Text

Reload Rept,Catawba Unit 2,Cycle 6
	ML20126F014
Person / Time
Site:	Catawba
Issue date:	12/15/1992
From:	; DUKE POWER CO.
To:	;
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ML20126E996	List: ML20126E995; ML20126F014; ML20126F026;
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	NUDOCS 9212300078
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- _ _ - _. - _ _ _,

RELOAD REPORT Catawba Unit 2 Cycle 6 Duke Power Company Nuclear Generation Department Nuclear Engineering Section 9212300078 921215 DR ADOCK 05000414 p

PDR u

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1 1

L 1-3 9

I Contents-i i

1 Pace 1.

INTRODUCTION AND

SUMMARY

1-1 i

2.

OPERATING HISTORY.........................................

2-l' i

4 3.

GENERAL DESCRIPTION.......................................

3-1 i

4 j

4.

FUEL SYSTEM DESIGN........................................

4-1 j

4.1.

Fuel Assembly Mechanical Design...................

4-1 4.2.

Fuel Rod Design..................

4-1 l

4.2.1.

Fuel Rod Cladding Collapse..............,,-. 4 4.2.2.

Fuel Rod Cladding Stress..................

4 !

4.2.3.

Fuel Rod Cladding Strain.................... 4-2 4.3.

Thermal Design...................................

4-2

}:

4.4.

Material Design...................................

4-2

)

4.5.

Operating Experience...............................-.4-2 j

5 NUCLEAR DESIGN...............,.............

5-l' 4-1 5.1.

Physics Characteristics...........................

5-1

}

5.2.

Changes in Nuclear Design.........................

5 t' 6.

THERMAL-HYDRAULIC DESIGN................................... 6-1 l

7.

ACCIDENT ANALYSIS.........................................

7-1 i

j 8.

PROPOSED MODIFICATIONS TO LICENSING BASIS DOCUMENTS.......

8-1 J

8.1 Changes to Technical Specifications............... B-6 i

8.2 Changes to Core Operating Limits' Report.......... 8-130 l_

8.3 Changes-to the Catawba FSAR....,,...............

8-154-4 l~

9.

REFERENCES................................................

9-1 e..

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_r-m _ _ __ _ _ m List of Tables Table Page 4-1 Mark-BW Fuel Design Parameters and Dimensions 4-3 5-1 Physics Parameters, Catawba 2 Cycles-5 and 6...........

5-2 5-2 Shutdown Margin Calculation for Catawba 2 Cycle 6....... 5-4 6-1 System Uncertainties Included in the Statistical Core Design Analysis..........-..........................

6, 6-2 Nominal Thermal-Hydraulic Design. Conditions, C a t awba 2 Cy c l e 6...................................... 3 6-3 DNBR Penalties 6-4 6-4 Flcw Anomaly Peaking Penalties 6.........................

8-1 Technical Specifications Changes............

8-3 8-2 Core Operating Limits Report Changes.................._., 8 List of Ficures Figure Page.-

3-1 Core Loading _ Pattern for' Catawba Unit 2 Cycle 6......... 3-2 3-2 Enrichment and-DOC Burnup Distribution for Catawba-Unit 2 Cycle.6

....................................3-3 3-3 Catawba Unit 2 Cycle 6 Burnable Absorber and Source Assembly Locations..........................

3-4 5-1 BOC (4 EFPD), Cycle 6 Two-Dimensional Relative Power Distribution - HFP, Equilibrium Xenon.............

5-5 ii

_m m_2--__.-

.= _

l 1.

INTRODUCTION-AND

SUMMARY

This report justifies the operation of the sixth cycle of Catawba Nuclear Station, Unit 2 at the rated core power level of 3411 MWth.

Included are the required analyses as outlined in the USNRC document

" Guidance for Proposed License Amendments Relating to Refueling,' July 1975.

Cycle 6 for Catawba Unit 2'is the third Catawba cycle for which the reload fuel is supplied by B&W Puel Company (BWFC).

The incoming Batch" 8 fuel assemblies are designated as Mark-BW.

To support implementation of Mark-BW. fuel in the McGuire and Catawba ' nuclear stations, Duke Power Company (DPC) has developed new methods and models to analyze the plants during normal and off-normal operation.

The thermal-hydraulic analytical models are documented in: topical report DPC-NE-3000P (Reference 11) for non-LOCA transients and BAW-10174 (Reference 13)-for LOCA. -Portions of the analytical methodology are documented in topical report DPC-NE-3001-PA-(Reference 12) and DPC-NE-2004PA (Reference 8).

The remaining Final Safety' Analysis Report (FSAR) Chapter 15 non-LOCA system transient. analysis methodology is dccumented in U 4-NE-3002-A (Reference 16).

The FSAR Chapter 15 LOCA system transient analysis methodology is documented in Reference 13. ' Approval of these topical-reports has been completed.

Section 2 of this report is the operating history for fue)

Catawba Unit 2.

Section 3 is a general description of the reacto, e, and the fuel system design is provided in Section 4.

Reactor e u system parameters and conditions are summarized in Sections.5, 6, and 7.

