Text

Amends 214 & 155 to Licenses DPR-57 & NPF-5,respectively, Revising TS to Accommodate Increase in Max Licensed Thermal Power Level from 2558 Megawatts Thermal (Mwt) to 2763 Mwt
	ML20155C653
Person / Time
Site:	Hatch
Issue date:	10/22/1998
From:	Collins S; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20155C657	List: ML20155C653; ML20155C660;
References
	NUDOCS 9811020309
	Download: ML20155C653 (32)
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9a UNITED STATES

f j

NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 2066Mm01 6

4....

4 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF D ALTON. GEORGIA DOCKET NO. 50-321 EDWIN 1. HATCH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 214 License No. DPR-57

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

A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 1 (the facility)

Facility Operating License No. DPR-57 filed by Southern Nuclear Operating Company, Inc. (Southern Nuclear), acting for itself, Georgia Power Company, Oglethorpe Power l

Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated August 8,1997, as supplemented by letters dated March 9, May 6, l

July 6, July 31, September 4, and September 11,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the l

Commission's rules and regulations as set forth in 10 CFR Chapter I; 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 set forth in l

10 CFR Chapter I; 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.

9811020309 981022 PDR ADOCK 05000321 P

PDR

I r

l 2. Accordingly, the license is hereby amended by page changes to Facility Operating License No. DPR-57 and page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 214 are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Enviromental Protection l

Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented prior to startup from the next refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION au i

irec r Office of Nuc ear Reactor Regulation

Attachment:

Technical Specification and Operating License Changes Date of issuance: October 22, 1998

m.

ATTACHMENT TO LICENSE AMENQMENT NO. 214 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Facility Operating License and the Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.

Facility Operating License Remove insert 3

3 Technical Specifications Remove Insert 1.1-5 1.1-5 3.3-2 3.3-2 3.3-5 3.3-5 3.3-7 3.3-7 3.3-8a 3.3-8a 3.3-27 3.3-27 3.3-28 3.3-28 3.3-29 3.3-29 3.4-25 3.4-25 3.4-26 3.4-26 3.4-27 3.4-27 5.0-16a 5.0-16a j

F i l the procedures and limitations set forth in this license; and the j

Georgia Power Company, the Oglethorpe Power Corporation, the Municipal Electric Authority of Georgia and the City of Dalttn, i

J Georgia to possess but not operate the facility in accordance with the procedures and limitations set forth in this license; (2) Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear m.terial as i

reactor fuel, in accordance with the limitations for storage and j

amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and i

70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation j

monitoring equipment calibration, and as fission detectors in amounts as required; i

I (4) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and l

70, to receive, possess and use in amounts as required any

{

byproduct, source or special nuclear material without restriction l

to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components;

}

i 3

(5) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

4 C.

This license shall be deemed to contain and is subject to the i

conditions specified in the following Comission regulations in i

10 CFR Chapter I:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50-54 and 50-59 of Part 50, and Section 70.32 of Part i

70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Comissign now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power levels not in excess of 2763 megawatts thermal.

l 3 The original licensee authorized to possess, use and operate the facility was Georgia Power Company (GPC).

Consequently, certain historical references to GPC remain in the license conditions.

Amendment No.

214

Definitions 1.1 1.1 Definitions (continued)

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a.

Described in Section 13.6, Startup and Power Test Program, of the FSAR; b.

Authorized under the provisions of 10 CFR 50.59; or c.

Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2763 MWt.

l REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM)

SDM shall be the amount of reactivity by which the reactor is suberitical or would be subcritical assuming that:

a.

The reactor is xenon free; b.

The moderator temperature is 68'F; and c.

All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance (continued)

HATCH UNIT 1 1.1-5 Amendment 'No.- 214

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMP!?: TION TIME s

C.

One or more Functions C.1 Restore RPS trip I hour with RPS trip capability, capability not maintained.

D.

Required Action and D.1 Enter the Condition Immediately

' associated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for B, or C not met.

the channel.

