Text

Amends 126 & 120 to Licenses NPF-2 & NPF-8,respectively, Revising Listed TSs & Associated Bases to Remove Certain cycle-specific Parameter Limits from TSs & Relocating to COLR
	ML20137E262
Person / Time
Site:	Farley
Issue date:	03/25/1997
From:	Berkow H; NRC (Affiliation Not Assigned)
To:	;
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ML20137E266	List: ML20137E262;
References
	NUDOCS 9703270129
	Download: ML20137E262 (54)
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UNITED STATES s

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666 4 001 i

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SOUTHERN NUCLEAR OPERATING COMPANY. INC.

AL6JAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.126 License No. NPF-2 1.

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

I A.

The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated September 30, 1996, complies with the standards and requirements of the Atomic Energy 3

Act of 1954, as amended (the Act), and the Commission's rules and j

regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations-l D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance witn 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

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

i NPF-2 is hereby amended to read as follows.

1 9703270129 970325 PDR ADOCK 05000348 i

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(2)

Technical Specifications The Technical Specifications contained in Appendices A and i

B, as revised through Amendment No.126, are hereby 1

incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall j

be implemented prior to entry into Mode 5 following the next scheduled refueling outage, which should begin in March 1997.

FOR THE NUCLEAR REGULATORY COMMISSION i

Her ert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical l

Specifications Date of Issuance: March 25, 1997 l

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l i

ATTACHMENT TO LICENSE AMENDMENT NO.126 TO FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Egmpve Paaes Insert Paael I

I XIX XIX l-2 1-2 3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-3 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 1-22 3/4 1-22 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 i

3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 6-19 6-19 6-19a

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1 IEXX 1

i j

1 1

1 DEFINITIONS

\\

SECTION g

1 1.0 DEFINITIONS i

1.1 ACTION........................................................ 1-1 1.2 AXIAL FLUX DIFFERENCE......................................... 1-1 a

j 1.3 CHANNEL CALIERATION........................................... 1-1 l

1.4 CHANNEL CHECK................................................. 1-1 1.5 CHANNEL FUNCTION TE5T.........................................

1-1 1.6 OONTAINMENT INTEGRITY........................................

1-2 l

1.7 CONTROLLED LEAKAGE............................................ 1-2

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1.8 CORE ALTERATION............................................... 1-2 1.se CORE OPEmATING LIMITS REPOAT.................................

1-2 l

1.9 DOSE EQUIVALENT I-131........................................

1-2 i

1.10 T-AVERAGE DISINTEGRATION ENERGY...............................

1-3 j

1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME...................... 1-3 1.12 FREQUENCY NOTATION............................................ 1-3 i

1.13 (Deleted 1-3 IDENTIFIED LEARAGE...............i........o)..................

1.14 1-3 1.15 (Deleted).................... 1-4 1.16 MheeR-3 (Deleted). 1-4 1.17 OFFSITE DOSE CALCULATION MANUAL (ODCM)........................

1-4 1.18 OPERAELE - OPERASILITY.......................................

1-4 1.19 OPERATIONAL MODE - MODE....................................... 1-5 1.20 PHYSICS TESTS...f............................................

1-5 1.21' PRESSURE SOUNDARY LEAKAGE..................................... 1-5 l

1.22 PROCESS CONTROL PROGRAM (PCP)................................. 1-5 l

1.23 PURGE - PURGING............................................... 1-5 i

1.24 QUADRANT POWER TILT RATIO..................................... 1-5 1.25 RATED THERMAL POWER........................................... 1-6 1

1.26 REACTOR TRIP SYSTEM RESPONSE TIME.............................

1-6 1.27 RE PORTABLE EV1 TNT.............................................. 1-6

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1.28 SHUTDOWN MARG1N...............................................

1-6 1.29 SetTDTFTCRTTON-(Deleted)...................................... 1-6 l

1.30 SOURCE CuRCK.................................................. 1-6 1.31 STAGGERED TEST BASIS.........................................

1-6 1.32 THERMAL POWER................................................

1-7 l

1.33 UNIDENTIFIED LEAKAGE.........................................

1-7 j

1.34 VENTILATION REMAUST TREATMENT SYSTEM..........................

1-7 1.35 VENTING......................................................

1-7 1

4 TABLE 1.1 OPERATIONAL MODES 1-8 J

TABLE 1.2 FREQUENCY NOTATION 1-9 1

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i FARLEY-UNIT 1 I

AMENDMENT NO. 126 4

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ADMINISTRATIVE CONTROLS 1

i sEcTION 2AE j

Review...................................................

6-10

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J Audits...................................................

6-11 Authority................................................

6-12 i

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Records...............................................'...

6-12 1

i 6.5.3 TECHNICAL REVIEW AND CONTROL Activities...............................................

6-12 a

g Records..................................................

6-13 i

6.6 masonTante EVENT ACTION.....................................

6-14 i

L7 a uRTY LYMTT vrotatioN

............5........................

6-14 i'

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6.8 PROCEDURES.AND PROGRAM 5.....................................

6-14 6.9 REPORTING REOUIREMENTS 4

6.9.1 ROUTINE REPORTS

j startup Report.......................................... 6-15a 1

Annual Report............................................

6-16 j

1 Annual Radiological Environmental Operating Report.......

6-17 i

j Annual Radioactive Effluent Release Report...............

6-17 1

i 1

i Monthly Operating Report.................................

6-19 Core Operating Limits Report.............................

6-19 l

Annual Diesel Generator Reliability Data Report..........

6-19a l

Annual Reactor Coolant System specific Activity Report... 6-20 Annual sealed source Leakage Report......................

6-20 6.9.2 SPECIAL REPORTS...........................................

6-20 6.10 RECORD RETENTION............................................ 6-20 6.11 RADIATION PROTECTION PROGRAM.......................

........ 6-21a 6.12 HIGH RADIATION AREA........................................

6-22 FARLEY-UNIT 1 XIX AMENDMENT mo.126

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' DEFINITIONS J

j CONTAINMENT INTEGRITY l

4 1.6 CONTAIMMENT INTEGRITY shall exist whens l

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All penetrations required to be closed during accident conditions are a.

either:

1)

Capable of being closed by an OPERABLE containment automatic i

j isolation valve system, or 2) closed by manual valves, blind flanges or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3, f

b.

All equipment hatches are closed and sealed, 4

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Each air lock is OPERABLE pursuant to specification 3.6.1.3, c.

1 i

j d.

The containment leakage rates are within the limits of specification l

3.6.1.2, and i

The sealing mechanism associated with each penetration (e.g., welds, j

o.

I bellows or 0-rings) le OPERABLE.

CONTROLLED LEAKAGE 4

1 1.7 CONTROLLED f* "_^0* shall be that seal water flow supplied to the reactor j

coolant pump seals.

CORI ALTERATION 1.8 CORE ALTERATION shall be the movement or manipulation of any component i

within the reactor pressure vessel with the vessel head removed and fuel in the vessel. suspension of CORE ALTERATION shall not preclude completion of movement j

of a component to a safe conservative position.

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l CORE OPERATING LIMITS REPORT 1

j 1.Sa The OORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in j

accordance with specification 6.9.1.11.

Unit operation within these operating i

l limits is addressed in individual specifications.

DOSE EQUIVALIET I-131 1

1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 I

l (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be s

those listed in Table E-7 of Regulatory Guide 1.109, Revision 1, 1977.

4 1

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FARLEY-UNIT 1 1-2 AMENDMENT NO.126

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'3/4.1 REACTIVITY CONTROL SYSTEMS i

3/4.1.1 BORATION CONTROL 1

1 SHUTDOWN MARGIN - Tava >200'F LIMITING CONDITION FOR OPERATION i

3.1.1.1 The SHUTDOFM MARGIN shall be greater than or equal to the limit specified in the OOLR for 3 loop operation.

l 5

APPLICABILI?Y:

MODES 1, 2*,

3, and 4.