Changes to the Technical Specifications, Core Operating Limits Report (COLR.), and FSAR are provided in Section 8.

All.of the accidents analyzed-in the FSAR (Reference 1) have been reviewed for Cycle 6 operation, and many of the FSAR Chapter 15 system thermal-hydraulic accident analyses sensitive to reload core physics parameters have been reanalyzed using Duke Power methodology.

These analyses are the same as those performed for the McGuire Unit 1,' Cycle B and Catawba Unit 1, Cycle 7 reloads.

Several bounding transients-were analyzed in detail to demonstrate the capability of DPC calculational techniques.

The results of these analyses were reported in DPC-ME-3001-PA.

For the other reanalyzed transients'.u -the-approved ~

methodology is documented in DPC-NE-3002-A.

.A further discussion of accident analysis is presented in Section 7.of this report. Other-reanalyzed transients.are included in Section 8 of this report.

-Amendment Number 74 (Unit 1) and Amendment Number 68-(Unit 2)'to the Catawha Nuclear Station Facility Operating License allow the removal of cycle-specific core parameter limits from Technical Specifications and require that these limits be included in a Core Operating Limits Report (COLR).

The Core. Operating. Limits Report is submitted to the NRC-uponD issuance and does not require-approval prior to implementation.

Changes to the operating limits are made via the Core Operating Limits-Report.

The Technical Specifications have'been-reviewed, and the modifications-for Cycle 6 are justified.in this report, Based <xt the-analyses performed,-it has been concluded-that' Catawba Unit 2 Cycle 6 can be safely operated at a core power level of 3411 MWth*

l-1

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2.

OPERATING HISTORY 1

i The current operating cycle for Catawba Unit 2 is Cycle 5, which achieved criticality on December 22, 1991 and reached 100% full power i

on December 28, 1991.

Cycle 5 is scheduled to shut-down in January 1993 af ter 3'15 EFPD.

Catawba Unit 2 Cycle 5 and previous cycles operated entirely with fuel assemblies of Westinghouse design.

j Catawba Unit 2, Cycle 6 is the first Catawba Unit 2 reload to-contain a-full reload batch of Mark-BW fuel assemblies (FAs).

Catawba Unit 1 has i

had two batches of Mark-BW fuel assemblies, the first installed for Cycle 6 and the second for Cycle 7_.

Catawba Unit 2,' Cycle 6 is s

scheduled to start up on April, 1, 1993 at a rated power level of 3411 t

MWt and has a-design cycle length of 380 EFPD. -No operating anomalies l

have occurred during previous cycle operations that would adversely affect fuel performance in Cycle 6.

1 I

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1 3.

GENERAL DESCRIPTION i

The Catawba Unit 2 reactor core is described in detail-in Chapter 4 of j

the FSAR (Reference 1)..

The core consists of 193 assemblies, each of which is a 17-by-17 array containing 264 fuel rods, 24 guide tubes, mnd 1 incore instrument tube.

There are 117 burned FAs in the core, all of l-the Westinghouse Optimized Puel. Assembly design, and 76' fresh FA's consisting of the Mark-BW design (Reference 2) : The fuel rod outside 1-diameters are 0.360 and 0.374 inch, and the clad thicknesses are 0.0225 i

and 0.024 inch for the OFA and Mark-BW designs, respectively.- 1he j

Mark-BW fuel consists of dished end, cylindrical pellets of uranium i-dioxide, (See Table 4-1-for-data!. -The nominal-fuel loadings are 423.5 kg of uranium per fuel assembly for the Westinghouse fuel in batches-a

?

5A, 6A and 7A; and 456.2 kg of uranium per fuel assembly for-the Mark-BW fuel in batch BA.

The initial enrichments of batches 5A, 6A and 7A 3

were 3.60, 3.50, and 3.75 wt% U235 The design enrichment of the fresh i-batch BA (Mark-BW) is 3.75 wt% U235-i l-The 8 batch SA, 33 batch 6A, and 76 batch 7A assemblies will be-i shuffled to new locations.

One batch 6A;FA will be inserted into the 1

center-assembly location. The 76-fresh batch 8A assemblies will be loaded into the core in a symmetric checkerboard pattern.

Figure-3-1 shows the locations of the fresh fuel: assemblies and the previous cycle

.1-location of the burned fuel assemblies.- Figure 3-2 is a quarter core l

map showing the burnup and region number with corresponding initial enrichments of each assembly at the beginning of. Cycle 6.

4 l

Cycle 6 will be operated in a feed-and-bleed mode.

Core' reactivity is.

i controlled by 53 rod cluster control assemblies (RCCAs). 640-Mark-BW burnable absorbers, and soluble boron shim.

Figure.3-3-shows the Cycle 6 fresh fuel locations with the Mark-BW BPRA clusters =and number of pins (loaded with 3.0 wt% B C in Al O ) in each location.

