ML18039A837
| ML18039A837 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/06/1999 |
| From: | Abney T TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML18039A838 | List: |
| References | |
| NUDOCS 9908130128 | |
| Download: ML18039A837 (168) | |
Text
REGULATu INFORMATION DISTRIBUTIO YSTEM (RIDS)
ACCESSION NBR:9908130128 DOC.DATE: 99/08/06 NOTARIZED: NO FACIL:50-260 .Browns Ferry Nuclear Power Station, Unit 2, Tennessee DOCKET 05000260 I
50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME AUTHOR AFFILIATION ABNEY,T.E. Tennessee Valley Authority RECIPIENT AFFILIATION 'ECIP.NAME Records Management Branch (Document Control Desk)
SUBJECT:
Forwards "BFN Unit 2 Cycle 10 ASME Section XI NIS-1 6 NZS-2 Data Repts," for NRC review. Corrected inservice insp summary rept for Unit,.3 cycle 8 operation, included in rept.
DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR 1 ENCL J SIZE:
TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code E
NOTES:
RECIPIENT, COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL LPD2-2 LA 1 1 LONG, W 1 1 INTERNAL: ACRS 1 1 LE CENTER 01 1 1 NUDOCS-ABSTRACT 1 1 OGC/RP 1 0 RES/DET/ERAB 1 1 RES/DET/MEB 1 1 EXTERNAL: LITCO ANDERSON 1 1 NOAC 1 1 NRC PDR 1 1 D'
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NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE.'TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION 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 COPXES REQUIRED: LTTR 11 ENCL 10
I Tennessee Valley Authority, Post Office Box-2000, Decatur, Afabama 35609-2000 August 6, 1999 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of ) Docket Nos. 50-260 Tennessee Valley Authority , ) 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 2 and 3 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI, INSERVICE INSPECTION AND REPAIR AND REPLACEMENTS PROGRAMS-
SUMMARY
REPORTS FOR CYCLE 10 OPERATION AND CORRECTED INSERVICE INSPECTION
SUMMARY
REPORT FOR UNIT 3 CYCLE 8 OPERATION In accordance, with paragraphs IWA-6220 and IWA-6230 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, TVA is submitting the BFN Unit 2 outage summary reports for NRC review. The, summary reports are for inservice inspection (NIS-1 Report, in accordance with 1986 Edition, no addenda), and repair and replacement (NIS-2 Report, in accordance with 1989 Edition, no addenda) activities for Unit 2 Cycle 10 operation. The examinations, repairs, and replacements for Unit 2 Cycle 10 operation (with the exception described below) were completed in accordance with applicable Code requirements except where specific'written .relief has been granted.
TVA has determined that the reactor pressure vessel support skirt weld, ASME Section XI Code Category B-H, Item B.8.10, is accessible for surface (i.e., magnetic particle) examination only from the outside surface. Consequently,
'TVA plans to submit a request for relief to address the inaccessible weld examination area.
TVA is also submitting corrected NIS-1 summary report data (Enclosure 1, Attachment 3) for the Main Steam System valve, 3-FCV-1-037, examined during Unit 3 Cycle 8 operation. The Unit Cycle3 8 NIS-1 Summary Report was submitted to NRC by TVA letter dated January 12, 1999. The valve received an internal examination, ASME Section XI, Code Category B-M-2,
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'tl'tl08130 050002b0 PDR 8 ,
ADQCK PDR
4> ~ i 4 U.S. Nuclear Regulatory Commission Page 2 August 6, 1999 Item B12.50, as documented on report R-223. This examination was actually a preservice examination. The Unit 3 Cycle 8 NIS-1 'Summary Report has been revised to delete the above mentioned component as an inservice inspection examination and is now listed as a preservice examination. contains the BFN Unit 2 Inservice Inspection Summary. Report (NIS-1) for Code Class 1 and 2 pressure retaining components and their supports. Enclosure 2 contains the Repair and Replacements Summary Report (NIS-2) for'ode Class 1 and 2 components and supports.
If you have any questions, regarding these me at (256) 729-2636.
reports, please contact Si cerel T. E. Abney Manager of z.cens'ng and In stry Af airs Enclosures cc (Enclosures):
Mr. Paul E. Fredrickson, Branch Chief U.S. Nuclear Regulatory Commission Region II 61 Forsyth Street, S.W.
Suite 23T85 Atlanta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. William O. Long, Senior Project Manager U-.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852
t
- 43 .4l 5
Distri72.txt 5'p ~~
Priority: Normal Distribution Sheet From: Stefanie Fountain Action Recipients: Copies:
RidsNrrPMWLong 0 OK RidsNrrPMRHernan 0 OK Internal Recipients:
RidsRgn2MailCenter 0 OK RidsNrrDipmEphp 0 OK RidsManager 0 OK IRO D Hagan FIL"E CENT.ER:01" Dennis Hagan
~ 1 0
1 Paper Copy Paper Copy-OK External Recipients:
NOAC Paper Copy Total Copies:
Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003699530:1
Subject:
TVA NUCLEAR (TVAN) RADIOLOGICALEMERGENCY PLAN (REP) REVISION 54 Body:
ADAMS DISTRIBUTION NOTIFICATION.
Electronic Recipients can. RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003699530.
A045-OR Submittal: Emergency Preparedness Plans, Implementing Procedures, Corre spondence Page 1 APR >0 M>>
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Distri72.txt Docket: 05000260 Docket: 05000296 Docket: 05000327 Docket: 05000328 Page 2
0 '4l Tennessee Valley Authortty, 1101 Market Street, Chattanooga, Tennessee 37402-2801 March 31, 2000 10 CFR 50 Appendix E 10 CFR 50.54'(q) 10 CFR 50.47 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:
In the Matter of ) Docket Nos. 50-260 50-327 Tennessee Valley Authority ) 50-296 50-328 TVA NUCLEAR (TVAN) RADIOLOGICAL EMERGENCY PLAN (REP) REVISION 54 In accordance with the requirements of 10 CFR Part 50, Appendix E, Section V, and 10 CFR 50.54(q), enclosed is Revision 54 dated March 21, 2000, to TVA's consolidated TVAN REP. TVA has determined that the changes do not decrease the effectiveness of the plan.
The plan, as changed, continues to meet the requirements of 10 CFR 50, Appendix E, and 10 CFR 50.47. Also enclosed is an effective page listing and revision log.
The enclosed information is being sent by certified mails The signed receipt signifies that you have received this information and will be taken as verification that the NRC copies of the plan have been updated, and the superseded material has been destroyed.
If you have any questions, please contact Rob Brown at (423) 751-7228 or Terry Knuettel at (423) 751-6673.
Sincerely,
/Mark J.. Burzynzki Manager Nuclear Licensing Enclosures cc: See page 2 pog
U.S. Nuclear Regulatory Commission Page 2 March 31, 2000 cc: Mr. W. O. Long, Senior Project Manager (Enclosures)
U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. R. W. Hernan, Senior Project Manager (Enclosures)
U.ST Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. R. E. Martin, Senior Project Manager (Enclosures)
U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr'. Luis Reyes, Regional Administrator (
Enclosures:
2)
U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303-3415 NRC Resident Inspector (
Enclosures:
Supplied by BFN's Browns Ferry Nuclear Plant Document Control) 10833 Shaw Road Athens, Alabama 35611 NRC Resident Inspector (
Enclosures:
Supplied by SQN's Seguoyah Nuclear Plant Document Control) 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379-3624 NRC Resident Inspector (
Enclosure:
Supplied by WBN's Watts Bar Nuclear Plant Document Control) 1260 Nuclear Plant Road Spring City, Tennessee 37381
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DOCUMENT RELEASE AND FILING INSTRUCTIONS Page 'I of 1 Release No.
To: Management Services/RIM/EDM Prepared By; Gait White Other Address; Extension: 751-2108 Date Submitted to Management Organization: AS&P Services/RIM/EDM: Address: LP 4D-C Date to Filed By: ASAP Attached are: (select one)
H QA Records/Documents Non-QA Records/Documents I
Release and Submitted for:
Distribution H Retention REC NO. ACCPT REMOVE INSERT DOCUMENT NUMBER REV PAGES N DATE PAGES PAGES Radiological Emergency Plan Title Page Title Page Title Page List of Effective Pages 03-21-00 Rev. Log 03-21-00 1-7
'able of Contents I - viii I vill List of Figures ix 54 188 AII 1 -188 Document No(s) Rejected/Reason: Acceptance:
Date X~u A'. f~meL 3->>~o Signature Date
Contact:
Exf, TVA 40074B (8-97) (08-97) Page1 of1 SPP-2.3-3 (08-29-97)
0 0 'I ill if'
S-~e Distri46.txt Distribution Sheet Priority: Normal From: Stefanie Fountain Action Recipients: Copies:
RidsNrr PMWLong 0 OK RidsNrrLABClayton 0 OK RidsNrrDlpmLpdii2 0 OK Internal Recipients:
.RidsRgn2MailCenter OK RidsOgcRp OK RidsNrrWpcMail 0 OK RidsNrrDssaSrxb 0 OK Rids Manager OK RidsAcrsAcnwMailCenter 0 OK
. IL CENTER 0 Paper Copy AC S Paper. Copy External Recipients:
NOAC Paper Copy Total Copies:
Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003700990:1
Subject:
Browns. Ferry,2 Revision 16 to Technical Requirements Manual Body:
ADAMS DISTRIBUTION NOTIFICATION.
Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession-.Number ML003700990.
