ML20199D244

From kanterella
Jump to navigation Jump to search
Amends 87 & 87 to Licenses NPF-72 & NPF-77,respectively, Changing TS Requirements for SG Water Level to Support SG Replacement
ML20199D244
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 01/15/1998
From: Dick G
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20199C170 List:
References
NUDOCS 9801300171
Download: ML20199D244 (14)
v • d • e


Text

_

t** C8 vg p'

t UNITED STATES

}

NUCLEAR REGULATORY COMMISSION f

WASHINGTON D.C. 3000H001

%,...../

COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-4M BRAIDWOOD STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 87 License No. NPF-72 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated February 18,1997, as supplemented by lette'

!sd September 22,1997, complies with the standards and requirt ents of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:

9801300171 900115 PDR ADOCK 05000454 P

PM

2 (2)

Iftnical Soecifications The Tecnnical Specifications contained in Appendix A as revin; Wough Amendment No.

87 and the Environmental Protection Pie contained in Appendix B, both of which are attached hereto, are hereby uncorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/

Geo e F. Dick, S ior Project Managr Project Directorate 1112 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of lasuance:

January 15, 1998 I

~

l

  • *tou 3

'4 UNITED 6TATES

]

NUCLEAR REGULATORY COMMISSION

't WASHINGTON, D.C. 30seH001

  • * *. *,G COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-457 RB&l.QL%)OD STATION. UNIT NO 2 AMENDMENT TO FACILITY OPERATING U.QfJGg Amendment No.87 License No. NPF-77 1.

The Nuctsar Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated February 18,1997, as supplemented by letter dated September 22, igg 7, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations eet forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is rease.; sbie assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:

I

4 2-(2)

Technical Soecificationg The Technical Specifications contained in Appendix A as revised through Amendment No. 87 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date ifits issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION 4

J-u p 7 George F. Dick, Se or Project Manager Project Directorate ill-2 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

January 15, 1998 4

1 i

l l

l l

AYTACHMENT TO LICENSE AMENDMENT NOS. 87 AND 87 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 ANDATE 50-457 Replace the following pages of the Appendix "A" Technicsl Specifications with the attached pages. The revised pages ar6 identified by amendment number ats.':entain verticallines indicating the area of change. Pages indicated by an asterisk are provided for convenience ony.

Remove Paoes insert Paaes a

25 25 3/4325 3/4325 3/4326 3/4 3-26

'3/4 4 1

  • 3/4 4-1 3/4 4-2 3/4 4-2
  • 3/4 4-3
  • 3/44-3 W444 W444 3/445 3/445
  • 3/4 4-6
  • 3/44-6

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLri VALUE

12. Reactor _ Coolant Flow-Low 190% of loop mini-289.3% of loop mini-zum measured flow num measured flow
13. Steam Generator Water Level Low-low a.

Unit 1 233.0% (prior to 231.0% (prior to cycle 8) cycle 8) 218.0% (cycle 8 216.1% (cycle 8 and after) of

'and after) of narrow range narrow range instrument span instrument span b.

Unit 2 236.3% of narrow 234.8% of narrow range instrument range instrument span span

14. Undervoltage.- Reactor 25268 volts -

14920 volts -

Coolant' Pumps each bus each bus 15..Underfrequency - Reactor 157.0 Hz 256.08 Hz Coolant Pumps

16. Turbine Trip a.

Emergency Trip Header 21000 psig 2815 psig Pressure b.

Turbine Throttle Valve 21% open 21% open Closure

17. Safety Injection Input N.A.

N.A.

[

from ESF

18. Reactor Coolant Pump N.A.

N.A.

Breaker Position Trip

  • Minimum measured flow = 92,850 gpm Ul!AIDWOOD - UNITS 1 & 2 2-5 AMEN 0 MENT NO.87 l

.n T7mfw 3.3-4 iContinuedi

~

~

.ENGI"***en SAFETY FEATURES ACTUATION SYSTEM IN51RUnEWTATION TRIP SETPOINTS TRIP ALLONhBLE FUNCTIONAL UNIT SETPOINT VAIRE 4.

