ML17228B242

Proposed Ts,Reflecting Relocation of Selected TS Requirements Re Instrumentation & Emergency & Security Plan Review Process,Per GL 93-07

ML17228B242
Person / Time
Site:	Saint Lucie
Issue date:	08/16/1995
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML17228B241	List: ML17228B240 ML17228B242
References
GL-93-07, GL-93-7, NUDOCS 9508210144
Download: ML17228B242 (54)

Text

St. Lucie Units 1 and 2 Docket No. 50-335 and 50-389 Proposed License Amendments, Relocation of Selected Technical ecification Re uiremen s Related to Instrumentation and NRC Generic Letter 93-07 Line-Item Technical S ecification Im rovement ATTACHMENT 1 St. Lucie Units 1 and 2 Marked-up Technical Specification Pages UNIT 1 UNIT 2 TS PAGES TS PAGES IV V 3/4 3-27 XXIII 3/4 3-30 XXIV 3/4 3-31 3/4 3-32 3/4 3-32 3/4 3-35 3/4 3-45 3/4 3-36 3/4 3-50 3/4 3-37 3/4 3-51 3/4 3-44 3/4 3-52 3/4 3-47 B 3/4 3-2 3/4 3-48 B 3/4 3-4 3/4 3-53 6-8 3/4 3-54 6-13 3/4 3-55 3/4 3-56 3/4 3.-60 B 3/4 3-2 B 3/4 3-3 B 3/4 3-4 6-8 6-13 9508210144 95081b PDR . ADOCK 05000335 P PDR

3/4.2.1 LINEAR HEAT RATE 3/4 2-1 3/4.2,2 DELETED 3/4 2-6 3/4.2.3 TOTAL INTEGRATED RADIALPEAKING FACTOR - F, 3/4 2-9 3/4.2 4 AZIMUTHALPOWER TILT - T, . 3/4 2-11 3/4.2.5 DNB PARAMETERS 3/4 2-13 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION .. ~ 3/4 3-9 3/4.3.3 MONITORING INSTRUMENTATION 3/4 3-21 Radiation Monitoring 3/4 3-21 ll Remote Shutdown Instrumentation . 3/4 3-33 Accident Monitoring Instrumentation 3/4 3<1 Lr.44.

3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4 4-1 3/4.4.2 SAFETY VALVES - SHUTDOWN 3/4 4-2 3/4.4.3 SAFETY VALVES - OPERATING 3/4 4-3 ST. LUCIE - UNIT 1 IV Amendment No. BP, BP. 56, 5P, 59, ~, ~, +34, +35, 1 36

BZ Pages 3/4 3-28 and 3/4 3-~ have been DELETED.

The next page is 3/4 3-30.

55 ST. LUG IE - UNIT 1 3/4 3-27 Amendment No. $ g, 135

NSTRUMENTATION M OROLOGICAL INSTRUMENTATION*

L IMITI CONDITION 'FOR OP ERAT ION 3 ~ 3.3.4 The eteorological monitoring instrumentation channels shown in Table 3.3-8 s 1 be OPERABLE.** I APPLICABILITY:, a 1 1 times.

ACTION:

a. With the numb z f OPERABLE meteorological monitoring chan-nels less than ired by Table 3.3-8, suspend all release of gaseous radio tive material from the radwaste gas decay tanks until the i ble channel(s) is restored to OPERABLE, status.

With one or more requi teorological monitoring channels inoperable for more tha days, prepare and submit a Special Report to the Commission suant to Specification 6.9.2 within the next 10 days ou ning the cause of the malfunction and the plans for restoring to OPERABLE status.

c ~ The provisions of Specification applicable.

~

canal(s) and 3.0.4 are not SURVEILLANCE REQUIREMENTS 4.3.3.4 Each of the. above. meteorological monitoring in rumentation channels shall be demonstrated OPERABLE by the performan of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the fr uencies shown in Table 4.3-5.

• The Meteorological Instrumentation system is shared between St. Luc Unit 1 and.St. Lucie - Unit 2.

• • The emergency power source may be inoperable in Nodes 5 or 6.

ST. LUC I E - UNIT 1 3/4 3-30 Amendment No 39 ~ 5 9

TABLE 3.3-3 t"ET=ORQLGC ICAL HQHITGp. HG . )STR<<a!=HT'TrOH LOS I R"if H I

'I

'i. H:;iU'a H IDIUM C'-;AHHELS iHSTRU . Ec EI EYATEOH ACCURACY OPERABLE

l. MIHOS a) 5)

H Hom' "0

'l Elev (10 meters)

El ev (57. 9 meters) il'

~ 0. 5 "moh

0. 5 "mph 1A =

H.A;

2. MIHO DIRECT a) Hominal v (10 eter s) 5o 1B b) Hominal E ( i.9 meters) 50 H. A.

3. A.'R T=HP=RATURE ( ta T) a) Hominal =lev ( me ar 0 1+Cxx lC b) Hominal Elev (5 me ers 0 1oCxx 1C c) Hominal Elav (33. etar H. A.

5 ar.ing speed oi anemometer sh 1 mph..

AT measuremient channels only.

"Tne 57. 9-me te r ch anne 1 'may be s ub s tut ti o&ihe 10-meter wind speed for up.

o 30 days in he event the l0-meter cha is inopera' e. Mind speed "cta from -he 57.9-meter elevation should be a "ad using the wind speed power law:

5 10 meters = 57.9 meters (0. 1727) where:

5 = wind spe d in mph n = 0.25 for Pasquill Vertical Stability Classes'A,,O.

n = 0.50 for Pasquill Vertical 5tability .Classes "-,F, 1..727 x 10" = constant = 10 meters/57.9 meters 2

The 57.9-meter channel may be subs ituted for the 10-meter win direct'on chcnnel >or uo to 30.days in the event, the '0-meter channel is cerable.

C The- 33. 5-meter chanre I may be substi tuted for one oi she l0 meiaf oi

57. -me: ar tar..perature c annels for up to 30 days ir one o -ha charm s

'no ei abl e. Tne data shoul'd always be ilorma1 l Bee det rmine -he v r-ical stcbili:y cia Ss.