4 23 t

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3-1 i

FIGURE 3-1 CORE LOADING PATTERN FOR CATAWBA UNIT 2 CYCLE 6 PREVIOUS CORE LOCATIONS REGION NUMBERS-J-04 Feed F-05 F-15 K-05 Feed G-04 7

8 7

7 7

8 7

L-04 K-13 Feed J-14 Feed H-09 Feed G-14 Feed F-13 E-04 7

7 8

7 8-7 8

7 8

7 7

M-05 H-05 Feed B-05 Feed D-13 H-15 M-13 Feed P-05 Feed E-08 D-05 7

7 8

7 7

7 8

7-8 7

7 C-06 Feed B-10 Feed G-10 Feed H-03 Feed J-10 Feed K-14 Feed N-06 7

8 6

8 7

M-07 Feed L-14 Feed R-11 Feed P-12 K-15 B-12 Feed L-01 Feed E-14 Feed D-0?

7 8

5 8

6 7

6 8

5 8

7 8

7 Feed B-07 Feed F-09 Feed J-01 Feed L-15 Feed A-07 Feed K-09 Feed P-07 Feed 8

7 8

6 8

5 8

6 8

6 8.

7 8

L-10 Feed C-12 Feed D-02 Feed R-07, Feed G-01 Feed M-02 Feed N-12 Feed E-10 7

8 7

8 6

l 8 6

8 6

8 7

A-O' G-08 A-08 N-08 A-10 A-11 Feeo B-06 Feed R-05 R-06 C-08 R-08 J-08 R-10 7

7 7

6 7

5 8

6 8

5 7

6 7

7 7

L-06 Feed C-04 Feed D-14 Feed J-15 Feed A-09 Feed M-14 Feed N-04 Feed E-06 7

8-7 8

6 8

-6 8

7 8

7-Feed B-09 Feed F-07 Feed R-09 Feed E-01 Feed 0-15 Feed K-07 Feed P-09 Feed 8

7-8 6

8 6

8 5-81 6

8 6.

8 7

-8 M-09 Feed L-02 Feed E-15 Feed P-04 F-01 B-04 Feed A-05 Feed E-02 Feed D-09 7

8 7

8 5

8 6

7 6

8-5 8

7-8 7

C-10 Feed F-02 Feed C-06 Feed H-13 Feed J-06 Feed P-06 Feed N-10 7

8 6

8-7 M-11 L-08 Feed B-11 Feed D-03 H-01 M-03 Feed P-11 Feed H-11 D-11 7

7 8

7 7

7 8

7 81 7

7 L-12 K-03 Feed J-02 Feed H-07 Feed G-02 Feed F-03 E-12 7-7 8.

7

-8.

7 8

-7 7

J-12 Feed F-11 K-01 K-11 Feed G-12 7

8-7 7

7 8

71 R

P N

M L.

K J

H G

F-E D

C B

A Z-ZZ CYCLE' 5 LOCATION YY REGION NUMBER 3-2

. ~ -.

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l FIGURE 3-2 i

ENRICHMENT AND BOC BURNUP DISTRIBUTION FOR CATAWBA 2 CYCLE 6 i

I H

G F

E D

C B

A-i 32359.1

.0 30617.5 12374.2 30334.0 13196.1 19342.0 12374.2 j

8 33803.6

.0 31884 8 16850.4 16850.4 17456.0 20531,8 16850.4 1

6A 8A 5A 7A 6A 7A 7A' 7A

.0 27803.5

.0 27963.7

.0 17603.9

.0 19729.6 i

i,.

9

.0 30897.8

.0 33401.6

.0 20787.6

.0; 21018.6 8A 6A 8A 6A BA 7A 8A

-7A 30617.5

.0 27803.5

.0 29046.5

.0 18824.9

.0 1

10 31884.8

.0 30897.8

.0 33395.0

.0 20794.5

.0.

4 5A BA 6A 8A 6A 8A 7A 8A i

12374.2 27963.7

.0 30617.5

.0 16960.3

.0 20161.3 i

11 16850.4 33401.6

.0 32466.5

.0 20604.2

.0 21234.5 j

7A 6A 8A 5A BA 7A BA 7A 30334.0

.0 29046.5

.0 32359.1

.0 20413.2 4

12 34090.8

.0 33395.0

.0 36717.9

.0.

21436.7 j_

6A BA 6A BA 6A 8A 7A 1

13196.1 17603.9 0

16930.3

.0 20054.3 20573.7 i

13 17456.0 20787 6

.0 20604.2

.0.

21143.9 21735.3 j

7A 7A BA 7A 8A

~7A 7A i

19342.0

.0 18824.9

.0 20413.2 20573,7 Average i

14 20531.8

.0 20794.5

.0 21436.7 21735.3 Maximum i

7A 8A 7A BA 7A 7A Region #

{

12374.2 19729.6

.0 20161.3 15 16850.4 21018.8

.0~

21234.5 i

7A 7A BA 7A i

i DIRICHMENT CYCLES NUMBER OF BOC BURNUP REGION w/o U-235 BURNED ASSEMBLIES FffiD/MTU i

5A 3.60 3

8.

