ML18096B494

Proposed Amendment to Facility Operating License DPR-67. Proposed Change Consists of Revision Regarding Appendix a Concerning the Proposed Pump and Valve Testing Program

ML18096B494
Person / Time
Site:	Saint Lucie
Issue date:	03/18/1977
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
L-77-88
Download: ML18096B494 (11)

Text

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Ao-0.38 FILE NUMBER U.S. NUCLEAR REGULATORY COMMISSION FROM:PPL Miami, 11.

33101 Robert E. Uhrig TO:

Mr. Stello DATE OF DOCUMENT O3-18-77 DATE RECEIVL=D 03~9-77 NUMBER OF COPIES RECEIVED C3NOTOBIZED NC LASS IF I E D e+.ETTE R

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DESCRIPTION Ltr. Notorized 03-18-77...Trans The POD.owing:

ENCLOSU RE Amdlt to OL/Change to Tech Specs:

Consisting of revisions regarding, Appendix A-concerning the proposed pump andLvalve teston program...With Attachments "A" 8 "B"...'.

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'. O. BOX 013100, MIAMI, FL 33101 llMIIUIIItMtf(HHLfll5Pll Director of Nuclear Reactor Regulation Attn:

Victor Stello, Jr., Director Division of Operating Reactors U.

S. Nuclear Regulatory Conmission Washington, D. C.

20555 FLORIDAPOWER & LIGHT COMPANY March 18, 1977 L-77-88 8

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Dear Mx. Stello:

Re:

St. Lucie Unit 1 Docket No. 50-335 Proposed Amendment to Facilit 0 eratin Li'cense'PR-67 In accordance with 10 CFR 50.30, Florida Power 6 Light: Company submits herewith three (3) signed originals and forty (40) copies of a request to -amend Appendix A of Facility Operating

, License DPR-67.

This letter concerns. the proposed pump and valve test program and associated Technical Specifications for St. Lucie Uni~ 1.

The proposed Technical Specification changes are described below and shown on the accompanying Technical Specification.

pages (Attachment A)'earing the date of this letter in the lower right hand corner.

The proposed pump and valve testing program is described in Attachment B.

The following Specifications are added, revised, or deleted to conform with the proposed pump and valve testing program.

S ecification 1.21 (revised)

. Pacae 1-5 Remarks Definition 1. 21 (Staggered Test Basis) is not applicable to components tested per the appli-cable ASME Code and Addenda.

4.0.5 (added) 3/4 0-2 pew Specifica ion 4.0.5 is added to specify the pump and valve test requirements.

4.1.2.l.a.l (deleted) 3/4 1-8 Boration Systems Shutdown Flow Paths:

cycling of power operated or automatic valves.

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'irector of Nuclear Reactor Regulation Page Two March 18, 1977 Specification 4.1.2.2.a.l (deleted) 4.1.2.2.c.l (deleted) 4.1.2.3.a, b,

(revised)

Page 3/4 1-10 3/4 1-11 c

3/4 1-12 Remarks Boration Systems Operating Plow Paths cycling of power operated or automatic valves.

P Boration Systems Operating Flow Paths:

cycling of power operated valves not testable during plant operation.

Boration Systems Charging Pump (shutdown):

charging and HPSI pump tests.

4.1.2.4.a s b (revised) 3/4 1-13 Boration Systems Charging Pump (operating):

charging pump tests.

4.1.2.5.a, b,

6 c 3/4 1-14 (revised)

Boration Systems Boric Acid Pumps (shutdown):

boric acid pump tests.

4.1.2.6.a, b,

(revised) c 3/4 1-15 Boration Systems Boric Acid Pumps (operating):

boric acid pump tests.

4.4.3 (revised) 3/4 4-3 Reactor Coolant System Safety Valves (Operating):

pressurizer'ode safety valve tests.

4.5 2 b 1, 2,

s 4

(deleted) 3/4 5-4@5 Emergency Core Cooling Systems-ECCS Sub-systems (Tavg

> 300 F):

HPSI pump, LPSI pump, and power operated valve tests.

4.5.2.e.l (deleted) 3/4 5-6 Emergency Core Cooling Systems ECCS Sub-systems (Tavg

> 300 F):

cycling of power operated valves not testable during plant operation.

4.6.2.l.a.l, 2, 3, 3/4 6-156 6

4 (deleted) 16 Depressurization and Cooling Systems Containment Spray System:

spray pump and power operated/automatic valve tests.

