Text

Amend 37 to License DPR-67,adding Requirements Associated W/Actions Taken to Satisfy Category a Lessons Learned Recommendations to Tech Specs
	ML17209A622
Person / Time
Site:	Saint Lucie
Issue date:	01/19/1981
From:	Clark R; Office of Nuclear Reactor Regulation
To:	;
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ML17209A623	List: ML17209A622; ML17209A624;
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20ddd FLORIDA POWER 5 LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE PLANT UNIT 1

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37 License No.

DPR-67 1.

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

A.

The applications for amendment by Florida Power and Light Company (the licensee) dated January 22 and October 31,

1980,

, comply 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 applications 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 activi.ties will be conducted in compliance with the Comission'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 Comission's regulations and all applicable requirements have been satisfied.

2.

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

DPR-67 is hereby amended to read as follows:.

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

37, 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.

Attachment:

Changes to the Technical Specifications Date of Issuance:

January 19, 1981 FOR THE NUCLEAR REGULATORY COMMISSION

&cC~

Robert A. Clark, Chief Operating Reactors Branch P3 Division of Licensing

ATTACHMENT TO LICENSE AMENDMENT NO.

37 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.

~Pa es IV XV 3/4 3-10 3/4 3-11 3/4 3-13 3/4 3-14 3/4 3-15 3/4 3-16 3/4 3-17 3/4 3-18 3/4 3-19 3/4 3-41 (new) 3/4 3-42 (new) 3/4 3-43 (new 3/4 3-44 new) 3/4 4-4 3/4 4-58 (new) 3/4 7-5 B 3/4 3-'1 B 3/4 3-3 B 3/4 4-2 B 3/4 4-14 (new)

B 3/4 6-3 6-4 6-5 (new) 6-21 6-22

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3 4.0 APPLICABILITY..................~........................

3/4 0-1 3 4.1=

REACTIVITY CONTROL SYSTEMS Shutdown Margin - Ta' 200'F.......

Shutdown Margin - T

< 200'F.......

Boron Dllut)on......................

Moderator Temperature Coefficient...

Minimum Temperature for Criticality.

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3/4.1.1 BORATION CONTROL.....................................

3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-4 3/4 1-5=

3/4 1-7 Flow Paths - Operating............

Charging Pump - Shutdown..........

Charging Pumps -. Operating........

Boric Acid Pumps - Shutdown.......

Boric Acid Pumps - Operating......

Borated Water Sources - Shutdown..

Borated Water Sources - Operating.

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3/4.1.2 BORATION SYSTEMS..................................,,.

Flow Paths - Shutdown.............,..................

3/4 1-8 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-13 3/4 1-14 3/4 1-15 3/4 1-16 3/4 1-4 3/4.1.3 MOVABLE CONTROL ASSEMBLIES......

Full Length CEA Position........

Position Indicator Channels.....

CEA Drop Time...................

Shutdown CEA Insertion Limit....

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~ ~ ~ ~

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Regulating CEA Insertion Limits......................

3/4 1-20 3/4 1-20 3/4 1-24 3/4 1-26 3/4 1-27 3/4 1-28 ST.

LUCIE - UNIT 1

Amendment No.27

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEIL'LANCE RE UIREMENTS SECTION PAGE 3/4.2 POMER DISTRIBUTION LIMITS, 3/4.2.1 3/4.2.2 3/4;2.3'/4.2.4, 3/4.2.5 LINEAR HEAT RATE...,........................;.'.

TOTAL PLANAR RADIAL PEAKING FACTOR - F.......

TOTAL INTEGRATED RADIAL PEAKING FACTOR - F....

AZIMUTHAL POWER TILT T........................

DNB PARAMETERS.................................

3/4 2-1 3/4 2-6 3/4 2-9 3/4 2-11 3/4, 2-13 3/4. 3 INSTRUMENTATION j

3/4.3.1'EACTOR PROTECTIVE INSTRUMENTATION............

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION...........;...................