ST. LUCIE - UHIT 1 3/4 3-31 Amendment Ho. 5 9

TA""L E 4. 3- 5

~%'/E'. '. NCE 2 'JU'...: Ni5 Ci AHHEL C'-:AHHEL

> H ST Ul'~ HT CHECK ~ CAL ""RAT:GH

'AIHD SPE=O a) Hominal El ev (= ts r SA b) Hominal Elev (57.. m ers) SA"

.I

2. MIH0 0<RECTiON a) Hcminal Elev (~0 -ieter s) SA b) Hominal Elev (57.o me-ers} SA" AIR, T=HPERATURE (DELTA T) a) Hominal Elev (10 meters) EA b) Hcmjnal Elev (57.9 meters) c), Hominal'Elev (33.5 meters) Q" Q PA"

".",eqvire" cnly ii tnesa channels are being svbstituteC =or one o Nin'.m m Channels Operable as per Table 3.3-3.

~ >

ST. LUCIE - UNIT 1 3/4 3-32 pmendtrent Ho. 5 9

Pages 3/4 3-46 through 3/4 3-49 (Amen n . 123) have been deleted from the Technical Specifications. Th n p is 3/4 3-50.

ST. LUCIE - ONIT 1 3/4 3-45 134

TRUMENTATION EXP VE GAS MONITORING INSTRUMENTATION LIMITING TrON FOR OPERATION 3.3.3. 10 The osive g onitoring instrumentation channels shown in Table 3.3-13 sha e OP B E with their alarm/trip setpoints set to ensure that the li '

Sp ci ication 3. 11.2.5 are not exceeded.

APPLICABILITY: As shown i T 4 3.3 13.

ACTION:

a. With an explosive gas m or n inst mentation channel alarm/trip setpoint less conservativ n requi e b the above Specification, declare the channel inopera e.

b. With less than the minimum numbe ex sive gas monitoring I instrumentation channels OPERABLE, the ACTION shown in Table 3.3-13. If the inoperable inst ts are n t turned to operable status within 30 days, prepar a d submit s ecial report to the Commission within 30 days to expla hy t o erability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.3 and 3.0. e o plicable.

SURVEILLANCE RE UIREMENTS 4.3.3.10 Each explosive gas monitoring instrumentation channel sha e demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CA TION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Tabl 4.3-9.

ST. LUG I E UNIT 1 3/4 3-50 Amendment No. 59, 69, l23

TA8LE 3.3-13 EXPLOSIVE GAS MONITORING INSTRUMENTATION MINIMUM CHANNELS INST T PE 8 i A~iii ~ ii JZI

1. WASTE GAS DECAY TANKS EXPLOSIVE GAS MONITORING SYSTE

a. Oxygen Monitors

• During waste gas system operation.

i ACTION 1 - With the number of channels OPERABLE one less than require the Minimum Channels OPERABLE requirement, operation of this sy e may contfnue for up to 30 days provided samples of Oz are analyze b the lab gas partitioner at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ST. LUCIE - UNIT 1 3/4 3-5I A ndment No. W,~ '34

X OS V GAS 0 0 G S U 0 SU AC U S CHANNEL HOOES I I H

/~SU~

CHANNEL

~c SOURCE

~cd CHANNEL

~II 8 A ~H FUNCTIONAL SURVE CE EO

1. WASTE GAS DECAY TANKS EXPLOSIVE GAS NNITORING SYSTEH

a. Oxygen Honitor a(t)

b. Oxygen Honitor (alternate) )

• Ouring wa e h dup system operation.

(I) The N CAK RATION shall include the use of standard gas samples c a ng a no nal:

One volume percent oxygen, balance nitrogen, and

2. Four volume percent oxygen, balance nitrogen.

(continued}

by the individual channels; and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 and NUREG-0737, "Clarification of TMI Action Plan Requirements,"

November 1980.

TIT~PE LITYof the meteorological instrumentation ensures that sufficient meteorological data is available for e tential radiation doses to the public as a result of routine or accidental release of radioactive matena os here. This capability is required to evaluate the need for initiating protective measures to pro e nd safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Programs," February 1972.

The OPERABILITYof 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 Amendment No. 59, +35, 1 36

TRUMENTATION BA 3 4.3.3.10 EXPLOSIVE GAS MO ORING INSTRUMENTATION-The explosive gas monitorin 'trumentation is provided to monitor the concentrations of potentially expl 've gas mixtures in the waste gas holdup system. The OPERABILITY and use of t instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

ST. LUCIE UNIT 1 B 3/4 3-4 Amendment No.-59, 123

6. 0 ADMINISTRATIVE CONTROLS

i.

-Renew-@'Ne-Security-f4an-and impkement4ng-procedures-and-subm-'(+he-ef recommended-changes t~4e-Gompany-NuRea~ev-tew-

-%oared-. got Os60 P g(k~

gC.

j. -Rev4ew-ofthe-Emergency-PRn-and-im~menting-pr ocedures-and-

-scbmi-'tt~f-recommended-'changes'-the-Company-Nu&ea~eview-

-Board-. Qor o,F60

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.

l. Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL and RADWASTE TREATMENT SYSTEMS.

m. 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.

n. Review of the Fire Protection Program and implementing procedures and submittal of recommended changes to the Company Nuclear Review Board.

AUTHORITY 6.5.1.7 The Facility Review Group shall:

a. Recommend in writing to the Plant General Manager, approval or disapproval of items considered under Specifications 6.5.1.6.a through d above.

b. 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.

C. Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the President-Nuclear Division and the Company Nuclear Review Board of disagree-ment between the FRG and the Plant General Hanager; however, the Plant General Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1 above.

RECORDS 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 Amendment No. . . , . . .126

.0 AOHIN STRAT V CONTR S

c. The Safety Limit Violation Report shall be submitted to the Coamis-sion, the CNRB, and the President - Nuclear Division within 14 days of the violation.

d. Critical operation of the unit shall not be resumed until authorized by the Commission.