E.

As required by E.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action D.1 to < 28% RTP.

l and referenced in-Table 3.3.1.1-1.

1 F.

As required by F.1 Be in MODE 2.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

G.

As required by G.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months t

Required Action D.1 and referenced in Table 3.3.1.1-1.

H.. As required by H.1 Initiate action to Immedit tely Required Action D.1 fully insert all and referenced in insertable control Table 3.3.1.1-1.

rods in core cells containing one or more fuel assemblies.

l l

(continued)

HATCH UNIT 1 3.3-2 Amendment N6.' f f4 i

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) sSURVEILLANCE FREQUENCY SR 3.3.1.1.7

NOTE-------------------

Only required to be met during entry into MODE 2 from MODE 1.

l Verify the IRM and APRM channels overlap.

7 days SR 3.3.1.1.8 Calibrate the local power range monitors.

1000 effective full power hours l

l SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST.

92 days SR 3.3.1.1.10

NOTE-------------------

For Function 2.a. not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST.

184 days l

SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 184 days Turbine Contrni Valve Fast Closure, Trip Oil Pressure - Low Functions are not I

bypassed when THERMAL POWER is 2: 28% RTP.

l l

l I

l SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.

18 months (continued) i HATCH UNIT 1 3.3-5 Amendment No. 214

_ - _ =.=.

- ~ _ -..

RPS Instrumentation 3.3.1.1 l

Table 3.3.1.1 1 (page 1 of 3)

Reactor Protection system Instrumentatlon l

\\

APPLICABLE CONDITIONS j

MOCES OR REQUIRED REFERENCE 0 OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE l

1.

Intermediate Range Monitor a.

Neutron Flux-Nigh 2

3 G

SR 3.3.1.1.1 s 120/125 sa 3.3.1.1.4 divisions of sa 3.3.1.1.6 futt scale sa 3.3.1.1.7 54 3.3.1.1.13 sa 3.3.1.1.15 5(a) 3 H

st 3.3.1.1.1 s 120/125 sa 3.3.1.1.5 divisions af SR 3.3.1.1.13 futt erste sa 3.3.1.1.15 b.

Inop 2

3 G

sa 3.3.1.1.4 NA st 3.3.1.1.15 5(a) 3 H

sa 3.3.1.1.5 NA st 3.3.1.1.15 2.

Average Power Range Monttor a.

Neutron Flux - Migh 2

3("I G

sR 3.3.1.1.1 5 20% RTP (setdown) st 3.3.1.1.7 sa 3.3.1.1.8 sa 3.3.1.1.10 SR 3.3.1.1.13 b.

simulated Thornet 1

3(C)

F SA 3.3.1.1.1 5 0.58 W +

Power - Migh SA 3.3.1.1.2 58% RTP and l

sa 3.3.1.1.8 s 115.5%

SA 3.3.1.1.10 RTP(b) sa 3.3.1.1.13 3 *I F

sa 3.3.1.1.1 5 120% RTP I

c.

Neutron Flux-Nigh 1

st 3.3.1.1.2 st 3.3.1.1.8 st 3.3.1.1.10 sa 3.3.1.1.13 d.

Inop 1,2 3(c)

G SR 3.'s.1.1.10 NA (continued)

(a) With any control rod withdrawn from a core cett containing one or more fuel assembtles.

(b) 0.58 W + 58%

0.58 AW RTP when reset for single loop operation per LC0 3.4.1, = Recirculation Loops l

cperating."

(c) Each APRM channet provides inputs to both trip systems.

HATCH UNIT 1 3.3-7 Amendment No. 214

RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (page 3 of 3)

Reactor Protectfon system Instrumentation APPLICASLE CONDIT10NS NODES OR REQUIRED REFERENCED OTHER CNANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS STSTEM ACTION D.1 REQUIREMENTS VALUE 8.

Turbine stop t 285 RTP 4

E sa 3.3.1.1.9 510s closed l

Vatwe - C1osure sa 3.3.1.1.11 I

st 3.3.1.1.13 l

SR 3.3.1.1.15 9.