3/4.3. 3 MONITORING INSTRUMENTATION.........

Radiation Monitoring...............

Incore Detectors.....................................

Seismic Instrumentation............

Meteorological Instrumentation.....

Remote Shutdown Instrumentation....

Chlorine Detection Systems........

Fire Detection Instrumentation.....

Accident Monitoring Instrumentation.................

3/4 3-1 3/4 3-9 3/4 3-21 3/4 3-21 3/4 3-25 3/4 3-27 3/4 3-30 3/4 3-33 3/4 3-36 3/4 3-37 3/4 3-41 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS.................................

3/4 4-1 3/4.4.2 SAFETY VALVES - SHUTDOMN.............................

3/4 4-2 3/4.4.3 SAFETY VALVES - OPERATING..............:..............

3/4 4-3 ST.

LUCIE - UNIT 1 IV Amendment No gf/, 37

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE OCCURRENCE ACTION...............................

6-12 6.7 SAFETY LIMITVIOLATION.....................................

6-13

.8 PROCEDURES.............................................'....

6-13 6.

6.9 REPORTING RE UIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES...............

6-14 6.9.2 SPECIAL REPORTS..........................................

6-18 6.10 RECORD RETENTION..........................................

6-19 6.11 RADIATION PROTECTION PROGRAM..............................

6-20 6.12 HIGH RADIATIONAREA........................,..............

6-20 6.13 ENVIRONMENTAL UALIFICATION...,...,.......,...,...,,...,.

6-21 6.14 SYSTEMS INTEGRITY........................................

6-21 6.15 IODINE MONITORING.........................................

6-21 6.16 BACKUP METHOD FOR DETERMINING SUBCOOLING MARGIN........... 6-22 ST.

LUCIE - UNIT 1

XV Amendment No.gg, 37

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY'FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instru-mentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY:

As shown in Table 3.3-3.

ACTION:

a

~

With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE RE UIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation.

The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES

RESPONSE

TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at'least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the "Total No. of Channels" Column of Table 3 ~ 3 3 ~

ST.

LUCIE - UNIT 1

3/4 3-9

TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUAT'ION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT SAFETY INJECTION (SIAS) a.

Manual (Trip Buttons) b.

Containment Pressure-High c.

Pressurizer Pressure-Low TOTAL NO.

OF CHANNELS CHANNELS TO TRIP 2

MINIMUM CHANNELS OPERABLE APPLICABLE MODES 1, 2, 3, 4 1, 2, 3

1, 2, 3(a)

ACTION 9¹ 9¹ 2.

3.

CONTAINMENT SPRAY (CSAS) a.

Manual (Trip Buttons) b.

Containment Pressure--

High - High CONTAINMENT ISOLATION (CIS) a.

Manual (Trip Buttons) b.

Containment Pressure-High c.

Containment Radiation-High d.

SIAS 2(b) 4 2

3

-,-----(See Functional Unit 1 above) 1, 2, 3, 4 1,

20 3

-1, 2, 3, 4 1, 2, 3

1, 2, 3, 4

10 9¹ 9¹ 4.

MAIN STEAM LINE ISOLATION (MSIS) a.

Manual (Trip Buttons) b.

Steam Generator Pressure

- Low 2/steam generator 4/steam generator 1/steam generator 2/steam generator 2/operating steam generator 3/steam generator 1,2,3,4 1, 2, 3(c)

FUNCTIONAL UNIT MINIMUM TOTAL NO.

CHANNELS OF CHANNELS OPERABLE CHANNELS TO TRIP TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION APPLICABLE MODES

,. ACTION 5.

CONTAINMENT SUMP RECIRCULATION (RAS) a.

Manual RAS (Trip Buttons) b.

Refueling Water Tank-Low 6.

LOSS OF POWER

. 4.16 kv Emergency Bus Undervoltage (Under-voltage relays) 7.

