Text

Amends 74 & 68 to Licenses NPF-35 & NPF-52,respectively, Changing Tech Specs by Replacing cycle-specific Parameter Limits in core-related Specs W/Ref to Core Operating Limits Rept
	ML20043A695
Person / Time
Site:	Catawba
Issue date:	05/17/1990
From:	Matthews D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20043A694	List: ML20043A693; ML20043A695; ML20043A697;
References
	NUDOCS 9005220409
	Download: ML20043A695 (49)
	v • d • e

.

~ -

i I(

[o UNITED STATES g~

NUCLEAR REGULATORY COMMISSION WASHINoToN. o. C. 20sts

$,....+-

DUKE POWER COMPANY-l

, NORTH CAROLINA ELECTRIC MEMBER $ NIP CORPORATION l

SALUDA-RIVER ELECTRIC COOPERATIVE. INC.

DOCKET NO. 50-413 l

CATAWBA NUCLEAR STATION, UNIT 1 u

AMENDMENT TO FACILITY OPERATING LICENSE

,i I

Amendment No.- 74 l

License No. NPF-35 j

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

i, L

A.

The application for amendment'to the Catawba Nuclear Station Unit'l i

(the facility) Facility Operating License No. NPF-35. filed by the>

Duke Power Company acting for itself, North Carolina Electric Membership Corporation:end Saluda River Electric Cooperative, Inc.,

(licensees) dated January 17, 1990, complies with the standards and requirements of'the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations as 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 regu12tions-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 set forth in 10 CFR Chapter I; D.

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

E.

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

been satisfied.

i I

L

.j

-9005220409 900517 PDR ADOCK 05000413 gl' r(

5 P

PDC

~)

~..

, _,3

r t l 2.

Accordingly, the license is hereby amended by page changes to the Technical l

Specifications as indicated in the attachment to this license amendment.-

1 and Paragraph 2.C.(2) of Facility Operating License No.'NPF-35'is hereby i

l amended to read as follows:.

Technical Specifications The Technical Specifications contained.in Appendix A, as revised through Amendment No. 74, are hereby incorporated into the license.

The licensee shall-operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance.

I FOR THE NUCLEAR REGULATORY COMMISSION.

14ct'T./

David B. Matthews, Director Project Directorate II-3 Division of. Reactor Projects - I/II-Office of Nuclear Reactor Regulation

Attachment:

i Technical Specification 1

Changes Date of Issuance: May 17, 1990 i

l-.

l 1

s i

1 i

D* Clo 1

8

'o UNITED STATES l'

~ ',,

NUCLEAR REGULATORY COMMISSION i

'f wAssiNotow. o. c. 20sss j

s*****/

DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414-

]

CATAWBA NUCLEAR STATION, UNIT 2 l

AMENDMENT TO FACILITY OPERATING 'ICENSE L

I Amendment No. 68

)

License No. NPF.

1.

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

A.

The application for amendment to the Catawba Nuclear Station, Unit 2 i

(the facility) Facility Operating Licente No. NPF-52 filed by the-Duke Power Company acting for itself, North Carolina Municipal Power Agency No. I and Piedmont Municipal Power Agency,:(licensees) dated January 17, 1990,- complies with the standards and requirements'of the Atomic Energy Act of 1954, as amended (the.Act),'and the Commission's i

rules and regulations as. 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 j

this amendment.can be conducted without endangering the health al.d safety of the public,~and (ii) that such activities.will-be conducted in compliance'with the Commission's regulations set forth in 10 CFR Chapter I; D.

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of g

the Commission's regulations and all applicable requirements have been satisfied.

t i

i I

I L

2-i 2.

Accordingly, the license is herehy amended by page changes to the. Technical i

Specifications as indicated in the attachment to-this license amendment, i

and Paragraph 2.C.(2) of Facility Operating License No. NPF-52 is hereby amended-to read as follows:

j; Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 68 c, are hereby. incorporated into the license.'

The licensee shall operate the facility in accordance with the Technical Specifications and the. Environmental Protection Plan.

l i

3.

This license amendment is effective as of its date of issuance.

FOR THE. NUCLEAR REGULATORY COMMISSION-4 l

David B. Matthews, Director i

Project Directorate'II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation.

Attachment:

Technical Specification Changes 4

Date of Issuance:

May 17, 1990 l

9 i

t i

i

?

f 4:

9 4

t d.mv+-w

-y-,

e m

e

r.

i a

ATTACHMENT T0 tTCENSE AR(NDMENT NO. 74 FACil1TY OFERATING LICENSE NO. NPF-35 DOCKET NO. 50 413 i

A.#E.

T0 tICENSE AMENDPENT NO. ' 68 FACitTTY OPERATING LICENSE N0. NPF-52 DOCKET NO. 50-414 Replace the followin the enclosed pages. g pages of the Appendix *A* Technical Specifications with The revised pages are identified by Amendment number and H

contain vertical lines. indicating the areas of change.

I Remove Pages Insert Pa0es-1 & 11 I & II f

111 111*

' l IV & V TV & V VI VI*

XIX XIX 1-1 1-1*

1-2 thru l-7 1-2 thru l-7 1-8 1-8*

3/4 1-4 & 1-5 3/4 1-4 & 1-5 3/4 1-14 & 1-15 3/4 1-14 & 1-15 3/4 1-20 thru l-22 3/4 1-20 thru 1-22.

3/4.2-1 3/4 2-l' 3/4 2-3 3/4 2-3.

3/4 2-5 thru 2-7 3/4 2-5 thru 2-7 l.

3/4 2-7a thru 2-7c 3/4 2-7a thru 2-7c:

I 3/4 2-8 thru 2-9 3/4 2-8 thru 2-9 0

3/4 2-9a 3/4 2-10 & 2-11 3/4 2-10 & 2-11 B 3/4 1-2 B 3/4 1-2 E 3/4 1-4 B 3/4 1-4 B 3/4 2-1 & 2-2 B 3/4 2-1:& 2-2 i

B 3/4 2-Pa B 3/4'2-2a B 3/4 2-4 8 3/4 2-4 B 3/4 2-5 B 3/4 2-5 j;

6-19 6-19 i

6-19a 6-19a 7

0verleaf page provided to maintain document completeness.

. i

..i.

4

...._.-s_, - -

m

?

i i

l DEFINITIONS

)

SECTION PAGE 1.0 DEFINITIONS 1.1 ACTI0N...........................................'.............

1-1 i

1.2 ACTUATION LOGIC TEST..........................................

1-1 i

1.3 ANALOG CHANNEL OPERATIONAL TEST..............................

1-1 I

1.4 AXIAL FLUX DIFFERENCE.........................................

1-1 1.5 CHANNEL CALIBRATION...........................................

1-1 1.6 CHANNEL CHECK.................................................

1 3 1.7 CONTAINMENT INTEGRITY................'.........................

1-2

1. 8 CONTROLLED LEAKAGE............................................

1-2 1.9 CORE ALTERATION.............................................'..

1-2 1.10 CORE OPERATING LIMITS REP 0RT..................................

1-2

'1.11 DOSE EQUIVALENT I-131...............................'.........

~ 1-2 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................

1 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................

1-3 s

1.14 FREQUENCY N0TATION...........................................

1-3 1.15 IDENTIFIED LEAKAGE...........................................

1-3

.?

1.16 MASTER RELAY TEST..................

1-3 1.17 MEMBER (S) 0F THE PUBLIC......................................

1-3 1,18 0FFSITE DOSE CALCULATION MANUAL.................'.............

1-3 1.19 OPERABLE - OPERABILITY.......................................

1-4 a

1 1.20 O P E R AT I ON AL MO D E - M0D E.......................................

1-4 1.21 PHYSICS TESTS................................................

14 e

1.22 PRESSURE BOUNDARY LE AKAGE..................................

1-4 1.23 PROCESS CONTROL PR0 GRAM.,....................................

1 ;

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

1-4 5

1.25 QUADRANT POWER TILT RATI0....................................

1-4 1.26 RATED THERMAL P0WER..........................................

1 ' -

1.27 ' REACTOR. BUILDING INTEGRITY...................................

1-5 1.28 REACTOR TRIP SYSTEM RESPONSE TIME............................

1 1,29 REPORTABLE EVENT.............................................

1-5 i

1.30 SHUTDOWN MARGIN..............................................

1-5 1.31 SITE B0VNDARY................................................

1 !

1.32' SLAVE RELAY TEST................................................

1-5 CATAWBA - UNITS 1 & 2 I

Amendment No. 74 (Unit 1) l Amendment No. 68 (Unit 2)

=;

-s.

--.~

m

- - -.i

-,,+w

+-

s t'i DEFINITIONS t

SECTION PAGE l

{

1.33 SOLIDIFICATION...............................................

1-5

.1.34 SOURCE CHECK!...........

1-6 1.35 STAGGERED TEST BASIS.........................................

1-6 I

1.36 THERMAL P0WER.......................

1-6 i

1.37 TRIP ACTUATING DEVICE OPERATIONAL TEST.......................

1-6

' 5 1.38 UNIDENTIFIED LEAKAGE.........................................

1-6 1.39 UNRESTRICTED AREA............................................

1-6 l

1.40 VENTILATION EXHAUST TREATMENT SYSTEM.........................

1-G 1.41 VENTING.......................................................

1-7 l

1.42 WASTE GAS HOLDUP SYSTEM......................................

1-7 TABLE 1.1 FREQUENCY N0TATION.......................................

