Text

Amends 112 & 53 to Licenses DPR-67 & NPF-16,respectively, Revising Tech Specs Re Administrative Matters
	ML17223B355
Person / Time
Site:	Saint Lucie
Issue date:	12/05/1991
From:	Berkow H; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17223B356	List: ML17223B355; ML17223B358;
References
	NUDOCS 9112110131
	Download: ML17223B355 (37)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FLORIDA POWER

& LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

112 License No.

DPR-67 The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Florida Power

& Light Company, et al. (the licensee),

dated February 26, 1991, complies with the standards and requirements 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.

91121l0131 911205 PDR ADDCN, 05000335' PDR

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2.

Accordingly, Faci lity Operating License No. DPR-67 is amended by changes to the Technical Specifications as indicated in the attachment'to'his license amendment, and by amending paragraph 2.C.(2) to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B

as revised through Amendment No. iip, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications'.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

gecemger 5

$ 99)

H bert N ~ Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE At1ENDMENT NO ~

2 TO FACILITY OPERATING LICENSE NO. DPR-67 DOCKET NO. 50-335 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 area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

Remove Pa es 3/4 3-28 3/4 3-42 3/4 4-4 3/4 8-1 3/4 8-2 3/4 8-3 3/4 8-4 3/4 8-6a 3/4 8-6b 3/4 10-2 6-8 3/4 3-28 3/4 3-42 3/4 4-4 3/4 8-1 3/4 8-2 3/4 8-3 3/4 8-4 3/4 8-6a 3/4 8-6b 3/4 10<<2 6-8

TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION INSTRUMENT CHANNEL SENSOR LOCATION MEASUREMENT RANGE MINIMUM CHANNELS OPERABLE 1.

STRONG MOTION TRIAXIAL ACCELEROGRAPHS a 0 b.

C.

d.

e.

SMR-42-1 SMR-42-2 SMR-42-3 SMR-42-4 SMR-42-5 R.B. Elev. 23.0>

R.B. Elev. 62.0+

R.A.B. Elev. -0.5<

R.A.B. Elev. 43.0+

R.A.B. Elev.

19.5<

O-l g O-l g 0-1 g

0-1 g O-l g 2.

PEAK RECORDING ACCELEROGRAPHS a.

SMR-42-6 b.

SMR-42-7 c.

SMR-42-8 3.

PEAK SHOCK RECORDERS a.

SMR-42-9 b.

SMR-42-10 R.B. Piping from S.I.T.1A2-c Elev.

46'0 9/16" R.B.

Equipment on S.I.T.1A2 R.A.B.-Sh.

Dn. Ht.

XCHR Supports R.B. Elev.

23.0'.B.

M.S. Pipe Restraints - S.G.lBl 0-2 g 0-2 g 0-2 g 4.

EARTHQUAKE FORCE MONITOR

a. SMI-42-11 5.

SEISMIC SMITCH Control Room 0-0.2 g

a.

SMS-42-12 R.B. Elev.

23.0'T.

LUGIE - UNIT 1

3/4 3-28 Amendment No.

112

fNSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a.'- Actions per Table 3.3-11.

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

ST LUCIE - UNIT 1

3/4 3-41 Amendment No. 37

TABLE 3.3-11 ACCIDENT NONITORING INSTRUMENTATION IC:

m INSTRUHENT 1.

Pressurizer Water Level 2.

Auxiliary Feedwater Flow Rate 3.

RCS Subcooling Hargin Honitor 4.

PORV Position Indicator Acoustic Flow Honitor 5.

PORV Block Valve.Position Indicator 6.

Safety Valve Position Indicator 7.

Incore thermocouples TOTAL NO.

OF CHANNELS 1/pump 1/valve 1/valve 1/valve 4/core quadrant HINIHUH CHANNELS OPERABLE ACTION

-1 1/pump 1/valve 1/valve 1/valve 2/core quadrant O

8.

Containment Sump Water Level (Narrow Range) 9.

Containment Sump Water Level (Wide Range)

10. Reactor Vessel Level Honitoring System ll. Containment Pressure 1*

4, 5

The non-safety grade containment sump water level instrument may be substituted.

- Definitionof OPERABLE: A channel is composed of eight (8) sensors in a probe, of which four (4) sensors must be OPERABLE.

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION ee %% %%~ %a MMES

~ 3.4.4 The pressurizer shall be OPERABLE with a steam bubble, and with at least 150 kw of pressurizer heaters capable of being supplied by emergency power.

APPLICABILITY: MODES 1 and 2.

ACTION:

With the pressurizer inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.4 In accordance with 4.8. 1. 1.2.

ST LUCIE - UNIT 1 3/4 4-4 Amendment No. %-,

112

)

3 4.8 ELECTRICAL POWER YSTEMS 3 4.8.1 A.C.

SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8. 1, 1 As a mintmum, the following A.C. electrical power sources shall be OPERABLE:

a.

Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and b.

Two separate and independent diesel generator sets each with:

1.

Engine-mounted fuel tanks containing a minimum of 152 gallons of fuel, 2.

A separate fuel storage system containing a minimum of 16,450 gallons of fuel, and 3.