Page 1
C Distri46.txt A001 - OR Submittal: General Distribution Docket:
05000260'age 2
II Qi j t
PROCEDURE AND INSTRUCTION CONTROL ONP DDSBP 011 TRANSMITTAL/RECEIPT ACKNOWLEDGMENT CTRA)'AGE Oi OF 01 4 ~ . 'O:
'ADDRESS:
W ~ 0 LONG> PROJ
~ MGR 11555 ROCKVILLE PIKE 1 WHITE FLT N HOLDER ¹: 005276 DCRM BFNP TRANSMITTAL NO: 000004192
'ROCKVILLE> MD 20852 TRANSMITTAL DATE: 03/30/00 SEE ATTACHED F IL'ING INSTRUCTIONS INFORMATION ONLY DCRM DOCUMENT REV REV (D) MANUAL NUMBER DATE LEVEL BFNP TR MANUAL TECHNICAL REQUIREMENTS UNIT 2 033100 STATUS: ACTIVE COPY ¹ 001 AS THE ASSIGNED DOCUMENT HOLDER FOR THE ABOVE CONTROLLED COPY NUMBER> YOU ARE RESPONSIBLE'OR FILING AND FOR MAINTAINING THESE DOCUMENTS. RECEIPT ACKNOWLEDGMENT IS NOT REQUIRED.
TVA 4 0 183 C NP. 5/90 )
- BfNP000004192'
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- R0800032980 March 29, 2000 Holders of Browns Ferry Nuclear Plant Unit 2 Technical Requirements Manual BROWNS FERRY NUCLEAR PLANT (BFN) REVISIONS TO BFN UNIT 2 TECHNICAL MANUAL (TRM) 'EQUIREMENTS Attached, is Revision 16 to the BFN Unit 2 Technical Requirements Manual which should be inserted into your controlled copy of the Unit 2 TRM in accordance with the attached instruction, sheet.
This revision applies to your "controlled" copy(ies) of the TRMs which means that each "controlled" copy must be maintained in an up-to-date condition:
If you have any questions, please call Diana Lee (256)729-7853.
~T.Manager E. Abney, of Licensing and Industry Affairs PAB 1G-BFN GMM BDL Attachment cc: EDMS, WT 3B<<K
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INSTRUCTIONS FOR UPDATlt4G BROWNS FERRY NUCLEAR PLANT UNIT 2 TECHNICAL REQUIREMENTS'MANUAL(REQUIREMENTS AND BASES)
(REVISION 16)
Remove Pa es ~lnserl Pa es TRM REQUIREMENTS EPL-1 through EPL-5, TRM Revision 14 EPL-1 through EPL-5, TRM Revision 16 3.3-55 3.3-55 3.3-56, 3.3-56 3.4-2 3.4-2 3.4-5 3.4-5 BASES EPL-1 through EPL-5, TRM Revision 15 EPL-1 thorugh EPL-5, TRM Revision 16 8 3.3-55 8 3.3-55 8 3.4-3 8'3.4-3 8 3.4-4 8 3A-4
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EFFECTIVE PAGE LISTING
.. =- .. BROV! FERRY NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL(REQUIREMENTS)
Implementation Date March 31, 2000
~Pa e Revision No. Effective Date No.'itle Page Initial Effective Page Listing Revision 16 03-31-2000 I 0 Initial II Revision 13 09-1 7-1 999 1.1-1 0 Initial 1.1-2 0 Initial 1.1-3'.1-4 0 Initial 8 04-1 3-1 999 1.1-5 0 Initial 1.1-6 0 Initial 1.2-1 0 Initial 1.2-2 0 Initial 1.2-3 0 Initial 1.3-1, 0 Initial 1.3-2 0 Initial 1.3-3 0 Initial 1.'3-4 0 Initial 1.3-5 0 Initial 1.3-6 0 In<t<al 1.3-7 0 Initial 1.3-8 .0 Initial 1.3-9 0 Initial
.1.3-10 0 Initial 1.3-11 0 Initial 1.3-12 0 Initial 1.3-13 '
Initial 1.4-1 0 Initial 1.4-2 0 Initial 1 4-3 0 In>tial 1.4-4 0 Initial
~ 1.4-5 0 Initial 3.0-1 0 Initial 3.0-2 0 Initial 3.0-3 0 Initial (continued)
BFN-UNIT 2 EPL-1 TRM Revision 16 March 31, 2000
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~ . '. - - BROW'FERRY NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL(REQUIREMENTS)
EFFECTIVE PAGE LISTING (continued)
Pa<ac No. Revision No. Effective Date 3 0Q 0 Initial 3.1-1 0 Initial 3.3-1 Revision 12 08-1 7-1 999 3.3-2 Revision 12 08-'I 7-'1 999 3.3-3 0 Initial 3.3-4 0 Initial 3;3-5 Revision 12 08-17-1999 3.3-6 Revision 12 08-17-1999 3.3-7 ~
Revision 12 08-1 7-1 999 3.3-8 Revision 12 08-1 7-1 999 3.3-9 Revision 14 12-15-1 999 3.3-10 0 Initial 3.3-1 1 0 Initial 3.3-12 0 Initial 3.3-13 0 Initial 3.3-14 0 Initial 3.3-15 0 Initial 3.3-16 0 Initial 3.3-17 0 Initial 3.3-18 0 Initial 3.3-19 0 initial 3.3-20 0 Initial 3.3-21 Revision 12 08-17-1999 3.3-22 8 04-1-3-1999 3.3-23 0 Initial 3.3-24 8 04-1 3-1 999 3.3'-.25 0 Initial 3.3-26 0 initial 3.3-27 0 Initial 3.3-28 0 Initial 3.3-29 0 Initial 3.3-30 0 Initial 3.3-31'.3-32 0 Initial 0 Initial 3.3-33 0' Initial 3.3-34 04-13-1999 3.3-35 0 Initial (continued)
BFN-UNIT 2 EPL-2 TRM Revision 16 March 31, 2000
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.... BROlAL FERRY. NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL(REQUIREMENTS)
EFFECTIVE PAGE LISTING (continued)
~Pe e No. Revision No. Effective Date 3.'3-36 0 Initial 3.3-37 0 Initial 3.3-38 0 Initial 3.3-39 0 Initial 3.3-40 0 Initial 3.3-41 0 Initial 3.3-42 0 Initial 3.3-43 8 04-1 3-1 999 3.3-44 8 04-1 3-1 999 3.3-45 0 Initial 3.3-46 0 Initial 3.3-47 0 . Initial 3.3-48 0 'Initial 3.3-49 0 Initial 3.3-50 0 Initial 3.3-51 0 Initial 3.3-52 0 Initial 3.3-53 8 04-13-1 999 3.3-54 0 Initial 3.3-55 Revision 16 03-'31-2000 3;3-56 Revision 16 03-31-2000 3.3-57 0 Initial 3.3-58 0 'Initial 3.3-.59 0 Initial 3.3-60 0 ~
Initial 3.4-1 0 'nitial 3.4-2 Revision 16 03-31-2000 3.4-3 0 Initial 3;4-4 -0 Initial 3.4-5 Rewsion 16 03-31-2000 I 3.4-6 0 Initial 3.4-7 0 Initial 3.4-8 0 Initial 3.5-1 0 Initial 3.5-2 0 'Initial (continued)
BFN-UNIT 2 EPL-3 TRM Revision 16 IVlarch 31, 2000
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- - BRO FERRY NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL(REQUIREMENTS)
EFFECTIVE PAGE LISTING (continued)
~Pa e No. Revision No. Effective Date 3.5-3 0 Initial 3.5-4 0 Initial 3.5-5 0 Initial 3.5-6 11-05-1998 3.5-7 11-05-1 998 3.6-1. 03-12-1999 3.6-2 8 04-1 3-1 999 3.6-3 0 Initial 3.6-4 0 Initial 3.6-5 0 Initial 3.6-6 0 Initial 3.6-7. 0 Initial 3.6-8 8 04-13-1 999 3.6-9 0 Initial 3.7-'1 0 Initial 3.7-2 0 Initial 3.7-3 0 Initial 3.7-4 0 Initial 3.7-5 0 Initial 3.7-6 0
- Initial 3.7-? 5 03-1 1-1 999 3.7-8 5 03-1 1-1 999 3.7-9 5 03-11-1 999 3.7-10 9 05-07-1 999 3.7-11 5 03-11-1999 3.7-12 5 03-11-1999 3.7-13'.7-14 5 03-1 1-1 999 5 03-1 1-1999 3.7-1 5 5 03-11-1999 3.7-1 6 5 03-11-1 999 3;8-1 0 Initial 3.8-2 0 Initial 3.9-1 0 Initial 3.9-2 0 Initial 3.9-3 Initial 3.9-4 0'evision 11 07-14-1999 (continued)
BFN-UNIT 2 EPL-4 TRM Revision 16
.March 31, 2000
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BROV% FERRY NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL(REQUIREMENTS)
EFFECTIVE PAGE LISTING (continued)
~Pe e No. Revision No. Effective Date 3.9-5 0 Initial 3.9-'6 0 Initial 3.9-7 0 Initial 3.9-8 0 Initial 4.0-'I 0 Initial 5.0-1 0' Initial 5.0-2 Initial 5.0-3 0 Initial 5.0-4 13 09-17-'I 999 App. A-1 0 Initial App. B (TVA-COLR-BF2C11),
Revision 0, Pages 1-22 04-13-1999 BFN-UNIT 2 EPL-5 TRM Revision 16 March 31, 2000
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. Offgas.Hydr n Analyzer Instrumentation t
TR 3.3.9 TR,3.3 INSTRUMENTATIQN TR 3.3.9 Offgas Hydrogen Analyzer Instrumentation LCO 3.3.9 There shall be at least one OPERABLE Offgas Hydrogen Analyzer instrument with alarm setpoint set to ensure the limit of TRM LCO 3.7.2 is not exceeded.
