ML20235S883
| ML20235S883 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 08/31/1987 |
| From: | Beth Brown, Mudlin J EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20235S866 | List: |
| References | |
| CON-FIN-D-6022 EGG-SD-7782, NUDOCS 8710090174 | |
| Download: ML20235S883 (47) | |
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TECHNICAL EVALUATION REPORT ON THE I
FIRST 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN:
DETROIT EDISON COMPANY, ENRICO FERMI ATOMIC POWEA PLANT, UNIT 2, <
DOCKET NUMBiiR 50-341 B. W. Brown i
J. D. Mudlin 1
Published August 1987 i
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Idaho National Engineering Laboratory J EG&G Idaho, Inc.
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Prepared for: I l U.S. Nuclear Regulatory Commission (
. Washington, D.C. 20555 under l
DOE Contract No. DE-AC07-761D01570 l ,
FIN No. 06022 (Project 5) ;
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, ABSTRACT This report presents the results of the evaluation of the Enrico Fermi j Atomic Power Plant,1: nit 2, First 10-Year Interval Inservice Inspection
, (ISI) Program Plan through Reviiion 0, Change 2, submitted April 2,1987, !
a including the requests for. relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements j which the Licensee has determined to be impractical. The Enrico Fermi
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Atomic Power Plant, Unit 2, First 10-Year Interval ISI Program Plan is {
evaluated in Section 2 of this report. The ISI Program Plan is evaluated i' for (a) compliance with the appropriate edition / addenda of Section XI,
)' (b) acceptability of examination sample, (c) exclusion criteria, and i
) (d) compliance with ISI-related commitments identified during the Nuclear Regulatory Commissian (NRC) review befcre granting an Operating License.
The requests for relief from the ASME Code requirements which the Licensee j has dettirmined to be impractical for the first 10-year inspection interval are evaluated in Section 3 of this report.
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This work was funded under:
I U.S. Nuclear Regulatory Commission FIN No. D6022, Project 5 l ;
Operating Reactor Licensing Issues Program, j ,
Review of ISI for ASME Code Class 1, 2, and 3 Components e
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SUMMARY
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a The Licensee, Detroit Edison C mpany, has prepared the Enrico Fermi Atomic Power Plant, Unit 2, First .10-Year Interval Inservice ~ Inspection (ISI)'
Program Plan, Revision 0, Change 2, to' meet the requirements of the 1960 Edition, Winter 1981 Addenda (80W81) of the'ASME Code.Section XI except that. i the extent of examination for Code Class 2 piping welds has been determined !
by the 1974 Edition through Summer 1975 Addenda (74575). y
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The information in the Enrico Fermi Atomic Pcuer Plant, Unit 2, First ' , ;
4 10-Year. Interval ISI Program Plan, through Revision 0, Change,2, submitted j April 2,1987,'was reviewed, including.the requests for relie'f from the ASME q
Code Section XI requirements which.the Licensee has determined to be' ~ l impractical. As a result of this review,.a Request for Additional !
Information (RAI) was prepared describing the information and/or clarification required from the Licensee in order to complete the review. i i
Based on the review of the Enrico Fermi Atomic Power Plant, Unit 2, First'
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10-Year Interval ISI Program Plan,' Revision 0, Change 2, the Licensee's i i
response to the NRC's RAI, and the recommendations for the granting of '
relief from the ISI examination requirements that have been determined to be impractical, it has been concluded that the Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval ISI Program Plan, Revision 0, Change 2, with ,
the exception of Request for Relief RR-C1, is arr.eptable and in compliance: l with 10 CFR 50.55a(g)(4).
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.. CONTENTS ;
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ABSTRACT.......................................... .
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SUMMARY
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INTR 000CTION.......................................................... 1-i 1
- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN....................... 3 1 1
2.1 Docume nt s Ev al u ated . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 !
2.2 Compliance with Code Requirements.................................. 3 l l 2.2.1. Compliance with Applicable Code Editions....................... 3
, 2.2.2 Acceptability of the Examination Sample. . . . . . . . . . . . . . . . . . . . . . . . 4
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i 2.2.3 Excli u s i o n C r i t e r i a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.0.4 Augmented Examination Commitments.............................. 4 1 2.3 r'
Conclusions......................................................... 5 i 1
- 3. EVALUATION OF RELIEF REQUESTS.........................................
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1.1 C l a s s 1 C omp on e n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.1.1 Reactor Pressure Vessel.......................................... 6 i 3.1.1.1 Request for Relief RR-A1, Examination Category B-A, 1
i Items 81.11, Bl.12, Bl.22, and B1.30, Pressure !
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Retaining Shell and Head Welds and Shell-to-Flan j Weld in the Reactor Pressure Vessel . . . . . . . . . . . . .ge ........... 6 i t
e 3.1.1.2 Request for Relief RR-A2, Examination Category B-A, i
Items Bl.21 and Bl.22, Pressure Retaining Circumferential J and Meridional Lower Head Weld in the Reactor Pressure Vessel........................s............................. 8 3.1.1.5 ' Request for Relief RR-A6, Examination Category B-D, l Item B3.90, Nozzle-to-Vessel Welds in the
, Pressure Vessel........................... Reactor ................ 10 4 3.1.1.4 Request for Relief RR-A7, Examination Category B-E, Items B4.12 and B4.13, Partial Penetration Welds in' !
Control Rod Drive Nozzles and Instrumentation Nozzles on the Reactor Pressure Vessel............................ 11 3.1.2 Pressurizer (Does not apply to BWRs) 3.1.3 Heat Exchangers and Steam Generators (No relief requests) i 1
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, 3.1.4 Piping Pressure Boundary...................................... 12 3.1.4.1 Request for Relief RR-A3, Examination Category B-J, Item B9.11, Pressure Retaining Circumferential Welds i n Cl a s s 1 Pi p i ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 l 3.1.4.2 Request for Relief RR-A8, Examination Category B-K-1, L Item B10.10, Inte l Piping...........grally Welded Attachments for Class 1
......................................... 14 3.1.4.3 Request for Relief RR-A9, Examination Category B-K-1, Item B10.10, Integrall Piping................y Welded Attachments for Class 1
....................................16-3.1.4.4 Request for Relief RR-A10. Examination Category B-K-1, Item B10.10, Inte Piping. . . . . . . . . . .grally Welded Attachments for Class 1
......................................... 17 3.1.5 Pump Pressure Boundary........................................ 19 3.1.5.1 Request for Relief RR-A4, Examination Category B-L-2, Item B12.20, Class 1 Pump Casings......................... 19 3.1.5.2 Request for Relief RR-All, Examination Category B-K-1, Item B10.20, Inte Pumps. . . . . . . . . . . .grally Welded Attachments for Class 1
......................................... 21 3.1.6 Valve Pressure Boundary....................................... 23 3.1.6.1 Request for Relief RR-A5, Examination Category B-M-2, Item B12. 50, Cl ass 1 Val ve Bodies. . . . . . . . . . . . . . . . . . . . . . . . . 23 3.1.7 General (No relief requests) l 3.2 Class 2 Components................................................ 25 3.2.1 Pressure Vessels..................
........................... 25 3.2.1.1 Request for Relief RR-A14, Examination Category C-C, Item C3.10, Integrall Pressure Vessels.....y Welded Attachments for Class 2
..................................... 25 3.2.2 Piping........................................................ 26 3.2.2.1 Request for Relief RR-A12, Examination Category C-C, Item C3.20, Inte Piping. . . . . . . . . .grally Welded Attachments for Class 2
.......................................... 26 3.2.2.2 Request for Relief RR-A13, Examination Category C-C, Item C3.20, Inte Piping..........grally Welded Attachments for Class 2 ,
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3.2.3 Pumps (No relief requests) v
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, 3.2.4 Valves (No relief requests) 3.2.5 General (No relief requests) I 3.3 Class 3 Components (No relief requests) i 3.4 Pressure Tests (No relief requests) I 3.5 General........................................................... 29 3.5.1 Ultrasonic Examination Techniques (No relief requests)
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3.5.2 Exempted Components (No relief requests)'
i 3.5.3 0ther......................................................... 29 l l
3.5.3.1 Request for Relief RR-B2, Class 1 2, and 3 Com Supports.........................,.............ponent
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3.5.3.2 Request for Relief RR-C1, Class 1, 2, and 3 Snubbers...... 31 l 3.5.3.3 Request for Relief RR-C2, Class 1, 2, and 3 Snubbers...... 34 i 4.