AUXILIARY FEEDWATER AUTOMATIC START Steam Generator (SG)

. Level Instruments 1/Bus

'4/SG 1/Bus 2/SG-1/Bus 2/SG 1, 2, 3, 4 8

1, 2, 3

1,2,3 1, 2, 3

gl 2/SG for.either steam generator will start one train of AFW.

TABLE 3.3-3 Contin'ued TABLE NOTATION (a)

Trip function may be bypassed in this MODE when pressurizer pressure is

< 1725 psia; bypass shall be automatically removed when pressurizer pressure is

> 1725 psia.

(b)

An SIAS signal is first necessary to enable CSAS logic.

(c)

Trip function may be bypassed in this MODE below 585 psig; bypass shall be automatically removed at or above 585 psig.

The provisions of Specification 3.0.4 are not applicable.

ACTION 8, ACT;ION, 9 ACTION STATEMENTS With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-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 .

With the number of-OPERABLE channels one less than, the Total Number of Channels, operation may proceed provided the. following conditions are satisfied:

a, The. inoperabl.e. channel is placed in either the bypassed or, tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

For the purposes, of testing, and maintenance, the inoperable

channel, may, be. bypassed. for up to, 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of; inittal, l,oss. of, OPERABILITY.; however, the inoperable

channel, shal,'1, then, be either: restored to OPERABLE status, or pl,aced in, the tripped condition.

b...

Wi;thin, one hour,, al,'1,'unct]onal uni,ts receiving an i,nput from, the. inoperable channel. are also placed i,n, the. same condition, (ei,ther. bypassed or tripped, as appli,cable),

as. that required by a.. above for the i,noperabl,e. channel~.,

c;.,

The Minimum, Channel;s.OPERABLE requi'rement is met; however,;, one. additi'onal~ channel'ay be bypassed for up to, 48: hours; whitl,e. performing tests; and maintenance on, that channel~

prov~i;ded; the;other inoperabl'e channel i:s placed, i,n; the tripped; conditi'on;,

ST(".. L'UCIE - UNIT[ l~

3/4i 3;12'.

Amendment No.

15

TABLE 3.3-3 Continued TABLE NOTATION ACTION 10-ACTION ll-With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition.

and the Minimum Channels OPERABLE requirement is demonstrated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Sepcifica-tion 4.3.2.1.1.

Instrument operability requirements are contained in the Reactor Protection System requirements for Reactor Trip on Steam Generator Level. If an Automatic Start channel is inoperable, operation may continue provided that the affected pump is verified to be OPERABLE per Specification 4.7.1.2.a within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and at least once per 7 days thereafter; and the Automatic Start channel shall be restored to OPERABLE status within 30 days or the reactor shall 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 .

ST.

LUCIE - UNIT 1

3/4 3-13 Amendment No. /@

37

TABLE 3.3-4 ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES FUNCTIONAL UNIT 1.

SAFETY INJECTION (SIAS) a.

Manual (Trip Buttons) b.

Containment Pressure - High c.

Pressurizer Pressure - Low 2.

CONTAINMENT SPRAY (CSAS) a.

Manual (Trip Buttons) b.

Containment Pressure High-High TRIP SETPOINT Not Applicable psig

> 1600 psia Not Applicable

< 10 psig ALLOWABLE VALUES Not Applicable

< 5 psig

> 1600 psia Not Applicable

< 10 psig 3.

CONTAINMENT ISOLATION (CIS) a.

Manual (Trip Buttons) b.

Containment Pressure - High c.

Containment Radiation - High d.

SIAS Not Applicable

< 5 psig Not Applicable

< 5 psig

< 10 R/hr

-(See FUNCTIONAL UNIT 1 above) 4.

MAIN STEAM LINE ISOLATION (MSIS) a.

Manual (Trip Buttons) b.

Steam Generator Pressure

- Low 5.

CONTAINMENT SUMP RECIRCULATION (RAS) a..

Manual RAS (Trip Buttons) b.

