ML17219A453

Proposed Tech Specs,Removing Outdated Matl,Making Minor Text Changes & Correcting Typos

ML17219A453
Person / Time
Site:	Saint Lucie
Issue date:	03/17/1987
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17219A452	List: ML17219A451 ML17219A453
References
NUDOCS 8703250096
Download: ML17219A453 (27)

Text

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 BORAT ION CONTROL SHUTDOWN MARGIN - T > 200'F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > ~~9+>>.

APPLICABILITY: MODES 1, 2*, 3 and 4. 34,60 pe~

ACTION:

With the SHUTDOWN MARGIN (~g-al+>>, imediately initiate and continue boration at > 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > ~~@f4:

a~ Within one hour after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoperable.

If the inoperable CEA is imovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the iamovable or un-trippable CEA(s).

b. When in MODES 1 or 2 , at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that CEA group withdrawal is within 'the Power Dependent Insertion Limits of Specification 3. 1.3.6. /

c When least in once MODE 2, at least once during IIII per hour thereafter until the CEA withdrawal and at reactor, is critical.

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d. Prior to initial operation above SX RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the CEA groups at the Power Dependent Insertion Limits of Specification 3.1.3.6.

870~2 5009b 870317 See Special Test Exception 3.10.1. DOCK AD 05000gg5 PDR PDR With K > 1.0.

W)th K ff eff < 1.0. P'/4 ST. LUCIE - UNIT 1 1-1 Amendment No. 27. 9g, ~3

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REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued

e. When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by con.=

sideration of the following factors:

1. Reactor coolant. system boron concentration,

2. CEA position,*

3; Reactor coolant system average temperature,

4. Fuel burnup based on gross thermal energy eneration,

5. Xenon concentration, and f044

6. Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be comp red to predicted values to demonstrate agreement within + ~~A+ at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.l.l.l.l.e.

above. The predicted reactivity values shall be adjusted (narmalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after .each fuel loading.

• For Modes 3 and 4, during calculation of shutdown margin with all C@'s verified'ully inserted, the single CEA with the highest reactivity worth need not be assumed to be stuck in the fully withdrawn position.

ST. LUCIE - UNIT 1 3/4 1-2 Amendment No. 45

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T ( 2OOoF LIMITING CONDITION FOR OPERATION 3.1.1,2 The SHUTDOWN MARGIN shall be:

>~~~, and in addition with the Reactor below the hot leg centerline, one charging Coolant System drained pump shall be rendered inoperable.*

APPLICABILITY: MODE 5.

ACTION:

If the SHUTDOWN MARGIN requirements cannot be met, immediately initiate and continue boration at > 40 gpm of 1720 ppm boron or equivalent until the required SHUTDOWN MARGIN is restor ed.

SURVEILLANCE REOUI REMENTS 4.1.1.2 The SHUTDOWN MARGIN requirements of Specification 3.1.1.2 shall be determined:

a. Within one hour after detection of an inoperable CEA(s) and at 1 ast once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the CEA(s) is inoperable.

I the inoperable CEA is immovable or untrippable, the above r=quired SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdraw'n worth of the immovable or untrippable CEA(s).

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following factors:

l. Reactor coolant system boron concentration,

2. CD, position,

3. Reactor coolant system average temperature, 4, Fuel burnup based on gross thermal energy generation, 5, Xenon concentration. and

6. Samarium concentration.

c. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, when the Reactor 'Coolant System is drained below the..hot leg centerline, by consid'er ation of the.

factors in 4.1;1:.2.b and by verifying at least one char ging pump is rendered inoperable.*

• Breaker racked-out.

ST. LUCIE - UNIT 1 3/4 1-3 Amendment No. < 8

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION e

3.1.1.4 The moderator temper ature coefficient (MTC) shall be:

a. Less positive than

'7pcN is < 70% of RATED THERMAL POWER, whenever THERMAL POWER

b. Less positive than whenever THERMAL POWER

+2 ta Cssi is > 70K of RATED THERMAL POWER, and

-g,/pete f c. Less negative than APPLICABILITY: MODES 1 and 2*II at RATED THERMAL POWER.