APPLICABILITY: During main condenser offgas treatment system operation NOTE TRM-LCO 3.0.3 is not applicable.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME, A. NoOPERABLEOffgas A.1 Installatemporary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Hydrogen Analyzer monitor instruments.
OR A.2.1 'ake grab samples 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from discover'y of no OPERABLE AND instrument AND Every 4.hours thereafter A.2.2 Analyze-the sample for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following explosive concentration grab sample of hydrogen.
BFN-UNIT 2 3.3-55 TRM Revision 0-, 16 March.31, 2000
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Offgas Hyd en Analyzer Instrumentation TR 3.3.9 TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.3.9.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> TSR 3.3;9.2 Perform CHANNEL FUNCTIONALTEST. 92 days TSR 3.3.9.3 NOTE Shall include use of standard gas samples containing a nominal zero volume percent hydrogen (compressed air), and a nominal one volume percent hydrogen, balance nitrogen.
Perform CHANNEL CALIBRATION. 92 days BFN-UNIT 2 3.3-56 TRM Revision 0-, 16 March 31, 2000
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Coolant Chemistry TR 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION'TIME D. Required Action and D.1 Initiate an orderly Immediately associated Completion shutdown.
Time of Conditions A,
.B, or C not met. AND OR D.2 Be in MODE 4. As rapidly as cooldown rate
)
Conductivity 10 permits pmho/cm at 25'C.
) 0.5 ppm.
OR Conductivity or chloride concentration limits of Table 3.4.1-1 Column A exceeded.
Coolant chemistry E.1 Initiate action to restore Immediately limits of Table 3.4.1-1 coolant chemistry within Column C or D limits.
exceeded.
BFN-UNIT 2 3.4-2 TRM Revision 0-, 16 March 31, 2000
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Coolant Chemistry TR 3.4.1 Table 3.4.1-1 Coolant Chemistry Limits "
COLUMN A COLUMN B COLUMN C COLUMN APPLICABLE APPLICABLE CONDITION APPLICABLE CONDITION CONDITION D"'PPLICABLE CHEMISTRY. CONDITION Steaming Rates Reactor Not Pressurized Noble Metal Chemical PARAMETERS Prior To Startup > 100,000 Ib/hr With Fuel In Reactor Application and Subsequent And At Steaming Vessel, Except During, Reactor Coolant Cleanup Rates Startup Condition
< 100 000 Ib/hr CHLORIDE 5 0.1 5 0.2 so.5 5 0.1 (ppm)
CONDUCTIVITY 6 2.0 5 1.0 5 10.0 5 20.0
,(pmho/cm at 25'C) 5.&8.6 5.&8.6 5.~.6 4.3-9.9
.(1)
When there is no fuel in the reactor vessel, Technical Requirement reactor coolant chemistry limits do not apply.
During the Noble IVletal Chemical Application and subsequent reactor coolant cleanup, CONDITIONS A, B, C, and D (including Required Actions and Completion Times) do not apply.
BFN-VNIT 2 3.4-5 TRM Revision 0-, 16 March 31, 2000
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. ~ BROMO FERRY NUCI EAR PLANT TECHNICAL REQUIREMENTS MANUAL(BASES)
EFFECTIVE PAGE L'ISTING Imptementatton Date. March 31, 2000 Pacae No. Revision No; Effective Date Title'Page 0 Initial Effective Page Listing, Revision 16 03-31-2000 0 Initial 0 Initial 8.3.0-1 0 Initial, 8 3.0-2 0 Initial 8 3.0-3 0 Initial 8 3.0-4 0 Initial 8 3.0-5 0 Initial 8 3.0-6 0 Initial 8 3.0-7 0 Initial 8.3.0-8 0 Initial 8 3.0-9 0 Initial 8 3.0-10 0 Initial
.8 3.0-11 0 Initial 8 3.0-12 0 Initial 8 3.0-13 0 Initial 8 3.1-1 0 ~
Initial 8 3.1-2 0 Initial.
8 3.1'-3 0 Initial 8 3.3-1 0 Initial 8 3.3-2 0 Initial 8 3:3-3 0 Initial 8 3.3-4 0 Initial 8 3;3-5 0 Initial 8 3.3-6 0 Initial 8 3.3-7 0 Initial 8 3;3-8 0 Initial 8 3.3-9 0 Initial 8 3.3-10 0 Initial 8 3.3-11 0 Initial 8 3.3-12 0 Initial 8 3.3-13 0 Initial 8 3.3-14 3 11-05-1 998 (continued)
BFN-UNIT 2 EPL-1 TRM Revision 16 IVlarch 31, 2000
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BROV FERRY NUCI EAR PLANT TECHNICAL REQUIREMENTS MANUAL(BASES)
EFFECTIVE PAGE LISTING (continued)
Pacae No. Revision No. Effective Date 8 3.3-15 Revision 14 12-15-1999 8 3.3-16 0' Initial 8.3.3-17 Initial 8 3;3-18 0 Initial 8 3.3-19 0 Initial B 3.3-20 0 Initial 8 3;3-21 0 Initial 8 3.3-22 0 Initial 8 3.3-23 0 Initial 8 3.3-24 0 Initial 8 3.3-25 0 Initial 8 3.3-26 0 Initial 8.3-27 0 Initial 8 3.3-28 0 Initial 8 3.3-29 0 Initial 8 3.3-30 0 Initial.
8 3.3-31 0 Initial 8 3.3-32 0 Initial 8 3.3-33 0 Initial 8 3.3-34 0 Initial 8 3.3-35 0 Initial 8 3.3-36 0 Initial 8 3.3-37 0 Initial 8 3.3-38 8 04-13-1 999 8 3.3-39 0 Initial 8 3.3-40 0 Initial 8 3.3-41 0 Initial 8 3.3-42 '0 Initial 8 3.3-43 8 04-13-.1999 8 3.3-44 0, Initial 8 3.3-45 0 Initial 8 3.3-46 0 Initial 8 3.3-47 0 Initial 8'3.3-48 0 Initial 8 3.3-49 0 Initial 8 3.3-50 0 Initial (continued)
BFN-UNIT 2 EPL-2'RM Revision 16 March 31, 2000
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BRO FERRY NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL(BASES)
EFFECTIVE PAGE LISTING (continued)
~Pe e No. Revision No. Effective Date 8 3;3-51 0 Initial 8 3.3-52 0 Initial 8 3.3-53 0 Initial 8 3.3-54 0 Initial 8 3.3-55 Revision 16 03-31-2000 8 3.3-56 0 Initial 8 3.3-57 0 Initial 8 3.3-58 0 Initial 8 3.3-59 0 Initial 8 3.3-60 0 Initial 8 3.4-1 0 .Initial 8 3.4-2 0 Initial 3.4-3 Revision 16 03-31-2000 8 34-4 Revision 16 03-31-2000 8 3.4-5 0 Initial 8 3.4-6 0 Initial 8 3.4-7 0 . Initial 8 3.4-8 0 Initial 8 3:4-9 0 Initial 8 3.5-1 0 Initial ~
8 3.5-2 0 Initial 8 3.5-3 0 . Initial 8 3.5-4 0 Initial 8 3.5-5 0 Initial 8 3.5-6 0 Initial
- 8 3;5-7 0 Initial 8 3.5-8 0 Initial 8 3.5-9 3 11-05-'I 998 8 3.'5-1 0 3 11-05-1 998 8 3.5-11 3 11-05-1 998 8 3.6-1 6 03-12-1999 8 3.6-2 6 03-12-1999 8 3.6-3 8 04-13-1999 8 3.6-'4 0 Initial 8 3.6-5 0 Initial 3.6-6 0 Initial (continued)
BFN-UNIT 2 EPL-3 TRM Revision 16 March 31, 2000
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BROV% FERRY NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL(BASES)
EFFECTIVE PAGE, LISTING (continued)
Parcae No. Revision No. Effective Date 8 3.6-7 0 Initial 8 3.6-8 0 Initial 8 3.6-9 8 04-1 3-1999 8 3;6-10 0 Initial 8 3.6-11 0 Initial 8 3.6-12 Revision 14 12-15-1999 8 3.6-13' Revision 14 12-15-1999 3.6-14 0 Initial 8 3.6-15 0 Initial 8 3.6-16 0 Initial 8 3.6-17 8 04-13-1999 8 3.7-1 Revision 15 . 02-08-2000 B 3.7-2 Revision 15 02-08-2000 8 3.7-3 ,0 Initial 8 3.7.-4 0 Initial 8 3.7-5 0 Initial 8 3.7 0 Initial 8 3.7-7 0 Initial 8 3.7-8 5 03-11-'I 999 8 3.7-9 5 03-1 1-1999 8 3.7-10 .5 03-11-1999 8 3.7-11 8 04-13-1 999 8 3;7-12 5 03-1 1-1999
,8 3.7-13 5 03-11-1999 8 3.7-14 5 03-11-1 999 8 3.7-15 5 03-1 1-1 999 .