1-8 1

t TABLE-1.2 OPERATIONAL MODES.............

4........................- 9 l

i L

r i-l

.i l-7 L

1 b

t

CATAWBA - UNITS 1 & 2 II' Amendment No. 74 (Unit 1)

L Amendment No.-68 (Unit 2)

+,,

4 w

6

+

e

4 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1

SECTION PAGE i

2.1-SAFETY LIMITS 2.1.1 REACTOR C0RE................................................

2-1 1

2.1.2 REACTOR COOLANT SYSTEM FRESSURE.............................

'2-1 FIGURE 2.1-1. REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION..

2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS................

2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM-INSTRUMENTATION TRIP SETPOINTS....

2-4 BASES I

SECTION l

2.1 SAFETY LIMITS

{

2.1.1 REACTOR C0RE................................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS................

B 2-3 i

?

-i 4

4 CATAWBA - UNITS 1 & 2 III

- L t

1

. i 1

4 i

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

SECTION PAGE 3/4.0 ' APPLICABILITY...............................................

3/4 0-1 t

3/4.1 REACTIVITY CONTROL SYSTEMS l

3/4.1.1 BORATION CONTROL f

Shutdown Margi n - T,yg > 200*F...........................

3/4 1-1 Shutdown Margin - T,yg 5 200*F...........................

3/4 1-3 l

Moderator Temperature Coefficient........................

3/4 1-4 l

Minimum Temperature for Criticality......................

3/4 1-6 3/4.1.2 BORATION SYSTEMS I

Flow Path - Shutdown.....................................

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

3/4 1-8 i

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

3/4 1-9 Charging Pumps 0perating...............................

3/4 1-10 Borated Water Source - Shutdown..........................

3/4 1-11 i

Borated Water Sources - Operating........................

3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................

3/4 1-14 5

TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R0D...................

3/4 1-16 Position Indication Systems - Operating..................

3/4 1-17 l

Position Indication System - Shutdown....................

3/4 1-18 l-l Rod Drop Time............................................

3/4 1-19 4

l Shutdown Rod Insertion Limit............,.................

3/4 1-20 j_

Control Bank Insertion Limits............................

3/4 1-21 l

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE....................................

3/4 2 l I

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z).....................

3/4 2-5' Q

CATAWBA - UNITS 1 & 2 IV Amendment No. 74 (Unit 1)

Amendment No. 6B(Unit 2) l,

+m.-v e

- -L-

l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE

[

t I

i 3/4.2.3= REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTMALPY RISE MOT CHANNEL FACT 0R..................................

3/4 2-9 I

3/4.2.4 QUADRANT POWER TILT RATI0................................

-3/4 2-12

~

3/4.2.5 DNB PARAMETERS...........................................

3/4 2-15

{

TABLE 3.2-1 DNB PARAMETERS........................................

3/4 2-16 1

i I

3f4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION......................

3/4 3-1 i

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION...................

3/4 3-2 i

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES....

3/4 3-7 TABLE 4.3-1 REACTOR TRIP SYSTEM. INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................

3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..........................................

3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.,.........................................

3/4 3-15 l

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS...........................

3/4 3"27 i

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESP 0NSE' TIMES.............

3/4 3-37 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-42

}

4 3/4.3.3 MONITORING INSTRUMENTATION Raciation Monitoring For Plant Operations................

3/4 3-51 TABLE 3-3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS.....................................

3/4 3-52 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS...-..................

3/4 3-54 Movable Incore Detectors.................................

3/4 3-55 l

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

3/4 3-56 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION....................

3/4 3-57 i

. REQUIREMENTS.............................................

3/4 3-58:

Meteorological Instrumentation...................-........

3/4 3-59 CATAWBA - UNITS 1 & 2 V

Amendment No. 74 (Unit 1)

Amendment No. 68(Unit 2)-

U i

6 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS f

SECTION PAGE TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION.............

3/4 3-60 j

TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE q

REQUIREMENTS..........................................

3/4 3 Remote Shutdown System...................................

3/4 3-62 TABLE 3.3-9 REMOTE SNUTDOWN MONITORING INSTRUMENTATION............

3/4 3-63 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMINTS................................

3/4 3-64 Accident Monitoring Instrumentation......................

-3/4 3-65 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION...............,..

3/4'3-66 i

TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..................

3/4 3-68 Chlorine Detection Systems...............................

3/4 3-70 Fire Detection Instrumentation...........................

3/4 3-71 TABLE-3.3-11 FIRE DETECTION INSTRUMENTATION......................

3/4 3-73 j

Loose-Part Detection System..............................

3/4 3-77 j

Radioactive Liquid Effluent Monitoring Instrumentation...

3/4 3.

TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3 79 TA6LE 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-81 Radioactive Gaseous Effluent Monitoring Instrumentation..

3/4 3-83 TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION..........................................

3/4 3-84 TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-88 3/4.3.4 TURBINE OVERSPEED PROTECTION..............................

3/4 3-91 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............................

3/4 4-1 Hot Standby...........................................

3/4 4-2 4

Hot Shutdown.............................................

3/4 4-3.

l Cold Shutdown -' Loops Fi11ed.............................

3/4 4-5 j

Cold Shutdown - Loops Not F111ed.........................

3/4 4-6 1

i i

- CATAWBA - UNITS 1 & 2 VI 4.

s i'S 1

g n.*"

i ADMINISTRATIVE CONTROLS-i SECTION.

PAGE

' A n n ua l R e p o r t s..... '.............. =.....................'.......

6-16 Ni4 Annual: Radiological Environmental Ope' rating' Report.........

6-16:

4

' --Semiannual Radioactive Effluent Release: Report............

6-17, Monthly Operating Reports...................-...............

16-19 Core. Operating Limits; Report.............;................'

m. _

6-19

'l-n i

-6.9.2 SPECIAL REP 0RTS............................................

6-20 o-J6.10 RECORD RETENTION...........-..-4...............................

6-20:

6.11 ' RADIATION PROTECTION PR0 GRAM...................................

6. ;

i:

6.12 HIGH RADIATION AREA...............,............:..............

.6-21 6.13 PROCESS CONTROL PROGRAM (PCP)..............................

'6-22 4

6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM).....................

6-23

~

6.15" MAJ'OR CHANGES T0' LIQUID.-GASEOUS, AND SOLID RADWASTE-T R E AT M E N T S Y ST E M S..................'..................... -.......

6-23 l

1r i

p 7.0 SPECIAL TEST PR0 GRAM........................................

7 4 i

1 i

-I f'

i l.

I' 1

n

!i'-

i i

i t

i 1 CATAWBA - UNITS 1 & 2 XIX Amendment No. 74 (Unit 1)

Amendment No. 6EF(Unit 2);

j 5,

and-accuracy.

- AXIAL FLUX' DIFFERENCE'

~

1.4 - AXIAL FLUX DIFFERENCE shall be the differencelin normalized flux signals between the top and bottom. halves of a two section excore neutron detector.

CHANNEL' CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment,~ as necessary, of the-l

~

p channel such~ thattit' responas within thetrequired. range and accuracy to known valuesJof input., The CHANNEL CALIBRATION:shall encompass.the: entire channel-including the sensors and alarm, interlock and/or.' trip functions..and may be:

}

-performed by any series of sequential, overlapping..or total channe1~ steps.

such that.the entire channel ~is calibrated.

i CJjANNEL CHECK 1.6 A CHANNEL CHECK shall be. the qualitative as'sessment-of channel behavior during operation by' observation.
This determination shall include, where' possible, comparison.of the channel indication.and/or status with'other-1
-indications and/or status derived from independent instrument channels measuring the.same parameter.
1 CATAWBA - UNITS 1 & 2 1-1 3
o l

