ML17265A494
| ML17265A494 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/18/1998 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9812290106 | |
| Download: ML17265A494 (19) | |
Text
CATEGORY j.
REGULAT KY'NFORMATION DISTRIBUTIOh SYSTEM (RIDS)
ACCESSION NBR:9812290106 DOC.DATE: 98/12/18 FACIL:50-244 Robert Emmet Ginna Nuclear Plant, NOTARIZED: NO Unit 1, Rochester G DOCKET 05000244 I
AUG'. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp.
RECIP'.NAME RECIPIENT AFFILIATION VISSING,G.S.
SUBJECT:
Requests approval for use of relief request number 36 concerning ASME Section Category B-F,to address surface examinations of identified Class 1 nozzle-to-safe end welds associated with reactor pressure vessel.
DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code GL-89-04 E
NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 0500024$
RECIPIENT COPIES RECIPIENT COPIES 0 ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 LA 1 1 PD1-1 PD 1 1 VISSING,G. 1 1 INTERNAL: AEOD/SPD/RAB 1 1 CENTER 01 1 1 NRR/DE/ECGB 1 1 NUDO ABSTRACT 1 1 OGC/HDS3 1 -
0 RES/DET/EIB 1 1 RES/DET/EMMEB 1 1 EXTERNAL: LXTCO ANDERSON 1 1 NOAC 1 1 NRC PDR 1 1 D 0
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NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRZBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415"2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 13 ENCL 12
AND
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ROCHE$ 1ER GAS ANDElEClFIC CORR784llON ~ 89 EAS1'VEN1JE, ROCHES1ER, N. Y 148494%1 AREA CODE71d 5'-270D ROBERT C. MECREDY Vice Presirtent Nvcleor 0 perotions December 18, 1998 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Guy S. Vissing Project Directorate I-1 Washington, D.C. 20555
Subject:
Inservice Inspection Program ASME Section XI Required Examinations Third 10-Year Interval Request for Relief Regarding Request No. 36 R.E. Ginna Nuclear Power Plant Docket No. 50/244
Reference:
(a) Letter from W.R. Butler (NRC) to R.C. Mecredy (RGE(E), dated September 8 1993r
Subject:
Relief Request No. 19 Dear Mr. Vissing The purpose of this letter is to seek approval for the use of Relief Request number 36 concerning ASME Section XI Category B-F, to address surface examination limitations associated with weld examinations of identified Class 1 nozzle-to-safe end welds associated with the Reactor Pressure Vessel.
This Relief is requested pursuant to the provisions of 10 CFR 50.55a(g)(5)(iii), the required examination coverage for the identified welds is impractical and would require redesign or replacement to obtain Code required surface examination coverage.
Justification and the proposed alternative are included in the attachment to this letter. It is requested that this relief request be expedited, and NRC reply obtained before the end of January, 1999, in order for it to be utilized at R.E. Ginna Nuclear Power Plant for the upcoming March 1999 outage.
Very truly yours, Robert C. Mecred 981229010b 981218 PDR ADQCK 05000244 PDR
Attachments: 3 XC ~ Mr. Guy S. Vissing (Mail Stop 14B2)
Project Directorate I-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector
ATTACHMENT 1 Rochester Gas and Electric Corporation Ginna Station Docket No. 50/244 Third 10-Year Interval Request for Relief No. 36 Reactor Pressure Vessel Nozzle-to-Safe End Butt Weld Surface Examination Limitations I. System/Component(s) for Which Relief is Requested:
This Relief Request is requested for six (6) Reactor Pressure Vessel Nozzle-to-Safe End Butt Welds. Inspection of these welds is addressed under Class 1, Category B-F, Item Number B5.10, Nozzle-to-Safe End Weld Surface Examinations as identified below.
Weld ID ISI Summar Covera e Obtained PL-FW-II 002100 744 PL-FW-V 002400 76%
PL-FW-IV 002700 704 PL-FW-VII 003000 76>o AC-1003-1 003300 0~~~ (*)
AC-1002-1 003600 0~~~ (*)
Note: (*) = welds embedded in concrete.
II. Code Requirement:
Under Category B-F, Item Number B5.10, volumetric and surface examinations shall be performed with essentially 1004 of the weld length to obtain code coverage. ASME Section XI Code Case N-460 states that if examination volume or area cannot be examined due to the entire interference by another component or part geometry, a reduction in coverage is acceptable provided that the (lack of) coverage is less than 10%. Previous Codes utilized did not include this 90% coverage requirement and examinations were performed to the extent obtainable.