ACTION 1 i

j With the SHUTDOWN MARGIN less than the limit specified in the COLR, launediately l

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initiate and continue boration at greater than or equal to 30 gpa of a solution j

containing greater than or equal to 7000 ppe boron or equivalent until the required SHUTDOWN MARGIN is restored.

l 1

SURVEILLANCE REQUIREMENTS i

j 4.1.1.1.1 The SHUTDOWN MARGIN shall be' determined to be greater than or equal to the limit specified in the.COLRs 3

1 Within one hour after detection of an inoperable control rod (s) and a.

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is imunovable or untrippable, the above

{

required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the imeovable or untrippable control rod (s).

j b.

When in MODE 1 or MODE 2 with K,gg greater than or equal to 1.0, i

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank position j

is within the insertion limits of Specification 3.1.3.6.

1'

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c.

When in MODE 2 with K gf less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to l

achieving reactor criticality by verifying that the predicted critical j

control rod position is within the limits of Specification 3.1.3.6.

4 d.

Prior to initial operation above 5% RATED THERMAL POWER after each j

fuel loading, by consideration of the factors of a below, with the j

control banks at the maximum insertion limit of Specification 3.1.3.6.

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"See Special Test Exception 3.10.1.

4 FARLI,Y-UNIT 1 3/4 1-1 AMENDMENT NO.126 y

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T s 200*F ava i

LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater.than or equal to the limit specified in the COLR.

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APPLICABILITY:

MODE 5.

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ACTIDM:

4 l

i With the SHUTDOWN MARGIN less than the limit specified in the COLR, immediately l

initiste and continue boration at greater than or equal to 30 gym of a solution l

containing greater than or equal to 7000 ppe boron or equivalent until the i

required SHUTDOWN MARGIN is restored.

I SURVEILLANCE REQUIREMENTS The SHUTDOWN MARGIN shall bh determiind to be greater than or equal 4.1.1.2 to the limit specified in the COLR Nithin one hour after detection of an inoperable control rod (s) and a.

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter whils the rod (s) is inoperable.

If the inophrable control rod is iemovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowanca for the withdrawn worth of the imeovable or untrippable control rod (s).

b.

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

1.

Reactor coolant system boron concentration, 2.

Control rod position, 3.

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

c.

FARLEY-UNIT 1 3/4 1-3 AMENDMENT No.126

REACTIVITY CONTROL SYSTEMS

)

l MODERATOR TEMPERATUB4 COEFFICIENT I

LIMITING CONDITION FOR OPERATION i

3.1.1.3 The moderator temperature coefficient (MTC) shall be within the j

beginning of cycle life (SOL) limit and the end of cycle life (BOL) limit specified 1n the COLR. The maximum upper limit shall be less than or equal to i

O.7 x 10"4 delta k/k/*F for power levels up to 70% THERMAL POWER with a linear j

ramp to O delta k/k/*F at 100% THERMAL POWER.

1 I

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l APPLICABILITY:

BOL limit - MODES 1 and 2* only#.

]

EOL limit - MODES 1, 2 and 3 only#.

t M Hz With the MTC more positive than the BoL limit specified in the COLR, l a.

operation in MODES 1 and 2 may proceed provided l

1.

Control rod withdrawal limits are established and maintained I

sufficient to restore the MTC to within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits

)

of specification 3.1.3.6.

2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

t 3.

A Special Report is prepared and submitted to the Commission pursuant to specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control red withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all j

rods withdrawn condition.

4 4

b.

With the MTC more negative than the EOL limit specified in the COLR, l be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i

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With K,gg greater than or equal to 1.0.

See special Test Exception 3.10.3.

FARLEY-UNIT 1 3/4 1-4 AMENDMENT No.126

l REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC snall be determined to be within its limits during each j

fuel cycle as follows:

i The MTC shall be measured and compared to the BOL limit specified a.

in the COLR, prior to initial operation above 5% of RATED THERMAL 3

POWER, after each fuel loading.

b.

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppe MTC surveillance limit specified in the COLR l

within 7 EFPD after reaching an equilibrium boron concentration of 300 ppe.

In the event this comparison indicates the MTC is more negative than the 300 ppe MTC surveillance limit specified in the j

COLR, the MTC shall be remeasured, and compared to the BOL MTC i

limit specified in the COLR, at least once per 14 EFPD during the remainder of the fuel cycle.

(1)

(1) once the equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) is 100 ppa or less, further measurement of the MTC in accordance with 4.1.1.3.b may be suspended, providing that the measured MTC at an equilibrium boron concentration less than or equal to 100 ppa is less negative than the 100 ppe MTC surveillance limit specified in the COLR.

l FARLEY-UNIT 1 3/4 1-5 AMENDMENT NO.126

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l REACTIVITY *hMTROL SYSTEMS 4

SHUTDOWN RCD INSERTIOff LIMIT 1

P 1

LIMITING CONDITION FOR OPERATION i

3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the COLR.

APPLICABILITY:

MODES 1* and 2*#.

ACTION:

With a maximum of one shutdown rod inserted beyond the insertion l'ait specified in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1 :, within one hour either Restore the rod to within the insertion limit specified in the a

COLR, or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specifiod in the,COLR:

Within 15 minutes prior to withdrawal of any rods in control banks a.

A, B, C or D during an approach to reactor criticality, and i

b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

1 4

See Special Test Exceptions 3.10.2 and 3.10.3.

with Twgg greater than or equal to 1.0.

FARLEY-UMI' 1 3/4 1-20 AMENDMENT NO.126

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UACTIVITY CONTROL SYSTEMS CGuinGL ROD INSERTION LIMITS 1

LIMITING CONDITION FOR OPERATION i

3.1.3.6 The control banks shall be limited in physical insertion as specified in the COLR.

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APPLICABILITY:

MODES 1* and 2*#.

4 ACTION:

4 l

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to specification 4.1.3.1.2, either i

Restore the control banks to within the limits within two hours, or a.

2 b.

Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position j

using the insertion limits specified in the COLR, or l

Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

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SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control tank shall be determined to be within i

the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals

)

within the Rod Insertion Limit Monitor is inoperable, then verify the individual l

rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

I j

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See special Test Exceptions 3.10.2 and 3.10.3.

With K,gg greater than or equal to 1.0.

9 FARLEY-UNIT 1 3/4 1-21 AMENDMENT NO.126

FIGURE 3.1-1 1

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(This Figure intentionally left blank.)

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FARLEY-UNIT 1 3/4 1-22 AMENDMENT No.126

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l 3/4.2 POWER DISTRIBUTION LIMTTs 3/4.2.1 arm. FLUX DIFFERENCE fAFD) i LIMITING CONDITION FOR OPERATION i

3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within i

the limits specified in the COLR*.

l i

APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER **.

l ACTION:

With the indicated AXIAL FLUX DIFFERENCE outside of the limits a.

specified in the COLR l

4 l

1.

Either restore the indicated AFD to within the limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes.

SURVIILLANCE REQUIREMENTS 4.2.1 The indicated ARIAL FLUX DIFFERENCE shall be determined to be within its limits by:

Monitoring the indicated AFD for each OPERABLE encore channel:

a.

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour with the AFD Monitor Alaza inoperabis.

The indicated AFD shall be considered outside of its limits when at least 2 OPERArtE excore channels are indicating the AFD to be outside its limits.

- See Special Test Exception 3.10.2.

FARLEY-UNIT 1 3/4 2-1 AMENDMENT NO.126

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FARLEY-UNIT 1 3/4 2-2 AMENDMENT No.126 l

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POWER DisTmIBUTION LIMITS 3/4.2.2 NEAT FLUY MOT CHANNEL FACTOR - FgfZ)

LIMITING OONDITION FOR OPERATION 3.2.2 Fg(E) shall be within the limits specified in the OOLR.

l

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j APPLICABILITY:

MODE 1.

