ML18227C704

Letter Request to Amend Appendix a of Facility Operating License

ML18227C704
Person / Time
Site:	Turkey Point
Issue date:	12/21/1976
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
Download: ML18227C704 (14)

Text

NRC FORM 195 U.S. NUCLEAR AEGULATORVCOMMISSION gI7SI'A'O'ISTRIBUTIONFoA PA T'50 DOCKET MATERIAL DOCKET NUMBER 0-2 0 FILE NUMOER Mr. Ste 1lo FROM: Flori~a Power R,Light 0o.

Miami Flori~a 33 101 Robert E. Uhrig I

DATE OF DOCUMENT 12-21-76 DATE RECEIVED u-27-.76 LETTER

'+ORIGINAL 0 COP V ONOTOR IZED

+UNCLASSIFIED PROP INPUT FOAM

~

~ ~

NUMBER OF COPIES RECEIVED 3 signg ~ org.

40 copes DEscAIPTIDN Ltr, w attache~... notorizea 1

~ ~.Ref their 10-04-76 ltr. ~ ~.A~t to OL/change to Appen~ ix A Tech Specs.

concering Unit 3 Cycle 4 Reloa~ Inform. an~ Revise~

Equations f Overtemperqture4 T an~ Overtemp++in Specif ica'ion 2 ~ 3 NCLOSURE ',

~

PLANT HA1CJBKEY,3POINT UNIT 3 I

DO NOT RZXOVZ KCKNovfT,+oppD SAFETY ASSXGNED AD:

~I~IQQ ROJECT M~KRa E

LXC ASST FOR ACTION/INFORMATION RANCID@

PROJECT MANAGER LXC ASST INTERNALDISTRIBUTION REG FILE NRC PDR I & E OELD GOSSICK & STAFF MIPC CASE HANAUER HARLESS PROJECT IfANAGEMENT BOYD P ~ COLLINS HOUSTON PETERSON MELTZ HELTFMES SKOVHOLT LPDR!

TIC!

NSIC!

ASLB!

ACRS CYS HOLDING/'E IIQC chai ~ onr. ls eau SYSTEMS SAFETY IIEINEMAN SCIKOEDER ENGXNEERING PJ CARRY KNIGIIT SXIWEXL PAIKXCK REACTOR SAFE ROSS NOVA K ROS7 TOCZY CHECK AT & X SALT7AaAN RUTBERG EXTERNAL DISTRIBUTION NA~TLA8:

REG V XE LA PDR CONSULTANTS:

PLANT SYSTEMS TEDESCO 0

XPPOLXTO F

'f OPERATING REACTORS STELLO OPERATING TECH EXSENHUT S+A0 h

BUTI.E WKIQIAVFBLM ULRIKSON OR SITE SAFE ERNS T BALLARD SPANGLER SITE TECH GAI,IM ILL STEPP HULIfAN SITE ANALYSIS VOLLMER BUNCH J

COLLINS KRFGER CONTROL NUMBER

- ~ )-

h

~

4

~

hh 4

~

~

p4'

~

~

~

h Lh

~

. p p EC' hh h

h

'late DOCke< FIiII

~cgaTtO g SNIIC iy>l I/i 6

0<

m 'L~Igf-U ~ > ~ ~<>iaNiaclOH poll ~c>lo C

FLORIDAPOWER & LIGHTCOMPANY L

December 21, 1976 L-76-430 Oigice of Nuclear Reactor Regulation

~

Af Attention:

Nr. Victor Stello, Director Division of Operating Reactor Cyg )97~ ',9 U. S. Nuclear Regulatory Commission Washington, D.

C.

20555 g:1'II@'go~>>'

Dear Hr. Stello:

-'g'(~

-.'gI~~li'lj ~

Re:

Turkey Point Unit 3 Docket No. 50-250 Proposed Amendment to Facility 0 crating License DPR-31 In accordance with 10 CFR 50.30 Florida Power 6 Light. Company hereby submits three (3) signed originals and forty (40) copies of a request to amend Appendix A of: Facility Operating License DPR-31.

This proposal is being submitted as a followup to our le ter L-76-347 of October 4,

1976 (Unit 3 Cyclo 4 Reload Information).

In our letter we indicated that we might elect to increase system operating pressure at some point during Cycle 4.