Refueling Water Tank - Low Not Applicable

> 485 psig Not Applicable 48 inches above tank bottom Not Applicable

> 485 psig Not Applicable 48 inches above tank bottom

TABLE 3.3-4 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES FUNCTIONAL UNIT 6.

LOSS OF POWER 4.16 kv Emergency Bus Undervoltage (Undervoltage relays) 7.

AUXILIARY FEEDWATER TRIP VALUE

> 3307 volts

> 30$ level ALLOWABLE VALUES

> 3307 volts

> 30Ã level

TABLE 3.3-5 I

ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATINGSIGNAL AND FUNCTION 1.

Manual a.

SIAS Safety Injection (ECCS)

Containment Fan Coolers Feedwater Isolation Containment Isolation b.

CSAS Containment Spray c.

CIS Containment Isolation 11 Shield Building Ventilation System d.

RAS Containment Sump Recirculation e.

MSIS Main Steam Isolation Feedwater Isolation 2.

Pressurizer Pressure-Low a.

Safety Injection (ECCS) b.

Containment Isolation c.

Containment Fan Coolers d.

Feedwater Isolation

RESPONSE

TIME IN SECONDS Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable'ot Applicable Not Applicable Not Applicable Not Applicable 3O O*/19 5**

3P.5*/2P 5**

3O P*/1 7 P**

< 60.0 ST,.

LUCIE - UNIT.. 1,'/4 3-16 Amendment No. J7, 37

TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION 3.

Containment Pressure-Hi h

a.

Safety Injection (ECCS) b.

Containment Isolation c.

Shield Building Ventilation System d.

Containment Fan Coolers e.

Feedwater Isolation 4.

Containment Pressure--Ki h-Hi h

a.

Containment Spray

RESPONSE

TIME IN SECONDS

< 30.0*/19.5**

< 30 5*/20 5~

< 30.0*/14.0**

< 30.0*/17.0**

< 60.0

< 30.0*/'I8.5**

5.

Containment Radiation-Hi h

a.

Containment Isolation b.

Shield Building Ventilation System 6.

Steam Generator Pressure-Low a.

Main Steam Isolation b.

Feedwater Isolation 7.

Refuelin Mater Stora e Tank-Low a.

Containment Sump Recirculation 8.

Steam Generator Level a.

Auxiliary Feedwater 30 5*/20 5**

< 30 0*/14 0**

< 6.9

< 60.0

< 9'l.5

> 180

< 600 TABLE NOTATION

Diesel generator starting and sequence loading delays included.

- Diesel generator starting and sequence loading delays not included.

~

Offsite power available.

ST

~ LUCIE - UNIT 1

3/4 3-'l7 Amendment No. p; 37

TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT-SAFETY INJECTION (SIAS) a.

Manual (Trip Buttons) b.

Containment Pressure - High c.

Pressurizer Pressure

'- Low d.

Automatic Actuation Logic 2.

CONTAINMENT SPRAY (CSAS) a.

Manual (Trip Buttons) b.

. Containment Pres'sure High - High c,

Automatic Actuation Logic CHANNEL

=

CHECK N.A.

S S

N.A.

S N;A.-

CHANNEL CALIBRATION N.A.

R R

N.A.

R N.A.

CHANNEL FUNCTIONAL

.,TEST R

M M

M(1)

M M(1)

MODES IN WHICH SURVEILLANCE RE UIRED N.A.

1, 2, 3

1, 2,:3 N.A.

1,2,3 1,2,3 3.

CONTAINMENT ISOLATION (CIS) a.

Manual (Trip Buttons) b.

Containment Pressure - High c.

Containment Radiation - High d.

Automatic Actuation Logic e.

SIAS N.A.

S N.A.

N.A.

R R

N.A.

R M

M M(1)

R N.A.

1, 2, 3

1, 2, 3, 4 1, 2, 3

N.A.

4.

5.

MAIN STEAM LINE ISOLATION (MSIS) a.

Manual (Trip Buttons)

N.A.

b.

Steam Generator Pressure - Low S

c.