III. Code Requirement from Which Relief is Requested:
Relief is requested from the surface examination requirements for the six (6) identified welds.
Request for Relief No. 36 Page 1 of 14
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Surface Examination of the first four (4) welds is limited due to Original Construction Code interferences of the floor and wall in the "Sandbox" where these welds are located.
The "Sandboxes" would have to be redesigned to enable the welds to be surface examined to obtain Code required coverage. (Volumetric examination of these welds is performed from the inside of the Vessel and is not a part of this Relief Request.)
Surface Examination of the last two welds is impractical.
The concrete surrounding the Reactor Pressure Vessel has embedded these welds. The concrete wall around the Reactor Pressure Vessel would have to be redesigned or replaced to enable the two (2) welds to be inspected with a surface examination. (Volumetric examination of these welds is performed from the inside of the Vessel and is 'not a part of this Relief Request.)
IV. Basis for Relief:
Relief is requested pursuant to the provisions of 10 CFR 50.55a(g)(5)(iii), the required examination coverage for the identified welds is impractical and would require redesign or replacement to obtain Code required surface examination coverage.
R.E. Ginna Nuclear Power Plant was designed and constructed to the B31.1, 1955 edition Construction Code. This code did not contain requirements to ensure that items be accessible for future examinations. The above noted piping welds were installed utilizing this construction code, which did not provide for accessibility for future ISI NDE. Due to the limited design accessibility, ISI surface examination coverage is below Code percentage requirements as identified within this Relief Request.
The first four (4) welds of this Relief Request are located in a "Sandbox" configuration. Within the "Sandbox", the welds are against the floor and one wall. The angled wall is joined to the floor and is against the weld. The surface examination of these welds is limited due to Original Construction Code interferences of the floor and wall of the "Sandbox". The "Sandboxes" would have to be redesigned to enable the welds to be inspected to obtain Code required coverage for the surface examinations. The attached sketch (Attachment 2) shows a representative weld with similar interferences.
Request for Relief No. 36 Page 2 of 14
4 The last two (2) welds of this Relief Request are embedded in concrete. This concrete structure is the wall that surrounds the Reactor Pressure Vessel.
ASME Section XI Class 1 system leakage examinations are performed. These leakage examinations demonstrate pressure boundary integrity and provide additional assurances in maintaining plant safety.
V. Alternate Examinations:
R.E. Ginna Nuclear Power Plant proposes that the surface examination coverage identified for the first four (4) welds above be acceptable in fulfilling the Code required examination coverage. The actual physical configuration of the "Sandboxes" is not conducive in obtaining the requirements specified within Code Case N-460 for acceptable coverage. Volumetric examination of these welds is performed from the inside of the Vessel, and will be performed during the 1999 outage.
For the last two (2) welds, the Code surface examination requirements are impractical and cannot be examined due to them being embedded in concrete. Volumetric examination of these welds is performed from the inside of the Vessel, and will be performed during the 1999 Outage.
VI. Justification for the Granting of Relief:
R.E. Ginna Nuclear Power Plant was designed and constructed to the B31.1, 1955 edition Construction Code. This code did not contain requirements to ensure that items be made accessible for future NDE examinations. Due to the original limited design accessibi.lity or lack of design accessibility, ISI surface examination coverage can not be obtained to the extent required by the ASME Code.
ASME Section XI Class 1 system leakage examinations are performed. These leakage examinations demonstrate pressure boundary integrity and provide additional assurances in maintaining plant safety. The identified examination coverage for these items should be acceptable in fulfilling ASME Section XI coverage requirements.
It should also be similar to RG&E s noted that Relief Request Number 36 is Relief Request Number 19, for which relief was previously granted. See Attachment 3.
Request for Relief No. 36 Page 3 of 14
VII. Implementation Schedule:
The surface examinations will be performed during the 1999 Outage on the first four (4) accessible welds. For the two (2) welds embedded in concrete, surface examinations can not be performed. The associated volumetric examinations will be performed during the 1999 Outage. Applicable Code credit shall be taken for the Third 10-year Interval inspection, upon approval of this Relief Request.
Request for Relief No. 36 Page 4 of 14
Attachment 2 fghiiii&LQL EXAMINATION ARE A LIMITATION ( IF NONE, SO STATE); .