Turbine Control Velve t 285 RTP 2

E SR 3.3.1.1.9 t 600 pois l

Fast Closure, Trip oil SA 3.3.1 :.11 Pressure - Low SR 3.3.1.1.13 SA 3.3.1.1.15 SA 3.3.1.1.16

10. Reactor Mode switch -

1,2 1

G st 3.3.1.1.12 NA shutdown PoeItIon SR 3.3.1.1.15 5(e) 1 N

sa 3.3.1.1.12 NA SA 3.3.1.1.15

11. Manuel seren 1,2 1

C sa 3.3.1.1.5 NA sa 3.3.1.1.15 5(e) 1 H

sa 3.3.1.1.5 NA sa 3.3.1.1.15 (a) With any control rod withdrawn from a core cett contelning one or more fuel assemblies.

I 4

HATCH UNIT 1 3.3-8a Amendment No. 214 1

l l

EOC-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 End of' Cycle Recirculation Pump Trip (E0C-RPT) Instrumentation LCO 3.3.4.1 a.

Two channels per trip system for each E0C-RPT ir.strumentation Function listed below shall be OPERABLE:

1.

Turbine Stop Valve (TSV) -- Closure; and 2.

Turbine Control Valve (TCV) Fast Closure, Trip 011 Pressure -- Low.

QB b.

LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for inoperable E0C-RPT as specified in the COLR are made applicable.

APPLICABILITY:

THERMAL POWER a: 28% RTP.

l ACTIONS

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more channels A.1 Restore channel to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable.

OPERABLE status.

QB A.2

NOTE---------

Not applicable if inoperable channel is the result of an inoperable breaker.

Place channel in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months trip.

(continued) l HATCH UNIT 1 3.3-27 Amendment No. 214 l

E0C-RPT Instrumentation 3.3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One or more Functions B.1 Restore EOC-RPT trip 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with EOC-RPT trip capability.

capability not maintained.

QB AND B.2 Apply the MCPR limit 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for inoperable MCPR limit for EOC-RPT as specified inoperable EOC-RPT in the COLR.

not made applicable.

C.

Required Action and C.1 Remove the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion recirculation pump Time not met.

from service.

Q8 C.2 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to < 28% RTP.

l SURVEILLANCE REQUIREMENTS

______________________________.------NOTE-------------------------------------

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains EOC-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST.

92 days i

(continued)

HATCH UNIT 1 3.3-28 Amenament No. 214~7

EOC-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS (continued)

S URVEILLANCE FREQUENCY S

SR 3.3.4.1.2 Verify TSV -- Closure and TCV Fast 184 days Closure, Trip 011 Pressure -- Low Functions are not bypassed when THERMAL l

POWER is at 28% RTP.

l l

SR 3.3.4.1.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:

TSV -- Closure: s 10% closed; and TCV Fast Closure, Trip Oil Pressure -- Low: a: 600 psig.

l SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.

i SR 3.3.4.1.5

NOTE-------------------

Breaker interruption time may be assumed from the most recent performance of SR 3.3.4.1.6.

Verify the EOC-RPT SYSTEM RESPONSE TIME 18 months on a is within limits.

STAGGERED TEST BASIS l

SR 3.3.4.1.6 Determine RPT breaker interruption time.

60 months i

i I

I HATCH UNIT 1 3.3-29 Amendment No. 2fs

)

l RCS P/T LIMITS 3.4.9 1400 A - SYSTEM

'i f

HYDROTEST UMIT fj INITIAL RTndt VALUES ARE

-20*F FOR BELTUNE, 1300

- WITH FUEL IN THE 20

,=

4o.F FOR UPPER VESSEL, VESSEL FOR j !

k y

AND HATCH 1 l!