1 CONCLUS10N........................................................... 37
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l 5. 1 REFERENCES........................................................... 39
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TECHNICAL EVAtVATION REPORT ON THE' !
FIRST-10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN: }
DETROIT EDIS0N COMPANY, ENRICO FERMI ATOMIC POWER PLANT, UNIT 2, DOCKET NUMBER 50-341
- 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) which are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class'2, and Class 3 meet j the requirements, except the design and access provisions and the preservice examination requirements, set forth _ in the ASME Code Section XI, " Rules for l
j Inservice Inspection of Nuclear Power Plant Components," (Reference 2) to l the extent practical within' the limitations of design, geometry, and i materials of construction of the components. This section of the regulations also requires that inservice examinations of components and i
system pressure tests conducted during the initial 120-month inspection interval shall comply with the requirements in the latest edition and j
addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the c te _of issuance of the operating license, j subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code which are incorporated by reference in. '
10 CFR 50.55a(b) subject to the limitations and modifications listed 1 therein. The Licensee, Detroit Edison Company, has prepared the Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval Inservice Inspection (ISI) Program Plan, Revision 0, Change 2 (Reference 3), to meet the requirements of the 1980 Edition, Winter 1981 Addenda (80W81) of the l ASME Code Section XI except that the extent of examination for Class 2 1' piping welds has been determined by the 1974 Edition through Summer 1975 Addenda (74S75).
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As required by 10 CFR 50.55a(g)(5), if the licensee determines that certain i
Code examination requirements are impractical and requests relief from them, q
the licensee shall submit information and justifications to the Nuclear Regulatory Commission (NRC) to support that determination.
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I Pursuant to'10 CFR_50.55a(g)(6), the NRC will evaluate the licensee's determinations under 10'CFR 50.55a(g)(5) that Code requirements are impractical. The NRC may grant relief and may impose alternative requirements that it determines are authorized by law, will not endanger I life or property or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
The information in the Enrico Fermi Atomic Power Plant . Unit 2, First (
10-Year Interval ISI Program Plan, through Revision 0, Cha.1ge 2, submitted April 2,1987, was reviewed, including the requests for relief from the ASME '
Code Section XI requirements which the Licensee has determined to be impractical. The review of the ISI Program Plan was perfo-med 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 March 27, 1987 (Reference 5), the NRC requested additional information that was required in order to complete the review of.the ISI Program Plan. The requested information was provided by the Licensee in a letter dated June 1,1987 (Reference 6).
The Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval ISI Program Plan, through Revision 0, Change 2, is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the
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appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) exclusion criteria, and (d) compliance with ISI-related
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commitments identified during the NRC's review before granting an Operating License. l 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, 1980 Edition including Addenda through Winter 1981. 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 J
This evaluation consisted of a review of the applicable program documents to l determine whether or not they are in compliance with the Code requirements and any license conditions pertinent to ISI activities. This section l
describes the submittals reviewed and the results of the review.-
2.1 Documents Evaluated Review has been completed on the following information provided by the L
Licensee:
1 (a) Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval ISI Program Plan, through Revision 0, Change 2, submitted i April 2, 1987; and (b) Letter, dated June 1,1987, Licensee's response to the NRC's RAI.
2.2 Comoliance with Code Requirements 2.2.1 Comoliance with Acolicable Code Editions 1
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 the Operating License date of July 15, 1985, the Code applicable to the first l
interval ISI program is the 1980 Edition with Addenda through Winter I 1981. As stated in Section 1 of this report, the Licensee has written the i Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval ISI Program Plan, through Revision 0, Change 2, to meet the requirements of the 1980 Edition, Winter 1981 Addenda of the Code except that.the extent of examination for Code Class 2 piping welds has been determined by the 1974 Edition through Summer 1975 Addenda.
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2 2.2 Acceptability of the Examination Samole Inservice volumetric, surface, and visual-examinations shall be performed en ASME Code Class 1, 2, and 3 components and their supports using sampling schedules described in Section'XI of the ASME Code ~ and 10 CFR 50.55a(b). Sample size and weld selection have been implemented in accordance with the Code and appear to be correct.
2.2.3 Exclusion Criteria 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 Licensee in accordance with the Code as discussed in the ISI Program Plan and appear to be correct.
2.2.4 Auamented Examination Commitments The following augmented examinations will be implemented during. the first 10-year inspection interval:
(a) The Licensee has committed to augmented ISI in accordance with NUREG-0313, Revision 1 (Reference 7). In addition to the specific IGSCC countermeasures implemented'at Fermi 2, the ISI program incorporates an 80-month examination cycle.
(b) In response to IE Bulletin 80-07, "BWR Jet Pump Assembly Failure" (Reference 8), the Licensee will ultrasonically inspect the Reactor Pressure Vessel (RPV) Jet Pump Hold Down Beams during each Reactor Refueling Outage until sufficient experience is gained to change the frequency of inspection. If a cracked beam is detected, it will be replaced prior to return to power operation.
(c) Augmented examinations will be performed in accordance with NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive' Return Line Nozzle Cracking" (Reference 9).
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- .(d) The welds in the~ Scram Discharge Volume piping will be examined in accordance with the requirements of Section XI 'and the recommendations of NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping" (Reference 10). I on 2.3 Conclusio.n_1 Based on the review of the documer.t: listed above, it is concluded that the- ,
Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year' Interval'ISI Program Plan, through Revision 0$ Change 2, is acceptable and in compliance with 10 CFR 50.55a(g)(4).
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- 3. EVALUATION OF RELIEF ^ REQUESTS l
The requests for. relief from the ASME Code' requirements which the Licensee has determined to be impractical for the first 10-year inspection interval-are evaluated in the following' sections.
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3.1.1.1 Reauest for Relief RR-A1. Examination Cateaory B-A. Items Bl.11. B1.12. B1.22. and B1.30. Pressure Retainino Shell and Head Welds and Shell-to-Flance Weld in the Reactor Pressure Vessel c
Code Requirement: Section XI, Table IWB-2500-1, Examination .
Category B-A, Items Bl.11 and B1.12 require a 100% volumetric l' examination of all circumferential and longitudinal Reactor Pressure Vessel (RPV) shell welds as defined by Figures IWB-2500-1 and -2. Item 81.22 requires a 100% volumetric examination of all meridional RPV head welds as defined by Figure IWB-2500-3. Item Bl.30 requires a 100% volumetric
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examination of the RPV shell-to-flange weld as defined by -!
Figure IWB-2500-4.
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Licensee's Code Relief Reauest: Relief.is requested from examining 100% of the Code-required volume of the following RPV welds: f Weld Item it of Weld Identification Number Examinable 9-307 B1.11. 26%-
1-313 Bl.11 35%
4-308A Bl .11 - 22%
4-308B Bl.11 25%
2-307A Bl.12 14%
2-307B Bl.12 62%
2-307C Bl.12 73%
1-308A Bl.12 72% i 1-308B Bl.12 72%
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Weld Item % of Weld Identification Number Examinable 1-308C Bl.12 72%
1-308D Bl.12 72% i 1-306A Bl.22 76%
1-306B Bl.22 76%
l-306C Bl.22 76%
l-306D Bl.22 76% !
l-306E Bl.22 76%
l-306F Bl.22 76% l l-306G Bl.22 76% '
l-306H Bl.22 76%
l-306J Bl.22 76%
l-306K Bl.22 76%
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- Bl.30 17%
See " Licensee's Proposed Alternative Examination" l below.