6.8 PROC A 6.8. I Mritten procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide ).33, Revision 2, February 1978, and those required for implementing the requirements of NUREG 0737.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. No< Q56o ~

g ~('! ace e. No< 0880

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. guality Control Program for effluent monitoring, using the guidance in Regulatory Guide 1.21, Revision 1, June 1974.

P guality Control Program for environmental monitoring using the guidance in Regulatory Guide 4.1, Revision 1, April 1975.

6.8.2 Each procedure of Specification 6.8.la through i. above, and changes thereto, shall be reviewed by the FRG and shall be approved by the Plant General Manager prior to implementation and shall be reviewed periodically as set forth. .4n-administrative procedures.

6.8.3 Temporary changes to procedures of Specification 6.8.1a through i.

above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

1 ST. LUCIE - UNIT 1 6-13 A 1 t N . S:MHaHs. 126

3/4.2.1 LINEAR HEAT RATE 3/4 2-1 3/4.2.2 TOTAL PLANAR RADIALPEAKING FACTOR - F~ 3/4 2-7 3/4.2.3 TOTAL INTEGRATED RADIALPEAKING FACTOR - F, 3/4 2-9 3/4.2.4 AZIMUTHALPOWER TILT .. 3/4 2<<13 3/4.2.5 ONB PARAMETERS 3/4 2-14 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION... ~ ~ ~ . ~ .. ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ . ~.... 3/4 3-1 1 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION . ~ . 3/4 3 24

-MmEeRe~eAL INSmVMEN~O della.

REMOTE SHUTDOWN INSTRUMENTATION ....... 3/4 3-38 ACCIDENT MONITORING INSTRUMENTATION ~ . ~.... 3/4 3<1 mcezeN-IN~RVMEN~e ...... 7 ge,(e4.

Game .. 6 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION 3/4 4-1 HOT STANDBY.............................. 3/4 4-2 HOT SHUTDOWN ............ 3/4 4-3 COLD SHUTDOWN (LOOPS FILI ED) . 3/4 4-5 COLD SHUTDOWN (LOOPS NOT FILLED)...................,,... 3/4 4-6 ST. LUCIE - UNIT 2 Amendment No. &tW,M, 75

INDEX LIST OF TABLES TABLE PAGE-FREQUENCY NOTATION.......-....... 1-8 1.2 OPERATIONAl MODES...-.

2. 2-1 REACTOR. PROTECTIVE INSTRUMENTATION TRIP SETPOINT L IMITS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3. 1-1 MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR ST. LUCIE-2.-..-..o................... ~..... 3/4 1-17

3. 2-1 D ELETED ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 3/4 2-11 3.2-2 DNB MARGIN LIMITS......................................... 3/4 2-15 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION........................ 3/4 3-2 3~3 2 D ELETED o e e o e ~ ~ ~ ~ ~ o o ~ ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

4. 3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-8 3~3 3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION........................................... 3/4 3-12 3.3-4 ENGINEERED SAfETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES............................... 3/4 3-17 3.3-5 D ELETED ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................. 3/4 3-22 3.3-6 RADIATION MONITORING INSTRUMENTATION...................... 3/4 3-25 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE R EQUIREMENTSoo ~ o ~ ~ ~ ~ .. ~ ~ ~ ~ ~ o ~ ~ oo ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ 3/4 3-28 PfcGTE~

3.3-8 WERM6 QG,BTP.D.

4.3-5 +meaeLG~LMe~~~maUMEm~e REgin REM ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ e ~ ~

ST. LUCIE-UNIT 2 XXIII Amendment No. W +. ~4

'FACE:(

.'ST CF .'BLES (Conti;.ued)

TABLE PAGE 3.3-9 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION...... 3/4 3-39 4.3-6 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE RFQUIREMENTS......................................... 3/6 3-40 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-42 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREHENTS............................... 3/4 3-43 3.3-11 DELETED 3.3-12 DELETED 4.3-8 DELETED Oe~evkb 3.3-13 ~A44G HRUHENTIR o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ o o ~ ~ o ~ o ~ o ~ ~ ~ ~ ~ ~ ~

ge:t.Et'Gb 4.3-9 . em-meme-mSWuxam~N

~RVH.t88~EQUSIEM 4.4-1 HINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION........................... 3/4 4-16 4.4-2 STEAM GENERATOR TUBE INSPECTION................-...... 3/4 4-17 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-21 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY............ 3/4 4-23 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS.......................................... 3/4 4-24 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE ANO ANALYSIS P ROGRAMo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-27 4.4-5 OELETEO 3.4-3 LOM TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE..... 3/4 4-37a 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP.... 3/4 4-37a 3.6-1 CONTAINMENT LEAKAGE PATHS............;.'............... 3/4 6-5 3.6-2 CONTAINMENT ISOLATION VALVES.................... 3/4 6-21 ST. LUCI E UNIT 2 XXIV Amendment No. 8, SQ,'~,73

3/4 3-33 and 3/4 3-A have been 3V'ages DELETED.

The next page is 3/4 3-P&.

38 ST. LUCIE - UNIT 2 3/4 3-32 Amendment No. /4

STRUMENTATION ME OLOG ICAL INSTRUMENTATION" LIMITING DITION FOR OPERATION 3.3.3.4 The me e ologic monitoring inqtrumentation channels shown in Table 3.3-8 shal OPE AB E.

APPLICABILITY: At a ACTION:

ao With the number of R me orological monitoring. channels less than required b ble 3. -,~suspend all release of gaseous radioactive material f th ra Paste gas decay tanks until the inoperable channel(s) is tor to OPERABLE status.

b. With one or more required me ological monitoring channels inoperable for mo'e than 7 day prepare and submit a Special Report to the Commission pursuant to Sp fication 9.2 within, the next 10 days outlining the cause of th 1 fun io and the plans for restoring the channel (s) to OPERABL t C. The provisions of Specifications 3.0.3 SURVEILLANCE RE UIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumenta on channels shall be demonstrated OPERABLE by the performance of the CHANNEL CK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.

The Meteorological Instrumentation system is shared between St. Lucie - U 1 and St. Lucie - Unit 2.