Steam Line Isolation a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic and Actuation Relays N.A.

N.A.

c.

Contairusent Fressure-High-2 58.2 psig 59.4 psig d.

' Steam Line Pressure-Low (Above P-11) 2640 psig*

2614 psig*

e.

Steam Line Pressure Negative Rate-High l

(Below P-11) 5100 psi +*

5165.3 poi ++

5.

Turbine Trip and Feedwater Isolation l

l a.

Automatic Actuation Logic and Actuation Relays N.A.

N.A.

b.

Steam Generator Water Level-High-High (P-14)

1) Unit I s81.4% (prior s83.4% (prior to cycle 8) to cycle 8) 588.04 (cycle 8 589.9% (cycle 8 and after) of and after) of narrow range narrow range instrument span instrument span
2) Unit 2 580.8%-of 582.8% of narrow range narrow range instrument span instrument span BRAIDWOOD - UNITS 1 & 2 3/4 3-25 AMENDMENT NO.87

'E"-

E'l

- ' ' ' - - - ' ' ' ^ ^ ' ' '

TABLE 3.3-4 fContinued1 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INsinUnshinTION TRIP sairvinis TRIP ALIstABLE FUNCTIONAL UNIT SETPOINT VALUE i

5.

Turbine Trip and Feedwater Isolation (continued) c.

Safety Injection See Itee 1. above for all Safety Injection Trip Setpoints and Allowable Values.

6.

Auxiliary Feedwater a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic and Actuation Relays M.A.

N.A.

c.

Steam Generator Water Level-Low-Low-Start Motor-Driven Pump and Diesel-Driven Pump

1) Unit 1 233.04 (prior to 231.04 (prior to cycle 8) cycle 8) 218.04 (cycle 8 216.1% (cycle 8 and after) of and after) of narrow range narrow range instrument span instrument span
2) Unit 2 t

236.3% of 234.8% of narrow range narrow range instrument span instrument span d.

Undervoltage-RCP Bus-25268 volts 24920 volts l

Start Motor Driven Pump and Diesel-Driven Pump e.

Safety Injection-Start Motor-1

(

Driven Pump and See Item 1. above for all Safety Injection Trip Setpoints and l

Diesel-Driven Pump Allowable Values.

i BRAIDNOOD - UNITS.1 & 2 3/4 3-26 AMENDMENT NO. 87

i 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND. COOLANT CIRCULATION STARTUP AND POWER OPERATION

-LIMITING CONDIT'40N FOR OPERATIONS 3.4.1.1 All reactor coolant loops shall be in operation.

APPLICABILITY:

MODES 1 and 2.*

ACTION:

With less than the soove required reactor coolant loops in operation, be in at least H0T STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"See Special Test Exceptions Specification 3.10.4.

BRAIDWOOD - UNITS 1 & 2 3/4 4-1

ll.

\\

f4!LCI91LSQ91&RT RHIM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with two reactor coolant loops in operation when the Reactor Trip system breakers are closed and one reactor coolant loop in operation when the Reactor Trip system breakers are opent*

a.

Reactor Coolant Loop A and its associated steam generator and reactor coolant pump, b.

Meactor Coolant Loop B and its associated steam generator and reactor coolant pump, c.

Reactor Coolant Loop C and its associated steam generator and reactor coolant pump, and d.

Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.

APPLICABILITY:

MODE 3.**

ACTION:

With less than the above required reactor coolant loops OPERABLE, a.

restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With only one reactor coolant loop in operation and the Reactor Trip system breakers in the closed position, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> open the Reactor Trip System breakers, c.

With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant system and immediately initiate corrective action to return the required reactor coolant loop to operation.

SURVEILLANCE REOUIREMENTS 4.4.1.2.1 At least the above required reector coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 18% (41% for Unit 1 prior to cycle 8) at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l 4.4.1.2.3 The required coolant loops shall be verified in operation and circu-lating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

  • All Reactor Coolant pumps may be doenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the Reactor Coolant system boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
    • see Special Test Exceptions Specification 3.10.4.

BRAIDWOOD - UNITS 1 & 2 3/4 4-2 AMENDMENT NO. 87

d 4

REACTOR-COOLANT SYSTEM HDT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation:"

a.

Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,**

b.

Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,**

c.

Reactor Coolant Loop C and its associated steam generator and reactor

solant pump,**

d..

Reactor Coolant Loop D and its associated steam generator and reactor coolant pump,**

e.

RHR Loop A, and f.

RHR Loop B.

APPLICABILITY: MODE 4.

ACTION:

a.

With less than the above required reactor coolant and/or RHR loops OPERABLE, immediately initiate corrective action to-return the required loops to'0PERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

With no reactor coolant or RHR loop in operation,-suspend all operations involving a reduction in bnron concentration of-the Reactor Coolant System and immediately initiate corrective action to 4

return the required coolant loop to operation.

"All reactor coolant pumps and RHR pumps may be deenergized for up to I hour provided:- (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

BRAIDWOOD - UNITS 1 & 2 3/4 4-3

REACTOR COOLANT SYSTEM

- I SURVEILLANCERE0V1Rf'iENis 4.4.1.3.1 The required reactor coolant pump (s) and/or RHR pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.3.2 The required steam generator s) shall be determined OPERABLE by verifying secondary side narrow range wa(ter levei to be greater than or equal to 18% (41% for Unit 1 prior to cycle 8) at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

BRAIDWOOD - UNITS 1 & 2 3/4 4 AMENDMENT NO. 87 i

REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LiMf TIl00 'NDITION FOR OPERATION 3.4,1.4.1 At least 'ne residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:

a.

One additional RNR loop shall be OPERABLEd, or b.

The secondar side narrow range water level of at least two steam generators shall be greater tt.an 18% (41% for Unit 1 prior to cycle 8).

APPLICABILITY: MODE 5 with reactor coolant loops filled ##.

AC1 ION:

a.

With one of the RHR loops inoperable and with less than the required staan generator level, launediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator level as soon as possible.

b.

With no idtR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate co rective action to return the required RHR loop to operation.

SURVEILLANCE REOUIREMENTS 4.4.1.4.1.1 The secondary side water level of at lee.st two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.4.1.2 At least one RHR loop shalt, be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

  • The RHR pump may be doenergised for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coelant System boron concentration,-and (2) core outlet temperature is maintained at least 10*F Lelow saturation taiperature.
  1. 0ne RHR loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RNR loop is OPERABLE and in operation.

j.

    1. A reactor coolant-pump shall not be started with one or more of the Reactor Coolant system cold leg temperatures less than or equal to 350*F unless the secondary water temperature of each-steam generator is less than 50*F above each of the Reactor Coolant system cc'1 leg temperatures.

RRAIDWOOD'- UNITS 1 & 2 3/4'4-5 AMENDMENT NO.

87 1

l REACTOR C0OLANT SYSTEN l

COLD $HUTDOWN-LOOPSNOTFILLEg

[]MITINGCONDITIONFOROPERATION

\\

3.4.1.4.2 Two residual heat removal tHR loops shall be OPERABLE

  • and at leastoneRHRloopshallbeinoperatjon.)*

l APPLICABILITY: MDOE 5 with reactor coolant loops not filled.

Ell.9!!!

1 With less than the above required RHR loops OPERA 8LC, immedistely a.

initiate corrective action to return the required RHR loops to j

OPERABLE status as soon as possible.

b.

With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

$URVEILLANCE REQUIRENENTS 4.4.1.4.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"One RHR loop may be inoperable for up to 2 holes for surveillance testing provided the other RNR loop is OPERA 8LE and in operation.

    • The RHR pump may be doenergized for up to I hour provided: (1) to operations are permitted that would cause dilution of the Reactor Coolant System boron concentration and (2) core outlet temperature is maintained at least 10*F belowsaturatIontemperature.-

BRAIDWD00 - UNITS 1 & 2 3/4 4-6 3

-v

-y

.w-9 v-

--rev e,-

...e--,,t=-a

,-.~.',3--,..-.,.

-i,-vw-_-

- - - - - - - - - - - - - - - - - -

Retrieved from "https://kanterella.com/w/index.php?title=ML20199D244&oldid=4089977"