A separate fuel transfer pump.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a

~

b.

With one offsite circuit of 3.8. 1. I.a inoperable, except as provided in Action f. below, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8. 1. 1. I.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If either EDG has not been successfully tested within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , demonstrate its OPERABILITY by performing Surveillance Requirement 4.8. 1. 1.2.a.4 separately for each such EDG within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Restore the offsite circuit to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With one diesel generator of 3.8. 1. I.b inoperable, demonstrate the OPERABILITY of the A.C. sources by performing Surveillance Requirement 4.8. l. 1. I.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; and if the EDG became inoperable due to any cause other than preplanned preventative maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE EDG by performing Surveillance Requirement 4.8. 1. 1.2.a.4 within 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s*; restore the diesel generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Additionally, verify within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months that:

This test is required to be completed regardless of when the inoperable EDG is restored to OPERABILITY.

ST.

LUGIE UNIT 1

3/4 8-1 Amendment No. 493-,

112

ELECTRICAL POWE STENS ACTION Continued I

all required

systems, subsystems,

trains, components and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also

OPERABLE, and

2. 'hen in NODE 1, 2 or 3, the steam-driven auxiliary feed pump is OPERABLE.

c.

With one offsite A.C. circuit and one diesel generator inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8. 1. 1. I.a within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; and if the EDG became inoperable due to any cause other than preplanned preventative maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE EDG by performing Surveillance Requirement 4.8. 1. 1.2.a.4 within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s*. Restore at least one of the inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following

.30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Restore the other A.C. power source (offsite circuit or

, diesel generator) to OPERABLE status in accordance with the provisions of Section 3.8. 1. 1 ACTION Statement a or b, as appropriate, with the time requirement of that ACTION Statement based on the time of the initial loss of the remaining inoperable A.C. power source.

Additionally, verify within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months that:

1.

all required

systems, subsystems,

trains, components and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also

OPERABLE, and 2.

When in NODE 1, 2 or 3, the steam-driven auxiliary feed pump is OPERABLE.

d.

With two of the required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by sequentially performing Surveillance Requirement 4.8. 1.1.2.a.4 on both diesels within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unless the diesel generators are already operating; restore one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Following restor-ation of one offsite source, follow ACTION Statement

a. with the time requirement of that ACTION Statement based on the time of the initial loss of the remaining inoperable offsite A.C. circuit.

This test is required to be completed regardless of when the inoperable,EDG is restored to OPERABILITY.

ST.

LUGIE UNIT 1 3/4 8-2 Amendment No. 483-,

112

ELECTRICAL POWER SYSTEMS

~CI0 <>,'

SURVEILLANCE RE UIREMENTS Continued 6.

Verifying the diesel generator operates for at least 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s****. During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded within a load band of 3800 to 3960 kWC and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this