006 tO Cnak&AhlieA E Flan'a&aQRll gg Qg~ ~V gpss QOTKJK)i(i& iQ QCMidG REVIEWED SY r~-/PE Fp TONS NO. Sr.R. Z iiOYN <7 II I Y. i ) Pal Request for Relief No. 36 Page 5 of 14
RG0.3.4 4 7 1 Current as of,1998/12/07
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I 0 0 UNITED STATES
~e r NUCLEAR REGULATORY COMMlSSlOhl Ovl Vr WASHINOTON, O.C. 206~at yO
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September 8, 1993 Y..~P<<~ +P~
Docket No. 50-244 ~i,M<~~ >g ~'Aa Or. Robert C. Mecredy
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Vice President, Nuclear Production Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649 Oear Or. Hecredy:
SUBJECT:
R. E. GINNA NUCLEAR POMER PLANT - THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAN PLAN AND ASSOCIATED REQUESTS FOR RELIEF (TAC NOS.
M84044 AND M86225)
Sy letter dated July 21, 1989, you submitted the R. E. Ginna Nuclear Power Plant Third 10-Year Interval lnservice Inspection Program Plan. The NRC concluded in its safety evaluation (SE) dated August 6, 1990, that the Program Plan, with the exception of Requests for Relief Nos. 10 and 13, was found to be acceptable and in compliance with the regulations.
Subsequent to the NRC's review above, you submitted two revisions of the Program Plan. Revision 1 was submitted in a letter dated August 10; 1992, and Revision 2 was submitted by letter dated January 25, 1993. The January 25, 1993, submittal of the Program Plan was reformatted for ease of use. The reformatted program consisted of eleven independent sections, each carrying its own revision number.
In your letter dated January 5, 1993, you submitted Relief Request (RR) No. 19 and notified the NRC of the intent to incorporate Code Cases N-460 and N-498 into the Program Plan.
The NRC staff, with the assistance of its contractor, Idaho .National Engineering Laboratory (INEL), reviewed and evaluated your submittal dated January 5, 1993, and concluded that pursuant to 10 CFR 55.55a{g)(6)(i), relief can be granted as requested for RR No. 19.
Regarding Revision 2 and the January 25, 1993, suhaittal of the Program Plan, the staff has concluded that your response regarding the removal of insulation during pressure testing at bolted connections in piping systems used for controlling boration is still considered unacceptable, and you should either withdraw Request for Relief No. 13 or acknowledge it as being unacceptable -by the staff in the Program Plan.
Therefore, due to inadequate VT-2 visual exaIIinations of the bolted connections in borated systems, and Request for Relief No. 13 not having been Request for Relief No. 36 Page 6 of 14
of 1998/12/07 RG0.14 4 73..
~ Current as Page 2 of 9 Robert C. Mecredy September 8, 1993 withdrawn, the staff concludes that the Revision 2 submittal of January 25, 19S3, is not in compliance with 10 CFR 50.55a(g) and Technioal Specification 4.2.1.5 and is therefore unacceptable. The staff's evaluation and conclusions are contained in the attached SE.
Sincerely, Malt R. Butler, Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
Safety Evaluation cc w/enclosure:
See next page Request for Relief No. 36 Page 7 of 14
Current 'ot 1998/12/07 of RGO'" 447<<L
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~ Page 3 9 Dr. Robert C. Hecredy R.E. Ginna Nuclear Power Plant CC:
Thomas A. Hoslak, Senior Resident Inspector R.E. Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road Ontario, New York 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Hs. Donna Ross Division of Policy Analysis 8 Planning New York State Energy Office Agency Building 2 Empire State Plaza Albany, New York 12223 Charlie Donaldson, Esq.
Assistant Attorney General New Yor k Department of Law 120 Broadway New York, New York 10271 Nicholas S. Reynolds Winston h Strawn 1400 L St. N.W.