ACTION:

With the moderator temperature coefficient outside any one of the above limits, be in HOT STANDBY within 6 hours.

SURVEILLANCE RE UIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

• With K ~ leOa PSee Special Test Exception 3.10.2.

ST. LUCIE - UNIT 1 3/4 1-5 Amendment No. gV, 63

REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONOITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths and one associated heat tracing circuit shall be OPERABLE:

a. Two flow paths from the boric acid makeup tanks via either a boric acid pump or a gravity feed connection, and a charging pump to the Reactor Coolant System, and

b. The flow path from the refueling water tank via a 'charging pump to the Reactor Coolant System.

APPLICABILITY: MOOES.1, 2, 3 and 4.

ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least.two boron i.njection flow paths to the Reactor Coolant System to OPERABLE status within 72

'ours or make the reactor'subcritical within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and borate to a SHUTDOWN MARGIN equivalent to 'at least at 2OO F; restore at least two flow paths to OPERABLE status within he next 7 days or be in COLO SHUTOOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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SURYEILLAHCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days 'by:

1. Cycling each testable power operated or automatic valve in th'e flow oath through at least one complete cycle of full travel.

ST. LUC IE - UNIT 1 3/4 1-10 Amendment Ho. g S

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 At least two of the following three borated water sources shall be OPERABLE:

a. Two boric acid makeup tanks and one associated heat tracing circuit with the contents of the tanks in accordance with Figure 3.1-1, and

b. The refueling water tank with:

l. A minimum contained volume of 401,800 gallons of water,

2. A minimum boron concentration of 1720 ppm,

3. A maximum solution temperature of 100'F,

4. A minimum solution temperature of 55'F when in MODES 1 and 2, and

5. A minimum sol'ution temperature of 40'F when in MODES 3 and 4,

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one borated water source OPERABLE, restore at least two borated water sources 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 make the reactor subcritical within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and borate to a SHUTDOWN MARGIN equivalent to at least at 200'F; restore at least two borated water sources to OPERABLE status within the next 7 .days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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~ ~ ~ 4 SURVEILLANCE RE UIREHENTS 4.1.2.8 At least two borated water..sources shall be demonstrated OPERABLE:

a. At least once per 7-:days by:

l. Verifying the boron concentration in each water source, ST. LUCIE - UNIT 1 3/4 1-'18 Amendment No. PP, g 8

INSTRUMENTATION INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The incore detection system shall be OPERABLE with:

a. At least 75% of all incore detector locations, and

b. 'A minimum of two quadrant symmetric incore detector locations per core quadrant.

An OPERABLE incore detector location shall consist of a fuel assembly contain-ing a fixed detector string with a minimum of three OPERABLE rhodium detectors.

APPLICABILITY: When the incore detection system is used for:

a. Recalibration of the excore axial flux offset detection system,

b. Monitoring the AZIMUTHAL POWER TILT,

c. Calibration of the power level neutron flux channels, or

d. Monitoring the linear heat rate.

ACTION:

'I With the incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of 3.0.3 and 3.0.4 are not applicable. 'pecifications SURVEILLANCE RE UIREHENTS 4.3.3.2 The incore detection system shall be demonstrated OPERABLE:

a. By performance of a CHANNEL CHECK within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to.its use and at least once per 7 days thereafter when required for:

1. Recalibration of the excore axial flux offset detection system,

2. Monitoring the linear heat rate pursuant to Specification

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4.2.1;3,

• Until- tober 1, 1 the in e detect system s 1 'be OPER E with:

a. Rt 1 50% of a incore d ctor loca ons, and

b. inimum of ree quadr symmetri incor e det ors for least thre evels.

. ST LUCIE - UNIT 1 3/4 3-25 Amendment No. 41

Thill.E 4.3-9 IIRD IOAC1 I VE GASEOIJS L'I I I.ttl:t(I tIDtl I lOll IIIG IHS I I(till(.HIRf ION SIJIIVEI I.I.AHCE Itf~(JlllEIIEHTS CIIAHNLI. MODI.S I II lllllCll CI IANHI:.L SOI) IICL CIIRNI'lLI. I UHC I I OHAL SilltVL I I.

