ML20199E070
| ML20199E070 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 01/12/1999 |
| From: | Ewing E ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20199E075 | List: |
| References | |
| W3F1-99-0001, W3F1-99-1, NUDOCS 9901200302 | |
| Download: ML20199E070 (2) | |
Text
._
!.. g Entngy Oper1tions,Inc.
)
Killona. LA 700f4 i
i Tel 504 739 6242 Early C. Ewing, lil oc Safety & Regulatwy Mairs W3F1-99-0001 A4.05 PR January 12,1999 U.S. Nuclear Regulatory Commission Attn: Document Control Desk l
Washington, D.C. 20555
Subject:
Waterford 3 SES i
Docket No. 50-382 i
License No. NPF-38 Technical Specification Change Request NPF-38-174, lievision 1 Reactor Trip Circuit Breaker Surveillance Interval Extension Gentlemen:
Entergy, on behalf of Waterford 3, originally provided Technical Specification Change Request (TSCR) NPF-38-174 by Letter W3F1-96-0004 dated July 17,1996. This request constituted a lead plant submittal, proposed by Entergy on behalt of the Combustion Engineering Owners Group (CEOG). CE NPSD-951," Reactor Trip Circuit Breakers Surveillance Frequency Extension," was submitted to the Staff for review by the CEOG. The Staff requested additionalinformation by Letter dated September 8,1998. Letter W3F1-98-0174, dated October 22,1998 provided answers to these questions.
On December 16,1998, Entergy discussed this change with Mr. Iqbal Ahmed of the
{ )i Staff. Based on thls conversation, Entergy is submitting the attached revised pages to replace the original TSCR in its entirety. The original TSCR extended the surveillar'ce frequency from monthly to quarterly, extended the allowed outage time (AOT) froro one hour to two hours, and extended the time allowed for reentry into the t}
ACTION (reentn/ Unne) from one hour to two hours. This change retains the surveillance frequency of the original change, but deletes the extension of the AOT and reentry time, t
Entergy concurs with the Staff position that the AOT extension should not be applied to the time to place the reactor trip circuit breakers in the tripped condition. This 9901200302 990112 DR ADOCK 0500o302Y PDR _
l Technical Specification Change Request NPF-38-174, Revision 1 Reactor Trip Circuit Breaker Surveillance Interval Extension W3F1-99-0001 Page 2 l
January 12,1999 change was not justified by CE NPSD-951 Since the time of the original suomittal, a revision to the Combustion Engineering Plants Standard Technical Specifications (NUREG-1432) deleted the note that allows reentry into the ACTION for testing.
Based on this revision, the requested time extension should not be applicable to the reentry time either. Therefore, Entergy desires to delete the reentry time extension from the proposed TSCR.
Entergy has concluded that this change is bounded by the No Significant Hazards Evaluation submitted in the July 17,1996 license amendment request. Therefore, the original No Significant Hazards Evaluation continues to be applicable.
Should you have any questions or comments concerning this request, please contact Early Ewing at (504) 739-6242.
Very truly yours, b Ewing Director Nuclear Safety & Regulatory Affairs ECE/ CWT /rtk
' Attachments:
Attachment A-Existing Specification Attachment B - Proposed Specification cc:
E.W. Merschoff, NRC Region IV, C.P. Patel, NRC-NRR,
- 1. Ahmed, NRC-NRR, J. Smith, N.S. Reynolds NRC Resident inspectors Office l
Administrator Radiation Protection Division (State of Loeisiana)
American Nuclear Insurers
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l NPF-38-174 Revision 1 1
ATTACHMENT B PROPOSED SPECIFICATIONS j
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t TABLE 4.3-1 s.
N REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREfENTS o"8 CH4fGIEL MODES FOR WICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED w
1.
Manual Reactor Trip N.A.
N.A.
R and S/U(1) 1, 2, 3*, 4*, 5*
2.
Linear Power Level - High 5
D(2,4),M(3,4), q 1, 2 Q(4) i 3.
Logarithmic Power Level - High S
R(4)
Q and S/U(1) 28, 3, 4, 5 4.
Pressurizer Pressure - High S
R q
1, 2 5.
Pressurizer Pressure - Low S
R Q
1, 2
[
6.
Containment Pressure - High 5
R q
1, 2 l
M 7.
Steam Generator Pressure - Low S
R q
1, 2 8.
Steam Generator Level - Low S
R q
1, 2 9.
Local Power Density - High 5
D(2,4),R(4,5) Q,R(6)
- 1. 2 t
10.
DM8R - Low S
SG),.D(2,4),
Q,R(6) 1, 2 l
M(8)s R(4,5) 11.
Steam Generator Level - High S
R q
1, 2 k
z 12.
Logic M.A.
N.A.
Q and S/U(1) 1, 2, 3*, 4*, 5*
(IO 5
i S
l
c TABLE 4.3-1 (Continued)
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REACTOR PROTECTIVE INSTRUMENTATION SURVEfLLANCE REWIREIElffS B8 CHAleIEL ORIDES FOR WICH CHANNEL CHANNEL FINICTIONAL
~ SURVEILLAIICE
! FUNCTIONAL UNIT
_ CHECK CALIBRATIOII TEST IS REWIRED
-4 w 13.
Reactor Trip Breakers M.A.
N.A.
M10hS/U(1) 1, 2, 3*
4*
14.
Core Protection Calculators S
D(2,4),R(4,5) Q(9),k6)
N (##> #
1, 2 15.
CEA Calculators S
R Q,R(E',
1, 2 16.
Reactor Coolant Flow - Low S
g a
1, 2 R
s g
sa 5
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