Washington, OC 20005-3502 Hs. Thelma Wideman Director, Wayne County Emergency Hanagement Office Wayne County Emergency Operations Center 7370 Route 31 Lyons, New York 14489 Hs. Vary Louise Heisenzahl Administrator, Honroe County Office of Emergency Preparedness 111 West Fall Road, Roon ll Rochester, New York 14620 Request for Relief No. 36 Page 8 of 14
1447,1 Current as of 1998/12/Q7 RGO
~ Page Enclosure 4 of 9
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vl r NUCLEAR REGULATORY COMM(SS) ON V/ Oy WASMINGTON, D.C. 205$ 5-0001
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SAF U 0 BY TH CL O U O OF D 10-Y I V S V CE INSP C ON PROG ROCH ST GAS AND C C CO T 0 R.. G I. ~RT Technical Specification 4.2.1.5 for the R. E. Ginna Nuclear Power Plant states that the fnservice inspection and testing of the American Society of Mechanical Engineers (ASHE) Code Class 1, 2, and 3 components shall be performed in accordance with Sectfon XI of the ASHE Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(f). The Code of Federal Regulations of 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed, alternatives would provide an acceptable level of quality and safety, or (if) compliance with the specified requfrements would result in hardship or unusual dffffculty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASHE Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASHE Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and systei pressure tests conducted during each 10-year interval comply with the requirements in the latest edition and addenda of Section XI of the ASHE Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month inspection interval, sub5ect to the limitations and modifications listed therein. The applicable Edition of Section XI of the ASHE Code for the R. E. Ginna Nuclear Power Plant Third 10-Year Interval is the 1986 Edition, no addenda. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASHE Code incorpor'ated by reference in 10 CFR 50.55a(b) sub5ect to the limitations and modifications listed therein.
Request fox Relief No. 36 Page .9 of 14
RGB>447' Current as oE 1998/12/07 'age 5 of 9 Gy letter dated July 21, ]989, Rochester Gas and Electric Corporation (the licensee) submitted the R. E, G1nna Nuclear Power Plant Third 10-Year Interval Inservice Inspect1on Program Plan. In a Safety Evaluat1on Report (SER) dated August 6, 1990, the staff found the Program Plan, with the exception of Requests for Relief Nos. 10 and 13, acceptable and in compl1ance with the regulations.
Revision 1 of the R. E. G1nna Nuclear Power Plant Third 10-Year Interval Inservice Inspection Program Plan was submitted in a letter dated August 10, 1992, and a subsequent rev1sion was submitted by letter dated January 25, 1993. The January 25, 1993, submittal of the Program Plan was reported to have been reformatted for ease of use, and supersedes the previous submittals in their entirety. The reformatted Program consists of eleven (11) independent sect1ons, each of which carries its own revision number and may be revised separately.
In a letter dated January 5, 1993, the licensee suhiitted Relief Request (RR)
No. 19 and notified the NRC of the intent to incorporate Code Cases N-460 and N-498 into the Program Plan. These items are also addressed in the following sect1on.
The staff, with technical assistance from 1ts Contractor, the Idaho National Engineering Laboratory (INEL), has evaluated the R. E. Ginna Nuclear Power Plant Third 10-.Year Interval Inservice Inspection, Program Plan, as submitted January 25, 1993, and the January 5, 1993, sutxaittal wh1ch 1ncludes RR No. 19.
The results are reported below.
2.0 ~VA QQQg The following are the major changes that have been incorporated into the January 25, 1993, rev1sion of the R. E. Ginna Nuclear Power Plant Thir'd 10-Year Interval Inservice Inspection Program Plan:
(a) R e ds 8ased on the Licensee's use of ASIDE Code Case N-481, "Alternative Examination Requirints for Cast Austenitic Pump Casings,Section XI, Division 1,'R No. 4 1s no longer required and was withdrawn 1n Section 2 of the revised Program Plan. Code Case N-481 is acceptable for general usage as it is referenced in NRC Regulatory Guide 1.147, Revision 9, "Inservice Inspection Code Case Acceptability ASHE Section XI, Division 1."
t In (b) a letter, A. Johnson (NRC) to Dr. R. C. Hecredy (RGKE), dated June 16, 1992, the NRC requested that the licensee confirm that all duties were be1ng performed by an ANII as required by the Code. In the response dated August 17, 1992, [Dr. R. C. Hecredy (RGKE) to Document Control Desk (NRC)], the 11censee coamitted to contract with the Request for Relief No. 36 Page 10 of 14
RGO 1 4 4 7 l ~ Current as of 1998/12/07 ~ Page 6 of 9 2
3 Hartford Steam Bo1ler Inspect1on and Insurance Company for services of an ANII for the Third 10-Year Inspection Interval. The licensee stated that the ANII will perform all required Code duties 1n accordance with IN-2110. Consequently, the licensee withdrew RR No. 3 (Use of an Authorized Inspection Agency to Provide Inspection Services) in Section 2 of the revised Program Plan.