a 'm m

m w
..m-.
2
.-e,,w,s, 4 -
y
~
v
. DEFINITIONS
CONTAINMENT INTEGRITY c./
1.7 : CONTAINMENT. INTEGRITY shall exist when:-
5
' a.
All' penetrations. required to be closed dur'ing accident conditions are either:
l 1);
Capable of being closed by an OPERABLE containment automatic'
isolation valve system,or.
3 U
2)-. Closed by manual: valves,. blind' flanges', or deactivated automatic valves secured..in their. closed' positions, except as'provided.in Table 3.6-2 of Specification 3.6.3.
y 2
b..
All equipment hatches are closed and sealed, c.
Each air ~ 1ock 'is in compliance with the: requirements of Specification =
3. 6.1. 3, -
y
d. -
2 The containment leakage-rates are within the limits of. Specification 3.6.1'2, and' e.
The sealing mechinism associated with each penetration (e.g., welds, L
bellows, or 0-rings) is OPERABLE.
q CONTROLLED LEAKAGE t
1.8 CONTROLLED LEAKAGE shall be that seal water flow. supplied toithe' reactor' j
coolant pump seals, j
' CORE ALTERATION 1.9 CORE ALTERATION shall be th'e 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 ecvement of a component to a safeLeonservative position.
CORE-OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR)-is the unit-specific document d
that provides' core operating limits ~for the current operating reload cycle.
~
These cycle-specific core operating limits shall be determined for each reload r
cycle in accordance with Specification 6.9.1.9.
Unit operation within these' L
- operating limits is addressed in individual specifications'.
i DOSE EQUIVALENT I-131 f
1.11 DOSE EQUIVALENT I-131'shall be that concentration of.1-131:(microcurie / gram).
u which alone would produce the same thyroid -dose as the quantity and: isotopic-R mixture of I-131, I-132, I-133, I-134,' and I-135 actually present.
The' thyroid t
dose conversion factors-used for this calculation shall:be those listed in 4
' Table III of TID-14844, " Calculation of Distance Factors ~for Power and Test Reactor Sites."
E - AVERAGE DISINTEGRATION ENERGY
' 1.12 E shall be the average (weighted in proportion to;the concentration o'f l
4 each radionuclide.in the sample) of the sum of the average beta and. gamma energies-per disintegration (MeV/d) for the radionuclides in the sample.
p CATAWBA - UNITS 1 & 2 1-2.
Amendment No. 74 (Unit 1)
L Amendment No. 68 (Unit.2) o
..f i
' : a 1
u
-~
, m N
1 s
DEFINITIONS l
ENGINEERED SAFETY FEATURES' RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME:shall.be that time.
l interval from when the monitored parameter exceeds.its ESF Actuation Setpoint-at the channel sensor until the_ESF equipment is. capable of performing its
. safety function (i.e'.'
the valves travel to their required positions, pump-(discharge' pressures re,ach their required values,.etc.).. Times shall' include
! diesel; generator' starting and sequence loading delays where' applicable.-
FREQUENCY-NOTATION
-1.14 -.The_ FREQUENCY NOTATION specified for the performance of Surveillance
[
]
> Requirements shall correspond.to the intervals defined in Table 1.1.
IDENTIFIED 1 LEAKAGE a
n 1.15 IDENTIFIED LEAKAGE shall be:
b L
'a.
' Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump.
- seal or valve packing leaks that are captured and conducted to a sump <
or collecting tank, or g
b.
Leakageintothecontainment.atmospherefromsourcesthatarebothi i
specifically located and known either not-'to interfere with the opera-tion of Leakage Detection Systems or-not to be PRESSURE BOUNDARY LEAKAGE, or c.
. Reactor Coolant System leakage through a. steam generator to the.
SecondaryLCoolant System.
1 i
MASTER RELAY TEST
1.16 A MASTER RELAY TEST shall be the energization of each master relay and:
l
)
m verification of. OPERAPILITY of each relay..The MASTER RELAY TEST shall include i
!9 a continuity check of each associated slave relay.
MEMBER (S) 0F THE PUBLIC 1.17 MEMBER (S) 0F THE PUBLIC shall include all persons who areinot occupa-l g
tionally associated with the plant.
This category does not include employees of the licensee, its contractors,.or vendors. 'Also excluded from this category are persons.who enter the site to service equipmer.t or to make deliveries..
i i-This category does include persons who;use portions of the. site-for recre-ational, occupationai, or other purposes not associated with the p1 ant.~
0FFSITE DOSE CALCULATION MANUALL 1.18 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology l
k and parameters used in the calculation of offsite doses due to radioactive-I gaseous and liquid effluents, in the calculation ~of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ -
. mental Radiological Monitoring Program.
L i
CATAWBA - UNITS 1 & 2 1-3 Amendment No. 74 (Unit 1) l Amendment No. 68 (Unit 2) e
[
1.
J.,.. _
~_
q q
DEFINITIONS
[
OPERABLE'--OPERABILITY
--1.'L9' A' system, subsystem.. train,' component or device shall be OPERABLE or l.
l l:
.hhve-OPERABILITY when it is capable of performing its specified. function (s),
j E
and ~when all; necessary attendant instrumentation, controls', electrical _ power, j
i
. cooling.or ' seal water, lubrication or other auxiliary equipment thatiare 1
L
. required for'the' system,' subsystem, train, component, or device to: perform its-l
. function (s) are also capable of performing their related support function (s).
2 l0PERATIONALMODE-MODE-a!
.1.20 An.0PERATIONALLMODE(i.e., MODE)shallcorrespondtoanponeinclusive; h
combination'of core reactivity condition, power: level, and. average ~ reactor
'i coolant temperature specified_in Table 1.2.
l PHYSICS-TESTS I
1 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental
[
(
nuclear' characteristics of the reactor core 'and related instrumentation:
(1) described in Chapter 14.0 of the FSAR, (2) authorized under the-provisions of 10 CFR 50.59, or'(3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE 1.22': PRESSURE BOUNDARY LEAKAGE sha11 4e leakage'(except steam generator tube.
I leakage) through.a nonisolable fault-in a Reactor Coolant System componentT body, pipe. wall, or vessel wall.
PROCESS CONTROL PROGRAM lL 1.23: The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, l.
sampling,' analyses, tests, and determinations to be madeito ensure that?.
t I
processing and packaging of solid radloactive wastes based on-demonstrated-processing of actual or simulated wat solid' wastes will: be accomplished in I
such a way as to assure compliance with 10 CFR Parts 20,:61and 71 and
}
Federal and. State regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste.
PURGE - PURGING d
l:
1.24 PURGE or PURGING shall be any controlled process of discharging air or gas l
l j
from a confinement to maintain temperature, pressure, humidity, concentration,.
0 or other operating condition, in such a manner that replacement air or. gas is
[
required to purify the confinement.
I
. QUADRANT POWER TILT RATIO h
1.25 QUADRANT POWER TILT RATIO shall be the ratio of-the maximum upper'excore l
detector calibrated output to the average af the upper excore; detector cali-J L,
brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of-the lower excore detector calibrated outputs, whichever.
.is greater.
With one excore detector inoperable, the remaining three detectors shall'be used for computing the average.
3 I
CATAWBA - UNITS 1 & 2 1-4 Amendment No. 74 (Unit 1)
Amendment No. 68 (Unit :2)-
J 4-t i
w'
-s
,o
.,.s,
, _ _, _,. p
Dt..'INITIONS RATED THERMAL-POWER o
1.26 RATED THERMAL POWER shall be a total reactor core heat transfer rate to.
l l
-the reactor coolant of 3411 MWt.-
REACTOR BUILDING INTEGRITY f
1
-1.27 REACTOR BUILDING. INTEGRITY!shall exist when:
l Each door in each access opening is closed except when the access a..
opening is' being used for normal transit entry and exit, then at least-one door shall be closed, b.
The Annulus Ventilation System is OPERABLE, and The sealing mechanism associated with:each pcnetratioa (e.g., welds,-
c.
bellows, or' 0-rings) is OPERABLE.
j u
REACTOR TRIP SYSTEM RESPONSE = TIME 1.28 The. REACTOR TRIP SYSTEM RESPONSE TIME shall be the. time interval from I
when the monitored parameter exceeds its Trip'Setpoint at the channel-sensor
-- j until: loss of stationary gripper coil voltage.
{
REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in l
Section 50.73 of 10 CFR Part 50.
SHUTDOWN-MARGIN t
1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity.by.which l
the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and. control) are fully inserted except-for the single rod cluster assembly of highest' reactivity worth which is assumed to be fully withdrawn.
}
SITE BOUNDARY i
1.31 The SITE BOUNDARY shall be that line beyond which the land is neither l-owned, nor leased, nor otherwise controlled by licensee.
j SLAVE RELAY TEST I
1.32-A SLAVE RELAY TEST shall be the energization of each slave-relay and-l verification of OPERABILITY of each-relay.
The SLAVE RELAY TEST shall include
.a continuity check, as a minimum, of associated testable actuation devices.
. SOLIDIFICATION i
1.33 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets I
i shipping and burial ground requirements.
CATAWBA - UNITS 1 & 2 1-5 Amendment No. 74 (Unit 1)
Amendment No. 68 (Unit 2) i
1 DEFINITIONS e
SOURCE CHECK
,j 1.34 A SOURCE CHECK shalltbe the qualitative assessment of. channel response.
l 1
when the channel sensor is exposed to a source of! increased' radioactivity, r
STAGGERED TEST BASIS' 1.35 A STAGGERED TEST BASIS shall consist of:-
l a.1 A test schedule for n' systems, subsystems, trains, or other designated-components obtained by. dividing the specified test' interval into' n -
= equal subintervals, and J
b.'.The testing of one system, subsystem, train, or other designated l
component at.the beginning of-each subinterval.
THERMAL POWER 1.36 : THERMAL POWER shal1 be' the total reactor core heat transfer rate to the I
~
reactor coolant.
TRIP ACTUATING DEVICE OPERATIONAL TEST 1.37 A-TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the-l-
4 Trip' Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions.
The TRIP. ACTUATING DEVICE OPERATIONAL TEST shal1Jinclude
{
adjustment,:as necessary, of the~ Trip Actuating Dcvice such.that it actuates at the required Setpoint within the required accuracy.
' UNIDENTIFIED LEAKAGE i
1.38 UNIDENTIFIED LEAKAGE shall be a11' leakage which is not' IDENTIFIED LEAKAGE l
or CONTROLLED LEAKAGE.
_ UNRESTRICTED ~ AREA 1.39 An UNRESTRICTED AREA shall be any. area at.or beyond the SITE BOUNDARY l
access to which is not controlled by the-licensee for purposes: of= protection of '
individuals from exposure to radiation and radioactive materials, orL any area i-within the SITE BOUNDARY used for residential quarters or for industrial,.
commercial, institutional, and/or recreational purposes.
\\.
VENTILATION EXHAUST TREATMENT SYSTEM 1.40 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and b
. installed to reduce gaseous radiciodine'or radioactive material in particulate' 1
form in effluents by passing ventilation or vent exhaust gases through activated carbon adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from-the gaseous exhaust stream prior to the' release to'the.envi-ronment. Such a system is.not considered to have any effect on noble gas effluents.
Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
CATAWBA - UNITS 1 & 2 1-6 Amendment No. 74 (Unit 1) l Amendment No. 68 (Unit 2) l j
i s
\\
DEFINITIONS q
VENTING 1.41 VENTING-shall: be' the controlled process of discharging air or ~ gas from a l~
confinement.to maintain temperature',- pressure, humidity, concentration or other operating condition, in such.a manner that replacement air. or gas is not pro-:
vided or required during.VENTINGi. Vent, used in system names, does not imply a: VENTING process.
q WASTE GAS HOLDUP' SYSTEM-L
-f 1.421 A WASTE' GAS HOLDUP SYSTEM'shall-be'any system designed'and' installed-to l
reduce radioactive gaseous effluents by collecting Reactor Coolant System
-offgases from-the Reactor Coolant System and providing for: delay or holdup'.
for the purpose of, reducing the: tota 1L radioactivity prior. to release to the' environment.'
1 d
L I
l
'l L
h-t I
i CATAWBA - UNITS 1 & 2 1-7 Amendment No. 74 (Unit 1)
)
l Amendment No. 68' (Unit 2) y 1
(^~
y, ll t
i TABLE 1.1' l
FREQUENCY NOTATION ~
.i
' NOTATION FREQUENCY S
At.least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .'.
5 D:
At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
1
W At