1,f -

32 10*F FOR BOTTOM HEAD J J l

1200 1

I

i i

HEATUP/COOLDOWN I

)

1100 RATE 20'F/HR I i j

t i

)

1000 t

BELTINE CURVES C

I

/

ADJUSTED AS SHOWN:

h

' lj

[

EFPY SHIFT ('F) n.

900 20 142 R

,' l l

/ !

800 BELTINE CURVES to

/

ADJUSTED AS SHOWN:

/

EFPY SHIFT (*F)

/ !

/

24 157 70o

+

{

U 5

/i

)

/

BEtriNE CURVES j

600

-i h

ADJUSTED AS SHOWN:

[

}

jf EFPY SHIFT (*F)

! //

28 167 500 BELTINE CURVES y

7; ADJUSTED AS SHOWN:

0:

EFPY SHIFT (*F)

"I 300

!312 PSIG1 32 180 l

Ft.ANGE 200

-- BELTINE LIMITS BELTLINE i

REGION AND i

76*F

- - BOTTOM HEAD 100 BOTTOM UMITS se UPPER VESSEL O

UMITS 0

50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

[ ACAD I F34911 1 i

Figure 3.4.9-1 (Page 1 of 1)

Pressure / Temperature Limits for Inservice Hydrostatic and Inservice Leckage Tests Hatch Unit 1 3.4-25 Amendment No. 214

l l

RCS P/T LIMITS 3.4.9 s

l 1400 e

(

B - CORE NOT l

1300 CRITICAL UMIT INITIAL RTndt VALUES ARE l

l 20*F FOR BELTUNE, FOR HATCH 1 40*F FOR UPPER VESSEL.

1200 l

g l

i 10*F FOR BOTTOM HEAD 1100 f

j f

j RATE 100*F/HR HEATUP/COOLDOWN 1000 I

o 5

1 8

x h

id e

800

?

m BELTINE CURVES 8

700 I

ADJUSTED AS SHOWN:

EFPY SHIFT (*F)

I s

32 180 5

/

E 600 E

l b

)

i 500 e

/

l

=

y 4%

A E

300 BELTINE UMITS I

200 mm

- - BOTTOM HEAD mo

/

UMITS UPPER VESSEL g

68'F UMIIS FLANGE REGON 76*F 0

O 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

[ acAo I F34921 1 Figure 3.4.9-2 (Page 1 of 1)

Pressure / Temperature Limits for Non-Nuclear Hectup, Low Power Physics Tests, and Cooldown Following a Shutdown Hatch Unit 1 3.4-26 Amendment No. 214

RCS P/T UWITS 3.4.9 1400 C-CORE 1300 -

CRITICAL UMIT INITIAL.RTndt VALUES ARE FOR HATCH 1

-20*F FOR BELTUNE.

1200 40'F FOR UPPER VESSEL, Ano 10*F FOR BOTTOM HEAD 1100 HEATUP/COOLDOWN g 1000 RATE 100*F/HR f

g 900 J

h 800 e

r 700 BELTINE CURVES u

ADJUSTED AS SHOWN:

EFPY SHIFT (*F) 32 180 g

w 500 I

400 W

a.

300

~*BELTUNE AND 200 NON-BELTUNE UMITS Minimum Cnticahty 100 Temperature 76 0

o 50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

[ ACADI F34931 1 Figure 3.4.9-3 (Page 1 of 1)

Pressure / Temperature Umits for Criticolity Hotch Unit 1 3.4-27 Amendment No. 214

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Technical So'ecifications (TS) Bases Control Proaram (continued) d.

Proposed changes that meet the criteria of b. above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.12 Primary Containment Leakaae Rate Testina Proaram A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, P., is 50.5 psig.

l The maximum allowable primary containment leakage rate, L., at P, is 1.2% of primary containment air weight per day.

Leakage rate acceptance criteria are:

a.