(c) The NRC's June 16, 1992, letter to the licensee also addressed the use of NIS-1 (Owner's Report for Inservice Inspections) and NIS-2 (Owner's Report for Repairs or Replacements) forms. These forms are specified fn Mandatory Appendix II of ASHE Code Section XI. IMA-6220(d)(10) states that the NIS-1 and NIS-2 forms shall be included fn the requ1red lnserv1ce Inspection Suaeary Report and that they include the signature of the ANII. Therefore, Section 1.6,1 of the Program Plan, fn the latest revision, has been rev1sed to include use of the NIS-1 and NIS-2 forms. Section 1.6.1 now states:
"An Inservice Inspection Report shall be generated to document applicable 1nservice inspection and associated repair, replacement and modification activities. ASME NIS-1 and NIS-2 forms shall be generated and included with1n the Inservice Inspection Report."
(d) m s in e testi : In the June 16, 1992, letter to the licensee, the staff did not agree w1th the 11censee's basis for limit1ng the extent of removal of insulation to inspections at bolted connections with ferrous steel fasteners. A non-fsolatable leak could occur anywhere fn the piping systems used for controlling borat1on regardless of fastener material types. Therefore, the licensee was requested to satisfy the Code requirements regarding VT-2 v1sual examinations at bolted connections.
In the response dated August 17, 1992, the licensee agreed with the staff's evaluat)on and stated that paragraph 1.10.3.2 would be revised to require the retloval of insulat1on for inspection of both ferrftic and austenft1c bolting. Sect1on 1.10.3.2 of the January 25, 1993, Program Plan submittal was revised to state, fn part, that:
"Insulat1on d Ig ~,rival tests are intended to dur1ng the VT-2 exam1nat1on is not required, however, fn accordance with IMA-5242(a), systems borated for the purpose of controlling reactivity shall have insulation removed at bolted connections during conduct of the VT-2 exam1natfon.
1s only applicable to those VT-2 examinations performed be 1 L kg,F tl 1 di Th1s'equirement non-intrusive type tests. At Ginna, this requirement 1s considered to be applicable to borated lines only fn the primary flow path of piping from the boric acid supply and CVCS Charging to the Reactor Vessel and return through CVCS Letdown, and Request for Relief No. 36 Page 11 of 14
RGO ~. 4471 Current as of 1998/12/07 ~ Page 7 of is not applicable to branch lines. connected to the primary flow path."
The staff considers this response unacceptable. The licensee 1s not intending to perform any hydrostat1c test1ng based on the use of ASME Code Case N-498, Alternative Rules for 10-Year Hydrostatic Pressure Testing for Class 1 and 2 Systems, Sect1on XI, Division 1.
Additionally, Sect1on XI of the Code requires the removal of insulation at bolted connections on all systems that contain borated ~ater during the conduct of a VT-2 visual examination. Th1s does not exclude VT-2 visual examinations during funct1onal or inservice tests. ASME Code Interpretation XI-1-89-38 supports th1s conclusion and should be referenced if further clar1fication of this requir'ement is necessary.
For the January 25, 1993, revision of the Program Plan to be considered acceptable, the licensee must meet the Code requirements regarding bolted connect1ons on systems containing borated water.
(e) As stated in Sect1on 1.0 of this report, the staff denied Requests for Relief Nos. 10 and 13 in the SER dated August 5, 1990. It 1s noted in the January 25, 1993, revision of the Program Plan that Request for Relief Ho. 10 has been withdrawn. However, Request, for Relief No. 13 has not been withdrawn and appears to be applicable for the current 10-year inspection 'interval. Request for Relief No. 13 should either be withdrawn, or acknowledged 1n the Program Plan as be1ng NRC unacceptable.
The follow1ng evaluations address the January 5, 1993, letter not1fying the staff of the licensee's intent to incorpqrate Code Cases N-460 and N-498 and the submittal of RR No. 19.