A Technical Specification change supporting such a pressure change has been prepared and is attached, for your review.

The proposed change 's discussed below and shown on the accompanying Technical Specification pages bearing the date of th's letter in the lower right hand corner.

Please

note, however, that this amendment is not required to conduct the eload or to return to full power following the reload.

Pacae 1-6 he Un-'

3 fuel residence time limit of Specification 1.16.1 is changed to 27,000 EFPH.

Pages 2.3-2 and 2.3-3 The equations for Overtemperature bi and Overpower bT in Specification 2.3 are revised to support operation at 2250 psia.

The limit for low pressurizer pressure is also revised.

~rp ~.j $

PEOPLE... SERVING PEOPLE

1 0

w II,

'-'- ~~~>..

lr v

I V

t J

l 1~"

'E

Office of Nuclear Reactor Regulation Attention:

Mr. Victor Stello, Director Page Two Pa e 3.1-7 The Unit 3 DNB parameters of Specification 3.1.6 are revised to support operation at 2250 psia.

The proposed amendment has been reviewed by the Turkey Point Plant Nuclear Safety Committee and the Florida Power 6 Light Company Nuclear Review Board.

They have concluded that it does not involve an unreviewed safety question.

A safety evaluation is attached.

Very truly yours, Robert E. Uhrig Vice President REU/MAS/cpc Attachments cc:

Mr. Norman C. Moseley Robert Lowenstein, Esquire

0

STATE OF FLORIDA

)

)

COUNTY OF DADE

)

SS

~

Robert E. Uhrig, being first duly sworn, deposes and says:

Tnat he is a Vice President of Florida Power

& Light Company, the Licensee herein; That he has executed the foregoing document; that the state-men s made in this said document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Licensee.

Robert E. Uhrig Subscribed and sworn to

/

day oZ (~) '~

before me this

,~9,(

):/

~/L~i,.I '"

/Ig'OTARY PUBLIC, in and for the County of Dade, State of Florida t4OTARY 7IJII'IC SThsc CF FIQIIIDA 47 I >~~c MY COhIMISSION H<7IIlKS JAN.

25, 1979 ENDED TI KD g N'gh INSURANCE UNCigqVr~)Ifg i~iy commission expires:

l.16 L'iTEREH LLGTS 1.16.1 Fuel Residence Tine Limit'he fuel residence time for Unit 3 shall be limited to 27,000 EFPH.

The fuel residence time for Unit 4 shall be limited to 30,000 EFPH.

1.16.2 Reactor Coolant Pumps Operation The reactor shall not be operated with less than three reactor coolant pumps in operation.

1.17 LOW POWER PHYSICS TESTS Low power physics tests are tests below a nominal 5% of rated power which measure fundamental characteristics of the reactor core and related instrumentation.

1-6 12/21/76

Unit No.

3 Reactor Coolant Temperature Overtempera-ture 4T

< 4T0 1 + vlS 1

2 1 + t S

(T-574.2) + K (P-2235) - f (4q) 3 4T0 T

~ Xndica ted 4T at rated

power, F

~ Average temperature, F

P

~ Pressurizer

pressure, psig Kl KZ K3 f(4q)

~ 1.095

~ 0.0107 per 'F

~ 0.000453 71

= 2S seconds

-v2 =

3 seconds

~ a function of thc indicated difference between top and bottom detectors of the power-range nuclear ion chambers; vith gains to be selected 1

based on measured instrument response during post-refueling startup tests such that:

\\

For (q q } wi'thin "+10 percent and -14 percent" t

b where q

and q

are the percent power in the'op t

b.

and bottom halves of the core respectively, and q

+ q is total c'ore.power in percent of rated.

t b

power, f(4q)

~ 0.

4 For each percent. that the magnitude of (q - q )

exceeds

+10 percent, the Delta-T trip set point shall be automatically reduced by 3.5 percent of" 0

\\

its value at.interim power.

For each percent that the magnitude of (q - qb) exceeds

-14 percent, the Delta-T trip set point shall be automatically reduced by 2 percent of;.

1 its value at interim power.

2 ~ 3~2 Unit Ho.

3 12/21/76

Unit No.

