ML20214P568

Amend 137 to License DPR-57,modifying Tech Specs to Permit Hydrostatic & Leak Testing W/Noncritical Reactor Core

ML20214P568
Person / Time
Site:	Hatch
Issue date:	05/26/1987
From:	Youngblood B Office of Nuclear Reactor Regulation
To:
Shared Package
ML20214P570	List: ML20214P568 ML20214P573
References
NUDOCS 8706030440
Download: ML20214P568 (9)

Text

..

,psuc%

3 e([k l

g UNITED STATES g

NUCLEAR REGULATORY COMMISSION y

• .r

/~ E WASHINGTON, D. C. 20555 e

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 137 License No. DPR-57 1.

The Nuclear Regulatory Comission (the Comission) 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 Georgia Power Company, acting for itself Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensee) dated March 4, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission'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 Comission; 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 Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have i

been satisfied.

l l

[ [ 05000321 870526 P

PDR

f.

. 2.

Accordingly, the license is amended by changes to the Technical Specifi-cations 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 Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 137, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

\\%

B. J. Youngblood, Director Project Directorate II-3 Division of Reactor Projects-I/II

Attachment:

Changes to the Technical Specifications Date of Issuance: May 26, 1987 k

PD#II-PD#II-3/DRP-I/II GC-yythesda PD#I RP-I/II MDunh'in(/DRP-I/II

/ rad LCrocker J'\\ a. PM +%

BJYo1 ood 05/ p[ /87 05/jq /87 05/

/87 05/

/ 87 L

ATTACHMENT TO LICENSE AMENDMENT NO. 137 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. The overleaf pages are provided for convenience.

Remove Insert Page Page 3.5-6 3.5-6 3.5-8 3.5-8 3.5-9 3.5-9 3.6-9 3.6-9 Figure 3.6-1 Figure 3.6-1 3.7-2 3.7-2 eup l

W LIMITING CON 0!tIONS FOR OPERATION SURVEllLANCE REOUIREwENTS 3.5.C.3.

Two p;q t incoerable 4.5.C.3.

Two Pumos inoceraDie 1.f two RHR service water pumps are When two RHR service water cumps' inoperacle,' the reactor may remain-are inoperable, tne remaining' in operation for a period not, to operable RHR serv. ice water exceed seven (7) days provided all subsystems and their associated redundant active components in both diesel generators shall be of the RHR service water subsystems demonstrated to be oDeraDie are operacle.

immediately and daily thereafter for seven (7) days or until the inoperable components are returned to normal operation.

4.

Shutdown Reauirements If Specifications 3.5.C cannot be met, the reactor shall be placed in the Cold shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

Mich Pressure Coolant Inieetion fHPcil 0.

High Pressure Coolant Iniection (MPCI) System g

1.

Normal system Availability 1.

Normal Onerational Tests f

HPCI' system testing shall be perforwed as follows:

lige Frequency a.

The HPCI System shall be a.

Simulated once/ Operating operable:

o automatic

' Cycle.

actuation 1.

Prior to reactor startup test from a cold condition, or i

b.

Flow rate at once/3. months 2.

When irradiated fuel is in normal reactor the reactor vessel and the vessel oper-i reactor pressure is greater ating pressure than 150 psig, except as and

.~1ow rate at once/ Operating stated in Specification

)

3.5.0.2.

• 150 psig Cycle reactor pressure

• HPCI is not required to be operable for performance of inservice hydrostatic or leak testing with reactor pressure greater than 150 psig and all control rods inserted.

MATCH - UNIT 1 3.5-6 Amendment No. 137

c..

t!NITING CONDITIONS FOR OPERATION SURVElltANCE REOUIREMENTS 3.5.E.1. Normal System Availability (Cont.)

4.5.E.1. Normal Coerational Tests (C:nt.)

a.(2) When there is irradiated

b. Vertfying that suc-Once/Operattnq fuel in the reactor vessel tion for the RCIC Cycle and the reactor pressure system is automati-is above 150 psig, except cally transferred as stated in Specification l

from the CST to the 3.5.E.2.*

suppression pool on a simulated low CST level or hign suo-pression pool level signal.

c. Flow rate at once/3.wntns normal reactor vessel operating pressure and Flow rate at once/ Operating 150 psig Cycle reactor pressure.

The RCIC pump shall deliver at least 400 gpa during each flow test.

d. Pump Operability Once/ month

s. Motor Operated once/ month valve operability

2. Ooeration with Inocerable 2.

Surveillance with Inonerable Cosmonents Components If the RCIC system is inoperable.