least once per 7 days.
M At least'once per 31 days..
-l
=Q' At least once..per.92 days.
j SA At least once per,184 days.
\\
R
, At least.once.per 18 months..
S/U-
. Prior.to each. reactor startup.
$.i N. A.
-Not' applicable.
e:
3 P-Complete.d prior:to each release.
j c
a I
f t
i-o 4
i 'l 11
'l i
CATAWBA - UNITS 1 & 2 1-8 e
s
+
..g e
0 REACTIVITY CONTROL SYSTEMS-1 MODERATOR TEMPERATURE COEFFICIENT a
LIMITING CONDITION FOR OPERATION o
3.1.1.3 The moderator temperature coefficient ~(MTC) shall be within the limits specified in the CORE-0PERATING LIMITS l REPORT (COLR). 'The maximum upper _limitishall be less than or equal to that shown in Figure 3.1-0.
a APPLICABILITY:
Figure 3.1-0 and_COLR Figure 1:Limitsc MODES l'and 2* only#.
End.of Cycle Life (EOL) Limit - MODES 1, 2, and 3 only#2 ACTION:
With'the MTC more-positive than the limit specified in Figure 1 of a.
the COLR,; operation in. MODES l'and' 2,may. proceed provided: ~
J 1.
Control rod withdrawal limits'are' established and maintained.
-sufficient to restore the MTC to:less positive than the i
limit specified in' Figure:1_ of-the =C010 sithin: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or'be
[
J in HOT STANDBY within the next.6 hot These withdrawal limits i
shall be in addition' to the'insertis limits of Specification 3.1.3.6; 4
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;'and-r 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 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.
a i
b.
With the MTC more negative than the EOL limit specified'in theLCOLR, L
be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
L 3
.t
'*With K,ffheaterthanorequalto1.
See Special Test Exceptions Specification 3.10.3.
A i
l CATAWBA - UNITS 1 & 2 3/4 1-4 An.endment No. 74 (Unit 1)
Amendment No. 68 (Unit'2)
,1
ly i
. REACTIVITY CONTROL SYSTEMS--.
4 SURVEILLANCE REQUIREMENTS' 4.1.1.3 -The MTC shall be determined to be within_its limitsiduring each fue cycle as follows:-
a.
The~ MTC shallib'e' measured and compared to the '80L limit specified in 4
the.COLR, prior;to initial operation above 5% of' RATED THERMAL POWER,
-after each-fuel loading;' and r
b.-
~The MTC shall be measured at'any THERMAL POWER and' compared.to the
300' ppm surveillance!11mit specified in'the COLR-(all-rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.
In the1 event this--
l comparison ine. cates the'MTC is more negative'than the 300 ppm surveillance. limit;specified.in the COLR, thelMTC shall be remeasured,-
~
t and compared to the EOL MTC limit specified in the COLR,Jat least ~once.
per 14' EFPD.during' the remainder of the fuel cycle.
1.
i
%1
~
L.. -
i; i
l e
1 f
.i 4
l:
z l
L p
y l
i y
CATAWBA
-UNITS 1 & 2 3/4 1-5 Amendment 'No. 74 (Unit 1).
J F-Amendment No. 68 (Unit' 2)'
f n
w--.
+.
y' R
J j
REACTIVITY CONTROL SYSTEMS j
l 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP' HEIGHT' LIMITING CONDITION FOR OPERATION i
' 3.1.3.1.i A11 full-length shutdown and control rods shall be OPERABLE and l
positioned within 112 stepsL(indicated position) of their group step counter demand' position.
~
-APPLICABILITY:' MODES:1* and 2*.
I ACTION:
. a.
With.onelor more full-length rods inoperable due to being immovable 4
as a result of excessive-friction' or mechanical interference or-known to be untrippable, determine that'the' SHUTDOWN MARGIN require-1 ment of Specification 3.1.1.1'is satisfied within I hour and be>in 4
HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b.
With more than one full-length-rod misaligned from the group-step.
counter demand position by more than 112 steps.(indicated position),.
Ebe in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
I
.c.
With>one full-length rod trippable but inoperable due to causes-other than-addressed by ACTION a.,.above, or misaligned from its_
group step counter demand height by more than 112 steps (indicated-position), POWER'0PERATION may continue provided that.within 1 hour:
1 1.
The rod is restored to OPERABLE status within the above alignment requirements, or l
2.
The rod is declared inoperable and the remainder of the rods in-L the group with the' inoperable rod are' aligned to.within;*-12 steps 1
i of the inoperable-rod.while maintaining the rod sequence and L
insertion limits of Specification 3.1.3.6.
The-THERMAL POWER l-l' level shall be restricted pursuant to' Specification 3.1.3.6 during-l l
subsequent operation, or j
3.
The rod is declared inoperable and the' SHUTDOWN MARGIN require-ment of Specification 3.1.1.1'is satisfied.
POWER OPERATION may then continue provided that:
a)
A reevaluation of each accident analysis of Table 3.1-1 1s-i
(_
performed within 5 days; this reevaluation shall confirm that the previously_ analyzed results of these accidents-4 remain valid for the duration of operation under these
'i conditions; b)
The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; t
See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
b l
CATAWBA - UNITS 1 & 2 3/4 1-14' Amendment No. 74
' Amendment No. 68 (Unit 1)
(Unit 2) l
~
J
,.-w-.,
.-. -. +
- ma m.
-m s.
m'
1
~
REACTIVIT CONTROL SYSTEMS 1
LIMITING CONDITION FOR OPERATION ACTION (Continued) c);
A power distribution map is obtained[from'the movable N
incore detectors.and F-(Z) and F are verified to'be; 9
- AH within their limits within'721:ours; and P
d)
The THERMAL POWER: level ~is reduced to less than or equal to 75% of. RATED THERNAL POWER within the next hour and within the following 4' hours the.High_ Neutron Flux-Trip:Setpoint is' reduced to less than or equal to 85%.
.of RATED THERMAL POWER.
N d.
With more than one full-length rod.trippable but inoperable due-to 1
causes other than addressed by ACTION a above,3 POWER OPERATION mayi continue provided that:
1.
'WithinEl hour, the remainder of!the rods-in the bank (s') with; i
the inoperable rodsLare-aligned to within'112-steps of the~
-inoperable rods while maintaining theirod sequence andcinser-1 tion'.11mits'of Specification 3.1.3.6.
-The THERMAL POWER level
(
d shall be restricted pursuant-to' Specification'3.1.3.6 during subsequent operation, and=
4 s
6 2.
The inoperable rods are restored to OPERABLE status-within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .
i a
SURVEILLANCE REQUIREMENTS t
l
'4.1.3.1.1 The position of each-full" length rod shall.be-determined to be o
i-within the group demand limit"by verifying theLindividua1Lrod positions l
at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except'during time intervals 1when the Rod Position l:
Deviation Monitor is inoperable,;then verify the group positions at least once per 4 hevrs.
o 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be L
determined to be OPERABLE by movement of at least 10 stepstin any one
[
direction at least once per 31 days.
y L
i i
i i
I q
i
CATAWBA --UNITS 1 & 2 3/4 1-15 Amendment No. 74 l
. Amendment No. 68(Unit 1)
(Unit 2)
L
~,
_.,m
==
a
___,e
.m miWe
.,,e,e' Jm,
-gv,s rP
l Y,'
} {if l
REACTIVITY' CONTROL 4 SYSTEMS X
SHUTDOWN ROD' INSERTION. LIMIT-LIMITING CONDITION'FOR OPERATION'
.l 3.1.3.5 All shutdown 1 rods shall be limited-in physical insertion as specified' in the CORE OPERATING LIMITS REPORT.(COLR).
a APPLICABILITY:' MODES 1* and 2*#,
-ACTION:.
I
With a' maximum;of one shutdown rod inserted beyond the insertion limit I
specified in the COLR, except for surveillance testing pursuant to 1
. Specification 4.1.3.1.2, wit M a l' hour either:-
a.
Restore-the rod to.within the insertion limit specified in the COLR-l q
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 i
limit specified in the COLR:
Within 15 minutes prior to withdrawal of any rods in Control Bank A, a.
B, 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.
See Special Test. Exceptions Specifications 3.10.? and 3.10.3.
With K,ff greater than or equal to 1.
k
. CATAWBA - UNITS 1 & 2 3/4 1-20 Amendment No. 74 (Unit 1) 3 Amendment No. 68(Unit 2) l
._ 7 g-
.i
~ REACTIVITY CONTROL SYSTEMS-CONTROL BANK INSERTION LIMITS LIMITING CONDITION FOR OPERATION o
3.1. 3. 6 The control' banks.shall be limited in physical insertion'as'specified in'.the CORE OPERATING LIMITS REPORT-(COLR).
APPLICABILITY:
MODES 1* and 2*#.
. ACTION:-
With"the control banks 11nserted beyond the insertion limits specified in the COLR, except-for surveillance-testing pursuant to -Specification 4.1.3.1.2:=
1 a.
. Restore the control-banks to within the' limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or
' b'.
Reduce-THERMAL. POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal:to that:
fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits'.specified inLthe COLR, or.
l
c.
Be in at least HOT-STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIREMENTS f1 i