-CALIllltAIION I'I'.ST Itl:Olt lltl.l)

LANCL'III'.CK I t(STIIIIIII.NT CIII.CK

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a. Nntil e Gas hc L i v i Ly Moni Loi It(3) q(2)

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li. In~line Salllt)1 el II tI. A. H.A.

2 I'arLic>>laLe Sai!!I)1er M N. h. H.h. H. h.

il. Sauil)ler f lou lt'ilLe Mo>>i Lor H. A.

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

APPLICABILITY: NODES 1, 2, and 3.

ACTION:

with one or more block valve(s) noperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve(s) to OPERABLE stat s or close the block valve(s) and remove power from the block valve(s)g otherwise, be'in at least HOT STAtlDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDO>lN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREHENTS 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.

ntil 0 ober 1, 81, in 1 eu of cl ing and oving p er to th lock valve -1403, t PORV, V 402, ma e deener ized in e closed osition suc that it i incapab of bein opened.

dnlafa gwHMe ST LUCIE - UNIT 1 3/4 4-58 Amendment No. P7, 42

TABLE 4.7.-2 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMEHT MINIMUM AND ANALYSIS FRf(RUEHCY

l. Gross Activity Determination 3 times per 7 days with a of maximum'fFile 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples

2. Isotopic Analysis for DOSE a) 1 per 31 days, when-E(UIYALEHT I-131 Concentration eve" the gross activity d rmi iodine concentrations greater than 10:l of the allowable limit.

b) 1 per 6 months, whenever

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the, gross activity deter-mination indicates iodine concentrations below 10.l of th allowable limit.

ST. LUCIE UNIT 1 3/4 7-8

PLANT SYSTKl1S SECONDARY 1/ATER CHEMISTRY LIHITIHG CONDITION FOR 0. ERATIOH 3.7.1.6 The secondary water chemistry shall be m tained wi'thin the limits of Table 3.7-3 by use of All 'Jolatile Treatment VT).

APPLICABILITY: 1100ES 1, 2 and 3.

ACTION'.

(To be determined in a manner se forth in the bases in approximately six months and to be imposed b a change to this Specification.)

SUR'lEILLANCE . EOUIR":i1EHTS 4.7. The secondary water chemistry shall be determined to be within th imits by analysis of'those parameters at the frequencies specified Table 4.7-3.

Q tS I'AgE MFT IAT aAWioAAu't BLA.AK.

ST. LUG IE - UNIT 1 3/4 7-10

TABLE 3.7-3 SECONDARY HATER C ISTRY LIMITS Hater-S p1e Lo lon arameters*

R8f4< Kc t<w~ on @~< f'~J ~

TH <s JAKE LsF TlQ'TadnadAU-Y I<A<<

TABLE 4.7-3 SECONDARY WATER C}IEMISTRY SURVEIL E RE UIREMENTS n

m Water Sam Loca n arameter s*

ge ~ lace 'f4 fag f'y~ ls papaw ws'4 %- Loo45

3 4.1 REACTIYITY CONIBOL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 8444 pea 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that ) the reactor can be made, from all operating conditions, 2) e reactivity transients 'ubcritical associated with postulated accident condition are controllable within acceptable limits, and 3) the reactor will maintained sufficiently subcritical to preclude inadvertent critic ity in the shutdown condition.