8 3.7-16 5 '03-11-1999 8 3.7-17 5 03-11-1 999 8 3.7-18 5 03-1 1-1 999 8 3.7-19 5 03-11-1999 8 3.7-20 5 03-11-1999 8 3.8-1 0 Initial 8 3.8-2 0 Initial 8 3.9-1 0 Initial 8 3.9-2 0 Initial 8 3.9-3 0 Initial (continued)
BFN-UNIT 2 EPL-4 TRM Revision 16 March 31, 2000
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BRO FERRY NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL(BASES)
EFFECTIVE PAGE LISTING (contin'ued)
~
~Pa e No. Revision No. Effective Date 8 3.9-4 0 Initial B 3.9-5 0 Initial B 3.9-'6 Revision 11 07-14-1999 B '3.9-7 0 Initial B 3.9-8 0 Initial-B 3.9-9 0 Initial B 3.9-10 0 Initial B 3.9-1'I 0 Initial B 3.9-12 0 Initial 8 3.9-13 0 Initial BFN-.UNIT 2 EPL-5 TRM Revision 16
.March 31, 2000
4l
.i i
/gi
Offgas Hydr n Analyzer Instrumentation B 3.3.9 BASES APPLICABLE The hydrogen concentration of the gases from the air ejector is SAFETY ANALYSIS maintained below the flammable limit by maintaining adequate steam flow for dilution at all times. The pressure of the steam supplied to the first and third stage steam jet.air ejectors is monitored. The steam jet air ejector inlet and effluent are automatically isolated on low steam supply pressure. The preheaters are heated with steam, rather than electrically, to eliminate presence of potential ignition sources and to,limit the temperature of the gases in the event of cessation of gas flow. The recombiner temperatures are monitored and an alarm is actuated to indicate any deterioration of performance. A hydrogen analyzer downstream of the recombiners provides an additional check on recombiner performance.
LCO 3.'3.9 These instruments are required.to alert the operator of explosive conditions within the offgas system, and prompt the operator to with Technical Requirements 3.7:2 'omply APPLICABILITY The hydrogen buildup in the offgas system will stop when the main condenser offgas system is removed from service. Hence, this requirement is only applicable during main condenser offgas treatment system operation.
The OPERABIL'ITY and use of this instrumentation is consistent with the requirements of General Design Criteria 63 of Appendix A to 10 CFR 50.
ACTIONS A.1 and A.2 Continued operation of. the main condenser offgas treatment system is allowed provided adequate backup information is obtained from grab samples or a temporary monitor as required by ACTION A.
BFN-UNIT 2 B 3.3-55 TRM Revision 0-, 16 March 31, 2000
il~
ii
Coolant Chemistry 8 3.4.1 BASES
~ ~
LCO 3 4.1 Since oxygen may be at higher concentrations at low steaming (continued) rates, the chloride concentration limit is lower than at higher steaming rates when-the oxygen content is lower.
However, the conductivity is allowed to be at a higher level provided it is not caused from chloride ions due to the fact that the dissolved gases may result in higher conductivity. During startup or hot standby conditions, the reactor water cleanup system may be more efficient since the makeup from feedwater is very low.
Steamin Rates > 100 000'Ib/hr At steaming rates greater than 100,000 Ib/hr, the boiling rates are significant enough to strip away dissolved oxygen, but high enough to, start concentrating dissolved ions.
Because the dissolved oxygen is being effectively removed, the
,chloride ion.limits are relaxed.
However, because the reactor is now acting as a concentrator for ionic impurities and particulates, the conductivity limits are made more stringent.
Reactor Not Pressurized With Fuel In Reactor Vessel Exce t Durin Startu These are the baseline chemistry limits for water in contact with fuel. They are the same as the spent fuel pool with the extra limitation of pH.
Noble Metal Chemical A lication. NlVICA and Subse uent Reactor Coolant Cleanu During NMCA, the. chemicals added to the reactor coolant (which contain the noble metals) will increase conductivity and affect pH.
Therefore, special chemistry parameter limits are used for the NMCA process and subsequent reactor coolant cleanup. The chloride limits for this condition are unchanged from the 'Steaming Rates (100,000 Ib/hr'ondition.-
BFN-UNIT 2 B 3.4-3 TRM Revision 0; 16 March 31, 2000
Il
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Coolant Chemistry B 3.4.1 BASES
~ r AP-PLICABILITY These. limits are applicable, as specified, at all times when fuel is in the reactor. vessel.
ACTIONS A.1 and 8.1 A two week per year allowance for exceeding the normal chemistry limits is allowed.to give the opportunity for the reactor water cleanup system to return. the water chemistry to normal after a transient chemical intrusion.
C.'I These chemistry limits take into account factors of corrosion that may not be. affected by the amount of chloride ion.
D.1 and D.2 The major benefit of Cold Shutdown is to reduce the temperature dependent corrosion rates and provide time for the cleanup system to reestablish the purity of the reactor coolant.
Immediate ACTIONS are taken to bring coolant chemistry within limits.
BFN-UNIT 2 B 3.4-4 TRM Revision 0-, 16 March 31, 2000
0 0 1 i
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Item: ADAMS Document Library: ML ADAMS"HQNTAD01
,ID: 003699874:1
Subject:
Browns Ferry 2 & 3, ASME Section XI, ISI, Reactor Pressure Vessel Support Skirt Weld-Requests for Relief 2-ISI-10 & 3-ISI-9 1
Body:
ADAMS DISTRIBUTION NOTIFICATION.
Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003699874..
g g 2000 Page 1
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II 0 Distri89.txt A047 - OR Submittal: Inservice/Testing/Relief from ASME Code Docket: 05000260
- Docket: 05000296 Page 2
II 0
Tennessee Valley Authority, Post Olfice Box 2000. Decatur, Atabama 35609 March 24, 2000 10 CFR 50.55a(a) (3) (i)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In 0he Matter of ) Docket Nos. 50-260 Tennessee Valley Authority ) 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) UNITS 2 AND 3 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI r INSERVICE INSPECTION r REACTOR PRESSURE VESSEL SUPPORT SKIRT WELD REQUESTS FOR RELIEF 2-ISI-10 and 3-ISI-9 (TAC NOS. MA6408 AND MA8423)
In accordance with 10 CFR 50.55a(a) (3) (i), TVA is requesting relief from specified inservice inspection requirements in Section XI of the ASME Boiler and Pressure Vessel Code.
Enclosure 1 to this letter contains request for relief 2-ISI-10 and Enclosure 2 submits 3-ISI-9 for NRC review and approval.
TVA is requesting NRC approval to use the alternative rules of the ASME Section XI, Code Case N-323-1, for examination of the reactor pressure vessel (RPV) support skirt weld for BFN Units 2 and 3. The ASME Section XI Code requires that both surfaces (inside and outside) of the RPV support skirt weld receive a surface examination. However, for the inside surface of the BFN Units 2 and 3 RPV support skirt welds, access is restricted because of obstructions and high radiation levels. The ASME Code Case N-323-1 prescribes a surface examination of the accessible surface only. Also, in addition to the Code Case requirements, TVA is proposing
l1
, U.S,'. Nuclear Regulatory Commission Page 2 March 24, 2000 to supplement the N-323-1 examination with> a best-effort volumetric (ultrasonic) examination from the accessible surface of the weld. TVA considers that the proposed alternative examinations, will provide an acceptable level of quality and safety..
The enclosed requests for .relief are consistent with alternate examination requirements accepted for use at Hatch Nuclear Plant as stipulated by NRC letter to Southern Nuclear Operating Company, Incorporated, dated February 11, 2000.
TVA requests review of these requests for relief by January
'2001, to support the Unit 2 Cycle 11 (Sprp.ng 2001) refueling outage and preparation for the Unit 3 Cycle 10'Spring 2002),
refueling outage. There are no commitments contained in this letter. If you have any questions, please telephone me at (256) 729-2636.
cer y, E.
Mana r of Lice so.ng d Industry ffairs Enc osures cc: e Page
U.S: Nuclear Regulatory Commission,
, Page 3 March 24, 2000 Enclosures cc: (Enclosures):
Mr. Paul E. Fredrickson, Branch Chief U.S. Nuclear Regulatory Commission Region II 61 Forsyth Street, S.W.
Suite 23T85 Atl'anta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. William 0. Long, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852
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ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 2 AMERICAN SOCIETY OF'ECHANICAL ENGINEERS ASME SECTION XI INSERVICE INSPECTION'ISI) PROGRAM (SECOND TEN YEAR INSPECTION INTERVAL)
REQUEST FOR RELIEF '2-ISI-10 (See Attached)
/p I
I-TENNESSEE VALLEY AUTHORITY BROWNS .FERRY NUCLEAR PLANT,(BFN)
UNIT 2 ASME SECTION XI INSERVICE INSPECTION PROGRAM (SECOND TEN YEAR INSPECTION INTERVAL)
REQUEST FOR RELIEF 2-ISI-10 Executive Summa During a review of the BFN Units 2 and 3 ASME Section XI programs for future outage inspections, TVA determined that access to the inside weld surface of the reactor pressure vessel (RPV) support skirt (see attachments 1 and 2) would be restricted. It was also determined that examination and support personnel would encounter high radiation levels. The enclosed request for relief seeks to provide an alternative examination that will provide an acceptable level of quality and safety.
The weld configuration for the BFN Unit 2 RPV support skirt weld requires a surface examination of the outside and inside weld surfaces. For the inside weld surface, access is restricted because of high radiation and obstructions due to uniquely fitted mirror insulation panels covering the inside weld surface.
Control .Rod Drive housings and high radiation levels also limit access by the examiner.
As an alternative, TVA is proposing to use the requirements of ASME Code Case N-323-1, which allows a surface examination of the accessible weld surface only. TVA will also perform a best-effort volumetric (ultrasonic) examination from the accessible surface of the weld to detect service related flaws in the inside weld surface.
TVA's use of the alternative requirements of Code Case N-323-1 in conjunction with the best-effort volumetric examination
f c
from the accessible weld surface .will provide reasonable- assurance of the structural integrity of the weld.
TVA's proposed alternat'ive is consistent with the alternative examinat:ion requirements, accepted for use at Hatch Nuclear Plant, as stipulated by NRC letter to Southern Nuclear Operating Company,, Incorporated, dated February 11, 2000.
