Text

Forwards Details Re Listed Items to Resolve Open SER Issues, Per 811130 Meeting
	ML20033D224
Person / Time
Site:	Clinton
Issue date:	11/30/1981
From:	Geier J; ILLINOIS POWER CO.
To:	John Miller; Office of Nuclear Reactor Regulation
References
	U-0351, U-351, NUDOCS 8112070406
	Download: ML20033D224 (15)
	v • d • e

U-0351 ILLINDIS POWER COMPANY Sy L30 81 (11 30)-6 500 SOUTH 27TH STREET, DECATUR, ILLINOIS 62525 l

November 30, 1981 g\\ Ut V,.

Mr. James R. Miller, Chief S

Standardization & Special Projects Branch Og s

g Division of Licensing

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Office of Nuclear Reactor Regulations t[

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U. S. Nuclear Regulatory Commission 4

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Washington, D. C.

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Dear Mr. Miller:

1 Clinton Power Station Unit 1 Docket No.

50-461 Attached are details related to the following items which were discussed with Wayne Hodges, Reactor Systems Branch, during a meeting of November 30, 1981 to resolve issues for the Clinton SER:

ISSUES RCIC Venting ECCS Venting Diversion of LPCI Thermal Power Monitor Calibration High Drywell Pressure Interlock Bypass system and high water level trip operability The above items are considered by the NRC and IP to be closed for CPS Licensing purposes.

Sincerely,

' O)~,hcu d.D. Geier Manager, Nuclear Station Engineering Attachments J.H. Williams, NRC Clinton Project Manager cc:

H.H. Livermore, NRC Resident Inspector

(

W. Hodges, NRC Reactor Systems Branch s o 8112070406 811130 PDR ADOCK 05000461 L

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Section 5.4.6 RCIC We require that the technical specifications include provisions for (a) periodic high point venting of the ECCS and RCIC discharge lines and (b) flow and functional tests of the RCIC.

Response

RCIC technical specifications vill be in accordance with NRC standard technical specifications: both provisions are included.

See attached copy of draft CPS technical specification.

Action Required None I

l f

l Issue Section 6'.'3'ECCS L

We require the removal of the high drywell pressure interlock on the l_

IIPCS injection. valve to preclude possible flooding of the steam lines l

and damage to the pressure relief valves.

The applicant; has indicated that this change will be made.

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Response

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Illinois Power Company connits to removal of the high drywell pressure interlock on the HPCS injection valve. 'This modification will be made prior "to fuel load. The FSAR will be changed to reflect this change in Amendment

11. This amendment will be submitted prior to 1/1/82.

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I Action Required See Above.

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r Section 6.3 ECCS Verify that the uppermost vent lines in the ECCS are to be c.hecked monthly for liquid flow to eliminate the possibility of air pocket formation.

Response

ECCS piping will be verified to be full, through high point venting, in accordance with NRC Standard Technical Specifications.

See attached copy of draft CPS technical specification.

I Action Required None

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3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING

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'LIMITINGCONDITIONFOROPERATIbH T,

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3.5.1 ECCS divisions 1, 2 and 3 and the automatic depressurization system.

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(ADS) shall be OPERABLE with: ~

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a.

ECCS division 1 consisting of:

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1.

The OPERABLE low pressure core spray (LPCS) system with a flow l

path capable of taking suction from the suppression chamber and

< b

' transferring the water through the spray sparger to the reactor

, vessel.

it.

37 22.

-The OPERABLE low pressure coolant injection (LPCI) subsyitem " " of ['

the RHR system with a flow path capable of taking suction from t

the suppression chamber and transferring the water to the reactor r-vessel.

3.

(At least) (7) OPERABLE ADS valves.

l

~~

b.

ECCS division 2 consisting of-1.

The OPERABLE low pressurs coolant infection (LPCI) subsystems' "B" and "C" of the RHR systam, each with a flow path capable of taking

' suction from the suppression chamber and transferring the water

.to the reactor vessel.

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v..g, 2.

. fat least) (7-J OPERAhLE ADS valves.

ECCS division 3 consistin'g of ti.e OPERABLE high pressure core spray

-1 c.