I J

ACTION:

i With F (E) exceeding its limits g

t j

Reduce THERMAL POWER at least 14 for each 14 Fg(E) exceeds the a.

limit within 15 minutes and similarly reduce the Power Range j

Neutron Flux-High Trip setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed pr;vided the Overpower delta T Trip 8etpoints have been reduced at least 1% for each 14 Fg(E) exceeds the limit.

b.

THERMAL POWER may be increased provided Fg(E) is demonstrated through incore mapping to be within its limit.

e SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of specification 4.0.4 are not applicable.

4.2.2.2 Fg(E) shall be evaluated to determine if it is within its limit bys Using the movable incore detectors to obtain a power distribution a.

map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b FARLEY-UNIT 1 3/4 2-3 AMENDMINT NO. 126

. - _ ~. - ~.... _ - - - -. - _ - - -.... _ -... -. - - -. -.

t POWER DISTRIBUTION LIMITS l

SURVEII. LANCE REgUIREMENTS (Centinued) 4 j

b.

Determining the computed heat flux hot channel factor Fg (g),

C j

follows:

l Increase the measured Fg(I) obtained from the power distribution may by 34 to account for manufacturing tolerances and further l

increase the value by 5% to account for measurement uncertainties.

i l

C Verifying that Fg (I), obtained in specification 4.2.2.2b above, c.

i satisfies the relationship in specification 3.2.2.

i j

d.

satisfying the following relationships 4

F "" x K (Z ) fo r P j

F e (Z ) 5 o

> 0.5 a

y x W (Z)

F "" x K (Z ) fo r P 5 i

F e (Z ) s O

0.5 o

0.5 x W (Z)

I C

Where Fg (E) is obtained in specification 4.2.2.2b above, Fg is the Fg limit, K(3) is the normalised Fg(I) as a function of core height, F I's the fraction of RATED THERMAL POWER, and W(I) is the j

cycle dependent function that accounts for power distribution i

transients encountered during normal operation.

L l

Fg

, K(E), and W(1) are specified in the COLR as per j

specification 6.2.1.11.

l e.

Measuring Fg(E) according to the following schedules i

i 1.

Upon achieving equilibrium conditions after eaceeding by 20%

i or more of RATED THERMAL POWER, the THERMAL POWER at which l

Fg(E) wee last determined *, or 2.

At l' east once per 31 Effactive Full Power Days, whichever occurs first.

4 I

i

During power escalation after each fuel loading, power level may be t

increased until equilibrium conditions at any power level greater than or i

equal to 50% of RATED THERMAL POWER have been achieved and a power distribution map obtained.

l 1

i FARLEY-UNIT 1 3/4 2-4 AMENDMENT NO.I26 s

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POWER DISTRIBUTION 12MITS 1

SURVIII.I.ANCE REQUIREMENTS (Continued) f.

With met.surements indicating j

' F c(7)w o

s maximum j

over(Z)

K(Z)j s

has increased since the previous determination of Fg (E) either of C

the following actions shall be taken:

I C

1)

Increase Fg (E) by the Fg (E) penalty factor specified in the COI.R and verify that this value satisfies the relationship in l Specification 4.2.2.2d, or 2)

Fg (1) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that 1

'F C(Z)'

n maximum is not increasing.

K(Z).j over(Z)

(

g.

With the relationships specified in specification 4.2.2.2d above j

not being satisfied:

1) calcylate the percent Fg(E) exceeds its limits by the following expression:

1 Fj (Z) x W (Z) 4 m axim um i

1 x 100 for P > 0.5 over Z F,1, 4

o P

(

J Fj(Z) x W (Z) m axim um 1

x 100 for P s 0.5, and over Z F,,,

q

.\\

. 0.5

)

4 l

2)

The following action shall be taken:

Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the COLR by 1% AFD for each percent Fg(E) exceeds its limits as determined in specification 4.2.2.2g.1.

Within c hours, reset the AFD alarm setpoints to these modified limits.

t FARLEY-UNIT 1 3/4 2-5 AMENDMENT NO.126

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FARI2Y-UNIT 1 3/4 2-7 NDMENT NO.126

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. POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEARENTHALPYHOTCHANNELFACTOR-Ph

)

LIMITING CONDITION FOR OPERATION t

1 l

3.2.3 Ph shall be within the limits specified in the COLR.

l 1

i

)

i j

APPLICABILITY:

MODE 1.

l l

ACTIQHz i

l With Ph exceeding its limits a.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

j 1.

Restore Fh to within the above limit, and demonstrate through in-core mapping that Fh is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of exceeding lthe limity or i

2.

Reduce THERMAL POWER to less than 50%'of RATED THERMAL POWER 4

and reduce the Power Range Neutron Flux - High Trip setpoints to 5 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and l

}

b.

Demonstrate through in-core mapping, if not previously performed j

per a.1 above, that Ph is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and Identify and correct the cause of the out of limit condition prior c.

to increasing THERMAL POWER above the reduced ilmit required by a or b, above; subsequent POWER OPERATION may proceed provided that Fh is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

FARLEY-UNIT 1 3/4 2-8 AMENDMENT No.126

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a l

'3.4.1 REACTIVITY CONTROL SYSTEMB aAssa 3/4.1.1 N TION CONTROL 1

f 3/4.1.1.1 na 3/4.1.1.2 SNUTDOWN MARcIN i

l A sufficient SHUTDOWN MARCIN ensures that 1) the reactor can be made l

suberitical from all operating conditions, 2) the reactivity transients j

associated with postulated accident conditions are controllable within i

acceptable limits, and 3) the reactor will be maintained sufficiently i

j suberitical to preclude inadvertent criticality in the shutdown condition.

4 5

SNUTDOWN MARGIN requirements vary throughout core life as a function of

{

fuel depletion, RCS boron concentration, and RCS T.,g.

The most restrictive condition occurs at BOL, with T.,9 at no load operating i

l temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN as specified in the COLR is required to control the l

l reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis l

assumptions. With T less than 200'F, the reactivity transients resulting avg from a postulated steam line break cooldown are minimal and a lower SEUTDOWN t

l MARGIN (specified in the COLR) provides adequate protection.

{

I l

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT I

The limitations on moderator temperature coefficient (MTC) are provided to-ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

l The MTC values of this specification are applicable to a specific set of

}

plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those

)

conditions a,n order to permit an accurate comparison.

i" l

The most negative MTC value equivalent to the most positive moderator j

density coefficierpt (MDC) was obtained by incrementally correcting the MDC j

used in the FSAR analyses to nominal operating conditions. These l

corrections involved (1) a conversion of the MDC used in the FSAR safety j

analyses to its equivalent MTC, based on the rate of change _of moderator l

density with temperature at RATED THERMAL POWER conditions, and (2) 1 subtracting from this value the largest difforences in MTC observed l

between EOL, all rods withdrawn, RATED THERMAL POWER conditions, and those j

most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and menon concentration that can occur in d

normal operation and lead to a significantly more negative EOL MTC at RATED THERMAL POWER. These corrections transformed the MDC value used in j

the FSAR safety analyses into the limiting MTC value specified in the COLR.