30618 6A 3.50 2

33 29141 1

7A 3.75 1

76-18201 i

8A 3.75 0

76 0

CORE 193 13419 i

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FIGURE 3-3 l

CATAWBA UNIT 2 CYCLE 6 8 -

BURNABLE ABSORBER AND SOURCE ASSEMBLY _ LOCATIONS i-I 1

0 0

t 2

4-12 12 4

i-I 3

4 16 16 4

j 4

4 12 12 12 12 4

i 5

4 12 8-8 12 4

i 6

0 16 8

8 8

8 16 0

7 12 12 8

8 8

12 12 t

8 SS 8

8 SS l

9 12 12 8-8 8

12 12 i

10 0-16 8

8 8

8 16 0

j 4-11 4

12 8

8 12 i

l 12 4

12 12 12~

12 4

i 13 4

16 16 4

2 14 4

12 12 4

1 15 0

0 1

R P

N M

L K

J H

G F

E D

C B

A l

UUMBER OF NUMBER OF MkBW-BP PINS / ASSEMBLY BACKPLATE ASSEMBLIES 4

16 8-20 12 24 16f 8-

. Total =-640 pins Total = 68 backplates i

I SS~= Secondary Source location All BP Pins are 3.0 wt% B C-Al 01 MkBW BP's.

4 2

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m-v.,..n,----

4.

FUEL SYSTEM DESIGN 4.1 Fuel Assemb]v Mechanical Desian The Catawba 2 Cycle 6 core will include 76 fresh Mark-BW fuel assemblies with an enrichment of 3.75 wt %U235 The re-inserted fuel assemblies in Cycle 6 will be 117 Westinghouse eptimized fuel assemblies.

The Mark-BW 17 x 17 Zircaloy spacer grid fuel assembly is similar in design to the Westinghouse standard fuel assemb]y, Reference 2.

The fuel rod outer diamete and guide tube top section, dashpot diameters, and instrument tube diameter are the same as the Westinghouse standard 17 x 17 design.

The unique features of the Mark-BW design include the Zircaloy intermediate spacer grids, the spacer grid restraint system, and the use of Zircaloy grids with the standard lattice design.

Mark-BW fuel design dimensions and parameters for Catawba 2 Cycle 6 are listed in Table 4-1, 1

i 4.2 Fuel Rod Desion Duke Power Company has performed generic Mark-BW mechanical analyses using the approved methodologies described in Reference 3. The generic analyses envelope the Cycle 6 design as discussed below.

4.2.1 Fuel Rod Cladding Collapse The fuel rods were analyzed for creep collapse using the CROV computer code, Reference 4, and the methodology described in Reference 3.

Internal pin pressures and clad temperatures used in CROV were calculated using the TACO 2 computer code, Reference 5.

A conservative power history which envelopes the predicted peaking for the Catawba 2 Cycle 6 fuel was analyzed.

The collapse time was conservatively determined to be greater than the maximum predicted residence time for the Mark-BW fuel (Table 4-1).

4.2.2 Fuel Rod Cladding Stress As described in Reference 3, Duke Power Company has performed a conservative generic fuel rod cladding stress analysis using the ASME pressure vessel stress _ intensity limits as guidelines.

The maximum cladding stress intensities were shown to be within the ASME limits under all loading conditions.

The generic Mark-BW cladding stress analysis includes the following conservatisms:

Conservative cladding dimensions, e

High external pressure.

e Low internal pin pressure, 4

e High radial temperature gradient through the clad.

e 4-1

4.2.3 Fuel Rod Cladding Strain 4

Diametral cladding strain resulting from a local power transient is limited to 1.01.

A generic cladding strain analysis was performed using TACO 2 to determine the maximum allowable local power change that the fuel could experience without exceeding the 1.0% limit.

The maximum calculated local power change resulting from a worst case core maneuvering scenario was compared with the maximum allowable power change.

This comparison demonstrated that margin exists to the 1.0%

strain limit.

4.3 Thermal Desinn The thermal performance of the Mark-BW fuel assemblies was evaluated using TACO 2 with the methodology given in Reference 3.

The nominal fuel parameters used to determine the generic Jinear heat rate to centerline melt (LHRTM) limits are given in Table 4-1. The LHRTM analysis included the following bounding conservatisms:

Maximum gap based on as-fabricated pellet and clad data.

4 Maximum incore densification based on resinter test results.

The maximum predicted Mark-BW assembly burnup at EOC 6 (in Batch 8) is 18,992 MWD /MTU and the maximum predicted fuel rod burnup (in Batch 8) is 20,230 MND/MTU.

The fuel rod internal pressure has been evaluated for the higher

burnup rod using TACO 2 and a conservative pin power i

history.

The..taximum internal pin pressure is less than the nominal Reactor Coolant System pressure of 2250 psia.

4.4 Material Desian i

The Mark-BW fuel is not unique in concept, nor does'it utilize different component materials.

Thus, the chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the fresh fuel is identical to that of the present fuel.

4.5 Doeratina Exnerience Experience with the Mark-BW 17 x 17 fuel assembly design started with the irradiution of four lead assemblies in McGuire 1 Cycle 5.

McGuire 1 Cycle 7 was the third cycle of irradiation for three of the assemblies and the maximum predicted assembly burnup is 42,756 MWD /MTb.

The lead assemblies were examined after each cycle and the fuel assembly bow, twist, growth, and holddown spring' set were all within nominal bounds.