ST. LUC IE - UNIT 2 3/4 3"35

TABLE 3. 3-8 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT MINIMUM MINIMUM CHANNELS INSTRUMENT 8( L VATION ACCURACY OPERABLE 1 ~ WINDS PEED a) Nominal v (10 meters) 2 0.5 "mph 1A b) Nominal e (57.9 rs) 2 0.5 "mph N.A.

2. WIND DIRECTION a) Nominal Elev e r s) 50 1B b) Nominal Elev (5 met s) x 5~ N.A.

3. AIR TEMPERATURE (Delta a) Nominal Elev (10 met ) 0 10C j(g 1C b) Nominal El ev (57. 9 me ) 0 10C<g 1C c) Nominal Elev (33.5 mete ~ 0 oCww N.A.

Starting speed of aremometer shall b 1 ph.

hT measurement channels only.

The 57.9"meter channel may be substituted fo the 10-meter wind speed for up to 30 days in the event the 10-meter channel inoperable. Wind speed data from the 57.9"meter elevation should be adjuste sing the wind speed power l aw:

(0. 1727) n S S 10 meters = 57.9 meters where:

S = wind speed in mph n = 0. 25 for Pasquill Vertical Stability Classes A,B,C,D.

n = 0.50 for Pasquill Vertical Stability Classes E,F,G.

1.727 x 10- = constant = 10 meters/57.9 meters The 57.9-meter channel may be substituted for the 10-meter wind di tion channel for up to 30 days in the event the 10-meter channel is inop ble.

C The 33.5-meter channel may be substituted for one of the 10-meter or 57.9-meter temperature channels for up to 30 days if one of the channels inoperable. The data should always be normalized to 'C/100 meters to determine the vertical stability class.

ST. LUCIE - UNIT 2 3/4 3"36

TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SUR EILLANCE RE UIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION 1~ WIND SPEED a) Nominal Elev (10 me SA b) Nominal Elev (57.9 me r SA"

2. WIND DIRECTION a) Nominal Elev (10 meters) 0 SA

. b) Nominal Elev (57.9 meters) D SA"

3. AIR TEMPERATURE (DELTA T) a) Nominal Elev (10 meters) A b) Nominal Elev (57.9 meters) D ~

SA c) Nominal Elev (33.5 meters) D)kl Required only if these channels are being substituted for one o t Minimum Channels Operable as per Table 3.3-8.

ST. LUCIE " UNIT 2 3/4 3-37

0 U

Pages 3/4 3-45 through. 3/4 amendment No, 55} have heen deleted from the Technical Specificat . The next page is 3/4 3>>47.

ST. LUCIE - UNIT 2 3/4 3-44 Amendment No.~ 7

f TRUMENTATION LOO ". RT DETECTION INSTRUMENTATION LIMITIN C DITION FOR OPERATION 3.3.3.8 The e-part detection system shall be OPERABLE.

APPLICABILITY: M 1 and ACTION:

With one or more 1 e pa detection system channels inoperable for more than 30 days, e and bmit a Special Report to the Commission pursuant ecif atio 6.9.2 within the next 10 days outlining the cause o unct'nd the plans for restoring the channel(s) to OPE RAB atus.

b. The provisions of Specificat n 3.0 and 3..0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.8 Each channel of the loose-part detection sy t hall demonstrated OPERABLE by performance of:

a. a CHANNEL CHECK at least once pel 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. a CHANNEL FUNCTIONAL TEST at least once per 31 days, n C. a CHANNEL CALIBRATION at least once per 18 months.

ST. LUCIE - UNIT 2 3/4 3-47

pages 3/4 3-49 through 3/4 3-5 Amendment No. 61) have been deleted from the Technical Specification The next page is 3/4 3-53.

6 ST. LUCIE. - UNIT 2 3/4 3-48 AMENDMENT NO. T8s @> 73

STRUM ENTATION E IVE GAS MONITORING NSTRUMENTATION LIMITING N ITION FOR OPERATION 3.3.3.10 Th e losive ga monitoring instrumentation channels shown in Table 3.3-13 s l be OP LE with their alarm/trip setpoints set to ensure that the limits o ec'fic tio 3.11.2.5 are not exceeded.

APPLICABILITY: As show s abl 3. 3-ACTION:

a. With an explosive gas setpoint less conservati

'ri n g in r mentation channel alarm/trip re ired by the above Specification, declare the channel inoper

b. With less than the minimum numb f explosive ga monitoring instrumentation channels OPERABL , ke the ACT N hown in Table 3.3-13. If the inoperable instrumen re not etur ed to operable status within 30 days, prepare and su s a r ort to the Commission within 30 days to explain wh t inoper lity was not corrected in a timely manner.

C. The provisions of Specifications 3.0.3 and 3.0. t pl i cabl e.

SURVEILLANCE RE UIREMENTS 4.3.3.10 Explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALI ION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.

ST. LUCIE UNIT 2 3/4 3-53 Amendment No. ~%, 61

IAA~~. -L3 EX OS V GAS 0 0 S U A 0 H INIHUH CHANNELS QS EUUgg 0 CA

1. WASTE GAS OECAY TANKS EXPLOSIVE GAS MONITORING SYSTEM

a. Oxygen Honitors TABLE NOT N

• During waste g'as system ope on.

ACTION 35 Mit nu er of channels OPERABLE one less than required by t ni annels OPERABLE requirement, operation of this em ay co tinue for up to 30 days provided samples of Oz are a zed by the lab gas partitioner at least once per 24 hours.

)I TABLE 4.3-9 X OS GAS 0 0 CHANNEL NODES IN Mlli(.U LSIfSKtK CHANNEL SOURCE CHEK.

CHANNEL FUNCTI

~Rf'l SURVEII.LANCE

1. MASTE GAS DECAY TANKS EXPLOSIVE GAS HONITORING SYSTEM

a. Oxygen Honitor Q(I)

b. Oxygen Honitor (alternate) Q(I)

• During waste oR p s eration.