test, the"diesel generator shall be loaded within a load band of 3300 to 3500 kWO. The generator voltage and frequency shall be 4160
+ 420 volts and 60 i 1.2 Hz within 10 seconds after the start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test.
7.
Verifying that the auto-connected loads do not exceed the 2000-hour rating of 3730 kW.
8.
Verifying the diesel generator's capability to:
a) b)
Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power.
I Transfer its loads to the offsite power source, and 9.
c)
Be restored to its standby status.
Verifying that with the diesel generator operating in a test mode (connected to its bus),
a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizes the emergency loads with offsite power.
10.
Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the engine-mounted tanks of each diesel via the installed cross connection lines.
ll.
Verifying that the automatic load sequence timers are operable with the interval between each load block within k 1 second of its design interval.
f.
At least once per ten years or after any modification which could affect diesel generator independence by starting****the diesel generators simultaneously, during shutdown, and verifying that the diesel generators accelerate to approximately 900 rpm in less than or equal to 10 seconds.
PThis band is meant as guidance to avoid routine overloading of the engine.
Variations in load in excess of this band due to changing bus loads shall not invalidate this test.
Thistest may be conducted in accordance with the manufacturer's recommendations concerning engine prelube period.
ST.
LUCIE - UNIT 1 3/4 8-6a Amendment No. $8-,
'I12
ELECTRICAL POWER
'TEMS SURVEILLANCE RE UIREMENTS Continued g.
At least once per ten years by:
1.
Draining each fuel storage
tank, removing the accummulated sediment and cleaning the tank using an appropriate cleaning
compound, and 2.
Performing a pressure test of those portions of the diesel fuel oil system designed to USAS B31.7 Class 3 requirements at a test pressure equal to.110X of the system design pressure.
4.8.1.1.3
~Re orts All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2.
Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1. 108, Revision 1, August 1977.
If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide
1. 108, Revision 1, August 1977.
4.8. 1. 1.4 The Class 1E underground cable system shall be.demonstrated OPERABLE within 30 days after the movement of any loads in excess of 80X of the ground surface design basis load over the cable ducts by pulling a mandrel with a diameter of at least 80X of the duct's inside diameter through a duct exposed to the maximum loading (duct nearest the ground's surface) and verifying that the duct has not been damaged.
ST.
LUCIE UNIT 1 3/4 8-6b Amendment No. ~ 112
3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).
APPLICABILITY:
MODE 2.
ACTION:
a 0 b.
With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at
> 40 gpm of 1720 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification
3. l. 1. 1 is restored.
With all full length CEAs inserted and the reactor subcritical
.by less than the above reactivity equivalent, immediately initiate and continue boration at
> 40 gpm of 1720 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification
3. l. 1. 1 is restored.
SURVEILLANCE RE UIREMENTS 4.10.1.1 The position of each full length CEA required either partially or,fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days: prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification
3. 1.1.1.
ST.
LUCIE - UNIT 1 3/4 10-1 Amendment No. 27. 87
t SPECIAL TEST Exct.
IONS GROUP HEIGHT INS RTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION QSSR XDSSSC QRC CSR RQRRQ RZKCR
3. 10.2 The group height, insertion and power distribution limits of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.2.3, and 3.2.4 may'6 suspended during the performance of PHYSICS TESTS provided:
a.
The THERMAL POWER is restricted to the test power plateau which shall not exceed 85X of RATED THERMAL POWER, and b.
The limits of Specification 3.2. 1 are maintained and deter-mined as specified in Specification 4. 10.2.2 below.
APPLICABILITY:
MODES 1 and 2.
ACTION:
With any of the'imits of Specification 3.2. 1 being exceeded while the requirements of Specifications
3. 1. 1.4, 3. 1.3. 1, 3.1.3.2, 3.1.3.5,
3. 1.3.6, 3.2.3 and 3.2.4 are suspended, either:
a.
Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or b.
Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIREMENTS QRDZIRDRR DQQ RCC5 CR R
S CSC CRRCDRRSQQQC RRQSRRRCCRCERCL R%RR
%25RSCE
4. 10.2. 1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications
3. 1. 1.4,
3. 1.3. 1, 3. 1.3.5,
3. 1.3.6, 3.2.3 or 3.2.4 are suspended and shall be verified to be within the test power plateau.
4. 10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2. 1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2. 1.3 and 3.3.3.2 during PHYSICS TESTS above 5X of RATED THERMAL POWER in which the requirements of Specifications 3.1. 1.4,
3. 1.3. 1, 3. 1.3.5, 3.1.3.6, 3.2.3 or 3.2.4 are suspended.
ST.
LUCIE - UNIT 1 3/4 10-2 Amendment No. i~49-,
112
ADMI N I STRATI VE CONTROLS ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the FRG Chair-man to serve on a temporary basis;
however, no more than two alternates shall participate as voting members in FRG activities at any one time.
MEETING FRE UENCY 6.5.1.4 The FRG shall meet at least once per calendar month and as convened by the FRG Chairman or his designated alternate.