Therefore, in accordance with 10 CFR 50.55a(a)(3)(i), TVA is request:ing relief from inservice inspection requirements in the 1986 Edition, no addenda,Section XI of the ASME Boiler and Pressure Vessel Code for Category B'-H, Integral Attachments For Vessels (RPV support skirt), Item No. B8.10.
~astern:
Unit: Two (2)
Integral Attachments for Vessels (RPV
,Support Skirt: )
ASME Code Class: ASME Code Class 1 Section XI Edition: 1986 Edition, no Addenda Code Table: IWB-2500-1 Examination
~Cate or B-H (Integral Attachments for Vessels)
Examination Item Number:. B8.10 '(Integrally Welded Attachments)
Code Re irement: The 1986 Edition, no Addenda, ASME Section XI, Table IWB-2500-1, Examination Category B-H, Item B8.10 requi;res a surface or volumetric examination as applicable based on the .configuration of the support skirt to. vessel weld. BFN Unit 2 RPV support skirt configuration
0 0 4f
is illustrated in ASME Section XI Code, Figure IWB-2500-13 which requires a surface examination of areas A-B (outside surface) and C-D (inside surface) .
Code Re irements From Which Relief Relief is requested from the requirement to perform a surface examination of the RPV support skirt weld examination area C-D (restricted access), as illustrated in Figure IWB-2500-13.
List'f Items Associated With The Relief Re est: RPV Support Skirt weld No:
RPV-SUPP-2-1-IA Basis For Relief Receuest: The examination area C-D of Figure IWB-2500-13 is not accessible for examination due to the location, configuration, and insulation covering the C-D weld area. The bottom head and support skirt weld inside surface (C-D area) are covered with mirror insulation.
The insulation fits uniquely around each control rod drive (CRD) penetration and in close proximity with the head, taking the contour/shape of the head. The only way to gain access inside the support skirt is through one eighteen-inch diameter access opening. Removal of the uniquely indexed insulation in such a limited space and then passing it through the 18-inch diameter access hole would require extensive time and personnel exposure.
Physical access by the examiner is limited because of hi;gh radiation levels and obstructions due to the CRD housings.
Magnetic particle examination (yoke) cannot be used due to the space restrictions. The use of dye penetrant examination would require a very thorough cleaning of the weld and adjacent base material to remove rust and scale. The El<
I
\
preparation of the weld would potentially
~
require using techniques such as manual wire brushing since power tools may not fit into the limited area.
Radiological Control (RADCON) has indicated that a dose rate in these areas would be approximately 150 to 200 millirem/hour. It is estimated that approximately 56 man-hours would be required (6 people at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to remove/install insulation, and 2 people at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to perform the examination).
A total of 11.2 REM could be received by all involved personnel.
Further, there are no industry bulletins or reported failures of the subject weld.
Thus, the hardship associated with the examination of the inside surface is unwarranted when industry experience and ALARA principles are considered.
Alternative Examination: TVA will comply with the requirements of ASME Section XI, Code Case N-323-1 for the configuration illustrated in Figure 1 of the Code Case. In addition to the Code Case requirements, TVA will perform a best-effort volumetric (ultrasonic) examination from the accessible side of the weld to detect service related flaws in the inside weld surface.
Justification For The Grantin Of Relief: Code Case N-323-1 which was approved December 31, 1996, by ASME permits an alternative to the requirements of the 1986 Edition of ASME Section XI, Table IWB-2500-1, Examination Category B-H, Item B8.10 when only one surface of the weld is accessible for examination.
Code Case N-323-1 permits a surface examination from the accessible side only of the attachment weld. A copy, of Code Case N-323-1 is provided as Attachment 3 to this request for reli.:e f .
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The proposed alternative Code Case examination requirements have been evaluated by the ASME Section XI Code Committee and have been judged technically acceptable. The Code Case was incorporated into the 1997 Addenda of. the ASME Section XI Code, not as an alternative, but as the ASME Code requirement.
In addition to the alternative Code Case .requirements, TVA will perform a best-effort volumetric (ultrasonic) examination from the accessible side of the weld to detect service related flaws in the inside weld surface.
Using the alternative examination methods stated above, TVA considers that an acceptable level of quality and safety will be achieved and public health and safety will not be compromised.
TVA's proposed alternative is consistent with the alternative examination requirements, accepted for use at Hatch Nuclear Plant, as stipulated by NRC letter to Southern Nuclear Operating Company, Incorporated, dated February 11, 2000.
Im lementation
.Schedule: This request for relief is applicable to the Second'en Year Inservice Inspection Interval for BFN Unit 2..
Attachments: 1. Sketch of BFN Unit 2 Reactor Pressure Vessel Assembly
- 2. Sketch of BFN Unit 2 support skirt to vessel weld configuration
- 3. Code Case N-323-1, Alternate Examination For Welded Attachments to Pressure Vessels
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2-ISI-10 ATTACHMENT 1
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2-ISI-10 ATTACHMENT 3
~I ATTACHMENT P3 CASE N-323-1 CASES OF ASME BOtLER hND PRESSURE VESSEL CODE Approval Date: December 31, 1996 See Numerical Index for expiration and eny reaffirmation dates.
Case N-323-1 Alternative Examination for Welded Attachments
.to Pressure VesselsSection XI, 'Division I Inquiry: What alternative to the, requirements of Ex-amination Category B-K of the 1995 Addenda or Ex-amination Category B-H from the Winter 1981 Ad-denda, through the 1995 Edition may be performed for welded attachments to pressure vessels as shown in Figs. 1 and 2 when only one side of the attachment weld is accessible for examination?
E Reply: It is the opinion of the Committee that as an alternative to the requirements of Examination Cat-egory B-K of the 1995 Addenda or Examination Cate-gory B-H from Winter '1981'ddenda to the 1995 Edition:
(a) for the configuration shown in Figs. 1 and 2, a surface examination from the accessible side of the attachment weld may be performed or, (b) for the configuration shown in Fig. 2, a volumetric examination of Volume A-B, C-D from the accessible side of the attachment weld may be performed.
0 Il CASE (continued)
N-323-1 CASES OF ASME BOILER AND PRESSURE VESSEL CODE pressure retaining component 1/2 in.
B 1/2 In.
A 1/2 in 1/2 in. Attachment IWB Bounda1V Surtsce Erramination Areas A-B or C-D FIG. 1 WELDED ATTACHMENT
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CASE (continued)
N-323-'t CASES OF ASME BolLER AND PRESSURE VESSEL CODE oe
<o i~0 go ie ge Cast, forged. or weld built.up attachment t12 in.
Circumferential weld 0 A ti, in.
IWB Boundary Surface Examhatlon Areas A-8 or C-D FIG. 2 WELDED ATTACHMENT
4l l'
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
'UNIT SOCIETY OF MECHANICAL'NGINEERS 3'MERICAN ASME SECTION XI INSERVICE INSPECTION (IS I) PROGRAM (SECOND TEN YEAR INSPECTION INTERVAL)
REQUEST FOR RELIEF 3-ISI-9 (See Attached)
41 4 ~
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 3 ASME SECTION XI INSERVICE INSPECTION PROGRAM (SECOND TEN YEAR INSPECTION INTERVAL)
REQUEST FOR RELIEF 3-ISI-9 Executive Summa During a review of the BFN Units 2 and 3 ASME Section XI programs for future outage inspections, TVA determined that access to the inside weld surface of the reactor pressure vessel (RPV) support skirt (see attachments 1 and 2) would be restricted. It,was also determined that examination and support personnel would encounter high radiation levels. The enclosed request for relief seeks to provide an alternative examination that will provide an acceptable level of quality and safety.
The weld configuration for the BFN Unit 3 RPV support skirt weld requires a surface examination of the outside and inside weld surfaces. For the inside weld surface, access is restricted because of high radiation and obstructions due to uniquely fitted mirror insulation panels covering the inside weld surface.
Control Rod Drive housings and high radiation levels, also limit access by the examiner.
As an alternative, TVA is proposing to use the requirements of ASME Code Case N-323-1, which allows a surface examination of the accessible weld surface only. TVA will also perform,a best-effort volumetric (ultrasonic) examination from the accessible surface of the weld to detect service related flaws in the inside weld surface.
TVA's use of the alternative requirements of Code Case N-323-1 in conjunction with E2-2
Ck the best-effort volumetric examination from the accessible .weld surface will provid'e reasonable assurance of the structural integrity of the weld.
TVA's proposed alternative is consistent with the alternative examination requirements, accepted for use at Hatch Nuclear Plant, as- stipulated by NRC letter to Southern Nucl'ear Operating Company, Incorporated, dated February 11, 2000.
'Therefore, in accordance with 10 CFR 50.55a(a),(3)(i),. TVA is requesting relief from inservice inspection requirements in the 1989 Edition, no addenda, S'ection XI of the ASME Boiler and Pressure Vessel Code for Category .B-H, Integral Attachments For Vessels (RPV support skirt)'., Item No. B8.10.
Unit: Three (3)
~Ss tern:. Reactor Pressure Vessel (RPV)
Integral Attachments for Vessels (RPV Support Skirt)
ASME. Code Class: ASME Code Class 1 Section XI Edition: 1989 Edition, no Addenda Code Table: IWB-2500-1 Examination CatecLoO B-H (Integral Attachments for Vessels)
Examination Item Number: B8.10 (Integrally Welded Attachments)
Code Re irement: The 1989 Edition, no Addenda, ASME Section XI, Table IWB-2500-1, Examination Category B-H, Item B8.10 requires a surface or volumetric examination as applicable based on the configuration, of the support skirt to vessel weld. BFN Unit 3 RPV.'upport skirt configuration E2-3
is illustrated in ASME Section XI Code, Figure IWB-2500-13 w'hich requires a surface examination of areas A-B (outside surface) and C-D (inside surface).