(HPCS) system with a flowi path capable o'f taking suction from the suppression chamber and trans, ferring the water through the spray sparger to the reactor vessel.

APPLICABILITY:

OPERATIONAL CONDITION 1, 2* # and 3*.

.."The ADS is not required to be OPERABLE when reactor steam dome pressure is

~_

less than or equal to p.13) psig.

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1 EMERGENCY CORE COOLING SYSTEMS,

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- LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

For ECCS division 1, provided that ECCS divisions 2.and 3 are OPERABLE:

. a.

-1.

With the LPCS system inoperable, restore the inoperable LPCS system to OPERABLE status within 7 days.

~

With LPCI subsystem "A" inoperable, restore the inoperable LPCI JP ' 2

~,,

2.

subsystem "A" to OPERABLE status within 7 days.

3.

^

.With the LPCS system inoperable anif LPCI subsystem "A" inoperable, restore at least the inoperable LPCI subsystem "A" or the inoperable LPCS system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4.

Otherwise, be in at least HOT-SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTCOWN.within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

.l b.

For ECCS division 2, provided that ECCS divisions 1 and 3 are OPERABLE:

1.

With LPCI subsystem "E' or "C" inopera'.le, restore the inoperable LPCI subsystem "B" or "C" to OPERABLE status within 7 days.

%,.(].

2.

With LPCI subsystems "B" and "C" inoperable, restore at least the inoperable LPCI subsystem "B" or "C" to OPERABLE status ~

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3.

Otherwise, be in at least ' HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

{

and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *.

+

'For ECCS division 3, provided that ECCS divisions 1 and 2 and the.

c.

'RCIC system are OPERABLE:

.[

W'ith ECCS division 3 inoperable, restore the inoperable division 1) to OPERABLE status,within 14 days.

2)

'Otherwise, be in at least HOT SHUTDOWN within the.next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN'within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l L,

d., For ECCS divisions 1 and 2, provided that ECCS division 3_ is OPERABLE:

.L 1)

With LPCI subsystem "A" and LPCI subsystem "B" or "C" inoper-

,, li i

able, restore ^at least the inoperable LPCI subsystem "A" or the I

inoperable LPCI subsystem' "B" or "C" to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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^Whenever two or more RHR subsystems are inoperable, if unable to attain COLD -

N.. N SHUTDOWN as required by this ACTION, maintain reactor coolant temperature O

as low as practical by use of alternate heat removal methods.

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ACTION-(Continued)

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With the LPCS system inoperable and LPCI subsystems "B" or "C" inoperable, restore at least the inoperable LPCS system or the inoperable LPCI subsystem "B" or "C" to OPERABLE status within

.72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

._y, 3)

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ~; ->

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l and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *.

e.

For ECCS divisions 1 and 2, provided that ECCS division 3 is OPERABLE and divisions 1 and 2 are otherwise OPERABLE:

1.

'With one of the above required ADS valves i.noperable, restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and -reduce reactor steam dome pressure to 5 /,.R5J psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Ico 2.

With two or more of the above required ADS valves inoperable, h.ej steam dome pressure to 5 (Mt} psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor s

sco f.

With an ECCS discharge line " keep filled" (pressure) -(;=;. fail. c)--

alarm instrumentation channel inoperable, perform Surveillance Requirement 4.5.1.a.1 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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With an ECCS header delta P" instrumentation channel inoperable,

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restora the inoperable channel to OPERAB E 4tatus with 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or declare tne associated EC,CS inoperable.r(lete wM W9 IwMnL w the k

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With the Surveillance Requirement of Specificatiod 4.5.1.d.2 not

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performed at the required, interval due to low reactor steam pressure, the provisions of Specification 4.0.4 ara not applicable provided the appropriate surveilla'nce 'is perforc=0 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to p9rform the test.

i.

In the event an ECCS system is cetuated and injects water into the Reactor Coolant System, a Special Report shall be prepared arid sub-mitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation, the total accumulated actuation cycles to date and the current value of the useage factor for each affected safety injection nozzle whenever its value exceeds 0.70.