1 l l Licensee's Proposed Alternative Examination: None. The 1
Licensee states that each listed weld will be examined to tne extent indicated. A supplemental manual ultrasonic examination will be conducted for shell-to-flange weld #13-308. This examination will be conducted from the flange seal surface and 1
~.All include 100% of the weld. Leakage testing and hydrostatic l
pressure tests will be conducted in accordance with the Code requirements.
Licensee's Basis for Reauestina Relief: Due to mechanized examination equipment limitations and nozzle interferences, a 100% volumetric examination is not possible for certain shell and lower head welds.
The Fermi Unit 2 RPV preservice nondestructive examinations were conducted in accordance with 71W71 of the ASME Code Section XI and 68S69 of Section III. Subsequent to the RPV preservice inspection (PSI), a pole track system was designed to allow remote ultrasonic (mechanized) examination of the RPV l welds. The design allowed for maximum weld coverage possible with the equipment available at that time. The Code in effect 7
at that time, 1974 Edition'of Section XI, required that 10% of
'. the . length of each longitudinal weld and 5% of the length of each circumferential weld be examined, the designed pole track system exceeded the examination requirements in effect at that I time.
Performance of the limited ultrasonic examinations and the leakage tests will provide the maximum assurance practicable i that the vessel is structurally sound for plant operation.
Evaluation: The Licensee's submittal has been reviewed, including the drawing of the RPV and the mechanized ultrasonic examination pole track system. Based on the design of the RPV-and the mechanized examination equipment limitations, the volumetric examination of the subject weids, to the extent required by the Code, is impractical. A significant percentage of the Code required volume can and, as stated by the Licensee, I will be examined. The RPV would have to be redesigned and i prefabricated in order to complete the remainder. I
Conclusions:
Based on the above evaluation, it is concluded that the limited Section XI volumetric examination of the subject welds, along with the Code-required leakage and hydrostatic pressure tests, ensures an acceptable level of
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inservice structural integrity and that compliance with the
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specific requirements of Section XI would result in hardship or i unusual difficulties without a ' compensating increase in the level of quality and safety. Therefore, it is recommended that ,
relief be granted as requested.
I 3.1.1.2 I Reauest for Relief RR-A2. Examination Cateaorv 8-A. Items B1.21 and B1.22. Pressure Retainino Circumferential and Meridional Lower Head Welds in the Reactor Pressure Vessel Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-A, Items Bl.21 and Bl.22 require a 100% volumetric 8
examination of the circumferential and meridional head welds in the Reactor Pressure Vessel (RPV) as defined by Figure IWB-2500-3.
Licensee's Code Relief Recuest: Relief is requested from performing the Code-required volumetric examination of circumferential bottom head weld 5-306 and meridional bottom head welds 2-306A through G in the RPV.
Licensee's Prooosed Alternative Examination: None. The Licensee states that leakage testing and hydrostatic pressure tests will be conducted in accordance with the Code requirements.
Licensee's Basis for Recuestino Relief: The Licensee states that the Control Rod Drive (CRD) housings in the lower head prevent access to vessel welds in this area. Due to the cluster of CRD housings in the lower head, access to the seven meridional welds and one circumferential weld is not possible.
1 Evaluation: The Licensee's submittal has been reviewed, including the drawing which shows the examination obstructions. Based on the design of the RPV bottom head assembly and the high radiation levels in the area, the Code-required volumetric examination of the subject welds is impractical. The RPV would have to be redesigned and prefabricated in order to perform the Code-required examination.
Conclusions:
Based on the above evaluation, it is concluded that the Section XI volumetric examination of the subject welds is impractical and that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief be granted as requested.
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3.1.1.3 Reauest for Relief RR-A6. Examination Catecory B-D. Item B3.90.
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Nozzle-to-Vessel Welds in the Reactor Pressure Vessel j
Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-D, Item B3.90 requires a 100% volumetric examination '
of the nozzle-to-vessel welds in the RPV as defined by Figure IWB-2500-7.
I Licensee's Code Relief Reauest: Relief is requested from J examining 100% of the Code-required volume of nozzle-to-vessel welds 4-316A (Feedwater nozzle at 30') and 4-316D (Feedwater nozzle at 210') on the RPV.
Licensee's Prooosed Alternative Examination: None. The Licensee states that each of the subject welds will be examined l to the maximum extent possible. . Leakage testing and hydrostatic pressure tests will be conducted in accordance with l the Code requirements.
j Licensee's Basis for Reauestino Relief: The Licensee states that two feedwater nozzle penetrations (N4 nozzles) have adjacent instrumentation nozzle penetrations (N11 nozzles) which will interfere with the mechanized ultrasonic scanning of the required examination volume. The limitations will prevent examination of 12.8% of the subject nozzle-to-vessel welds.
Evaluation: The Licensee's submittal has been reviewed, including the sketch which shows the examination limitations.
Based on the location of the RPV feedwater nozzles with respect to the instrumentation nozzles and the design of the as-installed mechanized ultrasonic scanning equipment, the Section XI volumetric examination of the subject nozzle-to-vessel welds, to the extent required by the Code, is impractical. A significant percentage of the Code-required l inservice volumetric examination can and will be performed.
Failure to perform a 100% inservice examination of these welds 10.
will not significantly affect the assurance of the structural l
'. integrity. I
Conclusions:
Based on the above evaluation, it is concluded that the limited Section XI volumetric examination of the subject welds, along'with the Code-required. leakage and -
hydrostatic pressure tests, ensures ~an acceptable level of inservice structural integrity and that compliance with the specific requirements of Section XI would result in hardship or' unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief be granted as requested.
L 3.1.1.4 Recuest for Relief RR-A7 Examination Cateoory B-E Items 84,12 and B4.13. Partial Penetration Welds in Control Rod Drive Nozzles and Instrumentation Nozzles on the Reactor Pressure Vessel Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-E, Items B4.12 and B4.13 require a 100% visual examination (VT-2) of the external surfaces of partial penetration welds in 25% of the Control Rod Drive (CRD) nozzles and instrumentation nozzles.
Licensee's Code Relief Reouest: Relief is requested from i performing the Code-required visual examination (VT-2) of external surfaces of the CRD and instrumentation' nozzles-on the RPV.
licensee's Pronosed Alternative Examination: The Licensee states that a visual (VT-2) examination will be conducted from below the housing location under the vessel lower head during the system leakage and hydrostatic tests.
i Licensee's Basis for Reouestino Relief: The Licensee states '
r that, due to the location of these components, these partial t
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l penetration welds are not accessible for direct visual j
'. examination. The welds, located in the tightly clustered l
instrumentation and CR0 housings inside the vessel, can not be directly accessed ins'ide or outside the vessel. Visual-(VT-2) examination for evid'ence of leakage will insure the integrity
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I of these components.
Evaluation: IWA-5241(b), "Noninsulated Components," states:
"For components' whose external surfaces are inaccessible for direct visual examination (VT-2), only the examination of surrounding area, including floor areas or equipment surfaces located underneath the components, for evidence of leakage j shall be required." Because a visual (VT-2) examination will l
be conducted from below the housing location under the vessel lower head during the system leakage and hydrostatic tests, the Licensee is in accordance with IWA-5241(b) of Section XI of the Code.
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Conclusions:
Based on the above evaluation, it is concluded '
that the Licensee's proposed alternative examination is in accordance with the Code requirements. Therefore, relief is I not required. (
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I 3.1.2 Pressuriz_qC (Does not apply to BWRs) 4 i
3.1.3 Heat Exchancers and Steam Generators (No relief requests) 3.1.4 P_toino Pressure Boundary 3.1.4.1 Reauest for Relief RR-A3. Examination Cateoory B-J. Item B9.11.