SHUTDOWN MARGIN requirements vary t oughout core life as a function of fuel depletion, RCS boron concentrat n, and RCS Tavg The most restrictive condition occurs at EOL, w h Tavo at no load operating temperature, and is associated with a ostuTaCed steam line break accident and resulting uncontrolled RCS coold . In the analysis of this accident, a minimum SHUTDOWN MARGIN of . is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN required by Specification 3. 1. 1.1 is based upon this limiting condi tion and is con-sistent with FSAR accident analysis assumptions. For earlier peri OOOO during the fuel cycle, this value is conservative. With T 00'F the reactivity transient resulting from a boron on even a partially drained Reactor Coolant equires a HUTDOWN MARGIN .and rest> ictions on g pump operation to provide adequate protection. A TDOWN MARGIN is conservative for Mode 5 operation with total RCS volume present, however LCO 3.1.1.2 is written conservatively for simplicity. ~

' 'cTfj 3/4.1.1.3 BORON DILUTION AND ADDITION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that changes will be gradual during boron concentration changes reactivity in the Reactor Coolant System.

A flow rate of at least 3000 GPM will circulate an equivalent Reactor

'will be within the cap 'ator Coolant System volume of 11,400 cubic feet in approximately The reactivity 'change rate associated with boron

+7P~/'F concentration

~/oF 26 minutes.

changes control.

3/4.1.1. 4 MODERATOR TEMP E COEFFICIENT

+2 ga, p essa d

f The limiting values ass d for the MTC used in the a dent and transient analyses were . fo

< 70K of RATED THERMAL POWER, HERMAL OWER levels I Tevels > 70K of RATED THERMAL for ERMAL'OWER and . at RATED THERMAL POWER. Therefore, these limiting values are included in this specification.

Determination of MTC at the specified conditions ensures that the maximum positive and/or negative values of the MTC will not exceed the limiting values.

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REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1,5 MINIMUM TEMPERATURE FOR CRITICALITY The MTC is=- expected to be slightly negative at operating conditions.

However, at the beginning of the fuel cycle, the MTC may be slightly positive at operating conditions and since it will become more positive at, lower temperatures, this specification is provided to restrict reactor operation when T is significantly below the normal operating temperature.'/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during .each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200'F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall racility sarety from in'ection system failures gaaaata during the repair period.

Tne boration capability of either system is sufficient rovide a SHUTOGWN MARGIN from all operating conditions of ter xenon decay and cooldown to 200'F. The maximum boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 7,925 gallons of 8.0/ boric acid solution from the boric acid tanks or 13,700 gallons of 1720 ppm borated water from the refueling water tank.

The requirements for a minimum contained volume of 401,800 gallons of borated water in the refueling water tank ensures the capability for borating the. RCS to the desired level. The specified quantity of borated water is consistent with the ECCS requirements of Specificagion 3.5.4; .Therefore, the larger volume of borated water is specified here '.'oo

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With the RCS temperatu're below 200'F, one injection system is acceptable without single failure consideration on the basis of the stable reacti vi ty condi ti on of the reactor and the additional .restric-tions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

ST. LUCIE - UNIT 1 B 3/4 1-2 Amendment No. 27, +, < '

PLANT SYSTEMS BASES 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALYES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within contain-ment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 SECONOARY WATER CHEMISTRY A test pro am will be conducted during proximately the first 6 months of oll>> tion after initial criticali to establish the appropriat limits .on t secondary water chemistry p amet rs and to determine the appropri frequencies for monitoring ese parameters. The results f this t lir'ogram will be submitted the Commission for review. e Coom sion will then issue a revis to this specification speci ing th imits on t6t.: chemistry para ers and the frequencies for nitoring ese parameters.

The test program will elude an a'nalysis of the che. cal con-stitutent oi'he makeup w er for the St. Lucie Plant. e analysis shall identify the vari s traces of ions which upon ncentration may have the potential fo inducement for stress corros' in the steam generator tubing. e test program shall also ev uate the efficiency of the water trea ent systems in the St. Lucie acility for removal of such ions and t potential for addition of o er ions resulting from the treatment ethod. The test program sha analyze concentration phenomena a the concentration rates in e steam generator and th secondary ater system and shall consid concentration in the re culat-ing coo ng water system.

3/4.7.2 STl:AM GENERATOR PRESSURE/TEMPERATURE LI lITATION

", The l imitation on steam generator pressure and temperature ensures that the pri:ssure induced stresses in the steam generators do not exceed the maximi;m allowable fracture toughness stress limits. The limitations of 70'f and 200-psig are based on a steam generator RTNDT of 50'F and are sufficient to prevent brittle fracture.