Primary containment overall leakage rate acceptance criterion is s 1.0 L,.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 1 0.60 L combined Type B and Type C tests, and s 0.75 E,for the for Type A tests; b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is s 0.05 L, when tested at 1 P,,

2)

For each door, leakage rate is s 0.01 L, when the gap between the door seals is pressurized to 110 psig for at least 15 minutes.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

i (continued) l HATCH UNIT 1 5.0-16a Amendment No. 214

l ps>Mouq\\

l UNITED STATES g

j NUCLEAR REGULATORY COMMISSION o

WASHINGTON, D.C. 206edHm1 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

1 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA DOCKET NO. 50-366 EDWIN 1. HATCH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.155 License No. NPF-5

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

A. The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility)

Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (Southern Nuclear), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated August 8,1997, as supplemented by letters dated March 9, May 6, July 6, July 31, September 4, and September 11,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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 l

Commission's regulations and all applicable requirements have been satisfied.

i

_ i

2. Accordingly, the license is hereby amended by page changes to Facility Operating License No. NPF-5 and page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License l

No. NPF-5 is hereby amended to read as follows:

(2) Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental i

Protection Plan contained in Appendix B, as revised through Amendment No.155 are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented prior to startup from the current refueling outage.

I FOR THE NUCLEAR REGULATORY COMMISSION

\\

l a

p ctor Office of Nuclear Reactor Regulation

Attachment:

Technical Specification and Operating License Changes Date of issuance:

October 22, 1998 I

i l

\\

I

- - - - -. ~.

l-l ATTACHMENTTO LICENSE AMENDMENT NO.155 I

FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the Facility Operating License and the Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.

Operating License 1

Bemove Insert 4

4 Technical Specifications Remove Insert 1.1-5 1.1-5 3.3-2 3.3-2 3.3-5 3.3-5 3.3-7 3.3-7 3.3-8a 3.3-8b 3.3-9 3.3-9 3.3-28 3.3-28 3.3-29 3.3-29 3.3-30 3.3-30 3.4-25 3.4-25 3.4-26 3.4-26 3.4-27 3.4-27 5.0-16a 5.0-16a l

l

, C.

This licenset shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40 Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders oftheCommissionnoworgereafterineffect;andissubjectto the additional conditions specified or incorporated below:

(1)

Maximum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power levels not in excess of 2763 megawatts thermal in accordance with the conditions I

specified herein.

(2)

Technical Soecifications The Technical Specifications in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 155 are hereby incorporated

{

in the license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications i

and the Environmental Protection Plan.

l l

l 1

l The original licensee authorized to possess, use, and operate the facility was Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in the license conditions.

Amehdment No.155 i

l

__..____.________m..

Definitions 1.1 1.1 Definitions MINIMUM CRITICAL POWER a)propriate correlation (s) to cause some point in RATIO (MCPR)

(continued) t1e assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive i

combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, j

lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function (s) are also capable of performing their related support function (s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a.

Described in Chapter 14, Initial Tests and Operation, of the FSAR; b.

Authorized under the provisions of 10 CFR 50.59; or c.

Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2763 MWt.

l REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS IIME trip setpoint at the channel sensor until de-energization of the scram pilot valve (continued)

HATCH UNIT 2 1.1-5 Amendment No.155

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME s

C.

One or more Functions C.1 Restore RPS trip I hour with RPS trip capability.

capability not maintained.

4 D.

Required Action and D.1 Enter the Condition Immediately associated Completion referen;ad in Time of Condition A, Table 3.3.1.1-1 for B, or C not met.

the channel.

E.

As required by E.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Required Action D.1 to < 28% RTP.

l and referenced in Table 3.3.1.1-1.

F.

As required by F.1 Be in MODE 2.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

G.

As required by G.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required Action D.1 and referenced in Table 3.3.1.1-1.

H.

As required by H.1 Initiate action to Immediately Required Action D.1 fully insert all and referenced in insertable control Table 3.3.1.1-1.

rods in core cells containing one or more fuel assemblies.

(continued)

HATCH UNIT 2 3.3-2 Amendment No.155

i RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

' SURVEILLANCE FREQUENCY SR 3.3.1.1.7

NOTE-------------------

Only required to be met during entry into MODE 2 from NODE 1.