(I) The CH CA ION shall include the use of standard gas samples cont g a ominal One volume percent oxygen, balance nitrogen, and 0

2. Four volume percent oxygen, balance nitrogen.

Pages 3/4 3-57 through 3/4 3-59 ( dment No. 61) have been deleted from the Technical Specifications. next page is 3/4 3-60.

ST. LUCIE - UNIT 2 3/4 3-56 Amendment No. 57,73

0 ST RUNE NTATION 3 ..4 TURBINE OVERSPEED PROTECTION LIHIT CONDITION FOR OPERATION 3.3.4 At st one turbine overspeed protection system shall be OPERABLE.

APPLICABILI 'ODES 1,," and 3."

ACTION:

With one s o val one control valve per high pressure turbine steam lea o able and/or with one reheat stop valve or one reheat inte t lve er low pressure turbine steam lead inop-erable, rest r the operable valve(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or at least one valve in the affected steam lead or isolate the tur l from the steam supply within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the above re red tu i overspeed protection system otherwise inoperab e, w in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> isolate the turbine from the steam supply.

SURVEILLANCE RE UIREHENTS

4. 3.4. 1 The provisions of Specifica i .0. are not applicable.

4.3.4.2 The above required turbine ove s ed protection system shall be demonstrated OPERABLE:

a. At least once per month by cycli ach o he ollowing valves through at least one complete cycle om e nning position.

l. Sour high pressure turbine stop a ves.

2. Four high pressure turbine control v e

~1

3. Four low pressure turbine .reheat sto a ve .

4. Four low pressure turbine reheat inter e t va es

b. At least once per 31 days by direct observatio f he movement of each of the above valves through one complete c from the running pos l tl on.

c. At least once per 18 months by performance of a CHA CALIBRATION on the turbine overspeed protection systems.

d. At least once per 40 months by disassembling at least on f each of the above valves and performing a visual and surface insp c ion of valve seats, disks and stems and verifying no unacceptable ws or corrosion.

With any main steam line isolation valve and/or any main steam line iso t'on valve bypass valve not fully closed. ~

- 3/4 3"60 Amendment No.

ST. LUCIE UNIT 2 y

,J individual channels; and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on seiected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess, Plant and Environs Conditions During and Following an Accident,"

December 1980 and NUREG-0737, "Clarification of TMI Action Plan Requirements,"

November 1980.

I PE data are available for es I 'e TY of the meteorological instrumentation ensures that sufficient meteorological

'diation accidentalreleaseofradioactivematerialstotheamo doses to the public as a result of routine or her~'bili isrequiredto evaluate the need for initiating protective measures to protect the health and sa e ublic The OPERABILITY of the remote shutdown system instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY 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 Criterion 19 of 10 CFR Part 50.

The OPERABILITY of the remote shutdown system instrumentation ensures that a fire will not preclude achieving safe shutdown. The remote shutdown system instrumentation, control circuits, and transfer switches are independent of areas where a fire could damage systems normally used to shut down the reactor. This capability is consistent with General Design Criterion 3 and Appendix R to 10 CFR Part 50.

ST. LUCIE - UNIT 2 8 3/4 3-2 Amendment No. W, 75

0 INSTRUMENTATION BASES 3 4.3.3.6 ACCIOENT MONITORING INSTRUMENTATION The OPERABIL'ITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monito" and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

~34.3.3.7 DELETED DELETED 3 4.3.3.8 RABILITY of the loose-part detection instrumentation ensures that

'vailable sufficient capa primary system and avoid or 'g allowable out-of-service times and surve to detect loose metallic parts in the damage to primary system components.

requirements are consistent The with the recommendations of Regulatory Guide 1.13 , - art Detection Program for the Primary System of Light-Water-Cooled Reactors, ST. LUCIE , UNIT 2 B 3/4 3-3 Amendment No. SS, 61

STRUMENTAT ION BAS 3 4.3. 0 EXPLOSIVE GAS MONITORING INSTRUMENTATION The losive gas monitoring instrumentation is provided to monitor the concentrat> s of potentially explosive gas mixtures in the waste gas holdup system. The RABILITY and use of this instrumentation is consistent with the requirements General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3 4.3.4 TURBINE 0 PEED PROTECTION This specificatio is provided to ensure that the turbine overspeed protection .instrumentat'nd ' the turbine speed control valves are OPERABLE and will protect the tur from excessive overspeed. Protection from turbine excessive overspee 's required since excessive overspeed of the turbine could generate pote 'ally damaging missiles which could impact and damage safety related compone , equipment or structures.

ST. LUG IE UNIT 2 B 3/4 3-4 Amendment No. 8S,

M T NG FR U NCY 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.

QUOR~U 6.5.1.5 The quorum of the FRG necessary for the performance of the FRG responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alter nate and four members including alternates.

SPONSIBIL T S 6.5.1.6 The Facility Review Group shall be responsible for:

a. Review of (1),@l procedures required by Specification 6.8 and changes thereto-,~(P) all programs required by Specification 6.8 and changes thereto, and {3) any other proposed procedures or changes thereto as determined by the Plant General Manager to affect nu-clear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

c. Review of all proposed changes to Appendix A Technical Specifica-tions.

d. Review of all proposed changes o} modifications to unit systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the President-Nuclear Division and to the Chairman of the Company Nuclear Review Board.

f. Review of all REPORTABLE EVENTS.

g. Review of unit operations to detect potential nuclear safety hazards.

'..".- Performance of special reviews, investigations or analyses and

~

reports thereon as requested by the Plant General Manager or the Company Nuclear Review Board.

1 bn%:

MORrd-.

9 ~P1 lt~ N~p t4~ w-II 1 d Hav osau ST. LUCIE - UNIT 2 6-8 Amendment No. 43,%-,47,65

ADMINISTRATI VE CONTROLS

6. 6 REPORTABLE EVENTS ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Coamfssfon shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the FRG, and the results of this review shall be submitted to the CNRB, and the President-Nuclear Division.