ST. LUCIF - UflIT 1 B 3/4 7-3

ADMINISTRAT I VE CONTROLS 6.5.2 COMPANY NUCLEAR REVIEW BOARD CHRB FUiiCTIOl(

6.5.2.1 The Company Nuclear Review Board shall function to provide indepen-dent review and audit of designated activities in the areas of:

a. nuclear power plant operations

b. nuclear engineering C. chemistry and radiochemistry

d. metallurgy XNs'el.~ &vis4 2 e.

f.

instrumentation and control radiological safety

[sk. sec Acbcf

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g. mechanical and electrical engineering

h. quality assurance practices COMPOS IT ION 6.5.2.2 The CNRB shall be composed of the following members:

Vice President,.Advanced Systems and Tec Member: 'ef Engineer, Power Plant En 9 Member: Group 'esident, nergy Member: Vice Presiden, r Operations Member:

Member:

Member er:

Director Ma

. a i ty

, Nuclear Fuel ower Plant Engineering Principal e

Power Plant Engineering Senior Project

'r Ma The Chairman shall be a member of the CNRB and shall be designated in writing.

ALTERNATES'.5.2.3 All alternate members shall be appointed in writing by the CNRB Chair-man to serve on a temporary basis; however, no more than two alternates shall participate as voting members in CNRB activities at any one time ST. LUCIE - UNIT 1 6-9 Amendment No. 75,$ $ ,69

Member: Group Vice President Member: Group Vice President - Nuclear Energy Member: Vice President - Engineering, Projects &,

Construction Member: Vice President Nuclear Operations Member: Director - Nuclear Licensing Member: Director Quality Assurance Member: Chief Engineer - Power Plant Engineering Member: Manager Nuclear Energy Services Member: Manager Nuclear Fuel Member: Senior Project Manager - Power .Plant Engineering

ADMINISTRATIVE CONTROLS (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for all- off-control point chemistry conditions, and (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.

d. Backu Method for Determinin Subcoolin Mar in A program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:

(i) Training of personnel, and (ii) Procedures for monitoring.

e. Post-accident Sam lin A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:

(i) Training of personnel, (ii) Procedures for sampling and analysis, and (iii) Provisions for maintenance of sampling and analysis equipment.

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6. 9 REPORTING RE UI REMENTS lo CFN 6'o.k ROUTINE REPORTS'.9.1 In addition to the applicable reporting requirements of Title 10, Code f F d 1 g g 1 1  ; 1 f 11 1 g p 1 11 P 1 1 << d 1 the NRC g

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase'n power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear.

thermal, or hydraulic performance of the plant.

ST. LUCIE - UNIT 1 6-15 Amendment No. 69

ADMINISTRATIVE CONTROLS 6,9.1.2 The star tup report shall address each of the tests identified in FSAR and shall include a description of the measured values of the operating

t rnnditions or characteristics obtained during the test program and a compari-son of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details'equired in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of cormercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does test not cover all three events (i.e., initial criticality, completion of startup program, and resumption or commencement of comnercial operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS 1/

6.9.1.4 Annual reports covering the activities of the unit as descr'ibed below

.for the previous calendar year shall be submitted prior to March 1 of each :.

year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

6.9.1.5 Reports required on an annual basis sha11 include a tabulation on an annual basis of the number of station,'tility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr an) their assoc'iated man-rem exposure according to work and job functions, reactor operations and surveillance, inservice inspection, routine '.g.,

maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not'e accounted for, In the aggregate, at least SO% of the total whole body dose received from external sources should be assign OPERATING REPORTS delcfienfo ceriForw 4 lo CFR5o.f,'ONTHLY 6,9.1.6 Routine reports of operating statistics and shu';oo::.". e::ooriencc, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the NRC, no later'han the 15th of each month following the calendar month covered by the report.

sing e su m>ttal may be made for a multiple unit station. The submittal should combine those sections that are common to all units af the station.