Verify the IRM and APRM channels overlap.

7 days SR 3.3.1.1.8 Calibrate the local power range monitors.

1000 effective full power hours SR 3.3.1.1.9 Perfor,rF 40 FUNCTIONAL TEST.

92 days SR 3.3.1.1.10

NOTE-------------------

For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST.

184 days SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 18 months Turbine Control Valve Fast Closure, Trip 011 Pressure -- Low Functions are not bypassed when THERMAL POWER is 1 28% RTP.

l SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST.

18 months (continued)

HATCH UNIT 2 3.3-5 Amendment No.155

RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (page 1 of 3) seector Protection system Instrumentation APPLICABLE CONDITIONS 8EDEs OR REQUlaED REFERENCED OTHER CNANNELS FRON SPECIFIED PER TalP REQUlaED SURVEILLANCE ALLOWASLE FUNCTION CONDIT!0NS SYsTEN ACTION D.1 REQUIREMENTS VALUE 1.

Intermediate aanse Monitor a.

Neutron Ftum - Nish 2

3 G

sa 3.3.1.1.1 s 120/125 sa 3.3.1.1.4 divisions of sa 3.3.1.1.6 futt scale sa 3.3.1.1.7 sa 3.3.1.1.13 sa 3.3.1.1.15 5(*)

3 N

sa 3.3.1.1.1 s 120/125 sa 3.3.1.1.5 divisions of SR 3.3.1.1.13 full scale sa 3.3.1.1.15 b.

Inop 2

3 G

Sa 3.3.1.1.4 NA sa 3.3.1.1.15

$(s) 3 y

gy 3,3,3,3,3 g4 SR 3.3.1.1.15 2.

Average Power aanse Monitor a.

Neutron Ftux - NIsh 2

3(8)

G SR 3.3.1.1.1 s 20% ATP (setdow-)

sa 3.3.1.1.7 sa 3.3.1.1.8 sa 3.3.1.1.10 sa 3.3.1.1.13 b.

Simulated Thernet 1

3(8)

F sa 3.3.1.1.1 s 0.58 W +

Power - Nish sa 3.3.1.1.2 58% RTP and l

sa 3.3.1.1.8 s 115.5%

sa 3.3.1.1.10 aTP(b) sa 3.3.1.1.13 c.

Neutron Flux - Nish 1

3(8)

F sa 3.3.1.1.1 s 120% RTP sa 3.3.1.1.2 Sa 3.3.1.1.8 sa 3.3.1.1.10 Sa 3.3.1.1.13 d.

Inop 1,2 3(8)

C sa 3.3.1.1.10 NA (continued)

(a) With any control rod withdrawn from a core cett containing one or more fuel assemblies.

(b) 0.58 W + 58% 0.58 AW RTP den reset for single loop operation per LC0 3.4.1, "socirculation Loops l

0,... tin...

(c) Each APaN channel provides inputs to both trip systems.

HATCH UNIT 2 3.3-7 Amendment No.155

i RPS Instrumentation 3.3.1.1 i

4 Table 3.3.1.1 1 (pose 3 of 3)

Reactor Protection system Instrumentation APPL!CAsLE CONDIT10Ns MfLES OR REGUIRED REFERENCED OTHER CNANNELs FRON SPECIFIED PER TRIP REGUIRED SURVEILLANCE ALLOWASLE PUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE 8.

Turbine step R 28s RTP 6

E SR 3.3.1.1.9 s IDE closed l

Velve - Closure at 3.3.1.1.11 st 3.3.1.1.13 st 3.3.1.1.15 SR 3.3.1.1.16 9.