^

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^Whenever two or more RHR subsystems are inoperable, if unable to attain COLD

@. / N'j SHUTDOWN as required by this ACTION, maintain reactor coolant t.emperature as

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EMERGENCY CORE COOLING SYSTEMS

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SURVEILLANCE REQUIREMENTS 4.5.1 ECCS division 1, 2 and 3 shall be demonstrated OPERABLE by:

At least once per 31 days for the LPCS, LPCI and HPCS systems:

a.

1.

Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2.

Performance of a ' CHANNEL FUNCTIONAL TEST of the:

~

a)

Discharge line " keep filled" fpressureppump-feilur@

alarm instrumentation, and knybneak b)

Header delta P instrumentation.

l 3.

Verifing that each valve, manual, power operated or automatic,.

in the flow path that is not locked, sealed, or otherwise secure in position, is in its correct position.

b.

Verifing that, when tested pursuant to Specification 4.0.5, each:

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-O 1.

LPCS pump develops a flow of at least S9&) gpm.against a test line pressure greater than or equa@l to (-4Y2T psig. S p.)

4 l

Nro 2.

LPCI pump develops a total flow of at least h66fr) gpm against a test line pressure greater than or equal to41+P) psig. c

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HPCS pump develops a flow of at least (-659) gpm against a test line pressure greater than or equal to 997) psig.

32 For the LPCS, LPCI and HPCS systems, at least once per 18 months:

c.,

3 t

t 1.

Performing a system functional test which includes simulated automatic actuation'of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its co.rrect position.

Actual injection of coolant into the reactor vessel may be excluded from this test.

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.c RN&YCORECOOLINGSYSTEMS (o%p SURVEILLANCE REQUIREMENTS (Continued) s

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2.

Performing a CHANNEL CALIBRATION of the:

a)

Discharge line " keep filled" (pressuref(p=p hihe) alarm instrumentation and verifying the:

1)

High pressure setpoint and the low pressure set-("

', 5 -point of the:

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(a) LPCS system to be 5 6tSC) phig and 1 ( M psig, respectively.

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(b') LPCI s1bsystems to be < GRAU psig and

> G O) psig, respectively.

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Low pressure setpoint of the HPCS system to be 1

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b)

Header delta P instrumeritation and verifying the setpoint of the:

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LPCS system and LPCI subsystegs to be + /et)~ psid.

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HPCS system to be F1 57 thenormalindicatedAP.;iO.lQpsidlessthan 3.

Verifying that the suction for ttie gPCS system is (automati-cally) transferred from the ~ " +=- storage tank to the sup-pression chamber on a condensate storage tank low. water level signal and on a suppression chamber high water level signal.

dfif d.

At least once per 18 month's for the ADS by:

1.

Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuaticn.

1

.s 2.

Manually opening each A05 valve (when the reactor steam dome pressure is greater than or equal to 100 psig) and observing f the expected change in the indicated valve position} fthat-

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Issue Section 6.3 ECCS Confirm that the emergency operating procedures contain adequate cautions to prevent the operator from premature low pressure coolant injection flow to drywell spray cooling or suppression pool cooling.

Response

See attached memo.

Action Required u.

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3.0 OPERATOR ACTIONS 3.1 SUPPRESSION POOL TEMPERATURE CONTROL NOTE

(

Perform step 3.6 concurrently with this procedure.

3.1.1 Close any SRV not required to be open.

'I Any SRV cannot be closed.

IF THEN

. Commence a normal plant shutdown I and cooldown per CPS No.

!10P3006.0lS, UNIT SHUTDOWN WITH CONDENSER, or CPS No. 10P3007.013, UNIT SHUTDOWN WITHOUT CONDENSER, as applicable.

CAUTION C 4 u r. u IJ If continuous'LPCI operation is

, Arecicroce_

c required to assure adequate core no gp cooling, inject thru the heat n,

exchangers as soon as possible.

and DO NOT divert the RRR pumps from the LPCI mode. >

(

3.1.2 WHEN

' Suppression pool temperat'ure '

reaches 95 F.