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-
-j when the Rod Insertion Limit Monitor i.s' inoperable,1then' verify the individual i
rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
i q
l f
?]
\\.
See Special Test Exceptions Specifications 3.10.~2 and 3.10.3.
With K,ff greater than or equal to 1.
-CATAWBA - UNITS 1&2 3/4 1-21 Amendment No. 74 (Unit 1)
Amendment No. 68 (Unit 2)
h 5 ' l $...
. i
'$ l t
4 is s
., w;.
r i
i I
'i j,,
a s-N s
,1-
,I,
' 41 s:
u'
!)
i
. j
8; b.-
w b
. o, L
I a.
- i
..q 1
a f
THIS PAGE' INTENTIONALLY,DELETEDh
3 1
.. z ' _
1
'. i, a-l
' 7
,' 5 2
'o,-
.)
1
)
1
):
')
-s i
-CATAWBA - UNITS 1&2 3/4 1-22' Amendment No. 74 (Unit 1)
. Amendment No. 68 (Unit' 2).
i l
t
3
a-g g
- p 324.2 POWER DISTRIBUTION' LIMITS 3/4.2.1 AXIAL FLUX-DIFFERENCE (AFD)'
i
.g
. LIMITING' CONDITION FOR OPERATION i
-3.2.1 Thein5tcated.AXIALFLUXDIFFERENCE(AFD)shallbemaintainedwithin:
i a.
the allowed ' operational ~ space.as specified:in.the CORE OPERATING ;
LIMITS REPORT =(COLR) for RAOC operation, or
b..
within'the; target' band specified in the COLR about the target flux-
-difference-during baseload operation.
APPLICABILITY:
MODE:1, above 50% of RATED THERMAL POWER
4 ACTION:-
' a. -
For RA00 operation ~with the indicated AFD~outside of the limits
. specifled in' the COLR, 1.
Either restore the indicated AFD to within the COLR: limits within
~
15 minutes, or 2.
Reduce THERMALLPOWER.to less'than'50% of' RATED THERMAL POWER within:-30 minutes 1and reduce the Power Range-Neutron Flux-High
.i Trip setpoints to less than or'equa11to155% of RATED. THERMAL-POWER within theTnext=4 hours.
b.
For Base-Load operation above APLND** with the indicated AXIAL FLUX DIFFERENCE outside of the applicable target' band about the target
' flux difference:
l 1-
-Either restore the indicated AFD to within the COLR!specified l
target band limits within 15 minutes, or ND 2.
Reduce THERMAL POWER to'less than APL of RATED THERMAL POWER and discontinue Base Load' operation within 30 minutes.
e c.
THERMAL POWER shall not be increased'above 50% of RATED THERMAL POWER unless the' indicated AFD is within the limits'specified in the COLR.
l i.
l 4
See Special Test Exceptions Specification-3.10.2.
e NDt
APL is the minimum allowable (nuclear design) power level for base load O
operation and is specified in the CORE OPERATING LIMITS REPORT per i
Specification 6.9.1.9.
l'
-CATAWBA ^
UNITS 1&2 3/4 2-1 Amendment No. 74 (Unit 1)-
Amendment No. 68 (Unit 2) v w--
--e--
--m,--
e---,
w ren es 4-a
7' a
m r.
' 'l 4
"h n
'I 7,
/1 r
J
?i
.t
'l i
1
,-I ll i-
'M 1,,
i
't I
3 t
f I.
-Ea t
I t
j t
V f
( 1 a
THIS PAGE INTENTIONALLY DELETED.
O
.a T
s
.g
.1-i i
t
!F
t.q r
l.
c: g.
it i
f
_{
q
^
t ll e
?
y
-j e
]
. :r-
.i
]
~!
t-
-- U
' CATAWBA - UNITS 1&2 3/4'2-3 Amendment No. 74 (Unit *1)
' Amendment No 68 (Unit 12) yf l
_'t' '
a e '_
ji e
s
.e.
u
.I
q n
POWER DISTRIBUTION LIMITS-HEAT'FLUhHOTCHANNELFACTOR,Fg 3/4.2.2 tLIMITING CONDITION FOR' OPERATION I
i
3.2.2 F (Z) shall_ be limited'by the following relationships:'
q t
RTP F (Z)-
9 K(Z) for P >;0.5 l
.i RTP F (Z) $ F 9
K(Z) for P 3,0.5:
j:
^
Where:
F TP = the F Limitiat' RATED THERMAL POWER (RTP) 2 q
specified-in-the. CORE OPERATING' LIMITS' REPORT-(COLR),
p ~ THERMAL POWER-
, and; 1
J RATED THERMAL POWER-
~;
K(Z) = the normalized F (Z) for a.given core height' E
specified in'the COLR.
4 APPLICABILITY:
MODE 1.
ACTION:
4 F
With F (Z) exceeding its limit:
~
9 a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit-q
-within 15 minutes and similarly reduce the Power' Range Neutron'-
F Flux-Hi0h Trip -Setpoints within the ne,:t 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; hours;< subsequent POWER OPERATION _
H may proceed provided the Overpower AT Trip Setpoints (value~of K )-have' 4
been reduced at least 1%~(in AT span) for each 1% F (Z) exceeds the-limit, and q
l i
r b.
Identify and correct the cause of the out-of-limit condition: prior 4
to increasing THERMAL 1 POWER above the= reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore-mapping to be within its~ limit.
q e
u l'
' CATAWBA - UNITS 1&2 3/4 2-5 Amendment No. 74 '(Unit ~ 1)
I Amendment No. 68 (Unit 2) i P
- ~
~
p 1:
m.
POWER' DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS a
4I2.2l'Theprovisions'of-Specification 4.0.4arenotapplicable.-
4.2.2.2. For.RA00: operation, F (z) shall be. evaluated to determine if F (z)'
1 is within its limit by-q 9
1
- a.'
Using the movable incore detectors to obtain.a power distribution:
)
map'at any: THERMAL. POWER greater.than 5% of RATED; THERMAL' POWER.
b;- ; Increasing the measured F (z): component of:the power distributioni q
map byc3% to account:for manufacturing tolerances and'furtherfin -
i
-creasing the value by 5% to. account for measurement uncertainties.
' Verify the requirements of Specification 3.2.2 are satisfied; c.
Satisfying-the following relationship:
RTP 1
F '4(z) < p0 x K(z)~for P > 0.5 q
P x W(z).
RTP' M
1 q (zy[F l
F Q
x K(z) for P < 0.5 W(z) x 0.5 where.F (z) is:the measured F (z) increased byzthe allowances for q
TP manufacturing tolerances and measurement uncertainty,;F is the i
Fq imit, K(2) is the normalized F (z) as a' function of core height, 1
q P isithe" relative THERMAL POWER, and.W(z) is theLcycle' dependent!
3 function;that' accounts.for power distribution-transients encountered l
-during norma.1. operation.
FhTP, K(z), and W(z) are specified in:the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9.
d.
' Measuring F (z) according to the following schedule 1.
Upon. achieving. equilibrium conditions.after exceeding _by 10% or i
more of RATED THERMAL POWER, the THERMAL POWER at which F-(z)
. as'last determined,* or 9-n w
2.
'At;1 east once per 31 Effective Full Power Days, whichever' occurs first.
I
During power escalation at the beginning'of each cycle, power level may'be-1 increased until a power level for extended operation has been achieved and a
power' distribution map obtained.
l t
CATAWBAL-UNITS 1&2 3/4 2-6 Amendment No. 74 (Unit 1)
Amendmont No. 68 (Unit 2)-
j
, xx I; h *,.,V
=;
POWER DISTRIBUTION LIMITS-b.
. SURVEILLANCE REQUIREMENTS'(Continued) ec 1
e.
Withmeasurementsindicating)
Ff(z) g?
~
maximum k
over 2-K(z) a
+
f has increased'since the previous' determination of F (z) either of x
.the'following' actions shall be.taken:
1)l if (z)'.shall be increase $ by 2% over that:specified-in J
Specification l4.2.-2.2c.,or-d
..)
2)
.F (z) shall.be= measured at least once:per 7 Effective Full 3
Power Days until two successive' maps indicate that.
+
Ff(z) is not; increasing, d
maximum over z K(z)-
d 1
lf.
With the relationships specified-in Specification.4.2.2.2c. above enot being_ satisfied:
1)
' Calculate. the percent F (z) exceeds its! limit by the following; l
3 expression:
q a
i L
3 j
Fh(z)xW(z),
maximum 1 x1100 for P > 0.F i:
over z RTP
/
I p
k 4
L
[g 1
Fh(z)xW(2)-
cmaximum-
1Ix 100 for P <!0.5-l (overz-p RTP
(2)
.5
['
2)
One.of the following actions shall be.taken:
x a)
Within 15 minutes, control the AFD to within new AFD limits:
1 which are determined.by reducing the AFD limits of Specification 3.2 1 by 1% AFD for each percent F (z); exceeds-1 4
q its limits as determined in Specification 4.2.2.2f.1)i.
Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to these mod-ified limits, or.
b)
' Comply with the' requirements of-Specification 3.2.2 for F (z) exceeding its limit by the percent calculated abov.e, or n
3 c)
Verify that the requirements of Specification 4.2.2.3 for i
Base Load operation are satisfied and enter Base Load operation.
CATAWBA' UNITS 1&2 3/4 2-7 Amendment No. 74 (Unit 1).
1 Amendment No. 68 (Unit 2)
.e,
' 9:
g i
..l L "Od".
t
/
q l
. POWER DISTRIBUTION:LIMITO SURVEILLANCE REQUIREMENTS-(Continued)-
p
~ g.1 The limits specified in Specifications '4.2.2.2c., 4.2.2.2e...and 1!
4.2.2.2f., above are not applicable.in the:following' core plane J
' regions:.
L 1.c
' Lower core. region from 0:to 15%, inclu'sive' <
L 2.
_ Upper core' region from 85 to-100%, inclusive.
,1 p
conditions are satisfied:
ND* if the following.-
b l
4'2.2.3 Base' Load operation is permitted at? powers above APL h4 Prior to: entering Base, Load operation, maintain' THERMAL POWER'above.
[
a..
ND l
-APL and[less than ob aqual-to that allowed. byL Specification 4.2.2.2 i
lJ for at 1eastLthe previous 124 hours0.00144 days 0.0344 hours 2.050265e-4 weeks 4.7182e-5 months . Maintain Base Load operation
' surveillance (AFD within the-target band about the' target flux differ-3
,?
ence of: Specification.3;2.1) during this time period.
Base' Load '
d operation:is then permitted providing. THERMAL POWER is" maintained j
between APLNDi BL and APL or between APLND.and100%(whi$heveris'most N
11mitini)andFQsurveillance