QUORUM 6.5.1.5 The quorum of the FRG necessary for the performance of the FRG respons-ibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates.
RESPONSIBILITIES 6.5.1.6 The Facility Review Group shall be responsible for:
a
~
b.
c ~
d.
e.
Review of (1) all procedures required by Specification 6.8 and changes
thereto, (2) all programs required by Specification 6.8 and changes
thereto, and (3) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.
Review of all proposed tests and experiments that affect nuclear safety.
Review of all proposed changes to Appendix A Technical Specifications.
Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety.
Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent, recurrence to'the President-Nuclear Division and to the Chaipman'of the Company Nuclear Review Board.
f.
Review of all REPORTABLE EVENTS.
g.
Review of unit operations to detect potential nuclear safety hazards.
h.
Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant Manager or the Company Nuclear Review Board.
ST.
LUCIE -'NIT 1 6-7 Amendment No.
$P, ~p 88s 1O7~
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ADMINISTRATIVE CON OLS mama's aztec xaam a
aa aaaa aaaaaaaaaaaaaaaaaaaa3 aaaa Review of the Security Plan and implementing procedures and submittal of recommended changes to the Company Nuclear Review Board.
IL Review of the Emergency Plan and implementing procedures and submittal of recommended changes to the Company Nuclear Review Bo'ard.
k.
Review of every unplanned on-site release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the President - Nuclear Division and to the Company Nuclear Review Board.
Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION HANUAL and RADWASTE TREATMENT SYSTEMS.
m.
\\
3 E\\
0 AUTHORITY Review and documentation of judgment concerning prolonged operation.,in
bypass, channel trip, and/or repair of defective protection channels of process variables placed in bypass since the last FRG meeting.
6.5. 1.7 The Facility Review Group shall:
a ~
b.
c ~
RECORDS Recommend in writing to the Plant Manager, approval or disapproval of items considered under Specifications 6.5. 1.6.a through d above.
Render determinations in writing with regard to whether or not each item considered under Specifications 6.5. 1.6 a, b, d, and e above constitutes an unreviewed safety question.
Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the President-Nuclear Division and the Company Nuclear Review Board of of disagreement between the FRG and the Plant Manager;
however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6. 1. 1 above.
6.5. 1.8 The Facility Review Group shall maintain written minutes of each FRG meeting that, at a minimum, document the results of all FRG activities performed under the responsibility and authority provisions of these Technical Specifications.
Copies shall be provided to the President-Nuclear Division and the Chairman of the Company Nuclear Review Board.
ST.
LUCIE UNIT 1 6-8 a
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FLORIDA POWER 8t LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST.
LUCIE PLANT UNIT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
53 License No.
NPF-16 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Florida Power 5 Light Company, et al. (the licensee),
dated February 26, 1991, complies with the standards and requirements 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.
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2.
Accordingly, Facility Operating License No.
NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment'te this license amendment, and by amending paragraph 2.C.2 to read as follows:
2.
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
5y, 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 the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Speci ficat ions Date of Issuance:
S~cember 5,
1991 rbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AMENDMENT-NO.
TO FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 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 area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Remove Pa es XXI XXII XXIV XXV 3/4 6-21 3/4 8-4 3/4 8-18 3/4 8-19 6-9 Insert Pa es XXI XXII XXIV XXV 3/4 6-21 3/4 8-4 3/4 8-18 3/4 8-19 6-9
\\
LIST OF FIGURES INDEX I
a c
RQRRRDR R
CC DDSCCCRRQ QSQRR RR RX RR CQQCQQDRRZRRRRRRCS FIC'jRE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING.....................
PAGE 2-3 2.2-1 LOCAL POWER DENSITY HIGH TRIP SETPOINT PART 1
(FRACTION OF RATED THERMAL POWER VERSUS QR~).............
2-7 2.2-2 2.2-3 LOCAL POWER DENSITY HIGH TRIP SETPOINT PART 2
( QR) VERSUS Yi)....................................
THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1
Y) VERSUS Ai)...........................................
(
2-8 2-9 2.2-4 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS QR1).............
2-10 B 2.1-1 AXIAL POWER DISTRIBUTION FOR THERMAL MARGIN SAFETY LIMITS...........................................
B 2-2 3.1-1 3.1-la 3.1-2 3.2-1 3.2-2 3.2-3 4.2-1 3.2-4 3.4-1 3.4-2 ELETEDo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~
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D 3/4 2-6 AXIAL SHAPE INDEX OPERATING LIMITS WITH FOUR REACTOR COOLANT PUMPS OPERATING.................................
3/4 2-12 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMITS VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY >I uCi/GRAM DOSE EQUIVALENT I-131........................................