THEN Place.the RHR system in suppres-sion' pool cooling as per CPS No.y 10P3312.0lS, RESIDUAL HEAT REMOVAL

..(RH). 7 3.1.3 IF Suppression pool temperature reaches 110 F.

AND In operational conditions 1 or 2 with thermal power less than 1%

THEN Scram the reactor and perform

\\

CPS No. 10N4100.01S, REACTOR SCRAM.

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O Rev. No.

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Ja I3g PLANT SYSTEMS 3/4.7.h REACTOR CORE ISOLATION COOLING SYSTEM h~i l

i LIMITING CONDITION FOR OPERATION a

1

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3.7,3.The reactor core isolation cooling (RCIC) system shall be OPERABLE with

[

an OPERABLE flow path capable of lautomatically) taking suction from the sup-pression pool and transferring the water to the reactor pressure vesse1.fand-41-1-OPERABEE-essentiel cc,cnent..).

APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, and 3 with reactor steam dome -..

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pressure greater than -(4+3-) psig.

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With a RCIC discharge line " keep filled" dpressurel (pump fcilure)- alarm

' a.

instrumentation channel inoperable, perfom Surveillance Requirement

~

4.7.4.a.1 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With tile RCIC system inoperable, operation may continue provided the HPCS b.

$ystem is OPERABLE; restore the RCIC system to OPERABL s

reduce reactor steam dome pressure to less than or equal to GOO 7 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

IS SURVEILLANCE REOUIREMENTS g,a

~

4.73, The RCIC system shall be demonstrated OPERABLE:

At least once per 31 days by:

a.

~

' Verifying that the system' piping from the pump dis h 1.

~

c arge valve to the system isolation valve is filled with water, t

-Q, 2.

Performance of a CNANNEL FUNCTIONAL TEST of the discharge line

" keep filled" fpressure') (pump feilsre) alarm instrumentation, and i

I 3.

Verifying that each valve, manual. power operated or automatic in the flow path that is not lac,xed, sealed or otherwise secured in position, is in its correct position.

a M ic.nv E

Whenjtested pursuant to Specificatica 4.0.5'by verifying that from b.

the-cold. condition the RCIC pump develops a flow of greater than or equal to (6003 gpm in the te:,t flow path with c :y tem h :d corre: pend-kg t: rec:ter ve::c! cper: ting precture when steam is being supp. lied to the turbine at normal reactor vessel operati,ng pressure, $1020.+ 40,.-

20fpsig.*

< p.fc&J p. 4 dQ "The provisions of Specification 4.0.4 are not applicable provided the survef1

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lance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate ~ N' to perform the tests.

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SURVEILLANCE REOUIREMENTS (Continued) i 1

1

.At least once per 18 months by:

c.

l.

Perfo ming a system functional test which includes simulated automatic actuation and verifying that each automatic valve in the flow path actuates to its correct position, but may exclude actual injection of coolant into the reactor vessel.

w,....a 2.

Verifying that the system will develop a flow [of greater than ~..

. c..

or equal to (600) gpm in the test flow path when steam is

~

supplied to the turbine at a pressure of $150) + {l5] -(0J psig.*

s.

f 3. - Verifying that the sucgiogor the RCIC system is automatically I

transferred from the c.._... ta storage tank to the suppression paci on a condensate. storage tank, uer.er level-low signal.} &ncl

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"The provisions of Specification 4.0.4 are not applicable provided th lance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate e surveil-to perform the tests.

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Issue Chapter 15 Safety Analysis We require that the time constant of the thermal power monitor be included in the plant technical specification and that it be tested periodically.

Response

CPS will modify the NRC standard technical specification to include surveillance of the time constant.

Attached is a proposed revision including that requirement.

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!l TABLE 4.3.1.1-1

.j REACTOR PROTECTION SYSTEll INSTRUMENTATI0t! SURVEILLANCE REQUIREMENTS I

Cl!ANNEL OPERATIONAL T

'8 CilA!nlEL FUllCTIONAL Ct!ANilEL CONDITIONS FOR Will:ll f

y, FUNCTIONAL UNIT CllELK TEST CALI3RATI0ffa)

SURVEILLANCE REQUIR.:D 1.