Pressure Retainino Circumferential Welds in Class 1 pinina l
Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-J, Item B9.11 requires a 100% surface and volumetric examination of circumferential welds in Class 1 piping, nominal pipe size equal to or greater than 4 inches, as defined by 1 12
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Figure IWB-2500-8. Note (1) states: " Examination shall 1
'. include the following: ....(b) All terminal ends and joints in each pipe or branch run connected to other components'where the stress levels exceed the following limits under loads '
1 associated with specific seismic events and operational conditions: (1) primary plus secondary stress intensity range I
of 2.4Sm.for ferritic steel and austenitic steel..."
i licensee's Code Relief Reouest: The Licensee requests relief l from performing the self-imposed surface and volumetric I 1
examination of circumferential piping weld SW-PS-2-X70-W1 ;
(pipe-to-flued hea'd). '
i licensee's Pronosed Alternative Examination: None. The Licensee states that leakage testing and hydrostatic pressure tests will be conducted in accordance with the Code requirements.
Licensee's Basis for Reauestina Relief: The Licensee states that, due to the design of the primary containment penetration assemblies, the circumferential weld at the process pipe-to-flued head is not accessible. The clearance available between the process pipe and guard pipe is not large enough to allow examination equipment access to the weld area. In addition, the open end side of the penetration assembly is obstructed by construction centering lugs and the. installed insulation.
The examination of the subject weld is a self imposed requirement beyond the requirements of the Code. Due to the design limitations which prevent reasonable access to the weld area, a significant design and development program would be required to fabricate the required examination equipment. A program of this nature is not feasible. Only one selected weld is affected by this situation. Because of these factors and the fact that the Code does not require the examination of this 13
I e
weld, relief is requested from the commitment made by Detroit Edison to examine this moderate stress weld. l Evaluation: The Licensee's submittal has been reviewed, l including the drawing which shows the examination obstructions. As stated above, the Code requires surface and 4 volumetric examination of welds where the stress level exceeds !
2.4Sm . The Licensee has supplemented this requirement to include those welds which exceed 2.lSm . Welds which meet }
this stress level, between 2.lS mand 2.4S m, are designated moderate stress welds. Because examination of this weld is
{
self-imposed and is not a Code requirement, the Licensee will !
be in compliance with the Code even if this weld is not l examined.
l l
.C_qnclusion s : Based on the above evaluation, it is concluded that the Licensee is in compliance with the Code requirements.
i Therefore, relief is not required.
3.1.4.2 Reouest for Relief RR-A8. Examination Cateoory B-K-1. Item B10.10. Intearally Welded Attachments for Class 1 Pioina Code Requirement: Section XI, Table IWB-2500-1, Examination l
Category B-K-1, Item 810.10 requires a 100% volumetric or l
surface examination of integrally welded attachments of Class 1 piping as defined by Figures IWB-2500-13, -14, and -15.
Licensee's Code Relief Reouest: Relief is requested from performing the Code-required surface examination of the following Main Steam piping lug welds: !
FW-PS-2-A4-gal FW-PS-2-C4-GCl
-GA2 -GC2
-GA3 -GC3
-GA4 -GC4 FW-PS-2-B4-GB1 FW-PS-2-D4-GD1
-GB2 -GD2
-GB3 -GD3
-GB4 -GD4
! 14 l
l
[
. i Licensee's Proposed Alternative Examination: None. The Licensee states that, should the support component be removed-.
.for any reason, the required surface examination will be.
. conducted on the exposed lug' welds.
)
Licensee's Basis for Reauestino Relief: The Licensee states )
! that the guide lugs located on main steam piping immediately preceding the inboard isolation valves are surrounded by support components! Access to the examination areas requires 1 l that significant labor be expended.in the removal of the associated support components. The estimated time for removal ,
of one support is 112 man-hours. The significant labor. 1 i
required, along with the accompanying radiation exposure, can not be justified. No structural change to the welds is anticipated due to inservice operation.
1 Evaluation: The Licensee's submittal has been reviewed, including the drawing which shows the examination obstructions. Because the subject piping lugs are surrounded by support components which require a major effort to remove in i
order to allow examination access to these lugs, the l Code-required surface examination of these main steam piping I lugs is impractical. However, the Licensee has committed to perform the required surface examination on the exposed lugs should the support component be removed for any reason.
Failure to perform the Code-required inservice surface examination of the subject welds will not significantly affect the assurance of the structural integrity. i l
Conclusions:
Based on the above evaluation, it is concluded that the Section XI surface examination of the subject welds is I impractical and that compliance with the specific Section XI requirements would result in hardship or unusual difficulties i
without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief be granted as requested.
l 15 )
1 3.1.4.3 Recuest for Relief RR-A9. Examination Cateoory B-K-1. Item B10.10. Inteorally Welded Attachments for Class 1 Pioino ;
i i
Code Requirement: Section XI, Table IWB-2500-1, Examination Category B-K-1, Item B10.10 requires a 100% volumetric or surface examination of integrally welded attachments of Class 1 piping as defined by Figures IWB-2500-13, -14, and -15.
(
Licensee's Code Relief Recuest: Relief is requested from
]
performing the Code-required surface examination of the following Class 1 piping lugs:
l FW-RD-2-A2-W8 SW-RD-2-B2-W8
-WS FW-RS-2-A2-AL1 )
-W9 -AL2 1
-W12 -W12 -AL3 I
-W13 -W13 -AL4
-W16 -W16
-W17 -W17
-W18 -W18
-W19 -W19 i
Licensee's Pronosed Alternative Examination: None.
Licensee's Basis for Recuestino Relief: The Licensee states that various abandoned lugs exist on piping components within the Class I system boundaries. There are no support components associated with these lugs and the lugs are not required for i any function. Examination of these lugs will increase radiation exposure without benefit to plant safety. The attachment welds are not full penetration welds and their integrity is not required as they serve no functional purpose.
The lugs are not loaded and no structural change in the welds '
is expected due to inservice operation. In addition, the Licensee states that ASME Code Case N-343, paragraph 2.0,
" Examination Category B-K-1," item (b) allows this exemption.
Evaluatioq: Because the subject piping lugs serve no functional purpose and there are no support components associated with these lugs, the Section XI surface examination of these lugs is impractical. In paragraph 2.0 of Code Case 16
N-343, one of the conditions'for requiring an. integrally welded attachment to be examined is that the attachment provides-component support as defined in NF-1110.- Since the subject-lugs are not load bearing, they are exempt from examination per
, Code Case N-343. Code Case N-343 is listed as NRC-approved in Regulatory Guide 1.147, " Inservice Inspection Code Case Acceptability, ASME Section XI Division 1" (Reference'11).
Failure to perform the Code-required inservice examination of these attachment welds will not significantly affect the assurance of the structural integrity.
Conclusions:
Based on the above evaluation', it-is concluded that the Section XI surface examination of the subject piping lugs is impractical and that compliance with the specific Section XT requirements would result in hardship or unusual difficulties without a compensating-increase in the level of -
quality and safety. Therefore, it is recommended that relief l be granted as requested.
3.1.4.4 Reauest for Relief RR-A10. Examination Catecorv B-K-1. Item B10.10. Intearally Welded Attachments for Class 1 pioina Code Reouiremenl: Section XI, Table IWB-2500-1, Examination l Category B-K-1, Item B10.10 requires a 100% volumetric or surface examination of integrally welded attachments of Class 1 1 piping as defined by Figures IWB-2500-13, -14, and -15.
Licensee's Code Relief Reouest: Relief is requested from performing the Code-required surface examination of the following Class 1. piping lugs in the primary containment penetrations:
SW-E51-2192-X10-W2A SW-N21-2336-X98-W2A
-W2B -W2B
-W2C -W2C SW-E41-2297-X11-W2A SW-E21-3053-X16A-W2A
-W2B -W2B
-W2C -W2C 17
_ _ - - _ - _ - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . - - _ .____-_-___a
(cont'inued) i SW-E11-2298-X138-W2A SW-E11-3519-X17-W2A
- i. -W2B -W2B j
-W2C -W2C ;
l SW-E11-2289-X12-W2A SW-X7A-W2A l
-W2B -W2B' !