Four other Mark-BW lead assemblies underwent their first cycle of irradiation in Trojan Cycle 13.

l Catawba 2 Cycle 6 will be the fifth complete reload batch of Mark-BW 17 x 17 fuel.

The first complete reload batch finished operation in Catawba 1 Cycle 6 in June 1992.

The second, third and fourth batch are l

operating in both McGuire units and in Catawba 1 Cycle 7.

1 1

I 4-2

Table 4-1.

Mark-BW Fuel Design Parameters _and_ Dimensions.

- Batch B'-

Nominal fuel rod-OD, in.

' 0.374-Nominal fuel rod ID, in.

0.326 Nominal active fuel length, in.

144,0*.

Nominal' fuel pellet OD, in.

0.3195 Fuel pellet initial density,_% TD-96.0' 3.75 Initial fuel enrichment, wt. %_U235 Estimated residence time EOC 6,=EFPH 9,120 Cladding collapse-time, EPPH-

>18,100-Nominal linear heat rate (LHR)i kW/f t-5.43

' Minimum LHR to melt, kW/ft-

_ deg F_

_ 1360' Ave. fuel temperature O' nom. LHR, 0-1000 MWD /MTU 21.5

_ >.1000-MWD /MTU 21,8

_.t t

y e

4-3 ec-1p*-

y w4 F***

ON-P 4

7 W

-'T'W"j-@W-+ e y

-p=t www v

?gWtpeve*

v w

=y(- - 'u p 4-T Mr g ' rW 'p w $ y $

+4

M - TrM -p - * * - g g -" g % p - gy4 y

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1 5.

NUCLEAR DESIGN i

j 5.1 Physics characteristics Table 5-1 provides the core physics parameters for Cycles 5 and.6.- The values for Cycle 6 were generated using-the methodology described in I

DPC-NF-2010A (Reference 6) and are valid for the design cycle length (380 EFPD i-10 EFPD).

The values for Cycle 5 were generated by.

j-Westinghouse.

Figure 5-1 illustrates a representative relative power distribution for the beginning of Cycle 6 at full power. 'This case was

{

calculated as part of the design depletion using-the PD007 methodology as described in DPC-NF-2010A-(Reference 6),

This case' assumed equilibrium xenon and rods in the All Rods Out (ARO) positit.4.

t l

During verification of the control rod insertion limits specified in the COLR, calculated ejected rod worths and.their adherence to j-acceptance criteria were considered.

The_ adequacy of the shutdown margin with Cycle 6 stuck rod worths is demonstrated in Table 5-2. The j

shutdown margin calculations include a 10% uncertainty on available rod j

worth. The shutdown margin calculation at-the end of Cycle 6 was i

analyzed at 390 EFPD; i

l t

5.2.

chances in nuclear Desian i_

No core design changes have been implemented in Cycle 6 which will i

impact the nuclear design parameters. The Cycle 6 physics parameters l-appearing in this report were calculated with the PD007 and EPRI-NODE-P codes.

These codes and methods were approved by the NRC as: documented

=

in Reference 6.

The PDQ07 calculations were performed in two-dimensions; the EPRI-NODE-P calculations _were performed in three dimensions. The Reactor Protection System -(RPS) limits =and operational-limits for the core-were verified by analyses for this-fuel cycle-using i

methodology approved by the NRC in Reference 7 and are provided in the Technical. Specifications and the COLR, Revisions to these documents _for 1

Cycle 6 are presented in Section 8.

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Table 5.1 Physics Parameters (a) Catawba 2 Cycles 5-and 6-i j

Cvele 5 cvele 6 i-Design cycle length, EFPD:

375-380-i i

Design cycle burnup, MWD /MTU 15650 -

15390; 3

4, i-Design average core burnup - EOC, MWD /MTU-28668~

28831

}

Design initial core loading,-MTU

- 81'.73 5$ 1 84.2207 f

Critical boron - BOC,ppmb, no Xe(D) i HZP,-ARO 1621 1713 l

- HFP, ARO :

1502 1569 s

Critical boron - EOC,ppmb HZP, No Xe, ARO 489 525

}

HFP, Eq Xe, ARO O

O-1 Total Control Rod Worths - HZP, pcm 6996 l'

BOC (c) 6700:

EOC 69B1 7495-l Max ejected rod worth (d)

HZP, pcm

<780-373 DOC (c)

- <900

- 552-EOC i

Max stuck rod worth - HZP, pcm 830 1194-i I-BOC (c)

EOC 850 1202.

l-4 4-Powet deficit - HZP-to HFP, pcm

!~

-1560

- -1690

[

BOC(C)

EOC

-2920

-3018 Doppler coeff - HFP, pcm/0F BOC

-0.91

-1.16 EOctcfoXe i

eq Xe

-2.90-

-1.45 i

j Moderator coeff - HFP, pcm/*F.

-2.93-BOC n

- -3.56 EOC c)o Xe i

eq Xe,'O PPMB

-33;06

-32.50 1

l-j; Boron worth - HFP, pcm/ppmb E

(c)

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52-

=m-+-

e - - *

m5 es - E-e u w v-

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-..,,.-ec.ar

-w+ew,,

a w--

y arr w= *-

vir v+

w-w* =ryw.y y

,, r, y -e vym-g,

w ww - + ~ ~ 9

ywr-< (fv

-m mw

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TabAo 5.1 Physics Parameters (a) Catawba 2 Cycles 5 and'6 (cont) 1 Cycle 5 cvele 6 j.