Intermediate Range Monitors:

b C

y a.

Neutron Flux - liigh S/UU.S S/U

,W R

2 j;

S W

R 3,4,5 2

1 b.

Inoperative NA W

NA 2,3,4,5 2.

Average Power Range Honitor: U) b C

I

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S/U kS S/U

,W SA 2

a.

Neutron Flux - liigh (Setdown)

S V

SA 3, 4, 5 L

b.

Flow Blased Simulated 9

b Thermal Power - liigh S

S/U(tr)C,W W(d)(e), SA 1

i w

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.e c.

Neutron Flux - liigh S

S/U

,W W

, SA 1

9 Y

d.

. Ir.aperative NA W.

NA 1,2,3,4,5

' 3.

Peactor Vessel Steam Dome

{

Pressure - lilah S

H R IU) 1, 2

,'4.l -

-4.

Reactor Vessel Water Level -

Low, Level 3 S

H R(g)

. (

1, 2 ila 5.

Reactor Vessel Water Level -

N liigh, Level 8 5

H RIU) 1, 2 6.

Main Steam Line Isolation

,,ti.4

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11 1

'; L 7.

Main Steam Line Radiation -

._ ',. fI Ilich -

S H

R 1, 2

'- 5 '

8.

oryweil Pressure - tii h 5

H NU) 0 1, 2 1

1 II

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U TABLE 4.3.1.1-1 (Continued)

Ok REACTOR PROTECTION SYSTEM INSTRUMENTATION SilRVEILLANCE REQUIREMENTS

.o m I

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CHANNEL OPERATIONAL I1;r CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH i:h FUNCTIONAL UNIT CHECK TEST CAllBRATION SURVEILLANCE REQUIRED

" W

' Scram Discharge Volume Water S

v 9.

R@

Level - High

,Nf M

1, 2, 5 I

10. Turbine Stop Valve - Closure NA M

R

'l 5

i.

11. Turbine control valve Fast i

Closure Valve Trip System 011 R @)

Pressure - Low NA H

A-1

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12. Reactor Mode Switch R

Shutdown Position NA R

HA 1,2,3,4,5

13. Manual Scram NA H

hA 1,2,3,4,5 m

N "h.

(a) Neutron detectors may be excluded.from CHANNEL CALIBRATION.

8 (b) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup,. if not performed within the previous 7 days.

.(c) The IRM and SRM channels shall be determined to overlap for at least & // )" decades during each 1

startupandtheIRHandAPRHchannelsshallbedeterminedtooverlapforatleastjp}: decades during each controlled shutdown, if not performed within the previous 7 days (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when TilERMAL POWER > 25% of RATED TiiERHAL POWER. Adjust the APRM. channel if the absolute difference greater than 2%. Any APRM channel

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gain adjustment made in compliance with Specification 3.2.2 si all not be included in determining the absolute difference.

-(e) This calibration shall consist of the adjustment of the APRM readout to conform to a j

-C calibrated flow signal.

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(f) The LPRHs shall be calibrated at least once per 1000 effective full power hours (EFPH) 3 using the TIP system.

%g Calibrate trip unit at least once per 31 days.)

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Issue Chapter 15 Safety Analysis We require that.the plant technical specification address the availability, setpoints, and. surveillance testing of the turbine bypas.s; system and level 8.high water level trip equipment used to

. mitigate a postulated.. excess feedwater transien '.

c v:,... -

Response-Clinton Power Station.will commit to technical specifications on the bypass system similar to those submitted by LaSalle County Station.

These specifications ~w~ill impose a MCPR limit when the

. bypass system is inoperable.

Technical specifications-will also be provided on the level 8 high water level scram that mitigates the excess feedwater transient; these speci'fications will be -

consistent with the NP.C Standard Technical Specification.

Note':

Discussed with Brad Hardin (RSB) on I1/16/81.

Technical specifications will also be provided for level.8 high.

water level trip of reactor feedwater pumps. This will be consistent-with standard technical -specifications discussed with Wayne Hodges (RSB)on,11/30/81.-

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ML20033D224

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