When the RCIC system is inoper-the reactor may remain in oper-able. the HPCI systes shall be ation for a period not to exceed demonstrated to be operable seven (T} days if the HPCI system insnediately and daily thereaf ter is operable during such time.

until the RCIC system is returned to normal operation.

l

3. 'If Specification 3.5.E.1. or l

3.5.E.2. is not met, an o derly l

shutdown shall be initiated and the reactor shall be depressurized to less than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• RCIC is not required to be operable for performance of inservice hydrostatic or leak testing with reactor pressure greater than '150 psig and all control rods inserted.

l l

HATCH - UNIT 1 3.5-8 Amendment No. 137 l

l

LIMITING CON 0!TIONS FOR OPERATION

$JRVEILLANCE RE0VIRE=ENTS 3.5.F.

automatie Deoressurization System 4.5.F.

Automatie Deeressurization System 11M1 1191.1.

1.

Normal Doerational Tests 1.

Nomal System Availability The seven valves of the Automatic a.

A simulated automatic actuation Depressurization System shall be test shall be performed on the operable:

ADS prior to startup af ter each refueling outage. Surveillance a.

Prior to reactor startup f rom a of all relief valves is covered cold shutdown, or in Specification 4.6.H.

b.

When there is irradiated fuel in b.

A leak rate test of each ADS

.the reactor vessel and the valve accumulator, check valve, reactor is above 113 psig except and actuator assembly shall be as stated in Specific.*. tion performed during each refueling

3. 5.F.2.

• I outage at a pressure of 90 18 psig. The leakage rate shall be verified to be 14.5 SCFM.

2.

Operation with inocerable 2.

Surveillance with Inocerable Comconents Components If one of the seven ADS valves is When it is determined that one of known to be incapable of automatic the seven ADS valves is incapable of operation, the reactor may remain in automatic operation, the HPCI system operation for a period not to exceed and the actuation logic of the other seven (7) days, provided the HPCI ADS valves shall be demonstrated to 1

system is operable. (Note that the be operable immediately and daily pressure relief function of these thereaf ter until all seven A05 valves is assured by Specification valves are capable of automatic 3.6.H.; Specification 3.5.F. only operation. '

applies to the AOS function).

e I

3.

Shutdown Reauirements e

If Specification 3.5.F.1. or 3.5.F.2.

cannot be met an orderly shutdown will be initiated and the reactor pressure shall be reduced to 113 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• The ADS valves are not required to be operable for performance of inservice hydrostatic or leak testing with reactor pressure greater than 113 psig and all control rods inserted.

HATCH - UNIT 1 3.5-9 Amendment No. 137 O

D #

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.6.H.1.

Relief / Safety Valves 4.6.H.1.

Relief / Safety Valves l

a.

When one or more relief / safety a.

fnd of coeratine tvele valve (s) is known to be failed an t

orderly shutdown shall be initiated Approxiteately one-half of all and the reactor depressurized to relief / safety valves shall-be less than 113 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

benchchecked or replaced with Prior to reactor startup from a a benchchecked valve each re-cold condition all relief / safety fueling outage. All 11 valves valves shall be operable.**

l will have been checked or re-placed upon the completion of every second operating cycle, b.

With one or more relief / safety b.

Each Goeratino Cvele valve (s) stuck open, place the reactor mode switch in the shutdown position.

Once during each operating cycle, at a reactor pressure

> 100 psig each relief valve shall be manually opened until thermocouples downstream of the valve indicate steam is flow-l ing from the valve.

c.

With one or more safety / relief valve c.

Intecrity of Relief Valve tailpipe pressure switches of a sellows=

safety / relief valve declared inoperable and the associated The integrity of the relief valve safety / relief valve (s) otherwise bellows shall be continuously indicated to be open, place the monitored and the pressure l

reactor mode swi.tch in the Shut-switch calibrated once per down position.

operating cyi:le and the accu-mulators and air piping shall be inspected for leakage once per operating cycle.

d.

With one safety / relief valve tailpipe d.

Relief Valve Maintenance pressure switch of a safety / relief i

valve declared inoperable and the asso-At least one relief valve shall ciated safety / relief valve (s) otherwise be disassembled and inspected indicated to be closed, plant operation each operating cycle..

may continue. Remove the function of that pressure switch from the low low e.

Operability of Tail pine

~

set logic circuitry until the next COLO Pressure Switenes SHUTCOWN. Upon COLO SHUTDOWN, restore the pressure switch (es) to OPERABLE The tail pipe pressure switch status before STARTUP.

of each relief / safety valve shall be demonstrated operable e.

With both safety / relief valve tailpipe by perforinance of a:

I pressure switches of a safety / relief I

valve declared inoperable and the asso-1.

Functional Test:

ciated safety / relief valve (s) otherwise indicated to be closed, restore at least 4.

At least once per 31 one inoperable switch to OPERA 8LE status days,, except that all, within 14 days or be in at least NOT portions of instrumen-SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> tation inside the pri-and in COLD SHUTDOWN within the mary containment may be following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

excluded from the functional test, and

• • The salief/Sofety valves are not required to be operable for perforusace of inservice hydrostatic or pressure testing with reactor pressure greater than 113 peig and all control rode Laserted. Overpressure protection will be provided es required by ASME code.