Code Re irements From Which Relief Relief is requested from the requirement to perform a surface examination of the RPV support skirt weld examination area C-D (restricted access), as illustrated in Figure IWB-2500-13.
List Of Items Associated With The Relief Re est: RPV Support Skirt weld No:
RPV-SUPP-3-1-IA Basis For Relief Receue st: The examination area C-D of Figure IWB-2500-13 is not accessible for examination due to the'ocation, configuration, and insulation covering the C-D weld .area. The bottom head and support, skirt weld inside surface (C-D area') are covered .with mirror insulation.
The insulation fits uniquely around each contro'l rod drive (CRD) penetration and.
in close proximity with the head,,taking the contour/shape of the head. The only way to gain access inside the support skirt is through one eighteen-inch diameter access opening. Removal of the uniquely indexed insulation in such a limited space and then passing it through the 1'8-inch diameter access hole would require extensive time and personnel exposure.
Physical access by the examiner is limited because of high radiation levels and obstructions due to the CRD housings.
Magnetic particle examination (yoke) cannot be used due to the space restrictions. The use of dye penetrant examination would'equire a very thorough cleaning of the weld and adjacent base E24
~I material to remove rust and scale. The preparation of the weld would. potentially require using techniques such as manual wire brushing since power tools may not fit into the limited'rea.
Radiological Control (RADCON) has indicated that a dose rate in these areas would be approximately 150 to 200 millirem/hour. It is estimated that approximately 56 man-hours would be required (,6 people at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to remove/install insulation, and 2 people at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to perform the examination).
A total of 11.2 REM could be received by all involved personnel.
Further, there are no industry bulletins or reported failures of the subject weld.
Thus, the hardship associated with the examination of the inside surface is unwarranted'hen industry experience and ALARA principles are considered.
Alternative Examination: TVA will comply with the requirements of ASME Section XI, Code Case N-323-1 for the configuration 'illustrated in Figure 1 of the Code Case. In addition to the Code Case requirements, TVA will perform a best-effort volumetric (ultrasonic) examination from the accessible side of the weld to detect service related flaws in the inside weld surface..
Justification For The Grantin Of Relief: Code Case N-323-1 which was approved December 31, 1996, by ASME permits an alternative to the requirements of the 1986 Edition of ASME S'ection XI, Table IWB-2500-1, Examination Category B-H, Item B8.10 when only one surface of the weld is accessible for examination.
Code Case N-323-1 permits a surface examination from the accessible side only of the attachment weld. A copy
P ~
of Code Case N-323-1 is provided as Attachment 3 to this request for relief.
The proposed alternative Code Case examination requirements. have been evaluated by t:he ASME Section XI Code Committee and have been judged technically acceptable. The Code Case was incorporated into the 1997 Addenda of the ASME Section XI Code,.
not as an al'ternative, but as the ASME Code requirement.
In addition to the alternative Code Case requirements, TVA will perform a best-effort volumetric (ultrasonic) examination from the accessible side of the weld.
Using the alternative examination methods stated above, TVA considers that an acceptable level of quality and safety will be achieved and public healt:h and'afety will not be compromised.
TVA'. s proposed alternative is consistent with alternative examination requirements,, accepted for use at Hatch Nuclear Plant, .as:stipulated by NRC letter to Southern Nuclear Operating Company, Incorporated, dated February 11, 2000,.
Im lementation Schedule: This request for relief is applicable to the Second Ten Year Inservice Inspection Interval for BFN Unit 3.
Attachments.: ,Sketch of BFN Unit 3 Reactor Pressure Vessel Assembly 2'. Sketch of BFN Unit 3: support skirt to vessel weld configuration
- 3. Code Case N-323-1, Alternate Examination For Welded Attachments to Pressure Vessels E24
Qi 3-ISI-9 ATTACHMENT 1
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ATTACHMENT P3 CASE.
N-323-1 ChSES OF hSME BOILER'hND PRESSURE VESSEL CODE 1
Approval Date: December 31, 1996 See Numerical Index for expiration and any reaflirmation dates.
Case N-323-1 Alternative Examination for Welded Attachments to Pressure VesselsSection XI, Division 1 E Inquiry: What altemauve to'the requirements of Ex-amination Category B-K of the 1995 Addenda or Ex-amination Category B-H from the Winter 1981 Ad-denda,,through the 1995 Edition may be performed for welded attachments to pressute vessels as shown in Figs. 1 and 2 when only one side of the attachment weld is accessible for examination?
E Reply: It is the opinion, of the Committee that as an alternative to the requirements of Examination Cat-egory B-K of the 1995 Addenda or Examination Cate-gory B-H from Winter 1981 Addenda to the 1995 Edition:
(a) for:the configuration shown in Figs. 1 and 2, a surface examination Gum the accessible sidh of the attachment weld may be performed or, (b) for the configuration shown in Fig. 2, a volumetric examination of Volume A-B, C-D from the accessible side of the attachment .weld may be performed.
Ii r,
CASE (Continued)
N-323-1 CASES OF ASME EO2LER AND PRESSURE VESSEI COgE Preaaur e retaining component tl2 In.
ti2 In 2ln Attachment IWB Boundary
~ Exllmlnetlon Areea FIG. 1 WELDED ATTACHMENT A~ or ~
CASE (continued)
N-323-1 CASES OF hSME BOKER AND PRESSURE VESSEL CODE oo oo i8 Cast, forged. or weld built.up attachment 112 in.
Circumferential weld 1/2 in.
IWB Boundary Surface Stamhatlon Areas A-B or C-D FIG. 2 WELDED ATTACHMENT
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~'
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From: Esperanza Lomosbog Action Recipients: Copies:
RidsNrrPMWLong 0 OK RidsNrrLABClayton 0 OK I'nternal Recipients:
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Item: ADAMS Document ML ADAMS"HQNTAD01 'ibrary:
ID: 003699861'1
Subject:
BROWNS FERRY - UNIT 2- AMERICAN SOCIETY OF MECHANICALENGINEERS (ASME
- FOR RELIEF, 2-ISI-9, REGA
) SECTION XI AND AUGMENTED INSPECTIONS REQUEST RDING REACTOR PRESSURE VESSEL (RPV) CIRCUMFERENTIALSHELL WELDS Body:
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Distri63.txt A047 - OR -Submittal: Inservice/Testing/Relief from ASME Code Docket:,05000260 Page 2.
0 Tennessee Valley Authority. Post Office Box 2000, Decatur, Alabama 35609 March 24, 2000 10 CFR 50.55a (a) (3) (i)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 2055S Gentlemen:
In the Matter of Docket No. 50-260 Tennessee Valley Authority BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI AND AUGMENTED INSPECTIONS REQUEST FOR RELIEF r 2 IS I 9 r REGARDING REACTOR PRESSURE VESSEL (RPV) CIRCUMFERENTIAL SHELL WELDS, (TAC NO.
MA8424 )
ln accordance with 10 CFR 50.55a(a) (3) (i), TVA is requesting permanent relief from inservice inspection requirements of 10 CFR S0.55a(g) for the volumetric examination of the BFN Unit 2 reactor pressure vessel circumferential welds. This relief is for the remaining term of operation under the existing license. The alternative in TVA's request for relief provides an acceptable level of quality and safety and is consistent with the guidance and criteria described in NRC Generic Letter (GL) 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-OS Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds."
NRC issued GL 98-05 on November 10, 1998, which stated that licensees of BWRs may request permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential reactor pressure vessel shell welds by demonstrating that: (1) at the expiration of the operating license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the NRC staff's safety evaluation (SER) of the BWRVIP-05 Report dated July 28, 1998, and (2) licensee has implemented operator training and established procedures that limit the frequency
Ck U. S. Nuclear Regulatory Commission Page 2 March 24, 2000 of cold over-pressure events to the amount specified in the staff's July 28, 1998, SER. The enclosed request for relief demonstrates that TVA meets the guidance in GL 98-05 for permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of the BFN Unit 2 RPV circumferential welds.
TVA requests approval of this request for relief by December 31, 2000. This is. to allow for resource planning for the Unit 2 Cycle 11 (Spring 2001) refueling outage to support scheduled ASME Section XI outage activities.
This request for relief is consistent with one submitted to NRC for BFN Unit 3 by TVA letters dated June 25, 1999, and October 22, 1999. NRC letter to TVA dated November 18, 1999, approved the BFN Unit 3 request for relief.
There are no new commitments contained in this letter. If you have any questions, please telephone me at (256) 729-2636.
cerely, Manager of Lic and Indu y Affa rs Enclosure cc: See Pa 3
0 U.S. Nuclear Regulatory Commission Page 3 March 24, 2000 Enclosure cc: (Enclosure):
Mr. William O. Long, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville,. Maryland 20852 Mr. Paul E. Fredrickson, Branch Chief U..S. Nuclear Regulatory Commission Region Xl 61 Forsyth Street, S.W.
Suite 23T85 Atlanta, georgia 30303 NRC Resident inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611
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ENCLOSURE'ENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNI.T' AMERICAN SOCIETY OF MECHANICAL ENGINEERS,(ASME)
II SECTION XI I INSERVICE ( S ) AND AUGMENTED'NSPECTION'ROGRAM (SECOND TEN YEAR INSPECTION INTERVAL)
REQUEST FOR, 'RELIEF 2-ISI-9'(SEE ATTACHED);
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TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT'BFN)
UNIT 2 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)
II SECTION XI I I'NSERVICE ( S ) AND .AUGMENTED INSPECTION PROGRAM (SECOND 'TEN YEAR INSPECTION'NTERVAL)
REQUEST. FOR RELIEF '2-ISI-9 Executive Summa TVA is requesting permanent relief from the inservice inspection requirements for volumetric examination of reactor pressure vessel (RPV) circumferential shell welds.