The j

surveillance requirement MTC value specified in the COLR vapresents a conservative MTC value at a core condition of 300 ppe equilibrium boron concentration, and is obtained by making corrections for burnup and soluble j

boron to the limiting MTC value specified in the COLR.

l l

4 FARLEY-UNIT 1 B 3/4 1-1 AMENDMENT NO. 126

-m.___

i l

REACTIVITY CONTROZ, SYSTEMS l

mASEs 1

i MODERATOR TEMPERATUR4 COEFFICIENT fContinued) once the equilibrium boron concentration falls below 100 ppe, MTC

{

measurements may be suspended provided the measured MTC value at an i

equilibrium boron concentration s 100 pp;a is less negative than the 100 ppe MTC i

surveillance limit specified in the ccLR. The difference between this value j

and the limiting EOL MTC value conservatively bounds the maximum change in MTC between the 100 ppm equilibrium boron concentration (all rods withdrawn, RATED j

THERMAL POWER condition) and the licensed end-of-cycle, including the effects of j

boron concentration reduction, fuel depletion, and end-of-cycle coastdown.

The surveillance requirements for measurement of the MTC at the beginning

}

and near the end of the fuel cycle are adequate to confirm that the MTC j

remains within its limits since this coefficient changes slowly due i

principally to the reduction in RCs boron concentration associated with i

fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE POR CRITICAI.I]

t l

This specification ensures that the reactor will not be made critical with l

the Reactor Coolant system average temperature less than 541*F.

This j

limitation is required to ensure 1) the moderator temperature coefficient j

is within its analysed temperature range, 2) the protective instrumentation j

is within its normal operating range, 3) the F-12 interlock is above its l

setpoint, 4) the pressuriser is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor prassure vessel is above its minimum i

RTNDT temperature.

I 3/4.1.2 BORATION SYSTEMS l

The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required i

to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) ht.;ic acid transfer pumps, and 5) an j

emergency power supply from OPERABLE diesel generators.

i With the RCs average temperature above 200*F, a minimum of two boron i

injection flow paths are required to ensure single functional capability in i

the event an assumed failure renders one of the flow paths inoperable. The l

boration capability of either flow path is sufficient to provide the required SHUTDOWN MARGIN from expected operating conditions after menon decay and t

l cooldown to 200'F.

The maximum expected boration capability requirement occurs i

at BOL from full power equilibrium menon conditions and requires 11,336 gallons of 7000 ppe borated water from the boric acid storage tanks or 44,826 gallons of 2300 ppe borated water from the refueling water storage tank.

I i

4

=

1 FARLEY-UNIT 1 B 3/4 1-2 AMENDMENT NO.126 4

l

. REACT 2VITY CONTROL SYSTEMS l

BASES BQEATIDN SYSTEMS (Continued) t i

filth the RCs temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection l

system becomes inoperable.

}

l The limitation for a maximum of one centrifugal charging pump to be l

OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OFERABLE pump to be inoperable below 180*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single RER relief valve.

il J

j The boron capability required below 200*F is sufficient to provide a l

EMUTDOstN MARGIN as specified in the COLR after menon decay and cooldown from-l l

200*F to 140*F.

This condition requires either 2,000 gallons of 7000 ppe borated water from the boric acid storage tanks or 7,750 gallons of 2300 ppe

{

borated water from the refueling water storage tank.

I i

The contained water volume limits include allowance for water not available because of discharge line location and other physical 4

I characteristics.

j The limits on coEtained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated l

within containment after a LOCA. This pH band minimises the evolution of

]

iodine and minimises the offeet of chloride and caustic stress corrosion on d

mechanical systems and components.

i The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

j 3/4.1.3 MOvantes u w Assaynt.fas l

The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SEUTDOWN MARGIN is maintained, and (3) limit the potential affects of rod misalignment on j

associated accident analyses. OPERASILITY of the control rod position j

indicators is required to determine control rod positions and thereby ensure l

compliance with the conted rod alignment and insertion limits.

4 a

i 4

i FARLEY-UNIT 1 3 3/4 1-3 AMENDMENT No. 126 d

N d

j

'3/4.2 power DIsTRIaUTION L2MITS BASES i

The specifications of this section provide assurance of fuel integrity 1

during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) meeting the DNE design criterion during normal l

operation and in short term transients, and (b) limiting the fission gas 3

release, fuel pellet temperature and cladding mechanical properties to I

within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance j

criteria limit of 2200*r is not exceeded.

1 j

1 4

The definitions of certain hot channel and peaking factors as used in j

these specifications are as follows:

i

)

Fg(E)

Heat Flux Hot Channel Factor, is defined as the maximum local

{

heat flux on the surface of a fuel rod at core elevation E I

divided by the average fuel rod heat flux, allowing for

{

l manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

1 f

Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the f

i ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

I 3/4.2.1 AIT M. FLUI DIFFERENCE i

The limits on AIIAL FLUX DIFFERENCE (AFD) assure that the Fg(E) upper bound envelope of the Fg limit specified in the COLR times K(E) is not l

l exceeded during either normal operation or in the event of menon redistribution following power changes.

i Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OFERAELE escore detector j

outputs and provides an alarm message immediately if the AFD for 2 or more q

OPERAELE encore channels is outside the allowed AI operating space for RAOC d

operation specified in the COLR and the THERMAL POWER is greater than l

50% RATED THERMAL POWER.

t i

I FARLEY-UNIT 1 3 3/4 2-1 AMENDMENT No. 126

i i

. POWER DISTRIBUTION LIMITS j

l BASES t

{

3/4.2.2 and 3/4.2.3 MRaT FLUX HOT CHANNEL FACTOR. nut *fWaR EninALPY HOT CHANNEL FACTOR The limits on heat flux hot channel factor, and nuclear enthalpy rise hot

}

channel factor ensure that 1) the design limit on peak local power density is j

not exceeded, 2) the DNR design criterion is met, and 3) in the event of a LOCA j

the peak fuel clad temperature will not exceed the 2200*F ECCs j

acceptance criteria limit.

I Each of these is measurable but will normally only be determined i

periodically as specified in specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

Control rods in a single group move together with no individual a.

rod insertion differing by more than

12 steps, indicated, from the group demand position.

4 b.

Control rod banks are sequenced with overlapping groups as described in specification 3.1.3.6.

~

The control rod insertion limits of specifications 3.1.3.5 and c.

3.1.3.6 are maintained.

d.

The axial power distribution, expressed in terms rf AIIAL FLUX e

DIFFERENCE is maintained within the limits.

7 Fh will be maintained within its limits provided conditions a. through

d. above are maintained. The relaxation of Fh as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance taust be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.

The heat flux hot channel factor Fg(E) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(I), to provide assurance that the limit on the heat flux hot channel factor Fg(I) is set. W(E) Accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

The W(E) function for normal operation and the AFD limits are provided in the COLR per specification 6.9.1.11.

FARLEY-UNIT 1 5 3/4 242 AMENDMENT NO.126

i 3

' ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORT l

I 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall j

be submitted on a monthly basis to the Commission, pursuant to 10 CFR 50.4, no later than the 15th of mach month following the calendar month

)

covered by the report.

t 4

I CORI OPERATING LIMITS REPORT I

6.9.1.11 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of j

a reload cycle for the following:

1 1.

SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for Specification 3/4.1.1.1, 5

2.

SHUTDOWN MARGIN limit for MODE 5 for specification 3/4.1.1.2,

)

3.

Moderator Temperature Coeffici.ent BOL and ROL limits and 300 ppe and l

100 ppa surveillance limits for specification 3/4.1.1.3,

]

4.

shutdown Bank Insertion Limit for specification 3/4.1.3.5, A

5.

Control Bank Insertion Limits for specification 3/4.1.3.6, 6.

Axial Flux Diffegence limits for Specification 3/4.2.1, TP 7.

Heat Flux Hot Channel Factor Fg limits, K(I) figure, W(E) values, and l

Fg(E) Penalty Factors for Specification 3/4.2.2, j

8.

Nuclear Enthalpy Rise Hot Channel Factor limits, Fag

, and Power l

Factor Multipiter, PFag, for specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be

{

those previously reviewed and approved by NRC ins S

j 1.

