ML20046B427

Application for Amend to License NPF-8,modifying TS 3/4.4.6 SG & Associated Bases,Ts 3/4.4.7, Reactor Coolant Sys Leakage & TS 3/4.4.9, Specific Activity & Associated Bases

ML20046B427
Person / Time
Site:	Farley
Issue date:	07/29/1993
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20046B428	List: ML20046B427 ML20046B432
References
NUDOCS 9308040193
Download: ML20046B427 (7)

Text

-

s Southem Nuclear Operating Company s

Post Office Box 1295 Birmingham, Alabama 35201-1295 Telephone 205 868 5000 Southern Nuclear Operating Company

'"* * "'h " *#* '" 8 # "

July 29,1993 Docket No. 50-364 10 CFR 50.90 U. S. Nuclear Replatory Commission ATTN: Documens f.ontrol Desk s

Washington, D. C.

20555 Joseph M. farley Nuclear Plant - Unit 2 Technical Specification Changes Associated With Steam Generator Tube Suocort Plate Interim Pluaaina Criteria Gentlemen:

i By letter dated February 26, 1991, Alabama Power Company submitted a steam generator tube support plate alternate plugging criteria. As a result of delays in approving the full voltage alternate plugging criteria, Southern Nuclear Operating Company (SNC) submitted a reduced voltage interim plugging criteria (IPC) for both Unit I and Unit 2.

The Unit 2 IPC was approved on April 1, 1992 and the Unit I criteria was approved on October 8, 1992. As a result of further delays in approval of the full voltage alternate plugging criteria, SNC submitted a request to use the reduced voltage IPC for Unit 2 for the next operating cycle on May 28, 1993.

In early June 1993, the NRC placed draft NUREG-1477 dated June 1,1993, " Voltage Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report," in the public document room for comment.

Draft NUREG-1477 dated June 1, 1993 contains many new recommendations for obtaining approval of a 1 volt IPC.

In addition, it modifies several.NRC positions approved in the first two IPCs. SNC has reviewed the draft NUREG 1

and has incorporated the appropriate recommendations for Plant Farley.

The appropriate NUREG recommendations were confirmed by telephone conversation between Jack Woodard of SNC and Bill Russell of the NRC on July 13, 1993.

SNC's revised IPf. for Farley Unit 2 is attached.

NRC approval is' requested i

by Saptember 17,.1.993.

In addition to requesting use of the 1 volt criteria for leaving. tubes in service and rec ucing the allowable primary-to-secondary leak rate to 150 gallons per day per steam generator, SNC is also' requesting reductions in Farley's technical specification limits for specific activity of dose' equivalent I"* and for transient dose equivalent I"' reactor coolant specific activity by'a factor of four. The reduction in the limits for the allowable specific activity levels, coupled with the.NRC predicted' primary-to-secondary.

leakage (per draft NUREG-1477 dated June 1, 1993) in the event of a steam line break, will ensure that the site boundary _ dose levels do not exceed the appropriate Part 100 limits l

930004'0193 930729

)

PDR ADOCK 05000364 P

PDR

~-

j

1

' O. S. Nuclear Regulatory Commi.ssion Page 2 Since SNC's last interim plugging criteria -submittal, additional. data have been provided to the NRC Staff in other alternate / interim plugging criteria submittals.

These data points underscore the conservatism contained in a l' volt IPC fcr Farley Nuclear Plant.

Furthermore, two tubes were pulled from Farley Unit I during its last outage.

Although these two tubes had i

relatively high voltages' (3.3 and 3.2 volts), they exhibited no leakage under steam line break differential pressures.

In: addition, these same two. tubes burst at differential pressures of 6,600'psig and 8,100 psig, respectively.

These burst pressures are over twice the differential pressure expected-in a -

steam.line break.'

' contains' the proposed interim volt'a'ge plugging criteria;

-including a sampling program. Attachment' 2 contains the ~ safety analysis. to support the amendment raquest. Attachment-3 contains the proposed changed-technical specification pages in support of the IPC. Attachment 4 provides.a.

significant hazards evaluation for the proposed interim plugging criteria-in accordance with 10 CFR 50.92. ' Attachment 5 provides the proposed eddy current guidelines ~for use in the upcoming outages, assuming the interim-criteria is approved for use. Attachment 6 contains the supporting environmental evaluation.

The Plant Operations Review' Committee has reviewed and ' approved these proposed changes, and the Nuclear Operations Review Board will review these i

proposed changes at a future meeting.

1 A copy of these proposed changes has been sent to_Dr. D. E. Williamson, the Alabama State Designee, in'accordance with 10 CFR 50.91(b)(1).

j If there are any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY b

h Dave Morey Vice President Farley Project 1

l REM:maf-UNIT 21PC.R51 SWORN TO AND SUBSCRIBED BEFORE-ME I

Attachments THIS87M DAY OF, # / 11993 cc:

Mr. E. D. Ebneter ff L

Mr. 1. A. Reed

. 6mf i

Mr. E. F. Maxwell

- No g ublic j

Dr. D. E. Williamson My C mmission Expires: 7--/f-7'y l

~

Y

Voltage Plugging Criteria for the Steam Generator Tube Support Plate Interim Plugging Criteria-L r

)

Southern Nuclear Operatina Company Proposal:

Voltage Pluquina Criteria 1.

