.io Comroonwealth Edison Company, ~ ' - oa I
Braidwood Nuclear Power Station, Units 1 and 2, April 1991 a "* ca ca*Nt %m a Docket Numbers 50 456 and 50 457 r1N 06022 (Pro.i. 5) g 6 AWT*'ca t&' 6 f tPt of RiPoki a
Technical B.W. Brown, J.D. Mudlin t '* a co cov i a t o --- o-r ery,ea . . 6 .,. < u. . . .. . ,,c. ., . ..,,,,., . ,
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P. O. Box 1625 Idaho Falls, ID 83415 2209 w ..e,. s. .,. y . ,,c, gog g oamaaim~-e.au.a~oiooaissu,..c.. w . ., . . ,
Materials and Chemical Engineering Branch '
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Warhington, D.C. 20555
- 10. SUPPLBMINTA4Y hoTE$
11 A6&iP ACT stw ee.or eas This report presents the results of the evaluation of the Braidwood Nuclear Power ,
Station, Units l'and 2, First 10 Year Interval Inservice Inspection (151) Progrsm !
- Plan, through Revision 4, including the requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI t requirements that the Licensee has determined to be impractical. The Braidwood Nuclear Power Station, Units 1 and 2, First 10-Year Interval ISI Program Plan is evaluated in Section 2 of this report for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample,
-(c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during the Nuclear Regulatory Commission (NRC) review before granting an operating license.
The requests for relief from the ASME Code requirements that the Licensee has determined to be impractical for the first 10 year inspection interval are evaluated in Section 3 of-this report.
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