6.7 SAFETY LIMIT VIOLATION 6.7.l The following actions shall be taken fn the event a Safety Limit is violated:

a. The NRC Operatfons Center shall be notiffed by telephone as soon as possible and fn all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The President - Nuclear Ofvfsion and the CNRB shall be notified wfthfn 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. A Safety Lfmft Violation Report shall be prepared. .The report'hall be reviewed by the FRG. Thfs report shall describe (1) applicable circumstances preceding the violation, (2) effects of the vfol'ation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

c. The Safety Limit Violation Report shall be submitted to the Conefssion, the CNRB,.and the President -. Nuclear Ofvfsfon wfthfn"14 days of the vfolatfon.

d. Critical operation of the unit shall not be resumed until authorized by the Coaeission.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained coverfng the activities referenced below:

a. -=The applicable procedures r ecoamended fn Appendfx "A" of Regulatery

- Gufde 1.33, Revision 2, February 1978, and those required for implementing the requirements of NUREG 0737.

b. Refueling operations.

c. Surveillance and test activftfes of safety-related equipment.

e. Bmg N f l4 ST. LUCIE - UNIT 2 6-13

St. Lucie Units 1 and 2 Docket No. 50-335 and 50-389 Proposed License Amendments Relocation of Selected Technical S ecification Re uirements Related to Instrumentation and NRC Generic Letter 93-07 Line-Item Technical S ecification Im rovement ATTACHMENT2 SAFETY ANALYSIS Relocation of Selected Technical Specification Requirements Related to Insbumentation Introduction The staff of the Nuclear Regulatory Commission (NRC) issued a Final Policy Statement on Technical Specification Improvement for Nuclear Power Reactors, 58 FR 39132, July 22, 1993.

The purpose of this policy statement was to focus the Technical Specifications on those requirements that are of controlling importance to operational safety by screening each Technical Specification using the criteria provided in the Policy Statement. The criteria are intended to identify requirements derived from the analyses and evaluations included in the Updated Final Safety Analysis Report (UFSAR) that are of immediate concern to the health and safety of the public. Technical Specifications that meet one or more of the criteria must be retained. Those Technical Specifications that do not meet any of the criteria may be proposed for removal from the Technical Specifications. The selected Technical Specification Monitoring Instrumentation contained in this amendment do not meet any of the four Policy Statement criteria and will be relocated to the UFSAR. Relocating these Technical Specification requirements to the UFSAR will ensure the control of future changes are under the requirements of 10 CFR 50.59.

Discussion A review of the four criteria as they relate to the proposed changes to St. Lucie Unit 1 and 2 Technical Specification Sections 3/4.3.3.4, 3/4.3.3.10 and their associated bases, and Unit 2 Technical Specification Sections 3/4.3.3.8, 3/4.3.4 and their associated bases are discussed below:

1. TS 3/4.3.3.4: Delete the Meteorological Instrumentation TS and Bases Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Justification: The Meteorological Instruments are designed to provide a meteorological data base for use in planning of radioactive effluent releases and as a means of estimating the potential radiological consequences of hypothetical accidents. These instruments are not installed for, nor are they capable of, detecting reactor coolant leakage. Therefore, the Meteorological Instruments are not used to detect a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either 'assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

Justification: Parameters measured by these instruments do not represent process variables which are controlled during power operation to ensure that process values remain within the analysis bounds. The Meteorological Instrumentation system contains no active design features, nor does it require any operating restrictions, for the purpose of precluding unanalyzed accidents and transients. Therefore, initial conditions assumed in the Design Basis Accident and Transient Analyses evaluated in the PSL1 and PSL2 UFSARs do not involve the Meteorological Instrumentation.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Justification: The Meteorological Instrumentation does not contain any structures, sub-systems, or components that would be needed to function or actuate for the purpose of mitigating the consequences of a Design Basis Accident. Therefore, this instrument system is not part of a primary success path for the St. Lucie plant response to an accident.

Criterion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Justification: These instruments cannot be used to predict, prevent, or mitigate the consequences of Design Basis Accident. Consequently, the Meteorological Instrumentation would not be included among the structures, systems, and components typically evaluated in a probabilistic safety assessment to determine constraints of prime importance in limiting the likelihood or the severity of accident sequences that are commonly found to dominate risk. For these reasons, the Meteorological Instrumentation TS requirements proposed for relocation to the UFSARs are considered not significant to the protection of public health and safety.

2. TS 3/4.3.3.10: Delete the Explosive Gas Monitoring Instrumentation TS and Bases Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Justification: The Explosive Gas Monitoring Instrumentation is designed to detect the formation of potentially explosive mixtures of hydrogen and oxygen in the Gaseous Waste Management System (GWMS), Once detected, the remainder of the GWMS functions to preclude the formation of potentially explosive mixtures by an extensive combination of design features. The Explosive Gas Monitoring Instrumentation is not installed for, nor is it capable of, detecting

reactor coolant leakage. Therefore, the Explosive Gas Monitoring Instruments are not used to detect a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

Justification Parameters measured by these instruments do not represent process variables which are controlled during power operation to ensure that process values remain within the analysis

'bounds. The Explosive Gas Monitoring Instrumentation contains no active design features, nor does it require any operating restrictions, for the purpose of precluding unanalyzed accidents and transients. Therefore, initial conditions assumed in the Design Basis Accident and Transient Analyses evaluated in the PSLl and PSL2 UFSARs do not involve the Explosive Gas Monitoring Instrumentation.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

I~us ification: The Explosive Gas Monitoring Instrumentation does not contain any structures, sub-systems, or components that would be needed to function or actuate for the purpose of mitigating the consequences of a Design Basis Accident. Therefore, this instrument system is not part of a primary success path for the St. Lucie plant response to an accident.

~Cri erion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Justification: These instruments cannot be used to predict, prevent, or mitigate the consequences of a Design Basis Accident. Consequently, the Explosive Gas Monitoring Instrumentation would not be included among the structures, systems, and components typically evaluated in a probabilistic safety assessment to determine constraints of prime importance in limiting the likelihood or the severity of accident sequences that are commonly found to dominate risk. For these reasons, the Explosive Gas Monitoring Instrumentation TS requirements proposed for relocation to the UFSAR are considered not significant to the protection of public health and safety.

