ML18036B286
| ML18036B286 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/12/1993 |
| From: | Zeringue O TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9305190205 | |
| Download: ML18036B286 (16) | |
Text
ACCEI ERATO DOCUMENT DISTRIBUTIONSYSTEM REGULAT INFORMATION DISTRIBUTION+STEM (RIDE)
ACCESSION-NBR':9305190205 DOC.DATE: 93/05/12 NOTARIZED: NO DOCKET ¹ FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH.NAME 'UTHOR AFFIIIATION ZERINGUE,O.J.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards revised Relief Request SPR-5 from inservice pressure test program to accomplish repair of 12-inch weld two 2-inch socket'elds.Withdraws Relief Requests SPT-1, SPT-2
& SPT-3.
DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: Inservice/Testing/Relief from ASME ode NOTES:
D RECIPIENT ID CODE/NAME PD2-4 ROSS,T.
INTERNAL: ACRS NUDOCS-ABSTRACT OGC/HDS3 RES MILLMAN,G EXTERNAL: EG&G BROWN,B NRC PDR COPIES LTTR ENCL 1
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1 RECIPIENT ID CODE/NAME HEBDON,F NRR/DE/EMEB MB REG FI 01
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B EG&G RANSOME,C NSIC COPIES LTTR ENCL 1
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,D NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 504-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
D TOTAL NUMBER OF COPIES REQUIRED:
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Tennessee Valley Authority, Post Office Box 2000. Oecatur. Alabama 35609.2000 NY id l993 O. J. "Ike Zeringue Vice President. Browns Ferry Nuclear Plant 10 CFR 50.55a(g)(5)iii U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of Tennessee Valley Authority Docket No. 50-260 BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 CLARIFICATION OF INSERVICE SYSTEM PRESSURE TEST PROGRAM REQUEST FOR RELIEF SPT-5 AND WITHDRAWAL OF INSERVICE SYSTEM PRESSURE TEST PROGRAM REQUESTS FOR RELIEF SPT-1, SPT-2, AND SPT-3
Reference:
1.
Letter from TVA to NRC dated April 27,
- 1993, "Browns Ferry Nuclear Plant (BFN) American Society of Mechanical Engineers (ASME)Section XI Inservice System Pressure Test Program Request for Relief" 2.
Letter from TVA to NRC dated May 22,
- 1992, "Browns Ferry Nuclear Plant (BFN) Unit 2 American Society of Mechanical Engineers (ASME)Section XI Programs for the Second Inspection Interval" 3.
Letter from TVA to NRC dated April 8, 1993, "American Society of Mechanical Engineers (ASME)Section XI Inservice Pressure Test Program" The purpose of this letter is to provide clarifications and additional information relevant to NRC's consideration of BFN's Request for Relief (SPT-5) from the requirements of the 1986 Edition of Section XI of the ASME Boiler and Pressure Vessel Code submitted by Reference 1.
Pursuant to 10 CFR 50.55a(g)(5)iii, a revised Request for Relief SPT-5 is contained in the enclosure.
'1 An~
9305190205-'9305 k'2A',"
PDR ADOCK 05000260 I
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U.S. Nuc1ear Regulatory Commission NW %2 1993 Furthermore, this letter withdraws Requests for Relief SPT-l, SPT-2, and SPT-3 submitted in Enclosure 2 of Reference 2.
After further review of the pressure test requirements, TVA has determined Requests for Relief SPT-l, SPT-2, and SPT-3 are not required'nd that compliance with the applicable Code requirements can be achieved.
The aforementioned Requests for Relief were discussed in a May 10, 1993 teleconference between BFN and NRC personnel during which BFN agreed to provide a revised Request for Relief, SPT-5.
A Section XI hydrostatic test is necessary as a result of certain repairs and replacements which were made during the current Unit 2, Cycle 6 refueling outage.
BFN has determined that the performance of a Code hydrostatic test will require the removal of the main steam relief valves.
This results in the imposition of a substantial hardship on BFN while providing a minimal increase in the margin of safety.
The enclosed Request for Relief contains a revised Basis For Relief and redefines
.the proposed Alternate Testing.
The proposed Alternative Testing will subject the repair and replacement areas to a leakage test at a minimum pressure of 1034 psig and defers the required ASME Section XI hydrostatic test.
The deferred ASME Section XI code hydrostatic test will be performed at the scheduled ten-year interval reactor vessel hydrostatic test.
In addition, during the May 10, 1993 teleconference NRC notified BFN that Request For Relief SPT-4 submitted in the Reference 3 letter has been approved.
This letter supercedes the Requests for Relief provided in Enclosure 2 of the Reference 2 letter.
If you have any questions, please telephone
'Pedro Salas, Manager of Site Licensing, at (205) 729-2636.