This request applies to the remaining term of operation. under the existing license.
This request for relief will eliminate examination of the BFN Unit 2 RPV circumferential shell welds and is consistent with,the guidance provided in NRC Generic Letter 98-05,, "Boiling Water Reactor Licensees Use Of The BWRVIP-05 Report To Request. Relief From Augmented Examination Requirements On Reactor Pressure Vessel Circumferential Shell Welds" dated, November 10, 1998,.
The intent of 10 CFR 'SO.S5a rule change was to require licensees to perform an expanded RPV shell weld examination as specified in the 1989 Edition of the ASME Section XI Code, on an "expedited" basis.
Expedited in thi.s context effectively means during the inspection interval that the rule was approved or the first period of the next inspection interval. The final rule change was published in the Federal Register on August 6, 1992.
The examination schedule for,the RPV axially oriented welds shall continue as required by the ASME Section XI Code.
TVA is scheduled to .perform the RPV shell weld examinations required by the ASME Section XI Code and the expedited RPV E-2
0 Cl shell weld examinations in the third period (Spring 2001) of the Second Inservice Inspection Interval.
The BWRVIP-05 Report and the associated NRC SER supports exclusion of the examinations of the RPV circumferential shell welds provided certain limiting conditions regarding end of license vessel embrittlement and cold over-pressurization events are satisfied. TVA has satisfy';ed the limiting conditions specified in GL-98-05 for BFN Unit 2..
This request for relief is consistent with one submitted to NRC for BFN Unit 3 by TVA letters dated June 25, 1999, and October 22,, 1999. NRC letter to TVA dated November 18, 1999,, approved the BFN Unit 3 request for relief.
Therefore, in accordance with the guidance provided in- GL 98-05 and pursuant to 10 CFR 50.55a(a)(3)(i),, TVA requests that relief be granted from performing the volumetric examinations of the BFN Unit 2 RPV circumferential shell welds.
Uni t: Two (2)
Table 1 lists .the BFN Unit 2 RPV
,circumferential welds for which TVA is requesting, permanent relief from volumetric examination. The proposed relief is for the remaining term of operation under the existing license.
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TABLE 1 Cate~os Table Weld Descri tion and Exam IWB-2500-
,Method 1 Item Number Vessel Shel'1 to B-A, Flange Weld B1.11 No. C-5-FLG Volumetric Vessel Shell to B-A, Shell Weld No.C-4-5 Bl.ll Volumetric Vessel Shell to B-A, Shell Weld B1.11 No ~ C-3-4 Volumetric Vessel Shell to B-A, Shell Weld Bl. 11 No. C-2-3 Volumetric Vessel Shell to B-A, Shell Weld B1.11 No. C-1-.2 Volumetric (Located in Belt-line Region)
Vessel Shell to B-A',
Bottom Head Weld B1.11 No. C-BH-1 Volumetric ASME Code Class; ASME Code Class 1 Section XI Edition: 1986 Edition, no addenda Code Table: IWB-2500-1 Examination Cate~ox~: B-A (Pressure Retaining Welds in Reactor Vessel)
Examination Item Number: Bl.11 (Circumferential Shell Welds)
Code Re irement From Which Relief Is The inservice inspection requirements for the volumetric examination of RPV circumferential welds, ASME Section XI, Table IWB-2500-1, Examination Category
0 B-A, Item B1.11, Circumferential Shell Welds, and the (expedited) augmented examination requirements of 10 CFR 50.55a(g)-(6) (ii) (A) for vessel circumferential welds.
List Of Items Associated With The Relief Re est: See Table 1 Basis for Relief: The basis for this request for relief is outlined in the NRC SER for the BWRVIP-05 Report and the guidance outlined in GL 98-05. These documents provide the basis for the elimination of examinations of the BWR RPV circumferential shell welds. The BWRVIP-05 Report SER concluded that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. In addition, NRC conducted an independent risk-informed assessment of the analysis contained in the BWRVIP-05 Report SER. The NRC assessment and GL 98-05 concluded that the inspection of BWR RPV circumferential shell welds does not measurably affect the probability of failure. The industry examination results identified in the BWRVIP-05 topical report (Reference Electric Power Research Institute Report No. TR-105697), indicate that the necessity for performance of the circumferential shell weld volumetric examinations is not warranted based upon the low probability of failure of these welds.
TVA has addressed the two areas of concern outlined in the Permitted Action Section of Generic Letter 98-05: (1) the Unit 2 RPV level of embrittlement expected at the end of the period for which relief is requested in the most limiting RPV circumferential shell-weld areas, (2) the probability and expected frequency of the occurrence of a low temperature/high pressure transient on the Unit 2 RPV.
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4l II (1) Generic Letter 98-05 Permitted Action Item No. 1 Com arison Of The BFN Unit 2 RPV Brittle Fracture Information To The BWRVIP-05 And NRC Assessments Of The Probabilit Of Failure Of BWR RPV Circumferential Welds The BWRVIP-05 Report and the NRC Staff's independent risk-informed assessment of the initiative reports concluded that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. Additionally, the NRC assessment demonstrated that inspection of the RPV circumferential shell welds does not measurably affect the probability of failure.
The independent NRC assessment included a Probabilistic Fracture Mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM are: (1) the neutron fluence was that estimated to be the end-of-license mean fluence; (2) the chemistry values are mean values based on vessel types; and (3) the potential for beyond design basis events is considered. For plants with RPVs fabricated by Babcock and Wilcox (B & W), the mean end-of-license neutron fluence used in the NRC PFM analysis was 0.053 x 10'/cm . The highest fluence anticipated at the end of the period of 32 EFPY for BFN Unit 2 (in the RPV belt line region, weld C-1-2) is 0.11 x 10 n/cm'n the inside vessel surface. This fluence value was based on the BFN Unit 2 power uprate 32 EFPY operating curve information. The embrittlement for the BFN Unit 2 RPV due to fluence effects is less than the value obtained in the NRC limiting analysis for B & W RPVs shown in the SER (Table 2..6-4) for the BWRVIP-05 Report. A comparison of the limiting BFN Unit 2 RPV circumferential shell weld analysis versus the NRC limits;ng analysis for B & W RPVs is provided in Table 2 below.
~i The BFN Unit 2 beltline region circumferential shell weld .(C-1-2) was chosen for analysis to provide a basis for comparison to the NRC limiting analysis and as the Unit 2 RPV region where these calculated parameters would result in comparatively conservative values. The materials would also be representative of the Unit 2 RPV circumferential shell welds in general.
The informat'ion in Table 2 represents the beltline region circumferential shell weld C-1-2, located between Unit 2 RPV shells course 1 and course 2. As shown in Table 2, the RTND~ for BFN Unit 2 is much lower than the NRC limiting case.
Therefore, the conditional failure probability for BFN Unit 2 circumferential welds is bounded by the conditional failure probabilities in the NRC SER through the end of the current license period.
TABLE 2 PARAME TER BFN UNIT 2 LIMITING B&W Weld C-1-2 RPV .
Fluence (10 0.11 0.095 n/cm )
Initial RTND~ 40 F 20 F Chemistry 116. 8 196.7 Factor Cu (Wt %) 0. 09$ 0.31%
Ni (Wt 8) 0. 65% 0.59%
6RTNm 50.9 F 79.8$
Me an RTNDy 10.9 F 99.8 F
[Initial RTND~
+ hRTNDT]
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(2) Generic Letter 98-05 Permitted Action Item No. 2 Review Of BFN Unit 2 Procedural And Administrative Controls To Prevent RPV Low-Tem erature / Hi h-Pressure Transient Events The NRC staff stated in GL 98-05 that beyond design-basis events occurring during plant shutdown could lead to cold over-pressure events that could challenge vessel integrity. Although unlikely, the industry concluded that condensate and control rod drive pumps could cause conditions that could lead to cold over-pressure events that could challenge vessel integrity. For a BWR to experience such an event, the plant would require several operator errors. The NRC staff's assessment described several types of events that could be precursors to BWR RPV cold over-pressure transients. These were identified as precursors because no cold over-pressure event has occurred at a U.S.
BWR. The staff assessment identified one actual cold over-pressure event that occurred during shutdown at a non-U.S.
BWR. This event apparently included several operator errors that resulted in a maximum RPV pressure of 1150 psi with a temperature range of 79'F to 88'F. The operating procedures for BFN Unit 2 are sufficient to prevent a cold over-pressure event from occurring during activities such as the system leak test performed at the conclusion of each refueling outage.
Thus, the challenge to the BFN Unit 2 RPV from a non-design basis cold over-pressure transient is unlikely. The following discussion will provide further information to support TVA's conclusion.
BFN Operation procedures and administrative control processes are in
- place to minimize the potential for occurrence of RPV cold over-pressurization events. These processes include plant operating procedures, plant evolution planning and scheduling, administrative controls, and operator training.
0 0 Since cold over-pressurization events are most likely to occur during normal cold shutdown conditions, BFN operating procedures are written to require that RPV water level, pressure, and temperature are established and maintained in well controlled bands. Plant Unit Operators frequently monitor these parameters for abnormalities and indications of unwanted transients. Also, any plant evolution which requires changes in these critical parameters is performed under the oversight of the Shift Manager who is also notified immediately of any abnormalities in the indications. Therefore, any deviation of these parameters from the established bands are, promptly identified and corrected. In addition to these procedures, unit conditions for on-going activities which potentially can effect the maintenance of acceptable operating conditions and available contingency systems and plans are discussed by unit operations personnel at the time of shift turnover. These administrative controls and procedures provide assurance that activities which could adversely effect RPV water level, temperature, and pressure are precluded.