Automatic Actuation Logic N.A.

CONTAINMENT SUMP RECIRCULATION (RAS) a.

Manual RAS (Trip Buttons) b.

Refueling Water Storage Tank - Low c.

Automatic Actuation Logic N.A.

R N.A.

N.A.

R N.A.

R M

M(1)

M M(1)

'.A.

1, 2, 3

N.A.

1,2,3 1,2,3

TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 6.

LOSS OF POWER 4.16 kv Emergency Bus Undervoltage (Under voltage relays)

CHANNEL CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE RE UIRED 1, 2, 3

7.

AUXILIARY FEEDWATER a.

Auto Start b.

Steam Generator (See Surveillance 4.7.1.2.b) -

(See RPS Table 4.3-1)-

TABLE 4.3-2 Continued TABLE NOTATION (1)

The logic circuits shall be tested manually at least once per 31 days.

F 5T.

LUCIE - UNIT 1 3/4 '3-20

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 sha11 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 MONITORING INSTRUMENTATION INSTRUMENT 1;

Pressurizer Mater Level 2;

Auxiliary Feedwater Flow Rate 3.

RCS Subcooling Margin Monitor, 1/pump 1/pump MINIMUM TOTAL NO.

CHANNELS OF CHANNELS OPERABLE 3

1 ACTION 4.

PORV Position Indicator Acoustic Flow Monitor 5.

PORV Block Valve Position Indicat'or 6.

Safety Valve Position Indicator 1/val ve 1/valve 1/val ve 1/valve 1/valve 1/val ve

TABLE 3.3-11 Continued ACTION STATEMENTS ACTION 1

ACTION 2-ACTION 3-With the number of OPERABLE channels less than required by Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 30 days or be in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

'I With position indication inoperable, restore the inoperable indicator to OPERABLE status or close the associated PORV block valve and remove power from its operator within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 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 .

With. any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information once per shift to determine valve position.

ST LUCIE - UNIT 1

3/4 3-43 Amendment No. 37

TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS INSTRUMENT 1.

Pressurizer Water Level 2.

Auxil>ary Feedwater Flow Rate 3.

Reactor Coolant System Subcooling Margin Monitor 4.

PORV Position Indicator 5.

PORV Block Valve Position Indicator 6., Safety Valve Postition Indicator CHANNEL CHECK CHANNEL CALIBRATION R

REACTOR COOLANT SYSTEM SAFETY VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2500 PSIA+ 1X.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

" With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.4.3 Each pressurizer code safety valve shall be demonstrated OPERABLE with a lift setting of 2500 PSIA + lX, in accordance with Section XI of the ASME Boiler and Pressure Vessel

Code, 1974 Edition.

ST.

LUCIE - UNIT 1

3/-'.

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 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 RE UIREMENTS 4.4.4 In accordance with 4.8.1.1.

l ST LUCIE - UNIT 1

3/4 4-4 'mendment No.

37

REACTOR COOLANT SYSTEM PORV BLOCK VALVES LIMITING CONDITION FOR OPERATION 3.4.12 Each Power Operator Relief Valve (PORV) Block Valve shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

With one or more block valve(s) inoperable, within 1

hour either restore the block valve(s ) to OPERABLE status or close the block valve(s ) and remove power from the block valve(s); otherwise, 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 .

SURVEILLANCE RE UIREMENTS 4.4.12 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.,

ST LUCIE - UNIT 1

3/4 4-58 Amendment No.

37

PLANT SYSTENS SURVEILLANCE RE UIREMENTS Continued 3.

Verifying that each pump operates for at least 15 minutes.

4.

Cycling each testable power-.operated or automatic valve in the flow path through at least, one complete cycle of full travel.

5.

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

b.

At least once per 18 months during shutdown by cycling each power operated valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel, and l.

Verifying that each automatic valve in the flowpath actuates to its correct position upon receipt of the Auto Start actuation signal.

2.

Verifying that each auxiliary feedwater pump starts automatically as designed upon receipt of the Auto Start actuation signal.