3. TS 3/4.3.3.8 Delete the Loose-Part Detection Instrumentation TS and Bases Criterion I: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Justification The Loose-Part Detection Instrumentation is designed to detect loose parts in the Reactor Coolant System. These instruments are not installed for, nor are they capable of,

detecting reactor coolant leakage. Therefore, the Loose-Part Detection Instruments are not used to detect a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

Justification: Parameters measured by these instruments do not represent process variables which are controlled during power operation to ensure that process values remain within the analysis bounds. The Loose-Part Detection Instrumentation contains no active design features, nor does it require any operating restrictions, for the purpose of precluding unanalyzed accidents and transients. Therefore, initial conditions assumed in the Design Basis Accident and Transient Analyses evaluated in the PSLl and PSL2 UFSARs do not involve the Loose-Part Detection Instrumentation.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Justification: The Loose-Part Detection Instrumentation does not contain any structures, sub-systems, or components that would be needed to function or actuate for the purpose of mitigating the consequences of a Design Basis Accident. Therefore, this instrument system is not part of a primary success path for the St. Lucie plant response to an accident.

Criterion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Justification: These instruments cannot be used to predict, prevent, or mitigate the consequences of a Design Basis Accident. Consequently, the Loose-Part Detection Instrumentation would not be included among the structures, systems, and components typically evaluated in a probabilistic safety assessment to determine constraints of prime importance in limiting the likelihood or the severity of accident sequences that are commonly found to dominate risk. For these reasons, the Loose-Part Detection Instrumentation TS requirements proposed for relocation to the UFSARs are considered not significant to the protection of public health and safety.

4. TS 3/4.3.4: Delete the Turbine Overspeed Protection TS and Bases Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Justification: The Turbine Overspeed Protection system is designed to protect the unit generation set from damage. This system is not installed for, nor is it capable of, detecting reactor coolant

leakage. Therefore, the Turbine Overspeed Protection system is not used to detect a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

Justification Parameters measured by this system are not associated with process variables which are controlled during power operation to ensure that process values remain within the analysis bounds. The Turbine Overspeed Protection system contains no active design features, nor does it require any operating restrictions, for the purpose of precluding unanalyzed accidents and transients. Therefore, initial conditions assumed in the Design Basis Accident and Transient Analyses evaluated in the PSL2 UFSAR do not involve the Turbine Overspeed Protection system.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

~Justifies ion: The Turbine Overspeed Protection system does not contain any structures, sub-systems, or components that would be needed to function or actuate for the purpose of mitigating the consequences of a Design Basis Accident. Therefore, this instrument system is not part of a primary success path for the St. Lucie plant response to an accident.

Criterion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Justification: Operability and surveillance requirements for the Turbine Overspeed Protection system ensures that the system will function to protect the generation set at St. Lucie from damage. The NRC concluded in the Unit 2 Safety Evaluation Report, NUREG-0843 Supplement 1, dated December 1981 that the total turbine missile risk from high and low trajectory missiles for the St. Lucie Unit 2 design is acceptably low so that the plant structure, system, and components important to safety are adequately protected against potential turbine missiles.

Consequently, the Turbine Overspeed Protection system has been shown to not be significant to public health and safety by probabilistic safety assessment. For this reason, the Turbine Overspeed Protection TS requirements proposed for relocation to the UFSAR are considered not significant to the protection of public health and safety.

Summarur The selected Technical Specification Monitoring Instrumentation contained in this amendment do not meet any of the four Policy Statement criteria and will be relocated to the UFSAR.

Relocating these Technical Specification requirements to the UFSAR will ensure the control of future changes are under the requirements of 10 CFR 50.59. FPL will submit the revised UFSARs in accordance with 10 CFR 50.71(e) requirements.

NRC Generic Letter 93-07 Line-Item Technical Specification Impiuvement Introduction The NRC issued Generic Letter 93-07, "Modification of the Technical Specification Administrative Requirements for Emergency and Security Plans" on December 28, 1993. This Generic Letter was issued to provide guidance to licensees to remove the audit of the Emergency Plan, Security Plan and implementing procedures from the list of responsibilities of the company nuclear audit and review group. The basis of this change is that Parts 50 and 73 of Title 10 of the Code of Federal Regulations include provisions that are sufficient to address these requirements. In addition, the Generic Letter gives guidance to allow Technical Specifications changes to remove (1) the review of the emergency and security plans from the list of responsibilities of the unit review group and (2) the requirements for the unit review group to review procedures, and procedure changes, for the implementation of the emergency and security plans, provided the licensee relocates these requirements to the respective emergency and security plans. The selected Technical Specifications are being relocated to the Emergency Plan or Security Plan as appropriate. Relocating these requirements to the appropriate plan will ensure the control of future changes are under the requirements of 10 CFR 50.54, 10 CFR 73.55 and 10 CFR 73.56.

Discussion The proposed changes to St. Lucie Unit 1 and Unit 2 Technical Specification Sections 6.5.1.6i, 6.5.1.6j, 6.8.ld, and 6.8.le are discussed below:

1. TS 6.5.1.6 i: Delete the following wording from the TS on Facility Review Group responsibilities:

"Review of the Security Plan and implementing procedures and submittal of recommended changes to the Chairman of the Company Nuclear Review Board" Justification In accordance with Generic Letter 93-07, FPL proposes to remove the review of the Security Plan and implementing procedures from the list of responsibilities of the Facility Review Group. 10 CFR 73 includes provisions that are sufficient to address these requirements. Generic Letter 93-07 provides this line-item improvement, provided FPL relocates the review requirements to the Security Plan. Upon approval by the NRC of these proposed amendments, FPL will revise the Security Plan to include these review requirements.

S ... '. D I <<I @II i Bi .I I 1'S F F14'R i G responsibilities:

"Review of the Emergency Plan and implementing procedures and submittal of recommended changes to the Chairman of the Company Nuclear Review Board" Justification: In accordance'with Generic Letter 93-07, FPL proposes to remove the review of the Emergency Plan and implementing procedures from the list of responsibilities of the Facility Review Group. 10 CFR 50 includes provisions that are sufficient to address these requirements. Generic Letter 93-07 provides this line-item improvement, provided FPL relocates the review requirements to the Emergency Plan.