Sincerely,
- 0. J. Zering e
Enclosure cc:
See page 2
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U.S. Nuclear Regulatory Commission Ny i2 1993 cc (Enclosure):
NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637
- Athens, Alabama 35611 Mr. Thierry M. Ross, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555.Rockville.Pike Rockville, Maryland 20852 U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
Ik
ENCLOSURE REQUEST FOR RELIEF SPT-5 System:
Drawing:
Core Spray (75) 47E814-1 (FSAR Figure 7.4-4) 47E817-1 (FSAR Figure, 4.3-2a, Sheet 2)
Component:
(a) 12-inch weld 'repair (b) two 2-inch socket wel'ds Class:
Function:
(a)
Connects 12-inch core spray valve 'HCV-75-27 to 12-inch to 10 inch reducer at reactor vessel nozzle N5A (b)
Reactor pressure vessel bottom head drain manual valve 2-10-'505 to reactor water cleanup system Impractical Test Requirement:
IWA-4400(a) After repairs by welding on the pressure retaining boundary, a system hydrostatic test shall be performed in accordance with IWA-5000.
Basis for Relief:
The weld repaired areas are situated in their respective systems between the reactor pressure vessel (RPV) and the first isolation valve off the vessel.
These locations provide. no method to isolate the repaired areas for the RPV for the purpose of the hydrostatic test required by Article IWA-4000.
The performance of a hydrostatic pressure test of the repaired areas by pressurizing the RPV would require the removal and blanking of 8.of the 13 main steam relief valves cartridge assemblies to prevent the valve lifting and possible resulting valve seat damage.
The estimated exposure for the removal and replacement of the 8 relief valves is 0.48 man-rem.
In addition, performance of the hydrostatic test requires the reactor vessel high pressure SCRAM switch to be disabled.
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Basis for Relief (Continued)
.Although these activities can be performed as demonstrated during the performance of the first inspection interval hydrostatic test of the reactor coolant system performed in March 1991, they do require an extension of the current outage to perform.
TVA believes that the delay in the return of the unit to operation to perform the hydrostatic test is not commensurate with the small increase in,the margin of safety it provides compared,to the proposed alternative for the following reasons.
The integrity of the weld repairs have been ensured by the following nondestructive examinations:
1) the 12-inch core spray piping overlay repair was examined using volumetric (UT) and surface (PT) techniques; 2) the two 2-inch socket welds were surface examined (PT).
To further ensure the integrity of the repair
- welds, the reactor coolant. system will be pressurized to a minimum test pressure of 1034 psig (measured at the RPV dome) and a VT-2. visual examination for leakage.
This test will subject the repair welds to a minimum pressure of 29 psi (3%) above normal operating, versus the hydrostatic test pressure of 86 psi (8.6%):above normal operating pressure.
In addition, each of these areas will be.subject to a system leakage test performed't nominal operating pressure
,(1005 psig at the RPV dome) prior to startup following each refueling outage.
The decrease in the margin of safety due to the reduction in test pressure is minimal and.does not represent an undue risk to the plant or the public.
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Alternate Testing:
BFN, propose to defer the required hydrostatic test of the repair areas until near the end of the current inspection interval.
In our system pressure test program for the second inspection interval BFN. cited the use of ASME Code Case N498 which provides for a nominal operating pressure test to be performed in. lieu of the code required ten year hydrostatic test for class l,systems.
The invoking of Code. Case N498 is intended for use on systems which do not require hydrostatic testing due to welded repairs or replacements, or for which such repair/replacement hydrostatic tests have been performed.
In light the subject class 1 repairs and the resulting hydrostatic test requirement, BFH waive the use of Code Case N498 for the reactor coolant system.
Based on current regulatory requirements, a full code hydrostatic test will be performed on the reactor coolant system during the second inspection interval.
A leakage test of the reactor coolant system will be performed at the highest pressure
.which can be attained while ensuring that the main steam safety valves will'ot be challenged (1034 psig at the RPV dome).
This test pressure is based on the following considerations:
1) the lowest main steam relief valve has a
setpoint value of 1105 psi and a setpoint tolerance of g l%%d; 2) the main steam safety valves are located 41 feet below the RPV dome; 3) a 20 psi'est pressure range is necessary due to expected perturbations during the test; 4) a 20 psi margin is deemed essential to prevent the commencing of weeping through the main steam safety valve.
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In addition, as required by ASME Section. ZI, the reactor coolant system will receive a leakage test at full system operating pressure (1005 psig) prior to unit startup following each refueling outage.
A hydrostatic pressure test of these, repairs will be performed in conjunction with the reactor coolant system hydrostatic test near the end of the current inspection interval..
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