Amendment No. 137 MATCH - UNIT 1

.1.6-9 i

L T:11 L i.'

i s

4 i i

. 6..

i.

l l

l l'

Env 410121416 e

i 6

a 1400

.,.c 1-1.,

.i. 1,...,..

I e

. i

• It W I 11 e

I I 6p.r4i I e

i#

a r

'Ie I-I

.F i

n e

e e 6

1 g.

y i r if.

e 6

71 -,

.I 6 f.

i s

i.

p p. Ic 1*

1'1 B

• It i

s b

I I d ',I*

1 I-1 I

1200 P

I

(-

It I

e i

e e

e i

I.

1 Ir a ir e e.

/. I6 ;

.I 1-

. 5 i 6 t gg

.t 6

i.

i e

i F

• I Je I t

AtuusTED CORE StLTunt. x. t

,,,.,,,,r 1M T FLAuf

. x r ie i.i.

1-s e

,s.

6

.,o. r s* w. i t

r I6 I ie i

'T

.6 i

i 6 i

..,II

.J

.I.

d}

6.

s 6

.y

.I 1 yaA g

swww f.

y y.p of...

I 4 6

e A F I-J' I

1,

fi /4 #1 II i i N

6 I J 4/ /I (t i.

t 6

l I i

(- 1'f

• i y

. I I J 1

1*

b f

e

,1 1-

/

I# '

e l

e I./

/

I /- i

• 3 t. i e

Q N

6 e

6 e

/.

I-f 1 I-

• t a

3 i

I *I F /* I

• r

+

y 800 i 11.s i

s.

, e I

fi J K-)

e

. i i,

e,

II

/

f. f i i i I e

I fi

,r

/ fi

.. i

, i.

6 I I /

I #

t !

6 i i

-A

/ / i/ I6 W

/ 6/ / /1 /.

e i

/ / / / f-n

, e

/ /6 / / /

i i l

/ / / / r y

gy i

g.

, i t

I www t

4 t.,

2rf/

/ fi 6

6 e

i i..

t

'/

/ / / /-

O

./ // / /

t 6

ha

, e r

g 1 1111,

//#

6 i

. t I F

,g F##

1 9

t be i

v 4 i

4.

s VERTICAL UaMT uset POR PRt33unt i 400 3Astc ose 10 cpm 00 APPtMO4X G meculftth88NT OP (RTNOT + es*91 PUWo0E A004088 RTaeOT* 18"'

i 96 9

f i 4

1 1

. i t

t

+

.s SOLT PRSLOAO TEMPERATumE op i

i e

i mee sAsto om amto 200 IRTNo? + eo**1 som tL264ss. PLAar e

• RTnor. ies,LAmsa neoton.

we cLosunt P i

i f

i I

t t i

0 0

100 200 300 317 Metal Temperature ('F) l l

Figure 3.6-1 Pressure versus Minimum Temperature for Pressure Tests.

Based on Surveillance Test Pesults Amendment No. 137 l-..-...

^

'.._,.,L.__=.____*___

q u! p e, ;,'N0!rIONS FOR OPERATION SURVEILLANCE RE0VIRE*ENTS i

2.

Primary Containment Inteceity 4.7.A.2.

Leak Testine to Verifv primary Containment Intecrity Primary containment integrity is required:

Primary containment integrity shall be demonstrated by the a.

Prior to withdrawing following test procedures:

control rods for the purpose of going critical.

4.

Tyne A Tests - Intecrated leak Rate Test (ILRT)*

b.

Whenever the reactor is P

critical.

Primary containment integrity is confirmed if the leak rate does c.

Whenever the reactor water not en:eed the maximum allow-temperature is above 212*F able leak rate La Of I 2 and fuel is in the reactor l

weight percent of the contained vessel.**

I air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the peak test pressure.

M. ca:aption is made while per-forming low power physics tests (1) Type A tests shall be performed at atmospheric pressure at power under the program estaelisned levels not to exceed 5 MWt. during in Appendix J of 10 CFR which time primary containment Part 50. (Reference 1).

Stegrity is not required.

l La - Maximum allowable peak pressure test leak rate - 1.2 weight percent per day i

Lt - Maximum &1lowable reduced pressure test leak rate Lam - Measured peak pressure test leak rate - values are subject to change with each ILRT perforised L m - Measured reduced pressure test leak rate - values are subject to t

change with each ILRT performed Lao - Allowable operational leak rate for peak pressure tests - values are subject to change with each ILRT performed L o - Allowable operational leak rate for reduced pressure tests - values t

are subject to change with each ILRT performed

( All leakage rates measured in weight percent of contained air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

.P

- Peak test pressure - 59 psig a

P

- Reduced test pressure - 29.5 psig t

• • Primary Containment Integrity is not required for perfonnance of inservice hydrostatic or leak testing with reactor coolant temperature greater than 2120F and all control rods inserted.

HATCH - UNIT 1 3.7-2 Amendment No. 137

- -