ST.

LUCIE - UNIT 1

3/4 7-5 Amendment No. 37

PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank shall be OPERABLE with a minimum contained volume of 116,000 gallons.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

With the conde'nsate storage tank inoperable, restore the condensate storage tank to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-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 .

-SURVEILLANCE RE UIREMENTS 4.7.1.3 The condensate storage tank shall be, demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the water level.

ST; LUCIE - UNIT 1

3/4 7-6

3/4.3 INSTRUMENATION BASES 3 4.3.1 and 3 4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES ESF INSTRUMENTATION The OPERABILITY of 'the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combi-nation thereof reaches its setpoint,

2) the specified coincidence logic is maintained,

3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from.

diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient con-ditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The survei llance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards'.

The periodic surveillance tests per-formed at the minimum frequencies are sufficient to demonstrate this capability.

The measurement of response time at the specified frequencies pro-vides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the-analyses for those channels with response times indicated as not applicable.

Response

time may be demonstrated by any series of sequential, over-lapping or total channel test measurements provided that such tests demonstrate the total channel response time as'efined.

Sensor response

'time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

The Safety Injection Actuation Signal (SIAS) provides direct actuation of the Containment Isolation Signal (CIS) to ensure containment isolation in the event of a small break LOCA.

3 4.3. 3 MONITORING INSTRUMENTATION 3 4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that

1) the radiation levels are continually measured in the areas served ST.

LUCIE - UNIT 1

B 3/4 3-1 Amendment No.

g7>>

37

INSTRUMENTATION BASES RADIATION MONITORING INSTRUMENTATION Continued by the individual channels and 2) an alarm is initiated when the radiation level alarm setpoint is exceeded.

3 4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore-detectors with the specified minimum complement of equipment ensures that the measurements obtained from 'use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3 4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that suffi-cient capbility is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capability is required to'permit comparison of the measured response to that used in the design basis for the facility.

3/4.3.3.4.

METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs",

February 1972.

3 4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of.

HOT SHUTDOWN of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

ST.

LUCIE - UNIT 1

B 3/4 3-2

INSTRUMENTATION BASES 3 4.3.3.6 CHLORINE DETECTION SYSTEMS The operability of the chlorine detection systems ensures that an accidental chlorine release will be detected promptly and the necessary protective actions will be automatically initiated to provide protection for control room personnel.

Upon detection of a high concentration of chlorine, the control room emergency ventilation system will automatically isolate the control room and initiate its operation in the recirculation

'mode of operation to provide the required protection.

The chlorine detection systems required by this specification are consistent with the recommendations of Regulatory Guide. 1.95, "Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine.Release",

Febr uary 1975.

3 4.3.3.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.

This capability is required in order to detect and locate fires in their early stages.

Prompt detection of fires will red'uce'the poten-tial for damage to safety related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4. 3. 3. 8 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.

This capability is consistent with the recommendations of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and -Short-Term Recommendations."

ST.

LUCIE - UNIT 1

B 3/4 3-3 Amendment No. W 37

3 4.4 REACTOR COOLANT SYSTEM BASES 3 4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients.

STARTUP and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for the Power Level-High, Reactor Coolant Flow-Low, and Thermal Margin/I ow Pressure trips have been reduced to their specified values.

Reducing these trip setpoints ensures that the DNBR will be maintained above 1.30 during three pump operation and that during two pump operation the core void fraction will be limited to ensure parallel channel flow stability within the core and thereby prevent premature DNB.

A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure consi-erations require plant cooldown if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

3 4.4.2 and 3 4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

Each safety valve is designed to relieve 2 x 10 lbs per hour of saturated steam at. the valve setpoint.

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating shutdown ooling loop, connected to the RCS, provides overpressure relief capa-bility and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.

The combined relief capacity of these valves is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e.,

no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.