ismaintainedpursuanttoSpecification-3 0 k s defined as:
4.2.2.4.
APL i
'RT F :P APLBL, minimum 7
.) x 100%:
d 0
x K(Z)
V'I 2 l-F (Z) x W(Z)BL.
~
l L
where:
F (z) is the measured F (z): increased by the' allowances for 9
TP manufacturing tolerances and measurement ' uncertainty. -
.F is:t W l
i F limit, K(z).is the normalized F (Z) as-a function of core height.1 q
q l-W(z)BL is the cycle dependent. function that accounts _ for -limited power i,
distribution transients-encountered during Base' Load 1operat' ion.'
s y
Ff,K(z),-andW(Z)BL are specifiedBin the CORE OPERATINGILIMITS f
L REPORT per, Specification 6.9.1.9.
j l
b.
During Base Load operation,"if the-THERMAL' POWER is decreased below l
ND APL then the conditions ~of 4.2.2.3afshall be satisfied before re-entering Base Load operation.
3 F
4.2.2.4 During Base Load Operation F (Z) shall be evaluated to determine if-1 F (Z)'.is within its limit by:
9
[
q j
a.
Using the movable incore detectors to obtain a power distribution HD map at any THERMAL POWER above APL b.
Increasing the measured F (Z) companent of the power distribution map i
q by 3%:to account'for manufacturing tolerances and further. increasing.-
a the.value by 5% to account for. measurement uncertainties.
Verify the requirements of Specification:3.2.2 are satisfied.
)
ND
APL is the minimum allowable (nuclear design) power level for. Base Load j.
operation in Specification -3.2.1.
[
t CATAWBA - UNITS 1 & 2 3/4 2-7a Amendment No. 74(Unit 1)
Amendment No. 68(Unit 2) 3 j
m r
.s-y, y-
. ~
r POWER' DISTRIBUTION L1MITS SURVEILLANCE REQUIRENENTS'(Continued) j q
c..
. Satisfying the.following relationship::
]
RTP a.
0 1
ND
F (Z) 5:
ifr.P>!APL p
x M -
TP where:..F(7)51s;themeasuredF(Z).. -
is the F limit..
q q
K(Z) is;the normalized F P is 4the-y
. relative 3 THERMAL POWER.'q(2) as'a' function of core height.
. W(Z) -is the; cycle dependent function that accounts forclimited power dibribution transients' encountered during, d
BaseLoad(operation.;.Fff,;'K(Z)',-andW(Z)BL are specifiedfin the '
. CORE:0PERATING LIMITS ; REPORT per Specification 6.9;1.9.
,w M
d.-
Measuring F (Z) in conjunction with' target flux difference; deter-
~
n mination according to the following schedule:
g li Prior.to entering. Base Load: operation after satisfying surveil-'
i lance 4.2.2.3 unless a' full l core flux map hasf been taken in-the;
-previous'31 EFPD with the relative thermal power.having been.
ND maintained above APL for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior:to: mapping,_ and-3, 2.
Atleastoncepor31effecAive_fullpower. days.
d e.
With measurements'iiidicating f
~Fh(5) maximum E K(z) 3 over z 4
has increased since the: previous determinationL F (Z) either of the following' actions shall be taken:
1.
F (Z) shall be increased by 2 percent over tha't specified'in 4.2.2.4c, or M
2.
F (Z) shall be measured ~at'least once per 7 EFPD uatil 2 q
successive maps indicate that-d Ff(z) maximum K(z) ) 1s not increasing.
over z f.
With the relationship specified in 4.2.2.4c above not being satisfied, either of the. following actions shall be 'taken:
1.
Place the core in an equilibrium condition where the limit in 4.2.2.2c is satisfied, and remeasure F (Z), or q
CATAWBA - UNITS.1 &'2 3/4 2-7b Amendment No. 74 (Unit 1)
Amendment No. 68(Unit 2)-
l!
k i N
l POWER DISTRIBUTION LIMITS
)
i SURVEILLANCE REQUIREMENTS (Continued) 2.
Comply with.the requirements of Specification.3.2.2 for F (2) exceeding its limit by the percent calculated with 9
the following expression:-
1 Fh(Z)xW(2)BL'))-11]x100'forP>APL ND
[(max. Lover;z of [
FhTP x:K(Z)
P O
I g.
The limits : specified in 4.' 2. 2.4c., 4. 2. 2.4e... and 4.2. 2.4f. -
y above are not applicable;in the:following. core plan regions:
l 1.
Lower ' core region 0 to;15 percent, inclusive.
2.
Upper core: region 85'to 100. percent',-inclusive.
4.2.2.5,When F (Z) is measured for reasons'other_ than meeting the requirements 9
[
of Specification 4.2.2.2 an overall measured F (z).shal1~be obtained.from a power.
j q
distribution' map.and increased by.3% to account for_ manufacturing l tolerances and further increased by. 5% to-account.for measurement. uncertainty.
i-I i
f a
t 4
k 4
~
i i.
CATAWBA - UNITS 1 & 2 3/4 2-7c Amendment No. 74(Unit.1).
Amendment No. 68(Unit 2)
Y i
,I/Ih
,n'
(
g-
-i 3
d t
c t-'
3 j
5
-t..
/
gr-ti, !
.t e
I I
....5
). r t
h THIS PAGE INTENTIONALLY! DELETED.
mj a
i
(
5
l 3
4 1
T' i.
1
i{
h:
u
'l 3,
1 5
a
. o: q i
a.
4f
\\
5.'
.r r
-7 l
l
.j
.1 CATAWBA - UNITS 1 & 2 3/4 2-8 Amendment No. 74(Unit 1)
" i
-. Amendment No. 68(Unit 2)'
t
.:p (q
..: t
POWER DISTRIBUTION LIMITS-3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR-LIMITING CONDITION FOR OPERATION' g-3.2.3 ^The combination of indicated Reactor Coolant System total flow rate and
.R shall be maintained within the region of permissible.' operation specified in the'COR5.0PERATING LIMITS REPORT (COLR) for.four loop operation.
Whail:
N.
AH a,
.R=
'FRTP[1.0 + MFAH (3.0 - P)]
THERMAL POWER if b*
P
=
RATED. THERMAL POWER 1 Ffg = 14easured values of Ffg obtained by using the' movable incore-i
~
c.
detectors to'obtain a power distribution map'.
The measured valuesof!FfH shall be used to. calculate R since the figure.
specified in the COLR includes penalties for undetected feed-water venturi fouling of 0.1% and for measurement. uncertainties of 2.1% for flow.and 4% for incore measurement of F$H' L
d.
F P= The FfHlimit at RATED..THE' MAL POWER.(RTP) specified in the R
~
COLR, and-l e.
MFAH= The power factor multiplier specified in the COLR.
APPLICABILITY:
MODE 1.
i l
ACTION:
't With the combination of Reactor Coolant System total flowirate and R within a.
.the region of restricted operation within 6' hours reduce the Power Range l
4 Neutron Flux-High Trip Setpoint to below the nominalisetpoint by the same amount (%.RTP) as the power reduction required by the figure specified in
.the COLR.
b.
With the combination.of Reactor Coolant System total flow rate and R withini L
the region of prohibited operation specified in the COLR:
1.
Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:
a)
Restore the combination of Reactor Coolant System total flow t
rate and R to within the region of permissible' operation, or b)
Restore the combination of Reactor Coolant. System total flow i
i rate and R to within the region of restricted operation and 4
comply with action a. above, or i
. CATAWBA - UNITS 1 & 2
.3/4 2-9 Amendment No. 74(Unit 1)
Amendment No. 68(Unit 2)'
C
4
. POWER DISTRIBUTION LIMITS ~
3/4.2.3 ~ REACTOR COOLANT-SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT.
CHANNEL FACTOR-LIMITING CONDITION FOR OPERATION!
0 ACTION (Continued) i c)
Reduce THERMAL; POWER to less than 50% of RATED THERMAL POWER' f
and. reduce the Power Range Neutron Flux,- High Trip Setpoint'
to less than on.equalfto 55% of RATED THERMAL' POWER within.
1 Jthe'next 4~ hours.'
2. -.
Within 24~ hours 'of initially being within the ~ region of prohibited -
' operation specified in the COLR, verify through incore flux' mapping-l.
and Reactor Coolant System total-flow rate ~ comparison that'the com-i
.bination-of R and Reactor Coolant System total flow rate are' restored.
to within the' regions of restricted or permissible' operation, 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 , 4
F
-r i
i t
1 6
1
I l
I i
i 1
i
-r j'
i l'
W
\\
CATAWBA - UNITS.1 & 2-3/4 2-9a Amendment No,' 74 (Unit 1)
(
Amendment No. 68 (Unit 2)-
~
p.'a 5
7
-e_.
.,,t i
POWER DISTRIBUTION'LIMITSi i
LIMITING CONDITION FOR OPERATION 0
q
' ACTION-(Continued)-
i
?
3.
Identify:and. correct the cause of the out-of-limit condition-prior; c
Lto. increasing THERMAL POWERJabove'the. reduced THERMAL POWER'11mit-1 required by ACTION.b.1.c) and/or.b.2., above; subsequent POWER OPERA-
1 TION may proceed