WCAP-9272-P A, " Westinghouse Reload Safety Evaluation 1

Methodology," July 1985 (M Proprietary).

1 (Methodology for Specifications 3.1.1.1 - Shutdown Margin - Tavg > 200'F, 4

3.1.1.2 - Shutdown Margin - Tavg 4 200*F, 3.1.1.3 - CNorator Temperature Coefficient, 3.1.3.5 - shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank l

Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot channel

{

Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

i 1

i d

j FARLEY-UNIT 1 6-19 AMENDMENT No. 126 2

ADMINISTRATIVE CX)NTROLS 4

2.

WCAP-10216-P-A, Rev. lA, " Relaxation Of Constant Axial Offset Control /

Fg Surveillance Technical Specification," February 1994 (H Proprietary).

i (Nethodology for specifications 3.2.1 - Axial" Flux Difference and 3.2.2 - Beat Flux Bot Channel Factor.)

l 3.

WCAP-10266-P-A, Rev. 2, "The 1981 Version Of Westinghouse Evaluation Model Using BASH Code," March 1987 (M Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Bot Channel i

Factor.)

i The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS i

limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

1 i

The OORE OPERATING LIMITS REPORT, including any'mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to i

the NRC Document Control Desk with copies to the Regional Administrator and j

Resident Inspector.

ANNUAL DIESEL GENERATOR RELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually.

This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.

O FARLEY-UNIT 1 6-19a AMENDMENT No.126 l

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1 UNITED STATES l

g

,j NUCLEAR REGULATORY COMMISSION t

WASHINGTON, D.C. 30eeHWO1

\\...../

4 SOUTHERN NUCLEAR OPERATING COMPANY. INC.

ALABAMA POWER COMPANY DOCKET NO. 50-364 i

JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE i

j Amendment No. 120 License No. NPF-8 1.

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

A.

The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated September 30, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the i

Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be i

conducted in compliance with the Comission's regulations; D.

The issuance of this license amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-8 is hereby amended to read as follows:

I 4

j l

].

(2)

Technical Specifications The Technical Specifications contained in Appendices A and i

B, as revised through Amendment No.120, are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical j

Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented prior to entry into Mode 5 following the refueling outage

.i scheduled to begin in March 1998.

FOR THE NUCLEAR REGULATORY COMMISSION i

e rbert N. Berkow, Director roject Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation j

Attachment:

Changes to the Technical j

Specifications Date of Issuance:

fiarch 25, 1997 5

i s

t i

4 4

i i

4 ATTACHMENT TO LICENSE AMENDMENT NO.120 TO FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Insert I

I XIX XIX l-2 1-2 3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-3 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 1-22 3/4 1-22 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 B 3/4 1-1 B 3/4 1-1 B 3/4 1-2

.B 3/4 1-2 B 3/4 1-3 8 3/4 1-3 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 6-19 6-19 6-19a

.. ~.

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1 I

l 4

i t

DEFINITIONS SECTIDM 1

PAgg

{

1.0 DEFIhlTIONS 1.1 ACTION...............................................

1-1 1.2 AIIAL FLUX DIFFERENCE.........................................

1-1 1.3 CHANNEL CALIBRATION........................................... 1-1 1.4 CHANNEL CHECK.................................................

CHANNEL FUNCTION T587........................................

1-1 1.5 OONTAINMENT INTEGRITY........................................

1-1 1.6 CONTROLLED LEAKAGE...........................................

1-2 1.7 1-2 1.8 OORE ALTERATION......................................

........ 1-2 1.Sa CORE OPERATING LIMITS REPORT.................................

1-2 l

1.9 DOSE EQUIVALENT I-131........................................

1-2 1.10 T-AV2 RAGE DISINTEGRATION ENERGY..............................

1-3 1.11 ENGINERAED SAFETY FEATURES RESPONSE TIME..............

1 FREQUENCY NOTATILN...........................................

1-3 1.12

........ 1-3 1.13 (Deleted................

1-3 IDENTIFIED LEARAGE................*........)...................

1.14 1-3 1.15 (Deleted)....................

1-4 1.16

"'ar ""* """? O "'2 M'""M "'fr """'OCMD -SUSWENS (Deleted ) 1-4 1.17 OFFSITE DOSE CALCULATION MANUAL (ODCM)......................

OPERAELE - OPERABILITY.......................................

1-4 1.18 1-4 1.19 OPERATIONAL MODE - M005....................................... 1-5 1.20 PNYSICS TESTS...p............................................. 1-5 1.21 PRESSURE BOUNDARY LEAKAGE..................................... 1-5 1.22 PROCESS CONTROL PROGRAM (PCP)...............................

PURGE - PURGING..............................................

1-5

. 23 1-5 1.24 QUADRANT PONER TILT RATIO..................................... 1-5 1.25 RATED THERMAL PONER..........................................

1-6 1.26 REACTOR TRIP SYSTEM R3SPONSE TIME............................

1-6 1.27 REPORTABLE EVENT.............................................

1-6 1.28 SHUTDONN MARGIN..............................................

1-6 1.29 SeE,5MMeheteN ( De leted )...................................... 1-6 1.30 SOURCE CERCK.................................................. 1-6 1.31 STAGGERED TEST BASIS.........................................

1-6 1.32 THERMAL PONER................................................

1-7 1.33 UNIDENTIFIED LEAKAGE.........................................

1-7 1.34 VENTILATION RERAUST TREATMENT SYSTEM.......................... 1-7 1.35 VENTING.......................................................

1-7 TABLE 1.1 OPERATIONAL MODES 1-8 TABLE 1.2 FREQUENCY NDYATION 1-9 FARLEY-UNIT 2 I

ANENDMENT NO. 120 e

ADMINISTRATIVE CONTROLS

+

i 4

a ARCIIEE FKs1 Review...................................................

6-10 Audits...................................................

6-11 Authority................................................

6-12 i

Records.................................................,

6_12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities...............................................

6-12 1

Records..................................................

6-13 i

6.6 REPORTAnfE EVENT ACTION.....................................

6-14 1

4 6.7 mm*ETr t rutt vim mTION..................................... 6-14 6 B_2RQCEDllBER_AIGLIRQGEAME..................................... 6-14 a

6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS f i

Startup Report.......................................... 6-15a 1

Annual Report............................................

6-16 I

Annual Radiological Environmental Operating Report.......

6-17 4

Annual Radioactive Effluent Release Report...............

6-17 Monthly Operating Report.................................

6-19 Core Operating Limits Report.............................

6-19 l

Annual Diesel Generator Reliability Data Report..........

6-19a l

A Annual Reactor Coolant System Specific Activity Report... 6-20 Annual Scaled Source Leakage Report......................

6-20 6.9.2 SPECIAL REPORTS...........................................

6-20 6.10 RECORD RETENTION...........................................

6-20 6.11 RADIATION PROM CTION PROGRAM................................ 6-21a

)

J i

6.12 HIGN RADIATION AREA........................................

6-22 FARLEY-UNIT 2 XIX AMEND 3 TENT NO.120 e

_. _ ~

.OEFINITIows CONTAIMMENT INTEGRITY j

1.6 CONTAINMENT INTEGRITY shall suist when All penetrations required to be closed during accident conditions a.

are eLthers 1)

Capable of being closed by an OPERAELE containment automatic isolation valve system, or 2) closed by manual valves, blind flanges or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of specification 3.6.3, l

l b.

All equi,ument hatches are closed and sealed, Each air lock is OPERAELE pursuant to Specification 3.6.1.3, c.

d.

The containment leakage rates are within the limits of specification 3.6.1.2, and The sealing mechanism associated with each penetration (e.g.,

e.

i l

welds, bellows or 0-rings) is OPERAELE.

l l

CONTROLLED LEAEAGE l

1.7 CONTROLLED LEAEAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.8 CORE ALTERATION s' hall be the movement or manipulation of any component l

within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.Sa The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current reload cycle. These cycle-specific core operating limits shall be determined for each reload cycis in accordance with specification 6.9.1.11.