Turbine control Velve R 2SE RTP 2

E st 3.3.1.1.9 R 600 pois l

Fast Closure, Trip ott SR 3.3.1.1.11 Pressure - Low SR 3.3.1.1.13 st 3.3.1.1.15 st 3.3.1.1.16

10. Reactor Mode switch -

1,2 2

G SR 3.3.1.1.12 NA shutdown Position SR 3.3.1.1.15 5(e) 2 N

st 3.3.1.1.12 NA SR 3.3.1.1.15

11. Manuel scram 1,2 2

G st 3.3.1.1.5 NA st 3.3.1.1.15 5(e) 2 N

st 3.3.1.1.5 NA st 3.3.1.1.15 (a) With any control rod withdrawn from a core cett contelning one or more fust essenbtles, HATCH UNIT 2 3.3-9 Amendment I:0. 155

EOC-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a.

Two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE:

1.

Turbine Stop Valve (TSV) -- Closure; and 2.

Turbine Control Valve (TCV) Fast Cloture, Trip Oil Pressure -- Low.

QB b.

LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for inoperable EOC-RPT as specified in the COLR are made applicable.

APPLICABILITY:

THERMAL POWER a: 28% RTP, l

ACTIONS

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more channels A.1 Restore channel to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable.

OPERABLE status.

QB A.2

NOTE---------

Not applicable if inoperable channel is the result of an inoperable breaker.

Place channel in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months trip.

(continued)

HATCH UNIT 2 3.3-28 Amendment No. 155

EOC-RPT Instrumentation 3.3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME s

B.

One or more Functions B.1 Restore EOC-RPT trip 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with EOC-RPT trip capability.

capability not maintained.

Og AND B.2 Apply the MCPR limit 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for inoperable MCPR limit for EOC-RPT as specified inoperable EOC-RPT in the COLR.

not made applicable.

C.

Required Action and C.1 Remove the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion recirculation pump Time not met.

from service.

Q8 C.2 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to < 28% RTP.

l SURVEILLANCE REQUIREMENTS

NOTE-------------------------------------

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains EOC-RPT trip capability.

SURVEILLANCE FREQUENCY i

SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST.

92 days (continued)

HATCH UNIT 2 3.3-29 Amendment No. 155 s

-n-

_.._.._._____..m.._

l EOC-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS (continued) sSURVEILLANCE FREQUENCY SR 3.3.4.1.2 Verify TSV -- Closure and TCV Fast 18 months Closure, Trip 011 Pressure -- Low Functions are not bypassed when THERMAL POWER is at 28% RTP.

l 4

SR 3.3.4.1.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:

TSV -- Closure: s 10% closed; and TCV Fast Closure,' Trip 011 Pressure -- Low: a: 600 psig.

1 a

SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.

SR 3.3.4.1.5

NOTE-------------------

Breaker interruption time may be assumed from the most recent performance of SR 3.3.4.1.6.

t Verify the EOC-RPT SYSTEM RESPONSE TIME 18 months on a is within limits.

STAGGERED TEST l

BASIS i

4 SR 3.3.4.1.6 Determine RPT breaker interruption time.

60 months t

i o

HATCH UNIT 2 3.3-30 Amendment No. 155

~

. ~.

l RCS P/T LIMITS 3.4.9 1400 A* A l

INITIAL RTndt VALUES ARE 50'F FOR BELTLINE 1300 26*F FOR UPPER VESSEL, l

AND 50*F FOR BOTTOM HEAD l

1200 i

BELTINE CURVES ADJUSTED AS SHOWN:

1100 EFPY SHIFT (*F)

$2 127 i

1000

-o o.

900 j

o HEATUP/COOLDOWN g

RATE 20*F/HR FOR CURVE A 800 W

700 1

0 i

\\

A'-

CORE BELTLINE 600 E

A - NON-BELTLINE A - PRESSURE TEST WITH FUEL IN THE VESSEL l

a 500 uJ I

$g 4oo NON BELTLINE wg BELTLINE AT 32 EFPY 300 1312 PStGl CURVE A' IS VAllD UP TO 32 EFPY 200 0F OPERATION.

scLTup CURVE A 90*F 100 IS VALID UP TO 32 EFPY 0F OPERATION FOR BELTLINE AND EOL FOR NON-BELTLINE.