-W2C -W2C-1 I SW-E11-2327-X13A-W2A SW-X7B-W2A i
-W2B -W2B
-W2C -W2C )
1 SW-E21-3052-X16B-W2A SW-X7C-W2A I
-W2B -W2B l
-W2C -W2C j SW-N21-2336-X9A-W2A SW-X70-W2A I
-W2B -W2B i
-W2C -W2C Licensee's Proposed Alternative Examination: None.
Licensee's Basis for Recuestina Relief: The Licensee states j that, due to the design of the Primary Containment Penetration Assemblies, the centering lugs installed in the penetrations are not accessible for examination. The clearance between the guard pipe and process pipe is not large enough to allow inspection of these areas. The lugs are installed inside the l
1 penetrations and examinations can not be performed. The l attachment welds are not full penetration welds.and their I integrity is not required as they serve no functional purpose.
I The lugs are not loaded and no structural change in the welds is expected due to inservice operation, In~ addition, the Licensee states that ASME Code Case N-343, paragraph 2.0, l l " Examination Category B-K-1," item (b) allows this exemption. !
l l
Evaluation: Because the subject piping lugs are installed l inside the primary containment penetration assemblies and are inaccessible for examination, the Section XI surface- i examination of these lugs is impractical. Since the subject lugs are not loaded, failure to perform the Code-required t inservice examination of these attachment welds will not 18 !
l l
i
1 d
significantly affect the assurance of the structural integrity.
Conclusions:
Based on the above evaluation, it is concluded )
i that the Section XI surface examination of the subject piping lugs is impractical and that compliance with the specific l Section XI requirements would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief
}
be granted as requested.
3.1.5 Pumo Pressure Boundary
{
3.1.5.1 Reouest for Relief RR-A4. Examination Cateoory B-L-2. Item l
812.20 Class 1 Pumo Casinos Code Requirement: Section XI, Table IWB-2500-1, Examination l
' Category B-L-2, Item 812.20 requires a 100% visual (VT-3) examination of the internal surfaces of Class 1 pump casings. j l
Licensee's Code Relief Reouest: Relief is requested from performing the Code-required visual (VT-3) examination of the internal surfaces of the pump casings of recirculation pumps t
B31010001A and 8310100018.
i licensee's Procosed Alternative Examination: None. Whenever these pumps are disassembled for maintenance or for any other reason which would provide access to the internal surfaces, a visual examination of the internal surfaces shall be performed.
Licensee's Basis for Reouestino Relief: The Licensee states that access to the recirculation pump casing internal surfaces can only be accomplished when the pump has been disassembled.
At this time, there are no other requirements to disassemble the pump on a scheduled basis. Disassembly of a recirculation pump for performance of a VT-3 visual examination would require significant labor and radiation exposure. It is expected that 19 t
,s-e approximately 1000 man-hours and 50 man-rem exposure would be required to disassemble, inspect, and reassemble one pump.
The pump casing is primarily a single piece design of cast stainless steel (ASTM A-351, grade CF8H) and is widely'used in the nuclear industry. Operating histories of recirculation-pumps do not warrant concern for casing failure or inservice growth of manufacturing flaws. The presence of some delta ferrite (typically 5% or more) imparts substantially increased resistance to intergranular stress corrosion cracking (IGSCC).
The delta ferrite also results in improved pitting corrosion-resistance in chloride-containing environments.
The Licensee feels that adequate safety margins are inherent in the basic pump design and that the health and safety of the public will not be adversely affected by performing the visual examination of the pump internal pressure boundary surfaces only when the pumps are required to be disassembled for maintenance. In addition, both pumps will be VT-2 examined every refueli.ng outage during leakage tests and once in the 1
interval during hydrostatic testing.
Evaluation: The visual examination is to determine whether unanticipated severe degradation of the casing is occurring due to phenomena such as erosion, corrosion, or crackir.g. However, previous experience during examination of pumps at other plants has not shown any significant degradation of pump casings. The concept of visual examination if the pump is disassembled for maintenance is acceptable. The disassembly of the pumps solely for the purpose of inspection is a major effort and, in addition to the possibility of additional wear or damage to the intsrnal surfaces of the pumps, could result in large amounts of radiation exposure to personnel. However, if the pumps are-disassembled for maintenance, the internal surfaces would be examined, in which case relief would not be required for those particular pumps.
20
)
1 l
l
Conclusions:
Based on the above evaluation, it is concluded i 1
, that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Therefore, it is recommended that: (1) The Licensee's proposal 4 to perform the visual examination (VT-3) of the internal surfaces of the pumps, whenever they are made accessible due to i disassembly for maintenance purposes, should be accepted; and '
(2) Relief should be granted at the end of the interval if one
)
of the subject pumps, for which a visual examination is l 1
required, has not been disassembled for maintenance.
3.1.5.2 Reauest for Relief RR-All. Examination Cateoory B-K-1. Item B10.20. Intearally Welded Attachments for Class 1 Pumos )
l Code Reauiremen_t.: Section XI, Table IWB-2500-1, Examination Category B-K-1, Item B10.20 requires a 100% volumetric or surface examination of integrally welded attachments of Class I pumps as defined by figures IWB-2500-13, -14, and -15.
Licensee's Code Relief Reauest: Relief is requested from performing the Code-required surfa e examination of the following reactor recirculation pump insulation lugs and hanger brackets:
Pump A: Pump B:
SW-B31-5365-PUMPA-WA SW-831-5365-PUMPB WA
-WB -WB
-WC -WC Licensee's Procosed Alternative Examination: None. The Licensee states that, should the insulation be removed for other reasons, the required examinations will be conducted.
l Licensee's Basis for Recuestina Relief: The Licensee states I i
that the reactor recirculation pump insulation lugs and hanger !
brackets are covered by insulation. To gain access to the l
examination areas, the pump insulation requires removal.
1 Removal and replacement of the recirculation pump insulation 21
4 will result in an excessive amount of radiation exposure for plant personnel. A significant amount of labor would be required to remove the insulation needed to provide access to the examination areas.
Based upon discussions with the contract personnel responsible for the installation of the mirror insulation, a calculation of radiation exposure was made. It was estimated that a total of 12,800 mR would be expected for this task. This assumes a radiation field of 100 mR and 9 man-days for removal and replacement of insulation. The operating records of other i
plants do not show that pump lugs have a record of significant problems. In addition, industry field data do not indicate '
that ca'st stainless steel pump casings are su:ceptible to service-induced flaws. Also, the Licensee states that ASME Code Case N-343, paragraph 2.0, " Examination Category B-K-1,"
item (b), allows this exemption.
Evaluation: Because the removal and replacement of the reactor recirculation pump insulation solely for the purpose of inspection is a major effort and could result in large amounts of radiation exposure to personnel, the Code-requ. ired surfac'e examination of the subject attachment welds is impractical.
However, if the insulation is removed for other reasons, the Licensee has committed to perform the required examinations.
I
Conclusions:
Based on the above evaluation, it is concluded that the Section XI surface examination of the subject reactor recirculation pump insulation lugs and hanger brackets is impractical and that compliance with the specific Section XI requirements would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief be granted as requested.
f
(
f 22
f
- i )
3.1.6 Valve Pressure Boundarv
. \
3.1.6.1 Reauest for Relief RR-A5. Examination Cateoory 1-M-2. Item _ , f B12.50. Class 1 Valve Bodies E '[
Code Requirement: Sdetion XI, Table IWB-2500-1, Examination, ,1 k Category B-M-2, Item 812.50 requires'a 100% visual (VT-3) 'h' l examination of the internal surfaces of Class 1 valve bodies i exceeding 4 inches nominal pipe size.
.i Licensee's Code Relie' Reauest: Relief is requested fr\om performing the Code nquired visual (VT-3) examination -cf the internal surfaces of the Class 1 valve bodies exceeding '.
4 inches nominal pipe sizec The Fermi Unit 2 Class I systems contain 63 valves which are greater that four inches 'ncminal ,
pipe size. These valves vary in size, design, and manufacturer /
but all are manufactured from either cast or forged stainless I A
steel or carbon steel. i Neue of the valve bodies are selded.] {
Of these 63 valves, 23 are subject to visual examination /per Table IWB-2500-1, Examination Category B-M-2, Item B12.50.