Equilibrium Xenon worth _- HFP, pcm f

BOC (4 EFPD) 2654 2604 l

EOC 2990 2816' i

Effective delayed neutron' fraction - HFP 4

BOC 0.00609 0.006242 EOC

>0.00440-0.005228 L

(a)

Cycle 5 values obtained from Westinghouse analyses and cycle 6 l

values were obtained from Duke Power company analyses, j

(b)

HZP denotes hot zero power (core average 5570F Tavg); HFP denotes hot. full power (590.80F vessel Tavg).

3; (c)

EOC physics parameters calculated at design EOC plus 10 EFPD.

i t

(d)

Ejected rod' worth for banks D, C, and B inserted to HZP RIL.

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Table 5-2.-

Shutdown Margin Calculation for-Catawba 2 Cycle 6 Control Rod Worth BOC (PCM)

EOC(a)(PCH)

1. All rods inserted (ARI), HZP 6996 7495

2. ARI less most reactive stuck rod, HZP 5802 6293-

3. Less 10% uncertainty 5222 5664 Required Rod Worth

4. Rod insertion allowance (RIA) (b) 246-342

5. Power defect, HFP to HZP(b)

-1992-3320-1 6, Shutdown margin (total available worth 2984 2002 minus total required worth)-

t j

NOTE:

Required shutdown n.argin is 1300 PCM.

(a)

EOC physics parameters calculated at 390 EFPD, ~ i e., ds 'gn.EOC l --

plus 10 EFPD.

I (b)

The rod insertion allowance and power defect include penalties to -

account for the effects of transient xenon conditions.

i-i i

i l

l t--

V l

i i

i l'

i I

l I' 4 E'

..---,,-,J,-

..,4~.,~

_,h._,

%,n s

..s--_[--..,---,L..,,,-

y.,.

-n

., [,

w

-. + - +,,, -,,,,.

--,-7

,+,.-.,-,ra

~.

1 1

I Figure 5-1: BOC (4 EFPD), Cycle 6 Two-Dimensional Relative Power Distribution - HFPi Equilibrioum Xenon H

G F

E D

C B

A 0.9329 1.2658 0.9602 1.1429 0.9419

-1.2488 1.0444 0.6427 l

8 0.9460 1.3456 1.0004 1.2410 0.9611

-1.3421 1.1829 0.9584 i

i 1.2658 1.0063 1.2778 0.9751 1.2329 1.2135 1.1550 1.6180 9

1.3456 1.0775 1.3646 1.0887 1.3730 1.3149 1.3682 0.9024 f

0.9602 1.2778 1.0178 1.2759 0.9876 1.@580 1.0R89 0.7740 10 1.0004 1.3646 1.0796 1.3638 1.0668 1.3828 1.2459 1.0910 i

1.1420 0.9751 1.2759 0.9577 1.2251 1.2040 1.1410 0.4549 1

11 1.2410 1.0887 1.3638 1.0083 1.3686 1.3312 1.3799 0.8203 i

4 i

0.9419 1.2329 0.9876 1.2251 0.9121 1.2017 0.6499 12 0.9611 1.3738 1.0668 1.3686 0.9856 1.3729 0.9876 1,2488 1.2135 1.2580 1.2040 1.2017 0.7361 0.3372 13 1.3421 1.3149 1.1828 1.3312 1.3729 0.9976 0.6780 l

1.0444 1.1550 1.0889 1.1410 0.9466 0.3372 i

14 1.1829 1.3682 1.2459 1.3794 0.9876 0.6780 1

0.6427 0.6180 0.7740 0.4549 l

15 0.9584 0.9024 1.0910 0.8203

?

4 4

f 1

s I

XXXX P(AVG)

YYYY PEAK PIN 4

A A

S-S n

e

6.

THERMAL-HYDRAULIC DESIGN The generic and cycle-specific analyses supporting Cycle ~6 operation were performed'by Duke Power Company using the methodology described in Reference 8.

Cycle 6 is the first-Mark-BW transition cycle-for Unit-2 and is. analyzed using Duke's Statistical Core Design (SCD) methodology.-

Uncertainties on parameters that affect DNB performance'are statistically combined to determine a Statistical DNBR limit! (SDL).

Using the BWCMV correlation, Reference 9, a generic SDL of 1.40 was-calculated using a set of generic uncertainties given in Reference 8.

The system parameter uncertainties used in Reference 8 and given in Table 6-1 bound the uncertainties specifically calculated for. Catawba.

Reactor core safety limits for Cycle 6 are based on a full Mark-BW core and a design FAH of 1.50.

The Cycle 6 nominal thermal-hydraulic design-conditions are given in Table 6-2.

The Mark-BW fuel assembly was designed to be hydraulically compatible with Westinghouse Optimized Fuel-(OFA). -BWFC has performed a series of flow tests to verify the compatibility of the two designs.