0 0

50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

[ AcAD I r3491,)

Figure 3.4.9-1 (Page 1 of 1)

Pressure / Temperature Limits for l

Inservice Hydrostatic and Inservice Leakoge Tests I

Hatch Unit 2 3.4-25 Amendment No.155 i

RCS P/T LIMITS 3.4.9 1400 s

B B' 4

INITIAL RTndt VALUES ARE 50'F FOR BELillNE 1300 26*F FOR UPPER VE.SSEL, 1

AND 50'F FOR BOTTOM HEAD 1200 BELTINE CURVES ADJUSTED AS SHOWN:

1100 EFPY SHIFT (*F) f 32 127 I

1000 O

a.

900 o

HEATUP/COOLDOWN g

RATE 100*F/HR

_a FOR CURVE B f

800 m

f 700 J

B'-

CORE BEL 1LINE a:

600 i

B - NON-BELTLINE E

B - NON-NUCLEAR g

HEATUP/COOLDOWN a

500 CORE NOT CRITICAL E

3 3

4on

-- NON-BELTLINE f

BELTLINE AT 32 300 ]312 PSIGI EFPY CURVE B' 200 IS VALID UP TO 32 EFPY 0F OPERATION.

sw CURVE B 100

/

IS VALID UP TO 32 EFPY OF OPERATION FOR BELTLINE AND EOL FOR NON-BELTLINE.

0 0

50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

[ AcA31 r3o2 1 Figure 3.4.9-2 (Page 1 of 1)

)

Pressure / Temperature Limits for Non-Nuclear Heatup, Low Power Physics Tests, and Cooldown Following a Shutdown Hotch Unit 2 3.4-26 Amendment No. 155 i

l RCS P/T LIMITS 3.4.9 1400 C C' y

INITIAL RTndt VALUES ARE

-50'F FOR BELTLINE 1300 26*F FOR UPPER VESSEL, AND 50'F FOR BOTTOM HEAD 1200 BELTINE CURVES ADJUSTED AS SHOWN:

1100 J

EFPY SHIFT ('F) h 32 127 1000

-o 5

[

900 O

HEATUP/COOLDOWN RATE 100*F/HR a

FOR CURVE C 800 W

b 700 tg C'- CORE BELTLINE E

600 E

C - NON-BELTLINE C - NON-NUCLEAR HEATUP/COOLDOWN

.a 500 CORE NOT CRITICAL

$3g 4oo

- NON-BELTLINE E

BELTLINE AT 32 l

300

{312 PSIGl EFPY CURVE C' IS VALID UP TO 32 EFPY 200 OF OPERATION j

CURVE C 100 IS VALIO UP TO 32 EFPY l

/

OF OPERATION

/

FOR BELTLINE AND f

EOL FOR NON-BELTLINE O

50 100 150 200 250 300 350 400 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

[AcAOI Fj49) }

Figure 3.4.9-3 (Poge 1 of 1)

Pressure / Temperature Limits for Criticality l

Hatch Unit 2 3.4-27 Amendment NO. 155 l

l

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Technical Soecifications (TS) Bases Control Procram (continued) d.

Proposed changes that meet the criteria of b. above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.12 Primary Containment Leakaae Rate Testina Prooram A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated primary containment internal pressure for the design basis loss of coolant accident, P,, is 46.9 psig.

l The maximum allowable primary containment leakage rate, L, at P, is 1.2% of primary containment air weight per day.

Leakage rate acceptance criteria are:

a.

Primary containment overall leakage rate acceptance criterion is s 1.0 L,.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 L for the combined Type B and lype C tests, and 10.75 [, for Type A tests; b.

Air lock testing acceptance criteria are:

1)

Overall air lock leakage rate is s 0.05 L, when tested at s P,,

2)

For each door, leakage rate is s 0.01 L when the gap between the door seals is pressurized t,o 2 10 psig for at least 15 minutes.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

(continued)

HATCH UNIT 2 5.0-16a Amendment No. 155

ML20155C653

Text

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