ASME Code Cases H-460 and N-498 have both been approved for use by reference in Regulatory Guide I.147, Revision 9, "Inservice Inspection Code Case Acceptability ASME Section XI D1vis1on I", dated April 1992.
ief a C C I tl Xi. Tdl tllC-2500-l. E Category C-B, Item C2.21 requires a 100% surface and volumetric examination of nozzle-to-shell welds on nozzles in vessels with nominal wall thickness pl/2 inch. Item C2.22 requires a 100% volumetric examinat1on of the inside radius sections of the nozzles. These examinations are to be performed as def1ned by Figures IMC-2500-4(a) or (b) as applicable.
Request for Relief No. 36 Page 12 of 14
pGpg 4 4'7el Page 8 of 9 R11f\ df p 0 I t f volumetric examinations to the extent required by the Code for the following charging system pulsation dampener nozzle welds and inside radius sections:
~No CF-Nl
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RHUhifmd, Q~e29ft PT 66%
UT 65%
CF-N3 PT >90%
UT 80%
tin The pulsatfon dampener contains three (3) nozzles, fn line, located at the bottom of the unit. The outboard nozzle fs identified as CF-Nl. Between this nozzle and the nozzle (CF-N2) fs a support that covers from the edge of one 'iddle nozzle's weld heat affected zone to the edge of the other nozzle's weld heat affected zone. There is only 7/B inch between CF-N2 and the third nozzle (CF-N3) heat affected zone.
t te None. The Code-required surface and/or volumetrfc examinations will be performed to the maximum extent practical.
The R. E. Ginna Nuclear Power Plant was constructed to the 1955 Edition of ANSI 831.1. This Code dfd not contain requirements to ensure that items be accessible for future examinations. The pulsation dampener was constructed and installed in the early 1970s, and the construction code did not require provisions for accessibility for fnservice inspections. Due to the close proximity of the nozzles and/or the vessel support, the associated surface and volumetric examinations are fmpractfcal to perform to the extent require by the Code. The identified surface and volumetric examination coverage of 66% to >90%
should be considered acceptable for these nozzles at R. E. Gfnna Nuclear Power Plant.
8ased on the above evaluation, it is concluded that since the original construction code did not specify accessibility requirements for future ISI HDE, compliance with the Code for these nozzle examinations is impractical. Imposition of the surface and volumetric examinations, to the extent required by the Code, would necessitate redesign or replacement of the charging system pulsation dampener and result fn hardship or unusual difficulty without a compensating increase fn the level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), relief fs granted as requested.
Request for Relief No. 36 Page 13 of 14
RGS> 447/. Current as of 1998/12/07 Page 9 of 9 Paragraph 10 CFR 50.55a(g)(4) requires that components (including supports) that are classif1ed as ASHE Code Class 1, 2, and 3 meet the requirements, except design and access provisions and preservice requirements, set forth in applicable Editions of ASNE Section XI to the extent practical within the limitations of design, geometry, and materials of construction of the components.
Pursuant to 10 CFR 50.55a(g)(5)(111), the licensee determined that conformance with certain Code requirements is impractical for the1r facility and sulxlitted support1ng techn1cal justification. The staff has reviewed the licensee's submittal, dated January 5, 1993, and has concluded that pursuant to 10 CFR 55.55a(g)(6)(1) relief can be granted as requested for RR No. 19. Such relief is authorized by law and will not endanger life, property, or the common defense and security, and is otherwise in public interest. This relief is being granted giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Regarding Revision 2 and the January 25, 1993, submittal of the Program Plan, the staff has concluded that the licensee has adequately addressed the deficiencies cited tn the June 16, 1992, letter frea the NRC regarding the 11censee's use of an ANII and the NIS-I and NIS-2 forms. However, as addressed above, the licensee's response regarding the removal of insulat1on, during pressure testing, at bolted connections in piping systems used for controlling boration is st111 considered unacceptable. In addition, the licensee should e1ther withdraw Request for Relief No. 13 or acknowledge it as being unacceptable to the NRC in the Program Plan.
Based on inadequate VT-2 visual examinat1ons of bolted connections in borated systems, and Request for Relief No. 13 not hav1ng been withdrawn, the staff concludes that the R. E. Ginna Nuclear Power Plant Third 10-Year Interval Inservice Inspect1on Program Plan, Rev1sion 2, as submitted January 25, 1993, is not in compliance with 10 CFR 50.55a(g) and Technical Specif1cation 4.2.1.5 and is therefore unacceptable.
Principal Contributors: T. McLellan H. Khanna Oate:
Request for Relief No. 36 Page 14 of 14