The tests showed that the total pressure drop across the OFA fuel is=2~.4--% higher' than the pressure drop across the Mark-BW fuel, Reference 10; A

generic transition core analysis was performed to determine the DNBR impact of this difference.

Since the Mark-BW fuel has a lower overall pressure drop than the OFA design, a Mark-BW-assembly-in a mixed core will tend to have more flow.

through it and consequently more DNB margin than the same assembly in an all Mark-BW core.

Conversely, flow will be forced out of the OFA fuel in a mixed core; thus, the need to calculate a-DNBR penalty for the OFA fuel ~.

A generic transition core DNBR penalty was determined by modeling a conservative core configuration with one OFA1 assembly as the hot-assembly.

The rest of the core was modeled as Mark-BW' fuel. -A number of statepoints and peaking conditions were analyzed, yielding a maximum DNBR penalty of 3.8 % for the OFA fuel.

An anomalous flow condition has been observed in several Westinghouse.

plants, including both the Catawba units.

The anomaly is a vortex that

.forLa in the lower internals and re-distributes the~ flow into~the core.

The anomaly behavior was categorized based'on measured plant data and the impact on:DNBR'in the core evaluated.

As a result of.the anomaly, a penalty has been assessed to account for periods =during which the flow re-distribution occurs. -This penalty is in--terms of both'a' peaking penalty:and a DNBR penalty'and is applied to both the Catawba units.

~

To provide design flexibility, margin is added to the SDL to' determine a design DNBR limit (DDL).

For.the generic Mark-BW-and Catawba =2 Cycle 6 analyses, the-DDL is 1.55 (10.7 % ' margin above the SDL).

The DNBR penalties, such as the OFA transition core penalty, that-must be-assessed against the margin are given in Table-6-3.

6-1

__.____.m,__.,.

o______,_,.

.,.m 1

i ii i

4-j L

i 1

' l 1.

1 Table 6-1 l

System Uncertainties Included in the.

i Statistical Core Design Analysis n

l-a l

Reference 8

]

t l

Parameter Uncertaintv Distribution 1.

Core power

+/- 2-%-

Normal i

. Normal RCS flow

+/- 2,2 %

l Core bypass flow-

-+/- 1.5 %

Uniform i

Pressure

+/- 30 psi

_Uniforn l

Inlet' temperature

+/- 4-deg F Uniform i

y-1 I,

3 i

i 5-t 9

3 A

i 6-2 e

i 4

,---...,,,,c

.... -. ~... -. -.. - -. - -. -. -. ~.. - -,....... -. - -.... ~. - -.... -.

1 1

i ip i

Table 6-2.

Nominal Thermal-Hydrauliic Design Conditions Catawba 2_ Cycle 6 i

i i

5j-Core power, MWt 3411 4

I Core _ exit = pressure, pria 2280-1 j

Vessel ave. temperature, Deg F 590.8 T

(_

RCS flow,.gpm 385,000 f

Core bypass flow, %

7,5

.I Reference design FAH' 1.50

)

j Reference design axial-shape

-1.55' Cosine

~-

l CHF correlation

BWCMV i

l Statistical DNBR limit 1.40 2

4 1.55 Design DNBR limit 4

?:

4 i

f e

1 i

4 j.

i

~.

4 J

4 k

j 6-3 1

1

- ~.

..=. - - -. -... -

. ~..

l i

i Table 6-3.

DNDR Penalties i

1 i

Statistical DNBR limit 1.40 Design DNBR limit 1.55 DNBR margin 10.7 %

i i

i-j DNBR Penaltv Mark-BW QE6 Transition core 0%

3.8 %

Instrumentation / hardware 5.6 %

2.8 %

Rod bow 0%

3.5 %

l Flow anomaly 0.5 %

D 1.1 Total DNBR penalty 6.1 %

10.6 %

l l

Available DNBR margin 4.6 %

0.1 %

f l

i i

1 a

i i

i j

5 k

F i

l.

L 64 I

i

Table 6-4.

Flow Anomaly Peaking Penaltien The following penalties are applied'to.the Catawba maximum. allowable.

total peaking limits to account for the'RCS flow-anomaly. The

- penalties apply to all~ peak. magnitudes.

I Position-Penalty X/L

% Peak 0.01 1.5

.0.1 1.5 0.2

.1. 2 0.3 0.9 0.4 0.6 0.5 0.3 0.6-1,0-0.0-i.

8 a

i

(

k l

i

[

i l

t i

l-i t

65 L

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

~.

7 ACCIDENT ANALYSIS In order to determine the effects of this reload and to ensure that the thermal performance during hypothetical incidents is not degraded, each FSAR accident analysis sensitive to reload core physics parameters has been evaluated.

For the following FSAR Chapter 15 accidents, the licensing basis has been revised to reflect reanalysis by Duke Power Company of the thermal-hydraulic system transients:

e Steam line break 4

Turbine trip

- 1 e

e Feedwater line break

. Partial loss of forced reactor coolant flow Complete loss of forced reactor coolant flow e

Reactor coolant pump locked rotor i

e Uncontrolled bank withdrawal from subcritical e

or low power startup condition Uncontrolled bank withdrawal at power Dropped rod / rod bank e

Statically misaligned rod e

Single rod withdrawal e

Rod ejection j

e Steam generator tube rupture e

The analytical models and methodology for the statically misaligned rod accident are provided in approved topical reports, References 6, 8, and 16.