T. LUCIE - UNIT 1

B 3/4 4-1

REACTOR COOLANT SYSTEM BASES SAFETY VALVES Continued Demonstration of the safety valves'ift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASNE Boiler and Pressure Vessel

Code, 1974 Edition.

3 4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a

hydraulically solid system and is capable of acconmodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and power operated relief valve against water relief.

The power operated relief valve and steam bubble function to relieve RCS pressure during all design transients.

Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety Valves.

The required pressurizer.

heater capacity if capable of mainhining natural circulation subcooling.

Operability of the heaters, which are powered by a diesel generator

bus, ensures ability to maintain pressure control even with loss of offsite power.

3 4;4.5 STEAN GENERATORS One OPERABLE steam generator provides sufficient heat removal capa-bility to remove decay heat after a reactor shutdown.

The requirement for two steam generators capable of removing decay heat, combined with the requirements of Specifications 3.7.1.1, 3.7.1.2 and 3.7.1.3 ensures adequate decay heat removal capabi'lities for RCS temperatures greater than 325'F if one steam generator becomes inoperable due to single failure considerations Below 325'F, decay heat is removed'y the shutdown cooli.ng system.

The Surveillance Requirements for i.nspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for i.nservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1, Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there i's evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that

)ead to corrosion.

Inservice inspection of. steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

ST.

LUCIE - UNIT 1 B 3/4 4-2 Amendment No. M 37

REACTOR COOLANT SYSTEM BASES The nondestructive testing for repairs on components greater than 2

inches diameter gives a high degree of, confidence in the integrity of the system, and will detect any significant defects in and near the new welds.

Repairs on components 2 inches in diameter or smaller receive a

surface examination which assures a similar standard of integrity.

In each case, the leak test will ensure leak tightness during normal operation.

For normal opening and reclosing, the structural integrity of the Reactor Coolant System is unchanged.

Therefore, satisfactory performance of a system leak test at 2235 psia following each opening and subsequent reclosing is acceptable demonstration of the system's structural inte-rity.

These leak tests will be conducted within the pressure-temperature limitations for Inservice Leak and Hydrostatic Testing and figure 3.4-2.

1'he Safety Class 2 and 3 components will be pressure tested at least once toward the end of each inspection interval (10 years).

The Safety Class 2 components having a design temperature above 400'.F will be pressure tested at not less than 125 percent of the system design pressure while those components having a design temperature of 400'F and below will be pressure tested at 110 percent of design pressure.

The Safety Class 3 components will be pressure tested at the levels indicated in Specification 4.4.10.3b.

3/4.4:11 CORE BARREL MOVEMENT This specification 'is provided to ensure early detection of 'excessive core barrel movement if it should occur.

Core barrel movement will be detected by using four excore neutron detectors to obtain Amplitude Probability Distribution (APD) and Special Analysis (SA).

Baseline core barrel movement Alert Levels and Action Levels at nominal THERMAL POWER levels of 20Ã, 50/, 80/ and 100/ of RATED THERMAL POWER will be determined during the reactor startup test program.

A modification to the required monitoring program may be justified by an analysis of the data obtained and by an examination of -the affected parts during the plant shutdown at the end of the first fuel cycle.

T. LUCIE UNIT 1

'8 3/4 4-13

REACTOR COOLANT SYSTEM BASES 3/4.4.12 PORV BLOCK VALVES The opening of the Power Operated Relief Valves fulfills no safety related function.

The electronic controls of the PORVs must be maintained OPERABLE to ensure satisfaction of Specifications 4,5.l.d.l and 4.5.2.d.l.

Since it is impractical and undesirable to actually open the PORVs to demonstrate reclosing, it becomes necessary to verify operability of the PORV Block Valves to ensure the ca'pability to isolate a malfunctioning PORV.

ST 'LUCIE - 'UNIT 1

8 3/4 4-14 Amendment No. 37

CONTAINMENT SYSTEMS BASES 3 4.6.2. 2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the spray additive system ensures that suffi-cient NaOH is added to the containment spray in the event of a LOCA.