2/ This tabulation supplements the requirements of 520.407.of 10 CFR Part '20.

ST. LUCI E - UNIT 1 6-16 Amendment No. 28,69

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Jo(cia ou4ddd ADMINISTRATIVE CONTROLS

~~t MWenafS 6.13 PROCESS CONTROL PROGRAM PCP Licensee initiated changes to the PCP:

1. Shall be submitted to the Comnission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemerital information.

b. A determination that the change did not reduce the overall conformance of the dewatered bead resin to existing criteria for radioactive wastes; and

', c. Documentation of the fact that the change has been reviewed and found acceptable by the FRG.

2. Shall become effective upon review and acceptance by the FRG.

6.14 OFFSITE DOSE CALCULATION MANUAL ODCM

~4rQ Licensee initiated changes to the ODCM:

1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made effective. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses .

or evaluations justifying the change(s);

b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determina-tions; and

c. Documentation of the fact that the change has beep reviewed and found acceptable by the FRG.

2. Shall become effective upon review and acceptance by the FRG.

ST. LUCIE - UNIT 1 6-23 Amendment No. $ 9~6

ATTACHMENT 2 SAFETY EVALUATION INTRODUCTION The proposed amendment represents a broad administrative update of the St. Lucie Unit 1 Technical Specifications. The 21 affected pages can be put into five categories:

1 ~ Change terminology from 'elta-k/k" to "pcm" (8 pages).

2 ~ Remove outdated material (7 pages).

3 ~ Correct typographical errors (2 pages).

4 ~ CNRB update to reflect current organization (1 page).

5. Change addressee for certain reports to reflect revisions to 10 CFR 50.4 (3 pages).

DISCUSSION Cate or 1 The following eight pages pertaining to Technical Specification Section 3/4.1, "Reactivity Control Systems," are revised to change units of "Delta-k/k" to the equivalent "pcm:"

3/4 1-1 3/4 1-10 3/4 1-2 3/4 1-18 3/4 1-3 B 3/4 1-1 3/4 1-5 B 3/4 1-2 Cate or 2 The following seven pages are revised to remove requirements that are outdated. These requirements were either effective for a period of time that is now past, or they have become redundant to requirements in another section of the Technical Specifications:

3/4 3-25 3/4 7-12 3/4 4-58 B 3/4 7-3 3/4 7-10 6-23 3/4 7-11

Cate or 3 The following two pages are revised to correct typographical errors:

3/4 3-54 3/4 7-8 Cate or 4 Page 6-9 is revised to expand the composition of the Company Nuclear Review Board (CNRB) from eight to ten people, and to revise position titles to conform with the current, organization.

Cate or 5 A recent change to 10 CFR 50.4, effective January 5, 1987, directs that Part 50 reports be addressed to the Document Control Desk, Washington, DC, 20555. The following three pages are revised accordingly:

6-15 6-16 6-19

ATTACHMENT3 DETERMlNATION OF NO SIGNlFICANT HAZARDS CONSlDERATION The standards used to arrive at a determination that a request for amendment involves no significant hazards consideration are included in the Commission's regulations, 10 CFR 50.92, which states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (I) 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:

(I) 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 changes being proposed are administrative; they do not affect assumptions contained in plant safety analyses, nor do they affect Technical Specifications that preserve safety analysis assumptions. Therefore, the proposed changes do not affect the probability or consequences of accidents previously analyzed.

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

The changes being proposed by FPL are administrative; they will not lead to material procedure changes or to physical modifications. Therefore, the proposed changes do not create the possibility of a new or different kind of accident.

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

The changes being proposed by FPL are administrative; they do not relate to or modify the safety margins defined in and maintained by the Technical Specifications. Therefore, the proposed changes should not involve any reduction in a margin of safety.

Based on the above, we have determined that the amendment request does not (I) 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 accident previously evaluated, or (3) involve a significant reduction in a margin of safety; and therefore does not involve a significant hazards consideration.

E J W4/022/2

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