Unit operation within these operating limits is addressed in individual specifications.

DOSE EOUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid d ue as the quantity and isotopic mixture of I-131, I-132, I-133, 1-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of Regulatory Guide 1.109, Ravinion 1, 1977.

i i

,e FARLEY-UNIT 2 1-2 AMENDNENT NO.120

l i

3/4.1 M'"ffVITY hmGL SYSTitus 3/4.1.1 meinavraag u__i.st j

SNUTDOWN MARGIN - Tava >200'F i

LIMITING OONDITION POR OPERATION 1

3.1.1.1 The SNUTDOWN MARGIN shall be greater than or equal to the limit specified in the OOLR for 3 loop operation.

1 APPLICABILITY:

MODES 1, 2*,

3, and 4.

ACTION:

I WiththeSNUTDOWNMARGINlessthanthelimitspecifiedintheOOLR,immediatelyl initiate and continue boration at greater than or equal to 30 ggun of a solution containing greater than or equal to 7000 ppe boron or equivalent until the required SNUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or i

equal to the limit specified in the OOLRs 3

J Within one hour after detection of an inoperable control rod (s) and a.

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or j

untrippable, the above required SNUTDOWN MARGIN shall be verified l

acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s).

b.

When in MODE 1 or MODE 2 with K,gg greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank position is

}

within the insertion limits of specification 3.1.3.6.

}

c.

When in MODE 2 with K gg less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to 1

achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of specification

?

3.1.3.6.

d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of a below, with the control banks at the maximum insertion limit of specification 3.1.3.6.

see special Test Exception 3.10.1.

t FARLEY-UNIT 2 3/4 1-1 AMENDMENT NO.120

~ _ _

REACTIVITY CONTROL SYSTUL4 SNUTDOWN MARGIN - Tava 5 200'F LIMITING CONDITION POR OPERATION 3.1.1.2 The SNUTDOWN MARGIN shall be greater than or equal to the limit specified in the COLR.

f l

l l

APPLICABILITY:

MODE 5.

1 ACTION:

With the SHUTDOWN MARGIN less than the limit specified in the COLR, immediatelyl initiate and continue boration at greater than or equal to 30 gpo of a solution l

containing greater than or equal to 7000 ppe boron or equivalent until the required SMUTDONM MARGIN is restored.

{

SURVEILLANCE REQUIREMENTS g

.. _........ _. _ _ _ _ _. _ _ _ _. _ _ _ _ _.. _...... _ _ _.. _._...mmag. m 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to the limit specified in the COLRs l

Within one hour after detection of an inoperable control rod (s) a.

and at least onqn per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is l

inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with i

an increased allowance for the withdrawn worth of the immovable or l

untrippable control rod (s).

b.

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

1.

Reactor coolant system boron concentration, 2.

Control rod position, 3.

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

i I

1 1

i FARLEY-UNIT 2 3/4 1-3 AMENDMENT NO. 120

1 REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT l

LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the i

beginning of cycle life (BOL) limit and the end of cycle life (BOL) limit i

specified in the COLR. The maximum upper limit shall be less than or equal to

{

~

0.7 x 10 delta k/k/*F for power levels up to 70% THERMAL POWER with a linear j

ramp to 0 delta k/k/*F at 2004 THERMAL POWER.

i APPLICABILITY:

BOL limit - NODES 1 and 2* only#.

ROL limit - NODES 1, 2 and 3 only#.

&GIIDE8 With the MTC more positive than the BOL limit specified in the COLR, l

a.

operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within its limit within 24 i

hours or be in HOY STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertion limits of specification 3.1.3.6.

2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3.

A special Report is prepared and submitted to the commission pursuant to specification 6.9.2 within 10 days, describing the value of the measured NTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b.

With the NTC more :W ative than the BOL limit specified in the COLR, he in NOT SBUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With K,gg greater than or equal to 1.0.

see special Test Exception 3.10.3.

FARLEY-UNIT 2 3/4 1-4

$NENDMENT NO.120

REACTIVITY CONTROL SYSTEMS 1

SURVIILLANCE REQUIREMENTS l

4 4.1.1.3 The MTC shall be determined to be within its limits during each fu61 cycle as follows:

The MTC shall be measured and compared to the BOL limit specified a.

in the COLR, prior to initial operation above 5% of RATED THERMAL POtfER, after each fuel loading.

1 b.

The MTC shall be measured at any TERRMAL PotrER and compared to the 300 ppe MTC surveillance limit specified in the COLR l

within 7 EFFD after reaching an equilibrium boron concentration of 300 ppm.

In the event this comparison indicatas the MTC is more negative than the 300 ppe MTC surveillance limit specified in the COLR, the MTC shall be rameasured, and compared to the BCL MTC limit specified in the COLR, at least once per 14 EFFD during the 5

l remainder cf the fuel cycle.

(1) y

,I 4

E d

i (1)

Once the equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) is 100 ppe or less, further measurement of the MTC in accordance wita 4.1.1.3.b may be suspended, providing that the measured MTC at an equilibrium boron concentration less than or equal to 100 ppe is less negative than the 100 ppm MTC surveillance limit specified in the COLR.

l i

a f*

9 FARLEY-UNIT 2 3/4 1-5 AMENDMENT NO.120

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION EIMIT i

LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the COLR.

l APPLICABILITY:

N00ES 1* and 2*#.

ACIlGE8 With a maximum of one shutdown rod inserted beyond the insertion limit specified in the COLR, except for surveillance testing pursuant to l

Specification 4.1.3.1.2,'within one hour either:

i 1

Restore the rod to within the insertion limit specified in the I

a l

00LR, or l

l b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

l SURVEILLANCE REQUIREMENTS

~

l 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the coLR:

j l

a.

Within 15 minutes prior to withdrawal of any rods in control banks A, E, C or D during an approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

l i

1

See Special Test Exceptions 3.10.2 and 3.10.3.

With Keff greater than or equal to 1.0.

l l

l i

FARLEY-UNIT 2 3/4 1-20 AMENDMENT NO.120

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTIOl( LIMITS LIMITING OONDITION FOR OPERATION j

1 3.1.3.6 The control banks shall be limited in physical insertion as specified in the (X)LR.

l l

APPLICABILITY:

MODES la and 2*#.

M1125:

with the control banks inserted beyond the above insertion limits, except-for surveillance testing pursuant to specification 4.1.3.1.2, either:

Restore the control banks to within the limits within two hours, or a.

b.

Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the insertion limits specified in the OOLR, or l

c.

Be in at least NOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 2

4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals within the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1 l

I

see special Test Exceptions 3.10.2 and 3.10.3.

With K,gg greater than or equal to 1.0.

1 FARLEY-UNIT 2 3/4 1-21 AMENDMENT NO.120

I e

e FICURE 3.1 3 1

l I

1 i

i I

(This Figure intentionally left blank.)

l 1

1

]

e l

v' 8

FARLEY-UNIT 2 3/4 1-22 AMENDMENT No.120

1 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 aYT af. FI.UX DIFFERENCE (AFD1 LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AIIAL FLUX DIFFERENCF (AFD) shall be maintained within the limits specified in the COLR*.

l APPLICAEILITY:

MODE 1 above 50% of RATED THERMAL POWER **.

l ACTION:

With the indicated AEIAL FLUX DIFFERENCE outside of the limits a.

specified in the COLR*

l 1.

Either restore the indicated AFD to within the limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THEAMAL POWER within 30 minutes.

SURVIILLANCE REQUIRE"INTS 4.2.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits by:

t Monitoring the indicated AFD for each OPERAELE escore channels:

a.