provided that'the combination'of:R'and indicated; ieactor Coolant System total flow rate art demonstrated,' through-incore flux mapping-and Reactor Coolant-Syst o total flow rate <
..1.
comparison,- to be within the regions ofs restricted or permissible' i
L operation specified in the COLR prior;to' exceeding ^the:followingu
]
E THERMAL POWER levels:
1 i
a)
A nominal 50% of RATED. THERMAL 1 POWER,1 j
b)
A nominal 75% of' RATED THERMAL POWER, and j
c) :Within 24Lhours of attaining greater.t'han'or' equal'to 95%
4 i
.of RATED THERMAL POWER.
~
-t il SURVEILLANCE REQUIREMENTS
~
4.2.3.1.. The' provisions of SpecificationL 4.0.~4"are not applicable.
4.2.3.2 The. combination of-indicated Reactor Coolant System total, flow rate' h
determined by' process computer readings or. digital voltmeter' measurement and R:
.shall be determined to be within the regions of-restricted or permissible s
t f
operation specified.in the COLR:
y
-.ll E
a.
Prior to operation above'75% of RATED THERM L POWER after each fuel
[
loading, and
,q b.
At least once per 31 Effective. Full. Power Days.
1 4.'2.3.3 ~ 'The indicated Reactor Coolant System total flow rate shall!be verified; a
to'be within the' regions of restricted or permiscible operation specified in L
the' COLR at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently.;obtained.value of R, L-obtained per Specification 4.2.3.2, is assumed to: exist.
l 4.2.3.4 The Reactor Coolant System total flow rate indicators shall be subjected' to a CHANNEL CALIBRATION ~at-least once per 18 months.
The measurement i
L instrumentation shall be calibrated within'7 days prior to the performance of-
.l l--
the calorimetric flow measurement.
l 4
b 4.2;3.5 The Reactor Coolant System total flow rate shall b5 determined by j,
precision heat balance measurement at least once per 18 months.
p
.a u
l l
L 0
L CATAWBA - UNITS 1 & 2
~3/4 2-10 Amendment No. 74(Unit 1)
L Amendment-No. 68(Unit 2) 1
m..
e f Y 1
l
,l l
i i
j e
/
t I:
THIS PAGE INTENTIONALLY DELETED.
1 1
r
\\
l
- CATAWBA - UNITS 1 & 2 3/4 2-11 Amendment No.f/4 (Unit 1)
Amendment No. 68 (Un.it 2).
l t
REACTIVITY CONTROL SYSTEMS BASES-MODERATOR TEMDERATURE COEFFICIENT (Continued) l i
involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn' j
condition and, a conversion for the rate of change of moderator. density with temperature at kATED THERMAL POWER conditions.
This value of the M00 was then transformed into the limiting End of Cycle Life (E0L) MTC value.
The 300 ppm j
surveillance limit MTC value represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the' limiting EOL MTC value.
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 remains within its limits since this coefficient changes slowly due principally to the i
reduction in boron concentration associated with fuel burnup.
3/4.1.1.4 MINIM W TEMPERATURE FOR CRITICALITY j
This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551'F.
This limitation is required to ensure:
(1) the moderator temperature coefficient is within its analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the P-12 interlock is above its setpoint, l-(4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (5) the reactor vessel is above its minimum RT temperature, t
NDT 3/4.1.2 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 transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.
With the coolant average temperature above 200'r, 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 a SHUTDOWN
[
]
CATAWBA - UNITS 1&2 8 3/4 1-2 Amendment No. {4 (Unit 1)
Amendment No. 8 (Unit 2) 1
- ~ + - - " - ~ - -
_ ~
i l
REACTIVITY CONTROL SYSTEMS i
BASES MOVABLE CONTROL ASSEMBLIES (Continued)
The control rod insertion limit and shutdown rod insertion limits are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9.
The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the l
original design criteria are met.
Misalignment of a rod requires measurement of peaking f6ctors and a restriction in THERMAL POWER.
These restrictions pro-i vide assurance of fuel rod integrity during continued operation.
In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.
l The maximum rod drop time restriction is consistent with tha assumed rod l
drop time used in the safety analyses. Measurement with T,yg greater than or equal to 551*F and with all reactor coolant pumps operating ensures that.the i
measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.
Control rod positions and ON RABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more fre-quent verifications required if an automatic monitoring m nnel is inoperable.
These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.
For Specification 3.1.3.1 ACTIONS c. and d., it is incumbent upon the plant personnel to verify the trippability of the inoperable control rod (s).
This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping l
mechanism.
During performance of the Control Rod Movement periodic test (Specifica-tion 4.1.3.1.2), there have been some " Control Malfunctions" that prohibited l
a control rod bank or group from moving when selected, as evidenced by the demand counters and DRPI.
In all cases, when the control malfunctions were corrected, the rods moved freely (no excessive friction or mechanical inter-ference) and were trippable.
This surveillance ~ test is an indirect method of verifying the control rods are not immovable or untrippable.
It is highly unlikely that a complete.
control rod bank or bank group is immovable or untrippeble.
Past surveillance-i and operating history provide evidence of "trippability".
Based on the above information, during performance of the rod movement test, if a complete control rod bank or group fails to move when selected and can be attributed to a " Control Malfunction", the control rods can be considered
" Operable" and plant operation may continue while ACTIONS c. and d. are taken.
If one or more control rods fail to move during testing (not a complete bank or group and cannot be contributed to a " Control Malfunction"),.the affected control rod (s) shall be declared " Inoperable" and ACTION a, taken.
(
Reference:
W 1etter dated December 21,1984, NS-NRC-84-2990, E. P. Rahe to Dr. C. O. Thomas)
CATAWBA - UNITS 1&2 B 3/4 1-4 Amendment No. 74 (Unit 1)
Amendment No. 68(Unit 2)
I
=
I I
3/4.2 POWER DISTRIBUTION LIMITS
)
BASES The specifications of this section provide assurance of fuel integrity during condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:
(1) maintaining the ca".culated DNBN in the core greater than or equal to design limit DNBR,during normal operation and in short-term transients, and H
(2) limiting the fission gas release, fuel pellet temperature, and cladding"
(
mechanical properties to within assumed design criteria.
In addition, limiting the peak linear power density-during Condition I events provides assurance that i
the initial conditions assumed for the LOCA analyses are met and the ECCS 1
acceptance criteria limit of 2200'F is not exceeded.
l The definitions of certain hot channel and peaking factors as used i'n I
these. specifications are as follows*
F (Z)
Heat Flux Hot Channel F4ctor, is defined as the maximum local heat.
9 flux on the surface of a fuel rod at core elevation Z divided by the H
average fuel rod heat flux, allowing for manufacturing tolerances on-
)
fuel pellets and rods; FfH 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 AXIAL FLUX DIFFERENCE ~
1 The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper n
FhTP limit specified in the CORE OPERATING LIMITS REPORT.
bound envelope of the (COLR) times the normalized _ axial peaking factor.f s not exceeded during either normal operation or in the event of xenon redistribution following power changes.
-i Target flux difference is determined at equilibrium xenon conditions.
The full-length rods may be positioned within the core in accordance with.
their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.
The value of the target flux difference obtained under these coaditions divided by the fraction of RATED THERMAL POWER is the' target flux difference at RATED THERMAL POWER for.the associated core burnup conditions. Targat flux differences for other 4
THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.
The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.
i t
L
,5 CATAWBA - UNITS 1 & 2 B 3/4 2-1 Amendment No. 74 (Unit 1)
Amendment No. 68 (Unit 2) g
+-..~,,
+,w--~-
=
(
POWER DISTRIBUT10N LIMIT _S BASFS ND At power levels below APL
, the limits on AFD are defined in the COLR, j
1-e., that defined by the RAOC operating procedure and limits.
These limits were calculated in a manner such that expected operational transients, e.g.,
ibad: follow operations, would not result in the AFD deviating outside of those limits.
However, in the event such a deviation occurs,~the short period of time allowed outside of the limits at reduced-power levels will not result in signi-ficant xenon redistribution such that the envelope of peaking factors would change sufficiently to prevent operation in the vicinity of the APLND_p,,,7 level.
ND At power levels greater than APL
, two modes of operation are permis-sible; 1) RA00, the AFD limits of which are defined in the COLR, and 2) Base t-oad operation, which is defined es the maintenance of the AFD within a COLR specified band about a target value.
The RAOC operating procedure above HD APL is the same as that defined for operation below APLND However, it is possible when following extended load following maneuvers'that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee' operation with F (z) less-than its limiting value.
To allow operation 9
at the maximum permissible value,'the Base Load' operating procedure restricts-j.
CATAWBA - UNI % 1 & 2 B 3/4 2-2 Amendment No. 74 (Unit 3) l Amendment No. 68 (Unit 2) i l