A bobbin inspection of 100% of the hot and cold leg steam generator tube support plate intersections will be performed.

2.

Flaws within the bounds of the tube support' plate with bobbin voltage less than or equal to 1.0 volt will be allowed to remain in service.

3.

Flaws within the bounds of the tube support plate with a bobbin voltage greater than 1.0 volt'will be repaired or plugged except as noted in 4.

4.

Flaw indications within the bounds of the tube support plate with a bobbin voltage greater than 1.0 volt but less than or equal to 3.6 volts may remain in service if a rotating pancake coil probe (RPC) inspection does not detect a flaw.

Flaw indications with a bobbin voltage greater than 3.6 volts will be plugged or repaired.

5.

A sample RPC inspection of a minimum of 100 tube support plate intersections will be performed.

All intersections with a bobbin dent voltage exceeding 5 volts will be inspected. Other intersections in the sample population will be based on inspecting intersections with artifact indications and intersections with unusual phase angles.

Expansion of the sample plan, if required, will be based on the nature and number of the flaws discovered.

6.

RPC flaw indications not found by the bobbin due to masking effects (due to denting, artifact indications, noise) will be plugged or repaired.

End-of-Cycle (EOC) Voltage Distribution Acceptable methods for determining the EOC voltage distribution are as follows:

1.

The methodology described in WCAP-12871, Revision 2.

This involves sampling the cumulative probability distribution of NDE uncertainty and the voltage growth rate using Monte Carlo techniques and applying the results to the beginning-of-cycle (B0C) voltage distribution.

2.

A simplified approach may be used as an alternative (to the WCAP-12871, Revision 2 approach) provided it allows for a conservative treatment of the tails of the cumulative probability distributions of NDE uncertainty and the voltage growth rate to the 100% cumulative probability values.

1 l

1

4 Steam Line Break Leakage 1.

Steam line break (SLB) leakage as a function of EOC voltage (E0C-V) shall be determined in accordance with draft NUREG-1477 dated June 1, 1993, Section 3.3, " Recommended Method for Calculating Potential Leakage."

2.

Predicted SLB leakage will be limited to 22.8 gpm.

The limit of 22.8 gpm is derived as the product of 5.7 gpm x 4, where:

- 3.7 gpm is the leakage limit based on an SRP analysis submitted to the NRC by SNC's June 4, 1992 letter; and

- 4 is the multiplication factor due to the allowable increase in leak rate justified by reducing the limits for specific activity of the primary coolant contained in Farley Technical Specification 3.4.9. by a factor of 4.

The specific activity dose equivalent 1 " will be reduced 2

from 1 Ci/gm to 0.25 Ci/gm, and the transient dose equivalent I"'

will be reduced from 60 Ci/gm to 15 Ci/gm.

Additional Steam Generator Tube Fulls A total of 5 steam generator tube intersections from Farley Unit I and 13 from Farley Unit 2 have been' pulled and evaluated since 1985.

Included in the alternate plugging criteria databases are 5 intersections from Farley Unit 1 and 5 intersections from Farley Unit 2.

The existing tube pull results substantially confirm that the dominant flaw feature affecting the integrity at Farley is ODSCC.

Therefore, Farley does not intend to pull additional tubes for the next Unit 2 cycle.

2 l

l

y A

I

.r;

/

i, l

)

• l '. -

i L

l L

q i.

i 1

4 Safety Analysisi In'.Suppoftofthi

~

~

Technical ~ Specification Changes': Associated With:

.j

. Steam Generator Tube Support Plate' i

i l'

Interim Plugging Cricerits t

l-e a

F f

.4=

i r

1

'i gf

--.4 I

t

+

f' !

t h

I 1.

e a

a f

,.n-

?

i

w. v. -,

+r e-..

.,.. -. -- +

e e M.- e

+

~

i 4

Joseph M. Farley Nuclear Plant - Unit 2 l

Technical Specification Changes Associated With l

Steam Generator Tube Support Plate Interim Plugging Criteria Safety Analysis l

WCAP-12871, Revision 2, "J. M. Farley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tcbe Support Plates", provides the development of a 3.6 volt alternate plugging criteria. Use of a 1.0 volt threshold for application of the current technical specification is more conservative than the use of the 3.6 volt criteria.

Burst pressure for steam generator tubes with flaws within the bounds of the tube support plate will still meet the requirements of Regulatory Guide 1.121.

The Farley interim plugging criteria proposes that tubes within the bounds of the tube support plate with flaw indications with bobbin voltage less than or equal to 1.0 volt be allowed to remain in service.

Steam line break leakage must take into account the expected leakage from these flaws. This will be accomplished by using leakage estimates contained in Section 3.3 of draft l

NUREG-1477 dated June 1, 1993. Use of these limits for leaving tubes in service will ensure that leakage in the event of a steam line break will remain less than 22.8 gpm.

The use of a 22.8 gpm leakage limit is based on an additional change to the FNP technical specifications which will require l

tighter control over reactor coolant activity and gross activity.

As a result of the reduced coolant activity limits, the same level of protection to the public is provided.

I r

,