3/4 4-31 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 15
EFPY, HEATUP AND CORE CRITICAL................... 3/4 4-3la MINIMUM BORIC ACID STORAGE TANK VOLUME AS A FUNCTION OF STORED BORIC ACID CONCENTRATION......................
3/4 1-15 T
ALLOWABLE TIME TO REALIGN CEA VS INITIALF.............
3/4 1-19a CEA INSERTION LIMITS VS THERMAL POWER WITH FOUR REACTOR COOLANT PUMPS OPERATING.........................
3/4 1-28 ALLOWABLE PEAK LINEAR HEAT RATE VS BURNUP...............
3/4 2-3 AXIAL SHAPE INDEX VS FRACTION OF MAXIMUM ALLOWABLE POWER LEVEL PER SPECIFICATION 4.2.1.3........;..........
3/4 2-4 T
T ALLOWABLE COMBINATIONS Of THERMAL POWER AND F< > Fxy o
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~ 3/4 2 5 ST.
LUCIE UNIT 2 XXI Amendment No. 8, g3
LIST OF FIGURES (Continued)
FIGURE INDEX PAGE 3.4-3
'.4-4 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITATIONS FOR 15
EFPY, COOLOOWN AND INSERVICE TEST.
REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITATIONS FOR 15
EFPY, MAXIMUM ALLOWABLE COOLOOWN RATES..........
3/4 4-31b 3/4 4-32 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST.............
3/4 7-25 B 3/4.4-1 NIL-DUCTILITYTRANSITION TEMPERATURE INCREASE AS A FUNCTION OF FAST (E)l MeV)
NEUTRON FLUENCE (550'F IRRADIATION) FOR. REACTOR VESSEL BELTLINE MATERIALS....
B 3/4 4-10 5.1-1 5.6-1 6.2-1 6.2-2 SITE AREA HAP.
INITIAL ENRICHMENT VS.
BURNUP REQUIREMENTS FOR STORAGE OF FUEL ASSEMBLIES IN REGION II.......'........
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5-2 5-4a 6-3 6-4 ST.
LUCIE UNIT 2 XXII Amendment No. 8,'8-, 53
LIST OF TABLES TABLE INOEX PAGE 1.2
2. 2-1 FREQUENCY NOTATION........
~...............................
OPERATIONAL MODES.........
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT IMITS ~
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D 3/4 2-11 I
DNB MARGIN LIMITS....................
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~ 3/4 2 15 REACTOR PROTECTIVE INSTRUMENTATION........................ 3/4 3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES......... 3/4 3-6 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR ST. LUCIE-2.................................
3/4 1-17 REQUIREMENTS..............................................
3/4 3-8 3 ~ 3 3 3.3-4 3.3-5 4.3-2 3.3-6 4.3-3 3 ~ 3 7
4.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..............'.............................
3/4 3-12 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES...............................
3/4 3-17 ENGINEERED SAFETY FEATURES RESPONSE TIMES.................
3/4 3-19 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................
3/4 3-22 RADIATION MONITORING INSTRUMENTATION...................... 3/4 3-25 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.
3/4 3-28 SEISMIC MONITORING INSTRUMENTATION........................ 3/4 3-33 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............................................
3/4 3-34 3.3-8 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS o ~ ~
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LUCIE-UNIT 2 XXIII Amendment No. 8
LIST OF TABLES (Continued)
TABLE INDEX PA4E 4
8
~
f 3.3-9 4.3-6 3.3-10 4.3-7 3.3-11 3.3-12 4.3-8 3.3-13 4.3-9 4.4-1 4.4-2 3.4-1 3.4-2 4.4-3 4.4-4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION....
3/4 3-39 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.
3/4 ACCIDENT MONITORING INSTRUMENTATION..................
3/4 3-40 3-42 ACC IDENT MONITORING INSTRUMENTATION SURVEILLANCE REqUIREMENTS.........................................
3/4 3-43 FIRE DETECTION INSTRUMENTS...........................
3/4 3-45 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-49 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................
3/4 3-51 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.......................................
3/4 3-54 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................
3/4 3-57 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTEO DURING INSERVICE INSPECTION...........................
3/4 4-16 STEAM GENERATOR TUBE INSPECTION.......................
3/4 4-17 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-21 REACTOR COOLANT SYSTEM CHEMISTRY......................
3/4 4-23 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS..........................................
3/4 4-24 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS ROGRAH ~
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4-27 4.4-5 3.4-3 3.4-4 3.6-1 3.6-2 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHORAWAL SCHEDULE...................................
3/4 4-33 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE..... 3/4 4-37a MINIMUM COLO LEG TEMPERATURE FOR PORV USE FOR LTOP.... 3/4 4-37a CONTAINMENT LEAKAGE PATHS.............................
3/4 6-5 CONTAINMENT ISOLATION VALVES..........................
3/4 6-21 ST.
LUGIE UNIT 2 XXIV Amendment No.
8-,
L'I~~T OF TABLES (Continued)
INDEX TABLE 3.7-1
~PAG MAXIMUM ALLOWABLE LINEAR POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERA-TION WITH BOTH STEAM GENERATORS........................
3/4 7-2 4.7-0 4.7-1 4.7-2 3
0 7 3e 3.7-3b 3.7-4 3.7-5 4.8-1 4.8-2 3.8-1 4.1'1-1 4.11-2 3.12-1 STEAM LINE SAFETY VALVES PER LOOP...
SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM......................................
SNUBBER VISUAL INSPECTION INTERVAL....................