The limits on NaOH volume and concentration ensure a

pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The contained water volume limit'includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

These assumptions are consistent with the iodine removal efficiency assumed in the accident analyses.

3 4.6.2.3 CONTAINMENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when operated in conjunction with the containment spray systems during post-LOCA conditions.

3 4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmos-phere or pres'surization of the containment.

Containment isolation within the time limits specified'nsures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

This includes the containment purge inlet and outlet valves.

3 4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions.

Either recombiner unit is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions,

2) radiolytic decomposition of water and 3) corrosion of metals within containment.

The containment fan coolers are used in a secondary function to ensure adequate mixing of the containment atmosphere following a LOCA'.

This mixing action will prevent localized accumulations of hydrogen 'from exceeding the flammable limit.

ST.

LUCIE - UNIT 1

8 3/4 6-3

\\

Amendment No.'6',

37

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION ~

LICENSE CATEGORY APPLICABLE NODES SOL OL Non-Licensed Shift Technical Advisor "Does not include the 1'icensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS after the initial fuel loading.

dShift crew compositi'on may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accomodate unexpected absence of on duty 'shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

ST.

LUCIE '- UNIT 1

6-4 Amendment No. 37

ADMINISTRATIVE CONTROLS

6. 3 FACILITY STAFF VALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Protection Manager who shall meet or exceed the qualifications of Regulatory Guide 1.8.,

September

1975, and (2) the Shift Technical Advisor who shall have'a bachelor's degr ee or equivalent in a scientific or engineering discipline with specific training in plant design and in the response and analysis of the plant for'transients and accidents.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Fire Protection Administrator and shall meet or exceed the requirements of Section 27 of the NFPA Code-1975, except for Fire Brigade training sessions which shall be held at least quarterly.

6.5 REVIEW AND AUDIT

6. 5.1 FACILITY REVIEW GROUP FRG FUNCTION 6.5.1.1 The Facility Review Group shall function to advise the Plant Manager on all matters related to nuclear safety.

COMPOS ITION 6.5.1.2 The Facility Review Group shall be composed of the:

Member:

Plant Manager Operations Superintendent Operations Supervisor Maintenance Superintendent Instrument 5 Control Supervisor Reactor Supervisor Health Physics Supervisor Technical Supervisor Chemistry Supervisor guality Control Supervisor Assistant Plant Supt.

Mechanical Assistant Plant Supt. Electrical The FRG Chai'rman shall be designated in writing.

ST.

LUCIE - UNIT 1

6-5 Amendment No.

25~37

ADMINISTRATIVE CONTROLS

6. 1 3 ENVIRONMENTAL UALI FICATION 6.13.1 By no.later than June 30, 1982 all safety-related electrical equip-ment in the facility shall be qualified in accordance with the provisions of:

Division of Operating Reactors "Guidelines for Evaluating Environmental gualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental gualification of Safety-Related Electrical Equipment",

December 1979.

Copies of,these documents are attached to Order for Modification of License DPR-67 dated October 24, 1980.

6.13.2 By no later than December 1, 1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588.

Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

6.14 SYSTEMS INTEGRITY 6.14.1 The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

This program shall include the following:

1.

Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2.

Integrated leak test requirements for each system at a frequency not to exceed refuel.ing cycle intervals.

6.15 IODINE MONITORING 6.15.1 The licensee shall implement a program which will ensure the capability to accurately determine the airborne=iodine concentration in vital areas under accident conditions.

This program shall include the following 1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

ST LUCIE - UNIT 1

6-21 Order dated October24

, 1980 Amendment No 37

ADMINISTRATIVE CONTROLS 6.16 BACKUP METHOD FOR DETERMINING SUBCOOLING MARGIN 6.16.1 The licensee shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin.

This program shall include the following:

1.

Training of personnel, and 2.

Procedures for monitoring.

.'ST.

LUCIE - UNIT 1

6-22 Amendment No.

37

ML17209A622

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