.j,/
1 Licensee's Procosed Alternative Examination: None. Whenever these valves are disassembled for maintenance, or for any other reason which would provide access to the internal surfaces, a 1
visual examination of the internal surfaces-shall be performed. The Licensee states that leakage testing and- )
hydrostatic pressure tests will be conducted in accordance with I the Code requirements.
Licensee's Basis for Reduestino Relid: The Licensee states
- j that access to valve internal surfaces can only be accomplished j
when the valve is disassembled. Disassembly of valves for visual examination with no other iequir! ment' fo'r raainte'bance t
will increase exposure to plant personnel. In some cases, /
,q radiation fields as high as 10 R/hr could be experienced. In 9 l addition, further labor and radiation exposure would occur for 5
- s
,i ' !
23 <
A i
~ - - - - - - - - - - - - - - - - ~ ~ / h)
~
l 3 Aj those valves which gn not be isolated from reactor fluid.
3 This is the case for epprmfimately 20% of the subject valves 3
(12 total) and would require er. loading of the RPV core and
", vessel draining.
l The Licensee also states that performing these visual examinations on poor as cast surfaces provides little additional information as to valve body integrity, and that j , historically these materials have excellent service records and y
there is little reason to expect service-related flaws in these j
valve bodies. j l 1 Evaluation: The visual examination is to detere.ine whether p
unanticipated severe degradation if the valve body is occurring
, due to phenomena such as erosion, crirro:sion, or cracking.
However, previous experience during examination of valves at '
other plants has not shown any significant degradation of valve f bodies. The concept of visual examination if the valve is disassembled for maintenance is acceptable. The disassembly of J ;
the valves solely for the purpose of inspection is a major '
i 1 effort and, in addition to the possibility of additional wear or damage to the internal turfaces of the valves, could result in large amounts of radiation exposure to personnel. However, l
if the valves are disassembled for maintenance, the internal surfaces would be examined, in which case relief would not be I required for those particular valves.
QincTusions Based on the above evaluation, it is concluded that compliance with the specific requirements of Section XI L
pould result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Therefore, it is recommended that: (1) The Licensee's proposal to p.erform the visual examination (VT-3) of the internal surfaces of the valve bodies, whenever they are made accessible due to disassembly for maintenance purposes, sheuld be
) accepted; and (2) Relief should be granted at the end of the i
24 1 i
1 v .
interval if one of the subject valves, for which a visual examination is required, has not been disassembled for -
- maintenance.
-(
3.1.7 General (No relief requests)
+
3.2 Class 2 Comoonents 3.2.1 Pressure Vessels 3.2.1.1 'Reauest for Relief RR-A14. Examination Catenorv C-C. Item C3.10. Intearally Welded Attachments for Cigy 2 Pressure 1 Vessels 1
Code Requirement: Section XI, Table .IWC-2500-1, Examination -l l Category C-C, .Ikem C3.10 requires a 100*,' surface examination of.
the integrally selded attachments of Clas.s 2 pressure vessels , ,- !
as defined by Figure NC-2500-5-. ) '
.i-Licensee's Code Relief Recuest: Relief is requested from performing the Co' de-required surface examination of weld I
~
SW-E11-D2-HXS-08 on the bottom side of the lower support ring of the RHR heat exchanger. l;
)
Licensw's Proposed Alt.ernative Examination: None. Tne' Licensee states that the fillet welds on both sides of the top l 1
support ring, the stiffener plates connecting the top and e i / ;
. bottom support rings, and the weld on the top side of the lower t 4
1
.) support ring will be examined. Also, leakage testing and .
1 i
hydrostatic pressure. tests will be conducted in acet,rdance with i g the Code requirements. a !
t
/ l
' e s !
Licensee's Basis n.r Recuestino,Rgligi: The licensee states ;
I that the RHR heat exchanger has two 1-inch thick support rings '
t welded to the heat exchanger shell. The bottom support ring ,
W rests on, and is. bolted to, the structural support steel for
) ,
25 .
l l i
l .,
the heat exchanger. The weld on the bottom side of the support
'd . . ring is encased by structural steel, making the weld ;
l
' inaccessible for examination. The weld is a fillet weld and as i i
such not a full penetration weld. The fillet weld on the top l
side of the lower support ring (SW-E11-D2-HXS-07) will receive the Code-required surface examination. !
i I
Evaluation: The Licensee's. submittal has been reviewed, i including the sketch which shows the examination obstructions.
Based on the design of the support rings and structural steel
]
support for the RHR heat exchanger, the Section XI surface examination of the subject attachment weld is impractical. .
This weld is encased by structural steel and is totally inaccessible for examination. The heat exchanger supports i.
would have to be redesigned and prefabricated in order to perform the Code-required examination of the subject attachment ')
weld.
Conclusions:
Based on the above evaluation, it is concluded that the Section XI surface examination of the subject attachment weld is impractical and that compliance with the i I
specific Section XI requirements would result in hardship or
,k l
unusual difficulties without a compensating increase in the '
level of quality and safety. Therefore, it is recommended that
/ relief be granted as requested. i 3.2.2 Pioino i
3.2.2.1 Recuest for Relief RR-Al2. Examination Cateoorv C-C. Item l
, C3.20. Inteorally Welded Attachments for Class 2 Pioino i
p Code Requirement: Section XI, Table IWC-2500-1, Examination 1 Cate gory C-C, Item C3.20 requires a 100% surface examination of ;
the integrally welded attachments of Class 2 piping as defined by Figure IWC-2500-5.
26 I
l
Licensee's Code Relief Recuest: Relief is requested from performing the Code-required surface examination of 53 Class 2
)
piping lugs as listed in Request for Relief RR-A12, Revision 0,
}
Change 1 (page A-33 of 52 of the ISI Program Plan).
1 Licensee's Propose'd Alternative Examination: None.
Licensee's Basis for Recuestino Relief: The Licensee states )
l that various abandoned lugs exist on piping components within l the Class 2 system boundaries. There are no support components associated with these lugs and the lugs are~ not required for
. any function. Examination of these lugs will increase radiation exposure to~ plant personnel without benefit to plant safety. The attachment welds are not full penetration welds and their integrity is not required as they serve no functional purpose. The lugs are non-load bearing and no structural change in the welds is anticipated due to inservice operation. 1 Also, the Licensee states that ASME Code Case N-343, I paragraph 3.0, " Examination Category C-C," item (b), allows this exemption.
f Evaluation: Because the subject piping lugs serve no functional purpose and there are no support components associated with these lugs, the Section XI surface examination of these lugs is impractical. In paragraph 3.0 of Code Case N-343, one of the conditions for requiring an integrally welded attachment to be examined is that the attachment provides I
component support as defined in NF-1110. Since the subject lugs are not load bearing, they are exempt from examination per j 4 Code Case N-343. Failure to perform the Code-required inservice examination of these attachment welds will not significantly affect the assurance of the structural integrity, i
Conclusions:
Based on the above evaluation, it is concluded that the Section XI surface examination of the subject piping lugs is impractical and that compliance with the specific !
27
l
~
l !
Section XI requirements would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. - Therefore, it is recommended that relief )
l be granted as requested. l q
3.2.2.2 Reauest for Relief RR-A13. Examination Cateoory C-C. Item C3.20. Inteorally Welded Attachments for Class 2 Pinino t
Code Requirement: Section XI, Table IWC-2500-1, Examination Category C-C, Item C3.20 requires a 100%-surface examination of the integrally welded attachments for Class 2 piping as defined by Figure IWC-2500-5. l I
i Licens'ee's Code Relief Reauest: Relief is requested from j performing.the Code-required surface examination of the following-integrally welded pads on Class 2 piping:
PSFW-E11-3146-148A PSFW-E11-3151-145A
-1488 -145B l
-148C -145C. 'l'
-1480 -145D Licensee's Procosed Alternative Examination: None.