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FIRE HOSE STATIONS YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES DIESEL GENERATOR TEST SCHEDULE...................
BATTERY SURVEILLANCE REQUIREMENT.................
MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES...................................
RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS ROGRAM...............................................
P RADIOACTIVE GASEOUS WASTE SAMPLING ANO ANALYSIS ROGRAM ~
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P RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM....
3/4 7-3 3/4 7-8 3/4 7-22 3/4 7-26 3/4 7-27 3/4 7-36 3/4 7-38 3/4 8-8 3/4 8-12 3/4 8-18 3/4 11-2 3/4 11-8 3/4 12-3 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS ENVIRONMENTAL SAMPLES............................
IN 3/4 12-7 4.12-1 B 3/4.2-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS ~
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D B 3/4.4-1 REACTOR VESSEL TOUGHNESS..............................
3/4 12-8 B 3/4 2-3 B 3/4 4-9 5.7-1 6.2-1 MINIMUM SHIFT CREW COMPOSITION-TWO UNITS WITH TWO SEPARATE CONTROL ROOMS................................
6-5 COMPONENT CYCLIC OR TRANSIENT LIMITS..................
5-5 ST.
LUGIE UNIT 2 XXV Amendment No. Md-,
53
Valve Ta Number A) Containment Isolation I-HCV-15-1 I-HCV-18-2 I-HCV-18-1 I-FCV-25-5,4 I-FCV-25-2,3 V-6741 I-HCV-14-7 I-HCV-14-1 I-HCV-14-6 I-HCV-14-2 I-V-2516 I-V-2522 Penetration Number 10 14 23 24 26 TABLE 3.6-2 CONTAINMENT ISOLATION VALVES Function Primary Hakeup Water (CIS)
Station Air Supply Instrument Air Supply (CIS)
Containment Purge Exhaust (CIS)
Containment Purge Makeup (CIS)
Nitrogen Supply to Safety Injection Tanks (CIS)
Reactor Coolant Pump Cooling Water Supply (SIAS)
Reactor Coolant Pump Cooling Water Return (SIAS)
Letdown Line (CIS)
Maximum Testable During Isolation Plant 0 eratien
~Time Sec Yes Yes No No No Yes No No No I-V-5200 I-V-5203 28B I-SE-05-1A, 1B, lc, 1D,1E 28A Safety Injection Tank Sample Reactor Coolant System Hot Leg Sample (CIS)
Yes Yes I-V-5204 I-V-5201 I-V-5205 I-V-5202 29A 29B Pressurizer Surge Sample (CIS)
Pressurizer Steam Sample (CIS)
Yes Yes
C:nM ITl I-V-6718 I-V-6750 I-SE-03-2A,2B I-LCV-07-11A I-LCV-07-11B I-V-6341 I-V-6342 I-V-2524 I"V-2505 Penetration Number 31 41 42 43 TABLE 3.6-2 (Continued)
CONTAINMENT ISOLATION VALVES Function Containment Vent Header (CIS)
Safety Injection Tank Test Line (CIS/SIAS)
Reactor Cavity Sump Pump Discharge (CIS/SIAS)
RCDT Pump Suction (CIS)
RCP Controlled Bleed-off (CIS)
Maximum Testable During Isolation Yes Yes Yes Yes No O
I-FCV-26-1 I-FCV-26-2 I-FCV-26-3 I-FCV-26-4 I-FCV-26-5 I-FCV-26-6 I"FCV-25-26 I-FCY-25-36 I-FCV-25-20 I-FCV-25-21 52A 52B 52C 56 57 Containment Radiation Monitoring (CIS)
Containment Radiation Monitoring (CIS)
Containment Radiation Monitoring (CIS)
Cont.
Containment/Hz Purge Makeup Inlet (CIS)
Cont.
Containment/Hq Purge Exhaust (CIS)
Yes Yes Yes Yes Yes 10 10 10
,ELECTRICAL POMER SYSTEMS ACTION (Continued) d Mi tn two of th< requ i reo offs i te A ~ C ~ ci rcu i ts i rivperab 1 e, demonstrate cne OPERABILITY of two diesel generators by sequentially performing Surveillar.ce Requireme<<t 4.8.1.1.2a.4 on botn dsesels within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , unhSS tre dieSel generatOrS are alreaay Operating; reStOr~
One Of tne
><<operable cffsite scurces to OPERABLE status wstniri 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be iri at least HOT STANOBY v ithln the <<ex.
6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
Following restor-ation of una offsste scurce, fullvw ACTION Statement
a. with the tirrw raquiremer t of that ACTION Statemer:t based ori tl e t.ime oi initial loss of the remainirig inoperable offsite A.C. circuit.
Mith twv of the above requireo oiese I ge<<e~ ators inoperable, deoenstrate the CPERABIL;TY cf two off~lrc A ~ C. circuit; by per-ormirig Surveillance Reouireme<<t 4.8.1.1.1.a withi<<1 hour ono at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months therear'ter; restore one of tne inoparable diesel generators to OPERABLE stacus within
.". hours or he iri at least HOT STANOBY withi<< the rext 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months and i'OLO SHUTCOMN within the follvwirg 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
Following r estvraaion of one diesel gerlerator unit, follow ACTIOt'tatement b.
with the time requirement of that ACTION Statement based on the time of initial los~ of the remaining inoperable diesel generator',
f.
Mstn one Unit 2 startup transformer (2A or 2B) irioperab'ie ar.d with a Unct 1 star tup transformer (1A vr 1B) connected tv che somt A or' offsite power circuit and administratively available to botn units, then should L'nst 1 require tne use of the startup transformer aomin>StratiVely aVai!able tO both unitS, Unit 2 Shall demOriStrate the operability of the renaining A.C. sources by performing Surveillar;ce Requirement 4.8.1.1.la. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.
If either EOG has not been successfully tested within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , demonstrate its OPERABILITY by performing Surveillance Pequirement 4.8.1.1.2a.4 separately for each such EOG
Iithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
Restore the inoperable startup transformer to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLO SHUTOOMN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS
-'.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the onsite Class lE distribution system shall be:
a.
Oetermined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability; and b.
Oemonstrated OPERABLE at least once per 18 months by transferring (manually and automatically) unit power supply from the normal circuit to the alternate circuit.
4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
a.
In accordance with the frequency specified in Table 4.8-1 on a
STAGGEREO TEST BASIS BY:
ST.
LUC IE - UNIT 2 3/4 8-3
'Amendment No. 39, 43
ELECTRICAL POWER o
STEM SURVEILLANCE REQUIREMENTS (Continued) 1.
Verifying the fuel level in the engine-mounted fuel tank, 2.
Verifying the fuel level in the fuel storage
tank, 3.
Verifying the fuel transfer pump can be started and transfers fuel from the storage system to the engine-mounted
tank, 4.
Verifying the diesel starts form ambient condition and accelerates to approximately 900 rpm in less than or equal to 10 seconds**.
The generator voltage and frequency shall be 4160 2 420 volts and 60 k 1.2 Hz within 10 seconds after the start signal**. The diesel generator shall be started for this test by using one of the following signals:
a)
Manual/Local.
b)
Simulated loss-of-offsite power by itself.
c)
Simulated loss-of-offsite power in conjunction with an ESF actuation test signal.
d)
An ESF actuation test signal by itself.
5.
Verifying the generator is synchronized, loaded*** to greater than or equal to 3685 kW in less than or equal to 60 seconds, and operates within a load band of 3450 to 3685 kW at least an additional 60 minutes, and 6.
Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.
b.
By removing accumulated water:
1.
From the engine-mounted fuel tank at least once per 31 days and after each occasion when the diesel is operated for greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and 2.
From the storage tank at least once per 31 days.
The diesel generator start (10 sec) from ambient conditions shall be per-formed at least once per 184 days in these surveillance tests. All other diesel generator starts for purposes of this surveillance testing may be preceded by an engine prelube period and may also include warmup procedures (e.g.,
gradual acceleration) as recommended by the manufacturer so that mechanical stress and wear on the diesel generator is minimized.
Generator loading in less than or equal to 60 seconds shall be performed at least once every 184 days; timing for this loading test shall start upon the closing of the diesel generator breaker. All other loading for the purpose of this surveillance test may be performed according to manufacturer's recommen-dations.
The indicated load band is meant as guidance to avoid routine overloading. Variations in loads in excess of the band due to changing bus loads shall not invalidate this test.
ST.
LUGIE - UNIT 2 3/4 8-4 Amendment No. 89-,
53
PLFCTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL E UIPMENT PROTECTIVE DEVICES,, i,,
MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES LIMITING CONDITION FOR OPERATION 3.8.4 The thermal overload protection bypass
devices, integral with the motor
starter, of each valve listed in Table 3.8-1 shall be OPERABLE.
APPLICABILITY:
Whenever the motor-operated valve is required to be OPERABLE.
ACTION:
With one or more of the thermal overload protection bypass devices inoperable, declare the affected valve(s) inoperable and apply the appropriate ACTION Statement(s) for the affected valve(s).
SURVEILLANCE RE UIREMENTS 4.8.4 The above required thermal overload protection bypass devices shall be demonstrated OPERABLE.
a.
At least once per 18 months, by visually verifying the bypass switch to be in the bypass position for those thermal overload devices which are either:
1.
Continuously bypassed and temporarily placed in force only when the valve motors are undergoing periodic or maintenance
testing, or 2.
Normally in force during plant operation and bypassed under accident conditions.
b.
At least once per 18 months by the performance of a CHANNEL CALIBRATION of a representative sample of at least 25K of:
1.
All thermal over load devices which are not bypassed, such that each non-bypassed device is calibrated at least once per 6 years.
2.
All thermal overload devices which are continuously bypassed and temporarily placed in force only when the valve motors are undergoing periodic or maintenance
testing, and.thermal overload devices normally in force and bypassed under accident conditions such that each thermal overload is calibrated and each valve is cycled through at least one complete cycle of full travel with the motor-operator when the thermal overload is OPERABLE and not bypassed, at least once per 6 years.
ST.
LUCIE - UNIT 2 3/4 8"17 Amendment No. 53
TABL 3.8-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES VALVE NUMBER RCS:
V-1476 V-1477 CVCS:V-2508 V-2509 V-2514 V-2525 V-2553 V-2554 V-2555 V-2501 V-2504 FUNCTION LTOP ISOLATION LTOP ISOLATION BAMT ISOL.
BAMT ISOL.
BAMP DISCH.
PMW SUPPLY CHARGING PUMP BYPASS CHARGING PUMP BYPASS CHARGING PUMP BYPASS VCT ISOL.
RWT ISOL.
BYPASS YES NO YES YES YES YES YES YES YES YES YES YES YES SIS:
FCV-3301 FCV-3306 HCV-3512 HCV-3657 V-3456 V-3457 V-3517 V-3658 V-3540 V-3550 V-3523 V-3551 V-3656,3654 V-3659 V-3660 V-3615,25,35,45 V-3616,26,36,46 V-3617,27,37,47 V-3480 V-3481 V-3651 V-3652 V-3545 V-3664 V-3665 V-3536 V-3539 V-3614,24,34,44 V-3432 V-3444 SHUTDOWN COOLING SHUTDOWN COOLING SHUTDOWN COOLING SHUTDOWN COOLING SDC ISOL.
SDC ISOL.
SDC ISOL.