w t
POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) the indicated AfD to relatively small target band and power swings (AFD target band as specified in the COLR, APLND < p,,,7 < gptBL or 100% Rated Thermal Power.
l whichever is lower).
For Base Load operation 7 it is expected that the Units.will operate within the target band. Operation o sside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above.
To assure there is no residual xenon redistribution impact from past operation on the Base Load ND operation, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months wai'ing period at a powei Nel-above APL and allowed by RAOC is necessary.
During this time period imd changes and rod motion are restricted to that allowed by the Base Load procsdure.
After the waiting i
period extended Base Load operation is permissible.
The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at-least 2 of 4 or 2 of 3 OPERABLE excore channels are:
1) outside'the allowed AI power operating space (for RA00 operation), or 2) outside the allowed AI target band (for Base Load operation).
These alarms are active when l
power is greater than:
1) 50% of RATED THERMAL POWER (for RA00 operation), or
2) APLND (for Base Load operation).
Penalty deviation minutes for Base Load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed.
The limits on heat flux hot channel factor, coolant flow rata, and nuclear enthalpy rise hot channel facter ensure that:
(1) the design limits on peak local power density and. minimum DNBR are not exceeded and (2) in the event of.
a LOCA the peak fuel clad temperature will not exceed the 2200*F.ECCS acceptance criteria limit.
These limits are specified in the CORE OPERATING LIMITS REPORT' per Specification 6.9.1.9.
i i
Each of these is measurable but will normally only be thtermined 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:
a.
Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position; b.
Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; I
R l
.1 CATAWBA - UNITS 1 & 2 B 3/4 2-2a Amendment No. 74 (Unit 1)-
.l Amendment No. 68(Unit 2)
1 i
POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR. and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ERTilALPY RISE HOT CHANNEL FACTOR (Continued) c.
The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and l
d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
Ffg will be maintained within its limits provided Conditions a through d.
above are maintained.
As noted on the figure specified in the CORE OPERATING LIMITSREPORT(COLR),ReactorCoolantSystemflowrateandFfg may be " traded off" against one another (i.e., a low measured Reactor Coolant Sys_ tem flow rate N
is acceptable if the measured F is also low) to ensure that the calculated DNBR will not be below the design DNBR value.
TherelaxationofFfg-asa
[
function of THERMAL POWER allows changes in-the radial-power shape. for all permissible rod insertion limits.
R as calculated in Specification 3.2.3 and used in the figure specified intheCOLR,accountsforFfH lessthanorequaltotheFhP limit specified in the COLR.
This value is used in the various accident analyses where Ffg influences parameters other than DNBR, e.g.,. peak clad temperature, and thus is the maximum "as measured" value allowed.- The rod bow penalty as a function of f
burnupappliedforFfg is calculated with-the methods described in WCAP-8691,-
l Revision 1, " Fuel Rod Bow Evaluation," July 1979, and.the maximum rod bow penalty is 2. 3 DNBR.
Since the safety analysis is perforcad with plant-specific safety DNBR limits compared to the design DNBR limits, there is sufficient thermal l
margin available to offset the rod bow penalty of 2. M DNBR.
The hot channel factor F (z) is measured periodically and increased by a j
cycle ~and height dependent power factor appropriate to either RAOC or Base Load J
operation, W(z) or W(z)BL, to provide assurance that the limit on the hot channel factor, F (z), is met. W(z) accounts for the effects of normal oper-q i
etion transients and was determined from expected power control' maneuvers over the full range of burnup conditions in the core.
W(z)BL accounts for the more restrictive operating limits allowed by Base Load operation which result in I
l 1ess severe transient values.
The W(z) function for normal operation and the.
l W(Z)BL unction for Base Load Operation are specified in the CORE OPERATING-f LIMITS REPORT per Specification 6.9.1.9.
~ CATAWBA - UNITS 1 & 2' B 3/4 2-1 Amendment No. 74 (Unit 1)
Amendment No. 68 (Unit 2).
~
ri POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT,0R (Continued)
WhenReactorCoolantSystemflowrateandFfg are measured, no additional allowances are.necessary prior to comparison with the limits of the figure specified in the COLR, Measurement errors of 2.1% for Reactor Coolant System total flow rate and 4% for F have been allowed for in determination of the j
g design DNBR value.
The measurement error for Reactor Coolant System total flow rate is based upon performing a precision heat balance and using the result to calibrate the Reactor Coolant System flow ~ rate indicators.. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a nonconservative n.anner.
Therefore,'a penalty of 0.1% for undetected fouling of the feedwater venturi is included in the figure specified in the COLR.
Any fouling which might bias the Reactor Coolant System flow rate measurement greater than 0.1% can be detected by conitoring ar.d trending various 3
plant performance parameters.
If detected, action shall-be taken before per-j forming subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the Reactor Coolant System flow rate measurement or the venturi shall be cleaned to eliminate the
]
fouling.
The 12-hour periodic surveillance of indicated Reactor Coolant System 1
flow 15 sufficient to detect only flow degradation which could lead to opera-tion oatside the acceptable region of operation specified on the figure spec-ified in the COLR.
3/4.2.4 OVADRANT POWER TILT RATIO i
i l-The QUADRANT POWER TILT RATIO limit assures that the racial power distribu-l tion satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during STARTUP testing and.
periodically during power operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protectic; with x y plane power tilts.
A limit of 1.02 was selected to provide an allowance fer the uncertainty associated l_
with the indicated power tilt.
1 l
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification ano correction of a dropped or misaligned control rod.
In the event such. action does not i
correct the tilt, the margin for uncertainty on F is reinstated by reducing q
L the maximum allowed power by 3% for each percent of tilt in excess of 1.
For purposes of monitoring-QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used'to confirm that the normalized symmetric power distribution is consistent with th? QUADRANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore i
CATAWBA - UNITS 1 & 2 B 3/4'2-5 Amendment No. 74 (Unit 1)-
Amendment No. 68 (Unit 2)'
' l L
^
' ADMINISTRATIVE CONTROLS i
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)
The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active rat.erials in gaseous and liquid effluents made during the reporting period, i
The Radioacthe Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP).and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose. calculations and/or environmental monitoring identified by the land.
use census pursuant to Specification 3.12.2.
MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience.in-ciuding documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission.
Attn:
Document Control Desk, Washington, D.C. 20555, with a copy to the NRC Regional Office, no later than the 15th of.each month following the calendar month covered by the report.
5
+
CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before ecch reload cycle or any remaining part of a reload cycle for the following:
1.
Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2.
Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,
-l 3.
Control Bank Insertion Limits for Specification'3/4.1.3.6,
(
ND 4.
Axial Flux Difference Limits, target band, and APL for Specification 3/4.2.1, RTP ND 5.
Heat Flux Hot Channel Factor, F K(Z),W(2),APL and W(Z)BL forSpecification3/4.2.3,a,nd RTP 6.
Nuclear Enthalpy Rise Hot Channel Factor, F
, and Power Factor Multiplier, MFAH, limits 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 in:
1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"
July 1985 (W Proprietary).
j i
(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux l
1 u;BA - UNITS 1 & 2 6-19 Amendment No. 74 (Unit 1)
Amendment No. 68 (Unit 2)
R
~
e ADMINISTRATIVE CONTROLS i
CORE OPERATING LIMITS REPORT (Continued) j 1
Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -
Nuclear Enthalpy Rise Hct Channel FM tor.)
l 2.
WCAP-10216-P-A, " RELAXATION OF' CONSTANT AXIAL OFFSET CONTROL FQ i
SURVEILLANCE TECHNICAL' SPECIFICATION," June 1983 (W Proprietary).
I 3
(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial' Offset Control) and.3.2.2 - Heat Flux Hot Channel Factor (W(Z)' surveillance requirements for F Methodology.)
q 3.
WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL i
USING BASH CODE," March 1987, (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical _ limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are iet.
The CORE 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.
l l
l
[
l l
Amendment No. 74 CATAWBA - UNITS 1 & 2 6-19a Amendment No. 68 ((Unit 1).
Unit 2)
L
- a
.

ML20043A695

Contents

Text

Attachment:

Attachment:

Reference:

Navigation menu

ML20043A695

Text

Attachment:

Attachment:

Reference:

Navigation menu

Search