3 Over-power AT

< AT 5T T

, T-K<.(T-T') f (dq) 4 5 1+vSl 3g Indicated 5T at rated power, 'F Average temperature, F

Indicated average temperature at nominal conditions and rated power, F

K

~

1.11 4

K

~

0 for decreasing average temperature, 5

0.02 per 'F for increasing average temperature K6

~

0 for decreasing average temperature, 0.00068 per 'F for increasing average temperature f(Aq) ~

As defined above

= 10 seconds Pressurizer

~

Low-Pressux'izer pressure equal to or greater than 1835 psig.

High Pressurizer pressure - equal to or less than 2385 psig.

High Pressurizer water level - equal to or less than 92X of full scale.

Reactor Coolant Flow Low reactor coolant flow equal to or greater than

'90X of normal indicated flow Low reactor coolant p mp motor frequency - equal to or greater than 56.1 Hz Under voltage on reactor coolant pump motox bus equal to or greater than 60X of normal voltage Steam Generators Low-low steam generator water level - equal to or greater than 5X of narrow range instrument scale 2 ~ 3-3 Unit No.

3 12/21/76

~

~

~

~

UNIT NO.

3 6 ~

DNB PAKQKTERS The following DN3 related parameters limits shall be maintained during power operation:

a.

Reactor Coolant System Tavg

< 578.2 'F b.

Pressurizer Pressure

> 2220 psia*

c.

Reactor Coolant Flow > 268,500 gpm With any of the above parameters exceeding its limit, restore the pazameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 5X of rated thermal power using normal shutdown 'procedures.

Compliance with a., and b. is demonstrated by verifying that each of. the parameters is within its 'limits at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Compliance with c. is demonstrated by verifying that the parameter is within its limits after each refueling cycle.

• Limit not applicable during either a THERMAL POWER ramp increase in

, e-cess of (5/)

RATED THER?M POWER per minute or a THERMAL POWER step "increase in excess of (10Z)

RATED -THERMAL POWER.

3.1-7 UNIT NO.

3 12/21/76

~

~

SAFETY EVALUATION 1.0 Introduction This safety evaluation supports the following proposed changes to the Unit 3 Technical Specifications:

Primary system operating pressure is increased.

from 2100 to 2250 psia.

This affects the Overtemperature dT equation, the Overpower hT

equation, the limit for low pressurizer

pressure, and the limits for certain DNB parameters.

The fuel residence time limit for Unit 3 is increased from 23,500 EFPH to 27,000 EFPH.

2.0 Discussion S stem Pressure Increase As noted in our letter L-76-347 of October 4,

1976, no changes in the facility operating license or the technical specifications were anticipated as a result of the Unit 3 Cycle 4 refueling, which is currently in progress.

Ye intended at that time to continue operation with a system pressure of 2100 psia.

However, our letter also noted that the conclusions of the Cycle 4 Reload Safety Evaluation were also applicable to operation with a system pressure of 2250 psia.

So, now that the December, 1976 ECCS re-evaluation has led to a proposed technical specification change for F

we have decided to also propose operating Cycle 4

a8 2250 psia to take advantage of the increased fuel residence time afforded by the higher system pr ssure The clad flattening time, assuming 2250 psia system pressure throughout Cycle 4 is predicted to be 27,000 RRPH for the limiting region (Region 2

using the current Westinghouse evaluation madel.

Therefore at a system pressure of 2250 psia, Region 2 has a

nominal Cycle 4 allowed residence time of 9,900 EFPH (Cycle 2 lifetime was 6800

EFPH, and Cycle 1 lifetime was 10,300 EFPH).

The proposed Technical Specification changes are consistent with the design and safety evaluation conclusions of the previously submitted Reload Safety Evaluation (L-76-347).

.~

~

~

)

SAFETY EVALUATlON (Continued) 3.0 Conclusions Based on these considerations, (1) the proposed change does not increase the probability or consequences of accidents or malfunctions of equipment important to safety and coes not reduce the margin of sa=ety as defined in tne basis for any technical specificat'on, therefore, the change does not involve

a. sign'ficant hazards consideration, (2). there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, and (3) such activities will be conducted in compliance with the Commission's egulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the publica

George, R. A., et al "Revised Clad Flattening Aodel",

9/CAP 8377 (Proprietary) and NCAA 8331

(."on Proprietary),

July, 1974.