Licensee's Basis for Reauestino Relief: The Licensee states that several pads installed on Class 2 piping are inaccessible for examination. The pads -are located in floor penetrations and serve to restrict movement of the pr6 cess pipe in the l penetration. Access to the area of the pads is extremely limited. Examination of these pads is not possible due to !
their location in the floor penetrations. The pads are not i under constant loading and no structural change in the welds is !
anticipated due to operation. i
, Evaluation: The Licensee's submittal has been reviewed, !
including the sketch which shows the examination obstructions.
i Based on the design of the floor penetrations, the Section XI surface examination of the subject attachment welds is impractical. The floor penetrations would have to be -
28
i redesigned and prefabricated in order to perform the Code-required surface examination.
Conclusions:
Based on the above evaluation, it is concluded that the Section XI surface examination of the subject attachment welds is impractical and that compliance with the specific Section XI requirements would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief be granted as requested.
3.2.3 Pumos (No relief requests) 3.2.4 Valves (No relief requests) 3.2.5 General (No relief requests) 3.3 Class 3 Comoonents (No relief requests) 3.4 Pressure Tests (No relief requests) 3.5 General 3.5.1 Ultrasonic Examination Techniaues (No relief requests) 3.5.2 Exemoted Comoonents (No relief requests) 3.5.3 Other 3.5.3.1 Reauest for Relief RR-82. Class 1. 2. and 3 Comoonent Suocorts Code Requirement: Section XI, Paragraph IWF-2430(a) states:
"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. Also, the examinations shall be 29
I 1
extended to include additional supports equal in number and similar in type, design, and function to those initially j examined during the inspection." {
l i
Licensee's Code Relief Recuest: Relief is requested from the requirements of IWF-2430(a) for Class 1, 2, and 3 component ,
supports. I l
1 Licensee's Proposed Alternative Examination: The Licensee I states that an engineering evaluation of the failed component I support will be performed to determine whether additional l 1 component supports could also be subject to the same mode of l
failure. If additional component supports would be subject to l the same mode of failure, the engineering evaluation will specify an appropriate number of additional supports to be examined. I The engineering evaluation of the failed component support shall also consider the effect of the failed component support on other component supports within the piping subsystem or component anchorage. Any component supports that were i subjected to loads exceeding their upset condition design capacity, regardless of whether they are adjacent to the component support that failed, will be included in the scope of l additional examinations.
Licensee's Basis for Reouestino Relief: The Licensee states that the subject Section XI requirement is rather arbitrary l
since there is no sound technical basis for this requirement.
Using IWF-2430(a), there will be cases where unnecessary visual examinations will be performed on adjacent supports. There will also be cases where supports that are not adjacent to the failed component support should be included in the scope of additional examinations, yet these components are not even l
, considered by Section XI.
l 30
The proposed alternative criteria for additional examinations is a more technically complete method for determining whether additional supports in a piping subsystem should be examined.
Evaluation: The Licensee's submittal has been reviewed.
Because adjacent supports may not have the same failure characteristics as those that fail to satisfy the acceptance standards of IWF-3400 and because the additional examinations could result in large amounts of radiation exposure to personnel, the Code requirement to examine the component supports immediately adjacent to those requiring corrective action is impractical.
Conclusions:
Based on the above evaluation, it is concluded that the proposed alternative examination ensures an acceptable level of inservice structural integrity and that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety. Therefore, it is recommended that relief be granted as requested.
l 3.5.3.2 Reauest for Relief RR-C1, Class 1. 2, and 3 Snubbers i
Code Requirement: Component support examination boundaries are defined by IWF-1300 and Figure IWF-1300-1. and identify the l l boundary of a nonintegral attachment to the pressure retaining component as the contact surface between the component and the attachment. The IWF boundary of an integral attachment to the pressure retaining component begins where the IWB, IWC, or IWD boundary ends.
j Licensee's Code Relief Reauest: Relief is requested from removing the insulation from all nonexempt safety-related.
I snubbers on Code Class 1, 2, and 3 insulated lines solely for the purpose of performing a visual examination on the portion l of the integral or nonintegral attachment within the 31
insulation.
Licensee's Procosed Alternative Examination: The Licensee states that the visual examination _ (VT-3) will be limited at the pressure-retaining component boundary to the visually accessible portions of the nonintegral or integral attachment.
Penetrations of the component insulation by the nonintegral or integral attachment allow for a limited examination of the attachment to the pressure retaining component. In general, the component support boundary will extend from the surface of the insulation and include essentially 100% of the component support. Additionally, a check for damage to the insulation at the attachment penetration will provide indication of abnormal conditions that may exist beneath the insulation surface.
Licensee's Basis for Reauestina Relief: The Licensee states that the visual examination of the nonintegral or integral attachment is limited due to the installation of' insulation on the component. It is highly impractical to remove insulation from components solely for the purpose of performing a visual examination on a portion of the nonintegral or integral attachment within the insulation. Estimates of radiation exposure to plant personnel during removal and reinstallation of insulation are at least 10 times the exposure obtained from the performance of the visual examination alone. Therefore, removal of insulation would pose substantial ALARA concerns for plant personnel in radiation areas'without providing a significant increase in system reliability'or safety.
For some cases, when the mechanical connection of a nonintegral attachment is buried within the component insulation, Subarticle IWF-1300 allows the examination boundary to extend l from the surface of the component insulation. In the case of j
integral attachments, the critical area (i.e., .weldment and heat affected zone) of the integral attachment is within the IWB, IWC, and IWD boundary which requires and receives l
32 '
1
l examinations described in the ISI Program Plan. Thus, the Code !
, recognizes that the removal of insulation solely for the l purpose of performing the visual examination can impose undue difficulties, increase ALARA concerns or have no significant affect on the overall adequacy of the examination method in providing assurance of system integrity.
Portions of systems are noninsulated and, therefore, allow a I visual examination of the entire attachment portion. The support attachments and installation procedures used on the noninsulated components are essentially the same type of support attachments and installation procedures utilized on the insulated components. Therefore, the examiretions of the '
noninsulated components provide an adequate me.Las of identifying problems with a particular type of support or \
application. Since the visual examination is a general 1 l
) examination of structural and mechanical integrity, the !
examinations performed on the noninsulated and the limited examinations performed on the insulated components, along with a check of the insulation, srovide an adequate indication of the structural and mechanical integrity of the nonintegral or integral attachments.
In addition to the examinations required by ASME Section XI, augmented examinations of snubbers exempt from examination by l ASME Section XI are required to be performed per the Technical Specifications and greatly increase the number of supports examined. The examination methodology described in this request for relief ensures that a high degree of system safety
. and reliability is maintained while providing substantial ALARA benefits to plant personnel.
I Evaluation: IWF-1300(e) states: "Where the mechanical connection of a nonintegral support is buried within the component insulation, the support boundary may extend from the surface of the component insulation provided the support either 33
- carries the weight of the component or serves as a structural restraint in compression." ASME Code Section XI, Interpretation
- XI-1-86-11 (Interpretations No. 18),
Question (5), provides clarification on the Code requirement for those components not excluded based on IWF-1300(e). l Therefore, the Licensee should provide additional information describing the analyses performed to determine which of the component supports may be excluded based on IWF-1300(e), and evaluate the remaining supports to determine which supports may require relief from the Code-required examination along with l the technical justifications. The Licensee should not be requesting relief for the component supports which may be exempted based on IWF-1300(e).
For those supports remaining after the above exemption, relief would be considered for the following: (a) if the insulation i is required by other regulations or the Technical Specifications to be in place (e.g. fire stops); or (b) if the Licensee can demonstrate that the failure of the component !
support would be obvious should the support fail with the insulation installed. The method for determining item (b) I above should be described in the request for relief. If the above cannot be technically justified, the insulation should be removed for the examination.
Conclusions:
Based on the above evaluation, it is concluded that: (a) The Licensee should not be asking for relief for supports which may be exempted based on IWF-1300(e); and (b) The Licensee should provide the additional information as described above as part of the technical justification for the granting of relief. Therefore, it is recommended that relief should not be granted at this time.