Upon approval by the NRC of these proposed amendments, FPL will revise the Emergency Plan to include these review requirements.

3. TS 6 8 1 d and 6 8 1 e 'elete the following statements from the TS on Procedures and Programs:

d. Security Plan implementation; Emergency Plan implementation; Justification; In accordance with Generic Letter 93-07, FPL proposes to remove the requirements for the Facility Review Group to review procedures and procedure changes for the implementation of the Emergency Plan and Security Plan. 10 CFR 50 and 10 CFR 73 include provisions that are sufficient to address these requirements. Generic Letter 93-07 provides this line-item improvement, provided FPL relocates the review requirements to the Emergency and Security Plans. Upon approval by the NRC of these proposed amendments, FPL will revise the Emergency and Security Plans to include these review requirements.

Summaru FPL proposes to revise TS 6.5.1.6 i and j and 6.8.1 d and e, in accordance with Generic Letter 93-07. Generic Letter 93-07, "Modifications of the Technical Specifications Administrative Control Requirements for Emergency and Security Plans," issued December 28, 1993, provided guidance for changes to Technical Specifications to remove the audit of the emergency and security plans and implementing procedures from the list of responsibilities of the company nuclear audit and review group. The basis of this change is that Parts 50 and 73 of Title 10 of the Code of Federal Regulations include provisions that are sufficient to address these requirements. Also, the generic letter provides guidance to allow TS changes to remove (1) the review of the emergency and security plans from the list of responsibilities of the unit review group and (2) the requirements for the unit review group to review procedures and procedure changes for the implementation of the Emergency Plan and Security Plan, provided the licensee relocates these requirements to the respective plan. FPL commits to revise the Emergency Plan and Security Plan to include these requirements upon approval by the NRC of these proposed amendments.

The following pages have been deleted or modified to preserve the continuity of pagination.

The changes to these pages are administrative in nature.

UNIT I UNIT 2 TS PAGES TS PAGES 3/4 3-27 3/4 3-32 3/4 3-45 3/4 3-44 3/4 3-48 3/4 3-56

St. Lucie Units 1 and 2 Docket No. 50-335 and 50-389 Proposed License Amendments Relocation of Selected Technical S ecification Re uirements Related to Instrumentation and NRC Generic Letter 93-07 Line-Item Technical S ecification Im rovement ATTACHMENT3 DETERMINATIONOF NO SIGNIFICANT HAZARDS CONSIDERATION The standards used to arrive at a determination that a request for amendment involves a no significant hazards consideration are included in the Commission's regulation, 10 CFR 50.92.

10 CFR 50.92 states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Each standard is discussed as follows:

(1) Operation of the facility in accordance with the proposed amendments would not involve a significant incrr.ase in the probability or consequences of an accident previously evaluated.

The proposed changes to the Selected Technical Specification Requirements Related to Instrumentation are administrative in nature in that the specifications for operation and surveillance of the selected Technical Specification instrumentation will be relocated from Appendix A of the facility operating license to the Updated Final Safety Analysis Report (UFSAR) for each unit. Once relocated, future changes will be controlled by 10 CFR 50.59 and the UFSARs updated pursuant to 10 CFR 50.71(e). Relocation of these requirements to the UFSAR is consistent with the NRC "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" published in the Federal Register (58 FR 39132) dated July 22, 1993.

The selected Technical Specification instruments are not accident initiators nor a part of the success path(s) which function to mitigate accidents evaluated in the plant safety analyses. The proposed Technical Specification change does not involve any change to the configuration or method of operation of any plant equipment that is used to mitigate the consequences of an accident, nor do the changes alter any assumptions or conditions in any of the plant accident analyses. Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The Technical Specifications changes associated with Emergency Plan and Security Plan requirements are proposed in accordance with Generic Letter 93-07. The changes being proposed are administrative in nature and do not affect assumptions contained in plant safety analyses, the physical design and/or operation of the plant, nor do they affect Technical Specifications that preserve safety analysis assumptions. Therefore, operation

of the facility in accordance with the proposed amendments would not affect the probability or consequences of an accident previously analyzed.

(2) Use of the modified specification would not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed amendment to relocate the existing Technical Specification requirements for selected Technical Specification instrumentation to the UFSAR will not change the physical plant or the modes of plant operation defined in the Facility License. The change does not involve the addition or modification of equipment nor does it alter the design or operation of plant systems. Therefore, operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed amendments, in accordance with Generic Letter 93-07, change the Technical Specifications to remove the audit of the emergency and security plans and implementing procedures from the list of responsibilities of the Facility Review Group.

The changes being proposed are administrative in nature and will not change the physical plant or the modes of operation defined in the Facility License. The change does not involve the addition or modification of equipment nor does it alter the design or operation of plant systems. Therefore, operation of the facility in accordance with the proposed amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Use of the modified specification would not involve a significant reduction in a margin of safety.

The proposed changes are administrative in nature in that operating and surveillance requirements for the selected Technical Specification instrumentation will be relocated from Appendix A of the facility license to the appropriate Updated Final Safety Analysis Report for each unit. These selected instruments are not used to actuate safety-related equipment, provide interlocks, or otherwise perform plant control functions. Conditions evaluated in plant accident and transient analyses do not involve these selected instruments. The proposed changes do not alter the basis for any technical specification that is related to the establishment of, or the maintenance of, a nuclear safety margin.

Therefore, operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The proposed amendments, in accordance with Generic Letter 93-07, change the Technical Specifications to remove the audit of the emergency and security plans and implementing procedures from the list of responsibilities of the Facility Review Group.

The changes being proposed are administrative in nature and do not alter the bases for assurance that safety-related activities are performed correctly or the basis for any Technical Specification that is related to the establishment of or maintenance of a safety

margin. Therefore, operation of the facility in accordance with the proposed amendments would not involve a significant reduction in a margin of safety.

Based on the above, we have determined that the proposed amendments do not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the probability of a new or different kind of accident from any previously evaluated, or (3) 'involve a significant reduction in a margin of safety; and therefore do not involve a significant hazards consideration.