1.

At least once per 7 days when the AFD Monitor Alarm is OPERAELE, and 2.

At least once per hour with the AFD Monitor Alarm inoperable.

i i

The indicated AFD shall be considered outside of its limits when at least 2 OPERAELE encore channels are indicating the AFD to be outside its limits.

- see Special Test Exception 3.10.2.

FARLEY-UNIT 2 3/4 2-1

, AMENDMENT NO.120

l

)

1 1

(This page intentionally left blank.)

i 1

1 l

l FARLEY-UNIT 2 3/4 2-2 AMINDMENT No.120 l

~-. - _.......

. ~ -.

J POWER DISTRIBUTION LIMITS 3/4,2.2 atWAT FLUX HOT CHANNEL FACTOR - F;fE1 l

LIMITING CONDITION FOR OPERATION 3.2.2 Fg(E) shall be within the limits specified in the COLR.

l 1

APPLICABILITY:

MODE 1.

ACTION With Fg(E) exceeding its limits j

Reduce THERMAL POWER at least it for each 1% F (I) exceeds the a.

limitwithin15minutesandsimilarlyreducetbPowerRange a

)

Neutron Flux-Migh Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER l

OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed *provided_the Overpower delta T Trip Setpointo have been reduced at least 14 for each 1% F (E) exceeds g

the limit.

I b.

THERMAL POWER may be increased provided F (E) is demonstrated through ingora mapping to be within its 1 it.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F (E) shall be evaluated to determine if it is within its limit by g

a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

FARLEr-UNIT 2 3/4 2-3 AMENDMENT NO. 120

POWER DISTRIBUTION LIMTTS SURVEILLANCE REQUIREMENTS (Continued)

C b.

Determining the computed heat flux hot channel factor Fg {g),,,

fol1-s Increase the measured Fg(E) obtained from the power distribution map by 34 to ace:ount for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties.

Verifying that Fg (E), obtained in Specification 4.2.2.2b above, c.

i satisfies the relationship in specification 3.2.2.

d.

Natisfying the following relationship:

F c (Z ) s F'"" x K (Z) fo r P a

0.5.

P x W (Z) l F c (Z ) s F'*" x K (Z ) fo r ? s 0.5 a

0,5 x W (Z)

C Where Fg (E) is obtained in specification 4.2.2.2b above, Fg is the Fg limit, K(5) is the normalised Fg(E) as a function of core height, F 11s the fraction of RATED THERMAL POWER, and W(I) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

FgRD, K(E), and W(1) are specified in the COLR as per specification 6.9.1.11.

I e.

Measuring Fg(E) according to the following schedules 1.

Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERNAL POWER, the THERMAL POWER at which Fg(E) was last determined *, or 2.

At least once per 31 Effective Full Power Days, whichever occurs first.

l

During power escalation after each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or equal to 50% of RATED THERNAL POWER have been achieved and a power distribution map obtained, i

FARLEY-UNIT 2 3/4 2-4 AMENDMENT NO.120 t

I

POWER DISTRIBUTION LIMITS SURVEILLANCE REgGIREMENTS (Continued) i f.

With measurements indicating

' F, c(Z)'

maximum K(Z)j over(Z) s has increased since the previous determination of Fg (1) either of C

the following actions shall be taken:

C 1)

Increase Fg (E) by the Fg (E) penalty factor specified in the COLR and verify that this value satisfies the relationship in l specification 4.2.2.2d, or C

2)

Fg (E) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that

)

l

\\

' F C(Z)'

o 1

maximum is not incmasing.

K(Z).j over(Z) s g.

With the relationships specified in specification 4.2.2.2d above not being satisfied:

1)

Calcylate the percent Fg(E) exceeds its limits by the following expressions t

'r C

F (Z) x W (Z) o m axim um 1

x 100 for P > 0.5

,1, OV I l

x K(Z)

.t P

.)

y 5

Fc (Z) x W (Z) l o

m axim um 1, x 100 for P s 0.5, and over Z F,1, o

,\\

. 0.5

.s l

2)

The following action shall be taken Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the COLR by 1% AFD for each percent Fg(E) exceeds its limits i

as determined in Specification 4.2.2.2g.1.

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , l

reset the AFD alarm setpoints to these modified limits.

t FARLEY-UNIT 2 3/4 2-5 AMENDMENT NO.120

(This page intentionally left blank)

W W

I I

l FARLEY-UNIT 2 3/4 2-7 fMENDMENTNO.120

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEARENTHALPYNOTCHANNELFACTOR-Fh LIMITIP!G OONDITION FOR OPERATION 3.2.3 Ph shall be within the limits specified in the OOLR.

l j

i l

AEP11 7;1III:

MODE 1.

j ACTION:

With Fh exceeding its limits i

Reduce THERMAL POWER to leser than 50%.cf RATED THERMAL POWER a.

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux - High Trip Setpoints to 5 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

Demonstrate through in-core mapping that Ph is within its limit within 24 hpurs after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and Identify and correct the cause of the out of limit condition prior c.

to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that' P h is demonstrated through in-core mapping to be within its i

limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

4 i

1 e

FARLEY-UNIT 2 3/4 2-8 AMENDMENT NO.120 i

3.4.1

- WIVITY CunikGI, SYSTEMB EASES 3/4.1.1 noRaTION CONTROL 314.1.1.1 a3sn 3 /4.1.1. 2 SNUTDOWN MARGIN A sufficient SNUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within 3

acceptable limits, and 3) the reactor will be maintained sufficiently i

suberitical to preclude inadvertent criticality in the shutdown condition.

SMUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T vg.

The most a

restrictive condition occurs at BOL, with Tavg at no load operating i

temperature, and is associated with a postulated steam line break accic*ent and resulting uncontrolled RCS cooldown. In the analysis of this accideni-a minimum SNUTDOWM MARGIN as specified in the OOLR is required to control tr.e l

reactivity transier.t. Accordingly, the SNUTDOWN MARGIN requirement is based

{

upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T less than 200*F, the reactivity transients resulting ay, from a postulated steam line break cooldown are minimal and a lower SEUTDOWN MARGIN (specified in the OOLR) provides adequate protection.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on modgrator temperature coefficient (MTC) are provided to ensure that the value'of this coefficient remains within the limiting condition j

assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification af MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The most negative MTC value equivalent to the most positive moderator density coefficient (MDC) was obtained by incrementally correcting the MDC used in the PSAR analyses to nominal operating conditions. These corrections inveinds (1) a conversion of the MDC used in the FSAR safety analyses to itre equivalent MTC, based on the rate of change of moderator density with tecperature at RATED THERMAL POWER conditions, and (2) subtracting from thi: value the largest differences in MTC observed between EOL, all rods withdrawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon concentration that can occur in normal operation and lead to a significantly more negative EOL MTC at RATED THERMAL POWER. These corrections transformed the MDC value used in the FSAR safety analyses into the limiting MTC value specified in the COLR. The surveillance requirement MTC value specified in the OOLR represents a conservative MTC value at a core condition of 300 ppe equilibrium boron concentration, and is obtained by making corrections for burnup and soluble boron to the limiting MTC value specified in the COLR.

l i

l FARLEY-UNIT 2 E 3/4 1-1 AMENDMENT NO. 120 l

1

KEACTIVITY CONTROL SYSTEMS BASES MODERATOR T "*ERATURE COEFFICIENT (Continued)

Once the equilibrium boron concentration falls below 100 ppe, MTC measurements may be suspended provided the measured MTC value at an equilibrium boron concentration 5 100 ppe is less negative than the 100 ppe MTC surveillance limit specified in the OOLR.

The difference between this value and the limiting BOL MTC value conservatively bounds the maximum change in MTC between the 100 ppe equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) and the licensed end-of-cycle, including the effects of boron concentration reduction, fuel depletion, and end-of-cycle coastdown.