3.5.3.3 Recuest for Relief RR-C2. Class 1. 2. and 3 Snubbers Code Requirement: Section XI, Paragraph IWF-2430(a) states:
34
y
~
"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 1 action shall be examined. Also, the examinations shall be .
extended to include additional supports equal in number and similar in type, design, and function to those initially examined during the inspection."
Licensee's Code Relief Reauest: Relief is requested from the i requirements of IWF-2430(a) for Class 1, 2, and 3 snubbers. f l
Licensee's Procosed Alternative Examination: The Licensee states that, when corrective actions are required as a result ,
) of the Code-required visual examination not satisfying the f requirements of IWF-3400, the adjacent components will be f
examined if the adjacent components are generically susceptible to the same type of indication or mode of failure regardless of 1 support type. i
)i 4
Licensee's Basis for Reouestina Relief: The Licensee states }
that the additional examinations required by failure to satisfy (
the acceptance standards of IWF-3400 are not limited by Section XI to the type of indication identified, type of component being examined, or the types of ' adjacent components.
This would require that a documented visual examination be performed on adjacent supports which, in some cases, have completely different failure characteristics and would provide no meaningful results. Also, examination personnel would receive at least twice the radiation exposure obtained from the s initial examination in radiation areas. These actions cause undue hardships and increase ALARA concerns without significantly affecting the overall safety of the plant. l 1
Corrective actions are required for indications failing to-satisfy the acceptance standards of IWF-3400. The acceptance standards of IWF-3400 are both generic and specific in nature 35"
and therefore, in some cases, apply only to a particular type
,} of support. It is impractical to perform an examination on an adjacent component when the indication identified is limited to a particular type of component support and the adjacent '
components are not of that type.
If the indication requiring corrective measures can be generically applied to the adjacent component or components, the component or' components will be examined. Since additional examinations are required and performed in accordance with IWF-2430 and the components are reexamined per the requirements of IWF-2420 during the next inspection period, this provides' a high degree of system reliability and safety while providing ALARA benefits to plant personnel.
Evaluation:
The Licensee's submittal has been reviewed.
Because adjacent supports may not have the same failure characteristics as those that fail to satisfy the acceptance standards of IWF-3400 and because the additional examinations could result in large amounts of radiation exposure to personnel, the Code requirement to examine the component supports immediately adjacent to those requiring corrective action is impractical.
Conclusions:
Based on the above evaluation, it is concluded that the proposed alternative examination ensures an acceptable level of inservice structural integrity and that compliance with the specific requirements of Section XI would result i_n hardship or unusual difficulties without a compensating increase in the level' of quality and safety. Therefore, it is recommended that relief be granted as requested.
36
l k
1
- 4. CONCLUSION f
(
Pursuant to 10 CFR 50.55a(g)(6), it has been determined that certain
{
Section XI required inservice examinations are impractical. In all cases except Request for Relief RR-C1, the Licensee has demonstrated that either '
the proposed alternatives would provide an acceptable level of quality and 5
safety or that compliance-with the requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
This technical evaluation has not identified any practical method by which l the existing Enrico Fermi Atomic Power Plant, Unit 2, can meet all the specific inservice inspection requirements of Section XI of the ASME Code.
Requiring compliance with all the exact Section XI required inspections would require 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. Examples of components that would require redesign to meet the specific inservice examination prcvisions are: the reactor pressure vessel and a number of the piping and component .
support systems. 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 which are impractical .to implement.
The development of new or improved examination techniques will continue to be monitored. As improvements in these areas are achieved, the NRC may require that these techniques be incorporated in the next inspection interval ISI program plan examination requirements.
1 Based on the review of the Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval Inservice Inspection Program Plan, Revision 0, Change 2, the Licensee's response to the NRC's. Request for Additional Information, and the recommendations for granting relief from the ISI examination.
i requirements that have been determined to be impractical, it has been concluded that the Enrico Fermi Atomic. Power Plant, Unit 2, First 10-Year 37 l
l l
Interval Inservice Inspection Program Plan, Revision 0, Change 2, with the
. exception of Request for Relief RR-C1, is acceptable and in compliance with '
i l
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l l
l I
l .
l 1
38 1
. 5. REFERENCES
, 1. Code of Federal Regulations, Volume 10, Part 50.
- 2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Division 1:
1 1980 Edition through Winter 1981 Addenda
-1974 Edition through Summer 1975 Addenda {
q 4
- 3. Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval l Inservice Inspection Program Plan, Revision 0, Change 2, submitted April 2, 1987.
1 l 4. NUREG-0800, Standard Review Plans, Section 5.2.4, " Reactor Coolant i l Boundary Inservice Inspection and Testing," and Section 6.6, " Inservice l Inspection of Class 2 and 3 Components," July 1981.
- 5. Letter, dated March 27, 1987, J. J. Stefano (NRC) to B. R. Sylvia-
[ Detroit Edison Company (DEC)], " Request for Additional Information for
]
Review of the Enrico Fermi Atomic Power Plant, Unit 2, Fir t 10-Year Interval Inservice Inspection Program Plan." '
6.
Letter, dated June'1, 1987, F. E. Agosti (DEC).to Document Control Desk (NRC), " Additional Information on First 10-Year Interval ISI Program Pl an . " '
- 7. NUREG-0313, " Technical Report on Material Seldction and Processing
]
Guidelines for BWR Coolant Pressure Boundary Edoing," Revision 1, 1 July 1980.
- 8. IE Bulletin 80-07, Supplement 1, "BWR Jet Pump Assembly Failure,"
May 3, 1980.
- 9. NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line
' )
Nozzle. Cracking", November 1980.
I
- 10. NUREG-0803, " Generic Safety Evaluation Report Regarding ' Integrity of i BWR Scram System Piping," August 1981.
39 ,
I
l l
4 t
- 11. Regulatory Guide 1.147, Revision 5, " Inservice Inspection Code Case
+
j
, Acceptability, ASME Section XI Division 1," August 1986. j i
1-l 1
i l
l I
l l
l 1
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. j .
onc ,cau sas w s =wcts.n atowL.,cav cowess:ow at.o.Y %wws t . .. .,*e e. TG c es.r e ** < ean
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, Eo'.*EE BIBUOGRAPHIC DATA SHEET EGG-SD-7782 in ~1.ece.o , o~ ,. . .n
, , . r . . . ,; . . , r . , , . . . , . . . 8 Technical Evaluation Report on the First 10-Year Interval Inservice Inspection Program Plan: Detroit Edison Company, Enrico Fermi Atomic Power Plant, . o.re . ire.. co ..cio Unit 2, Docket Number 50-341 -o~,- u.a j l
i .e-c.<s. August 1987 j
. 2., ...o.n na s ,
~~'- l
- B.W. Brown, J.D. Mudlin l
's..
August 1987 i EG&G Idaho, Inc.
"o"e".'"*"".
P. O. Box 1625 '
Idaho Falls, ID 83415 FIN-06022 (Project 5)
'O SP a%5 .*G 0 3.% :.14 % %.vt .%C M.i..NG .0;.415 %,es. Le Case, tis r.eg gs .t.c.7 Materials Engineering Branch Technical 3 Office of Nuclear Reactor Regulation , , , , , , , , , , , _ , , , , _
j U.S. Nuclear Regulatory Commission j Washington, D.C. 20555 ll
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, ..s , . .c, 2x .-. . .
This report presents the results of the evaluation of the Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval Inservice Inspection (ISI) Program Plan through Revision 0, Change 2, submitted April 2, 1987, including the requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code l Section XI requirements which the Licensee has determined to be impractical. The !
Enrico Fermi Atomic Power Plant, Unit 2, First 10-Year Interval ISI Program Plan is l evaluated in Section 2 of this report. The ISI Program Plan is evaluated for l' (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) exclusion criteria, and (d) compliance with ISI related-commitments identified during the Nuclear Regislatory Commission (NRC) review before granting an Operating License. The requests for relief from the ASME Code requirements which 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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