SDC ISOL.
HOT LEG INJECTION HOT LEG INJECTION HOT LEG INJECTION HOT LEG INJECTION HPSI ISOL.
S I RECIRCULATION S I RECIRCULATION LPSI INJ.
HPSI INJ.
HPSI INJ.
SDC ISOL.
SDC ISOL.
SDC ISOL.
SDC ISOL.
SDC X-TIE SDC ISOL.
SDC ISOL.
SDC WARMUP SDC WARMUP SIT ISOL.
RWT ISOL.
RWT ISOL.
YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES ST.
LUCIE UNIT 2 3/4 8-18 Amendment No. gg
>>T, g-'>>)g)ggi'>>
TABLE 3.8.1'(Continued) h MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES VALVE NUMBER FUNCTION BYPASS YES NO MAIN STEAM:
HV-08-IA;}B MV-08-18A, 18B MV-08-19A,19B HV-08-12,13 HV-08-3 HV-08-14, 15, 16, 17 HSIV BYPASS YES A.D.V.
YES A.D.V, YES AFW TURBINE INLET YES AFW TURBINE INLET YES A.D.V. ISOL.
YES HAIN FEEDWATER:
HV-09-9,10,11,12 HV-09-13,14 ICW:
CCW:
HV-21-2,3 HV-21-4A,4B, HV-14-17,18,19,20 HV-14-9,10,11,12,13,14,15, 16 MV-14-1,2,3,4 AUX. FEED ISOL.
AUX. FEED X-TIE ICW ISOL.
ICW ISOL.
FUEL POOL ISOL.
CONT.
FAN ISOL.
CCW PUMP ISOL.
YES YES YES YES YES YES YES C.S.:
HVAC:
HV-07-1A, 1B MV-07-2A,2B MV-07-3,4 FCV-25-14, 15, 16, 17, 18, 19 FCV-25-24,25 FCV-25-11,12 FCV-25-35 FCV-25-29,34 FCV-25-30,31 FCV-25-32,33 RWT ISOL.
SUMP ISOL.
SYSTEM ISOL.
CRECS ISOL.
CRECS ISOL.
SBVS ISOL.
VENT ISOL.
H2 CONT.
PURGE SFP EXHAUST SBVS INLET YES YES YES YES YES YES YES YES YES YES ST.
LUCIE UNIT 2 3/4 8-19 Amendment No.
53
)4 ADMINISTRATIVE CONTROLS zsamesxmmaaaaaaaaaamaaaamaammmmm
=m===-a=aaaaamaamaaassaaaaaaammaaaaa
-m RESPONSIBILITIES (Continued) k.
Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the President Nuclear Division and to the Company Nuclear Review Board.
l.
Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL and RADWASTE TREATMENT SYSTEMS.
n.
AUTHORITY Review and documentation of judgment concerning prolonged operation in
bypass, channel trip, and/or repair of defective protection channels of process variables placed in bypass since the last FRG meeting.
Review of the Fire Protection Program and implementing procedures and submittal of recommended changes to the Company Nuclear Review Board.
6.5. 1.7 The Facility Review Group shall:
a ~
b.
C.
RECORDS Recommend in writing to the Plant Manager approval or disapproval of items considered under Specifications 6.5. 1.6a.
through d.
and
m. above.
Render determinations in writing with regard to whether or not each item considered under Specifications 6.5. 1.6a, b,
d and e above constitutes an unreviewed safety question.
Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the President-
'Nuclear Division and the Company Nuclear Review Board of disagreement between the FRG and the Plant Manager;
however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6. 1. 1 above.
6.5. 1.8 The Facility Review Group shall maintain written minutes of each FRG meeting that, at a minimum, document the results of all FRG activities performed under the responsibility and authority provisions of these technical specifica-tions.
Copies shall be provided to the President Nuclear Division and the Chairman of the Company Nuclear Review Board.
6.5.2 COMPANY NUCLEAR REVIEW BOARD CNRB)
FUNCTION 6.5.2. 1 The Company Nuclear Review Board shall function to provide independent review and audit of designated activities in the areas of:
'a ~
b.
C.
d.
nuclear power plant operations nuclear engineering chemistry and radiochemistry metallurgy ST.
LUCIE - UNIT 2 6-9 Amendment No. 43-,P3-,47-,
ADMINISTRATIVE CONTROLS FUNCTION (Conti nued) e.
instrumentation and control g,
COMPOS IT ION radiological safety mechagical and electrical engineering quality assurance practices 6.5.2.2 The President
- Nuclear Division shall appoint, in writing, a minimum of five members to the CNRB and shall designate from this membership, in writing, a hairman.
The membership shall function to provide independent review and audit in the areas listed in Specification 6.5.2.1.
The Chairman shall meet the requirements of ANSI/ANS-3.1-1987, Section:.4.7.1.'he members of the ORB shall meet the educ'ational requirements of the'ANSI/ANS-3.1-1987, Section 4.7,2,. and have at-least 5 years of professional level experience in one or more of the fields listed in Specification 6.5.2.1.
CNRB members who do not possess the educational tequirements'f ANSI/ANS-3'.1-1987, Section 4.7.2 (up to a maximum of 2'embers). shall. be evaluated, and have their; membership approved an'd documented, in writing, on a case-by-case basis by the President
- Nuclear Division, considering the alternatives to educational requirements of ANSI/ANS-3.1-1987, Sections 4.1.1 and 4.1.2.
ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the CNRB Chairman to serve on a temporary basis;
however, no more than two alternates shall participate as voting members in CNRB activities at any one 'time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the CNRB Chairman to provide expert advice to the CNRB.
MEETING FRE UENCY 6.5.2.5 The CNRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter and as convened by the CNRB Chairman or his designated alternate.
QUORUM 6.5.2.6 The-quorum of the CNRB necessary for review and'udit functions of these Technical the Chairman or. his designated alternate and members including alternates
. 'io more.than a
have line responsibility fear operation'f the the performance of the CNRB Specifications shall-consist of at least a major ity of CNRB minority; of. the quorom shall unit.
ST.
LUCIE - UNIT 2 6-10 Amendment No. 7$, D 88.
4~D 47,

ML17223B355

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