The surveillance requirements for measurement of the MTC at the beginning and near the end of the fue) ycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant system average temperature less than 541*F.

This

~

limitation is required to ensure 1) the moderator temperature coefficient is within its analysed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint, 4) the pressupiser is capable of being in an OPERABLE status with a steam bubble, and 5)*the reactor pressure vessel is above its minimum RTMDT temperature.

J/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) eeparate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide the required SHUTDOWN MARGIN from expected operating conditions after menon decay and cooldown to 200'F.

The maximum expected boration capability requirement occurs at EOL from full power equilibrium menon conditions and requires 11,336 gallons of 7000 ppe borated water from the boric acid storage tanks or 44,826 gallons of 2300 ppe borated water from the refueling water storage tank.

t FARLEY-UNIT 2 M 3/4 1-2 AMENDMENT NO.120

..- ~. - -

-- -. - ~._... _.__

- - -. - _.... - -.. _ _. -.. - - - - ~..-.

kEACTIVITY CONTROL SYSTEMS a

BASES i

RORATION SYSTEMS fContinued1 With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting OORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumpe except the required OPERABLE pump to be inoperable below 180'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single RHR relief valve.

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN as specified in the COLR after menon decay and cooldown from l

200'F to 140*F.

This condition requires either 2,000 gallons of 7000 pga borated water from the boric acid storage tanks or 7,750 gallons of 2300 ppe borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWCT also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a IDCA.

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

The OPERASILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVaefM Cu i GI. Asanwattas The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

(

I l

FARLEY-UNIT 2 B 3/4 1-3 AMENDMINT NO.120 l

l

3/4.2 POWER DISTRIBUTION LIMITS EASTS The specifications of this section provide assurance of fuel integrity during condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a). meeting the DNS design criterion during norcal operation and l

in short term transients, and (b) limiting the fission gas release, fuel l

pellet temperature and cladding mechanical properties to within assumed design i

criteria.

In addition, limiting the peak linear power density during l

condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the BCCS acceptance criteria limit of 2200*F is not exceeded.

1 The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

Fg(E)

Heat Flux Bot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation 5 divided by I

the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

Fh Nuclear Enthalpy Rise Hot channel Factor, is defined as the ratio of the integral of linear power along.the rod with the highest integrated power to the average rod power.

3/4.2.1 A1111. FLUX DIFFERENCE The limits on ARIAL FLUX DIFFERENCE (AFD) assure that the Fg(8) upper bound envelope of the Fg limit specified in the COLR times R(3) is not l

exceeded during either normal operation or in the event of menon redistribution following power changes.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE encore detector l

outputs and provides an alarm message isesediately if the AFD for 2 or more OPERABLE encore channels is outside the allowed AI operating space for RAOC operation specified in the 00LR and the THERMAL POWER is greater than 50%

l RATED THERMAL POWER.

l i

e FARLEY-UNIT 2 E 3/4 2-1 AMENDMENT NO. 120

PdWER DISTRIBUTION 12MITS 4

BASES 3/4.2.2 mad 3/4.2.3 NRa7 FLUX BOT r um_wNEL FAuum, nur, m_ inie=7*Y MOT CHANNEL FASTOR

-The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is not exceeded, 2) the DNB design criterion is met, and 3) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

Control rods in a single group move together with no individual rod a.

insertion differing by more than

12 steps, indicated, from the group demand position.

b.

Control rod banks are sequenced with overlapping groups as described in specification 3.1.3.6.

The control rod insertion limits of ipecifications 3 1.3.5 and c.

3.1.3.6 are maintained.

d.

The axial power distribution, expressed in terms of AX1hL FLUX DIFFERENCE is maintained within the limits.

Fh will be maintained within its limits provided conditions a. through d.

above are maintained. The relaxation of Fh as a function of THERMAL POWER

)

allows changes in the radial power shape for all permissible rod insertion limits.

l When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 34 allowance is appropriate for manufacturing tolerance.

The heat flux hot channel factor Fg(E) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(I), to provide assurance that the limit on the heat flux hot channel factor Fg(I) is not.

W(E). accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

The W(I) function for normal operation and the AFD limits are provided in the COLR per Specification 6.9.1.11.

FARLEY-UNIT 2 3 3/4 2-2 AMENDMENT NO. 120

. ~ -. _... - - -. -

l 1

i

. ADMINZsTRATIVE CONTROLS i

MONTELY OPERATING REPORT 3

i 6.9.1.10 1

Routine reports of operating statistics and shutdown esperience, including documentation of all challenges to the FORV's or safety valves, shall i

be. submitted on a monthly basis to the Commission, pursuant to i

10 CFR 50.4, no later than the 15th of each month following the calendar month j

covered by the report.

?-

j c0RE OPERATING LIMITS REPORT 4

6.9.1.11 Core operating limits shall be established and documented in the 1

CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the followings l

1.

SHUTDOWN MARGIN limit for MODES 1, 2, 3, and 4 for specification 3/4.1.1.1,

\\

2.

SHUTDOWN MARGIN limit for MODE 5 for specification 3/4.1.1.2, t

3.

Moderator Temperature coefficient BOL and BOL limits and 300 pga and I

100 ppo surveillance limits for specification 3/4.1.1.3, i

4.

Shutdown Bank Insertion Limit for specification 3/4.1.3.5, i

5.

Control Bank Insertion Limits for specification 3/4.1.3.6, 6.

Axial Flux Differ,ence limits for specification 3/4.2.1, RTP 7.

Heat Flux Bot Channel Factor Fg limits, K(5) figure, N(5) values, and j

Fg(E) Penalty Factors for specification 3/4.2.2, 1

s.

Nuclear Enthalpy Rise Hot Channel Factor limits, FgRTP, and Power j

Factor Multiplier, PFag, for specification 3/4.2.3.

1 The analytical methods used to determine the core operating limits shall be i

those previously reviewed and approved by NRC in i

1.

WCAP-9272-P-A, " Westinghouse Reload safety Evaluation j

Methodology," July 1985 (M Proprietary).

i (Methodology for specifications 3.1.1.1 - shutdown Margin - Tavg > 200'F, j

3.1.1.2 - Shutdown Margin - Tavg s 200'F, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank j

Insertion Limits, 3.2.1 - Asial Flux Difference, 3.2.2 - Heat Flux Mot Channel

Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot channel Factor.)

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1 1

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i FARLEY-UNIT 2 6-19 AMENDMENT NO.120 J

6

b MINISTRATIVE CONTROLS g

f 2.

WCAP-10216-P-A, Rev. 1A, " Relaxation Of Constant Axial Offset Control /

j Fg Surveillance Technical Sper:ification," February 1994 (g Proprietary).

4 (Methodology for specifications 3.2.1 - Axial Fluz Difference and 3.2.2 - Beat Fluz Not channel Factor.)

3.

WCAP-10266-P-A, Rev. 2, "The 1981 Version Of Westinghouse Evaluation

}

Model Using BASH Code," March 1987 (M Proprietary).

(Methodology for specification 3.2.2 - Neat Flum Not Channel s

Factor.)

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-eechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

?

l The 00RE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

anwnar. Drusar, ammmmmvan evt.rmart.rrr omva n.c,a i

j 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for eacA diesel generator shall be suhaitted to the NRC annt. ally.

l This report shall cont &in the information identified in Regulatory Position C.3.b l

of NRC Regulatory Celde 1.108, Revision 1, 1977.

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i FARLEY-UNIT 2 6-19a AMENDMENT NO. 120 l

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4

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v t e Letter Sequence Request
TAC:M97245, (Open)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML20140H687
MONTHYEAR1997-03-25 [Table View]

ML20137E262

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