Nuclear Experience reviews and industry operating histories have shown that inadequate work-control processes and procedures could precipitate a cold over-pressurization event. For BFN, outage work is controlled through planning and scheduling activities performed by the Outage Management and Work Control Team. Unit and system work activities are carefully reviewed and coordinated to avoid conditions which could adversely affect the unit's RPV water level, temperature, and pressure. Plant activities are routinely coordinated through the use of a plan-of-the-day (POD) which contains a list of activities to be performed and frequently contains cautionary notes on the activities.
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il These PODs are reviewed and discussed with station management and copies are maintained in appropriate locations.
Changes to these PODs are approved through the Operations Department Management and the Shift Manager. In addition, during outages, work on unit systems and components is coordinated through work control centers which provide an additional level of unit operations oversight.
In the main control room, the Shift Manager is required to maintain cognizance of any activity which could potentially affect reactivity, reactor water level, or decay heat removal. Unit Operators are required to provide positive control of reactor water level, temperature, and pressure within the specified bands, promptly report when operation outside the required bands occurs, and notify, the Shift Manager of any restoration corrective measures being taken. As part of the outage work control process, special procedures such as hydrostatic testing require pre-job briefings conducted with operations personnel for any activity which could potentially affect critical plant parameters. The pre-job briefing includes all cognizant individuals involved in the work activities. Expected plant system and component responses and contingency actions to mitigate unexpected conditions are also discussed. When the plant is in cold shutdown, plant procedures require that the RPV head vent valves be opened after the reactor has been cooled to less
,than 212'F. Administrative and plant operations control procedures for this evolution and for controlling reactor water level, temperature, and pressure are an integral part of operator initial and re-qualification training. Responses to abnormal water level and RPV conditions are also part of the operator's training.
In addition, unit-specific brittle-fracture operating pressure-temperature
limit curves and procedures have been developed to provide the appropriate guidance for compliance with the operating limits and the associated Technical Specification requirements.
Review of Hi h Pressure In ection Sources:
RPV water injection sources during cold shutdown conditions include three systems.
During normal cold shutdown, RPV water level and pressure are controlled through the Control Rod Drive (CRD) and the Reactor Water Cleanup (RWCU) Systems.
RPV conditions are controlled through a "feed and bleed" process using these two systems. The RPV and its piping system are not placed in solid water conditions and after the -plant is cooled below 212 F, the head vent valves are opened.
either one of the RWCU or CRD Systems If fail, the Unit Operator would adjust the other system to maintain the proper water level and pressure. In addition, BFN also has water level instrumentation with set-points for high and low water levels that alarm at 39 inches high and 27 inches low to alert operators that a level transient is in progress and action is required. During these plant activities the CRD System typically injects water at a rate of less than 60 gallons per minute (gpm). Injection rates at this level allow the operator sufficient time to compensate for unanticipated level and pressure changes. Therefore, the probability of an occurrence of a high-pressure/low temperature event from these two systems, that places RPV conditions outside the pressure-temperature curve limits is low.
In addition to the RWCU and CRD Systems, the Standby -Liquid Control System is another high-pressure source to the RPV.
For BFN, SLC System operation occurs only if the system is manually initiate'd by operator action in accordance with emergency operating procedures. Thus,
0 SLC operation will not occur during cold shutdown operations except under stringently controlled test conditions.
In the event of an inadvertent injection, the SLC injection rate (approximately 50 gpm) is sufficiently low to allow operators to intervene and control the reactor- pressure.
During cold shutdown periods following refueling, the RPV is pressure tested in accordance with the applicable ASME Section XI Code requirements. BFN hydrostatic tests of the RPV and the reactor coolant system are designated as complex and infrequently performed tests.
For these types of tests BFN requires a detailed pre-job briefing with all individuals participating in the test.
Also, BFN has a dedicated operator for RPV water level and pressure control.
RPV and reactor coolant system pressure testing is a carefully controlled plant evolution which receives special Operations management oversight and utilizes procedural controls to ensure that the test does not precipitate a transient outside the specified safety limits. These tests are also performed after the RPV and system are heated to the proper system inservice pressure test temperatures prior to increasing the system pressure. During these tests the RPV pressure, water level, and temperature are controlled through the CRD and RWCU Systems using the "feed and bleed" process. Increases (or decreases) in system pressure are limited to 50 pounds per square inch (psi) per minute. For example, if any RWCU valve fails, then the CRD pump is tripped and the RPV is depressurized. This practice minimizes the probability of exceeding the specified Technical Specification pressure-temperature limits during the system pressure test.
During plant startup following a cold shutdown, the High Pressure Coolant
il Injection (HPCI) and the Reactor Core Isolation Cooling (RCIC) pumps provide a possible means to over-pressurize the RPV. However, for BFN, these systems have high pressure steam-driven pumps which have automatic isolation set-points of 100 psi and 50 psi respectivel'y; and will not function when the plant is in cold shutdown.
Based upon the above evaluation the likelihood of a cold over-pressure transient event placing the Unit 2 RPV in non-design conditions is very low.
Therefore, the probability of an occurrence of a cold over-pressure transient is considered to be less than or equal to the probability used in the analysis described in the NRC independent evaluation performed in the assessment of the BWRVIP-05 Report.
Alternative Examination: As an alternative, TVA proposes to perform only the RPV longitudinal shell weld examinations during the third inspection period (Spring 2001) of the Second Ten-Year ISI Interval in conjunction with the scheduled ASME Section XI Code and augmented RPV Examinations.
Justification For The Grantin Of Relief: Based upon the previous stated technical justifications, performance of the examination of the Unit 2 RPV circumferential shell welds in accordance with the ASME Code requirements, is not warranted. This position is supported by actual industry inspection experience,
. industry initiatives, and their supporting calculations. Further, the additional costs and personnel exposure that would be incurred without any. apparent increase in safety does not warrant the performance of the examinations. These factors provide reasonable assurance of the continued structural integrity of the
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BFN Uni:t 2 RPV. Therefore, pursuant to 10 CFR 50. 55a (a), (3) (i),, TVA requests that permanent relief be granted from the inservice inspection and the augmented inspection requirements of 10 CFR 50.55a(g),(6)(ii)(A), for volumetric examination of reactor pressure vessel circumferential shell welds, ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item Bl.ll, Circumferential Shell Welds as permitted by GL '98-05.
Further, in accordance with the guidance specified in the NRC SER, Section 4.0 for the BWRVIP-05 Report, TVA intends to examine the RPV circumferential shell welds should axial weld examinations reveal an active mechanistic mode of degradation. The scope and schedule of these examinations would be submitted
.to NRC for approval.
This request for relief is 'consistent with one submitted to NRC for BFN Unit 3 by TVA letters dated June 25, 1999, and October 22, 1999.. NRC accepted TVA's request for relief by letter dated November 18, 1999.
Im lementation Schedule: This Request for Relief will be implemented during the Second Ten Year ISI Inspection Interval for Browns Ferry, Unit 2 and continue in effect for the remaining term of operation under the exist'ing license.
Attachment:
Brown Ferry Unit 2 RPV shell weld location, schematic drawing E-14
2-lsI,-9 Atcoachmen t,
II DRAT/INGS (GE)
SKETCHES RPV EXAUINATION PLAN ICE)
NOZZLE CROVP ) OISTAHCE TO UATINC SVRFACE SK-82001 SK-82005 SK 820'IO SK 8200$ SK 82007 NIX 66.5 -8 004 006 N12X 146 HIIX - 220 745'EFERENCE LEGEND N4X 246.5 47 VESSEt. NOZZLE HSX - 25S.S 3 FVLL PENETRATION NOZZLE TELO NS - 256.5'ltx
$ 75 ASME CC-7 (ET7UI VALENT)
H2X 564 NIX - 565 ~ 5 HDX - 610 UAT INC SVRFACE 0 560 270 160 560 I
F LANCE C-5-FLC M 550'0)
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210'5 80'IA 0 M COVRSE 5
(OQ 8 UQ (Qg C-4-5 Y-S-C 265'128 165'-5-A VESSEt. STABILIZERS H12A 547 C-5-4 V-4-C 227'-4-8 107'-4-A 45'OVRSE O ssr N4E Oo 0 N40 N40 O
p Hle N4A COVRSE Nse HSA NOTES:
Y-5-C PD NS-IR 1. REFER 10 RPY UANVAL FOR MATERIAL SPECIFICATION ANO UATERIAL THICKNESS.
Y-5-8 2 ~ NOZZLES H 11A ~ N 118 ~ N 12A ~ N 128 ~
Y-5-A N-16A. ANO N-168 ARE CATECCRT B-E.
C-2-5 265' 165' 45' 551.5'66 N168 COVRSE 140'56.5 BELI'LINE RECIOII F/iliEEiiliE/ii%PEÃEÃÃEÃiii, 216 Qo Qo Qo Qo Qo Qo Qo
'DID ACUs ss slOC N/A N A ss A 2
DtttltD Dtt SISSSD SKSS2 ~ D20DOD JDI SSS N68 V CSSSSSCt Sttt Cttt DS SD OSstt DDCSS HSIDSS 0-Bst-I ON08-IR 9 NDA-IR TENNESSEE YALLET AVIlsORITT TAtsCENT LINE/LOVER HEAO DELO BRODNS FERRT NVCLEAR PLANT VHIT 2 S REACTOR Pf(ESSURE YESSEL (RPY)
DO//DSS HEAt SHELL COURSE TELO llLE LOCATIONS (OU1SIOE YIB)
DRISlts N/4 DATEs ss/4 CssttsstDs ss/4 s/tssost 0 s tsctt 01 OF 02 tsttstsll(Ds N/4 CL8 2-CHII-2046-C DD ALL A/0 HISTORY RESEARCHEO AT R000 CCO