4.6.2.1.b.l (deleted) 3/4 6-16 Depressurization and Cooling Systems Containment Spray System:

cycling of power operated valves not testable during plant operation.

1 Director of Nuclear Reactor Regulation Page Three March 18, 1977 S ecification 4.6.3.l.l.a a b (revised)

Pacae 3/4 6-186 19 Remarks Containment Systems Containment. Isolation Valves:

valve tests.

4.6.3.1.2.b 6 c (deleted) 3/4 6-19 Containment, Systems Containment Isolation Valves:

valve tests.

4.7.1.1 (revised) 3/4 7-1 Plant Systems Turbine Cycle:

main steam line code safety valve tests.

4.7.1.2.a.l, 2, 3, 4 (deleted) 3/4 7-4@5 Plant Systems Auxiliary Feedwater System:

auxiliary feedwater pump and power operated/automatic valve tests.

4.7.1.2.b (deleted)

.3/4 7-5 Plant Systems Auxiliary Feedwater System:

cycling of power operated valves not testable during plant operation.

4.7.1.5.a s

b (revised) 3/4 7-9 Plant Systems Main Steam Isolation Valves:

MSIV tests.

4.7.3.l.a.l, 2,

3, 4,

6 5 (deleted) 3/4 7-14 Plant Systems Component Cooling Water System:

CCN pump and power operated/automatic valve tests.

4.7.3.1.b.l (deleted) 4 3/4 7-15 Plant Systems Component Cooling Water System:

cycling of power operated valves not testable during plant operation.

4.7.4.l.a.l, 2,

3, 4,

6 5 (deleted) 3/4 7-16 Plant Systems Intake Cooling Water System:

pump and power operated/automatic valve tests, 4.7.4.l.b.l (deleted) 3/4 7-17 Plant Systems Intake Cooling Water System:

cycling of power operated valves not testable during plant. operation.

4 1)

Director of Nuclear Reactor Regulation Page Four March 18, 1977 Pacae Remarks 3/4. 0. 5 B3/4 0-3 A new Bases section is added to correspond to new Specification 4.0.5.

3/4. 6. 1. 2 3/4.7.1.1 B3/4 6-1 B3/4 7-1 The Bases section on Containment Leakage is revised to include reference to surveillance testing performed in accordance with the applicable ASME Code and Addenda.

The Bases section on Safety Valves is revised to include reference to surveillance testing performed in accordance with the applicable ASME Code and Addenda.

The proposed Technical Specification changes and the pump and valve testing program are being submitted to satisfy a January 14, 1977 letter from your staff.

The January 14 letter states that any in-service tests of pumps and valves conducted on or after July 1, 1977, shall comply with those requirements in editions of the Code and Addenda referenced in 10 CFR 50.55a(b).

However, before the proposed program can be fully implemented after approval, numerous procedures will have to be written or modified.

Approximately 6

months will be needed to develop the required procedures and imple-ment ouz proposed pzogram.

Due to the magnitude of the effort in-volved, it will not be practical to commence the modification and preparation of the required procedures until the proposed

program, including Technical Specifications, has been accepted by your staff.

Therefore-, Florida Power

& Light Company respectfully requests, in accordance with 10 CFR 50.12, an exemption from the requirements of 10 CFR 50.55a(g)(4)(v) and the Staff's January 14, 1977 letter for the 6-month period beginning either on July 1, 1977, or on the date we receive NRC approval of our proposed pump and valve program, whichever occurs later.'ntil.that time, we plan to continue to comply with the existing Technical Specification surveillance re-quirements.

Our request for exemption has been reviewed by the St. Lucie Facility Review Group (FRG) and the Florida Power 6 Light Company Nuclear Review Board (CNRB) with the following conclusions:

1)

The proposed inservice examination and testing program provides some additional assurance over the already adequate requirements of the Facility Technical Speci-fications that certain safety-related components are functioning properly.

&1 II 1

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Director of Nuclear Reactor Regulation Page Five March l8, l977 2)

Delay in implementing this proposed program will not endanger life or property or the common defense and security of the public.

3)

Ensuring the careful preparation,

review, and approval of implementing procedures prior to implementation of the proposed program is in the public interest.

4)

The request for exemption, as stated above, is necessary, and the time period of the exemption is appropriate.

The FRG and CNRB have also reviewed the proposed pump and valve test, program and associated Technical Specifications.

They have concluded that they do not involve an unreviewed safety question.

Very truly yours, Robert E. Uhrig Vice President REV/mS/cpc cc:

Mr. Norman C. Moseley, Region II Robert Lowenstein, Esquire

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