For each of the remaining events, a single, generic, system thermal-hydraulic analysis is performed which bounds both Catawba Units 1 and 2, and McGuire Units 1 and 2.

Since a single set of generic analyses has been performed for these events, the results for Catawba are identical to those submitted in the approved McGuire 1 Cycle 8 (Reference 17), McGuire 2 Cycle 8 (Reference 18) and Catawba 1 Cycle 7 s

(Reference 19) reload reports.

The Catawba 2 Cycle 6 reload core physics parameter values have been reviewed with respect to the assumptions used in these analyses.

The analysis methodology for these events, except for the steam line break, the dropped rod / rod bank, and the rod ejection events, has been approved in References 11 and 16.

A minor change has been made to the operator action time value of 120 seconds presented in the feedwater line break analysis, Section 3.4J2.4 of Reference 16.

This is a conservative change in the value of the input assumption, and is not a change in the methodology.

The results of the analysis are within all acceptance criteria.

The analysis methodology for the steam line break, the dropped rod / rod bank, and the rod ejection events has been approved in References 11 and 12.

For the remaining FSAR Chapter 15 system thermal-hydraulic accident analyses sensitive to reload core physics parameters, e.g. LOCA, the current licensing basis is being retained.

In addition, the post-LOCA subcriticality evaluation and the boron precipitation evaluation have been performed by Duke Power Company as described in Chapter 15 and Chapter 6, respectively, of the Catawba FSAR, Reference 1.

The Catawba 2 Cycle 6 parameter values have also been reviewed with respect tua the assumptions used in the subcriticality analysis.

The radiological consequences for the following events are reanalyzed-due to differences between the Mark-BW fuel and OFA fuel fission 71 I.

product core inventorien, changes in the thermal-hydraulic analysis results, and changes in the dose analysis methodology.

Reactor coolant pump locked rotor o

Single rod withdrawal e

Rod ejection e

The above radiological consequence analyses are applicable to Catawba 2 Cycle 6.

These dose analyses and resulting FSAR changes were submitted in the approved Catawba 1 Cycle 7 reload submittal (Reference 19),

Catawba 2 Cycle 6 reload core physics parameters were found to be bounded by the accident analysis assumptions for all accidents which are sensitive to core physics parameters, thus demonstrating conservative results for the operation of Catastba 2 Cycle 6.

7-2

n.-.,

~. -. -

.n_

bi 4

i

?.

8.

PROPOSED MODIFICATIONS TO LICENSING BASIS DOCUMENTS Revisions to the Technical Specifications and Core Operating Limits

{

Report-(COUR) have been proposed for Cycle 6 operation to accommodate i

the influence of.the Cycle 6 core design on power peaking, reactivity, L

and control rod worths.

The Technical Specification limits and COLR limits also reflect changes in reload analysis methodology beginning 4

with this coro.. The Cycle 6 design analysis basis includes a low-leakage fuel cycle design and a mixed core containing both B&W Mark-BW-and Westinghouse OFA fuel assemblies.

I j

A cycle specific power distribution analysis of the-final core design-i j

was conducted to generate the f(AI) limits for the Overpower AT and-i Overtemperature AT trip functions and the Limiting Conditions _for l

Operation (centrol bank insertion and axial flux dif ference).

The'AT i

limits preserve the centerline' fuel melt and steady-state DNBR limits.

i The Limiting conditions for Operation preserve the maximum allowable i

LOCA and initial condition DNB peaking limits, ejected rod worth reactivity limits, and the shutdown margin reactivity limit.These i

limits were developed based on the NRC-approved methodology described-in Reference 7.

A peaking penalty for quadrant power tilt was taken in j

the analysis'so that_the resulting limits accommodate quadrant power tilt ratios up to a value of 1.02.

j The maximum allowable LOCA peaking limits shown in Figure 4 of the COLR-are based on the BNFC ECCS evaluation (References 13 and 14). A.

composite K(Z) limit'was developed based on both-large and-small break 1

analyses.

Separate composite limits applicable to Mark-BW and OFA fuel j

were used in the power distribution analysis, and are specified in the COLR.

These limits were used directly in determination of the control I:

rod insertion and axial flux difference operating limits given in Technical Specifications 3.1.3.6 and 3.2.1.

Technical Specification 3.2.2 provides the nuclear heat flux hot channel (Fo) peaking limit, i

TheinitialconditionDNBmaximumallowablepeaking(MAPIlimitsshown i

I in Table 4 of the COLR are based on core reference design peaking i

factors. -The MAP limits provide allowable combinations of peaking.

l factors that preserve DNER performance equivalent-to the design power

[

- distribution for _ a limiting loss ' of coolant' flow transier.t.

Tl'e 1

-initial condition MAPS are used as described in Refer'enceJ7 to calculate DNB peaking margins-for. determination,of.the control rod-position and axial flux-_ difference operating limits given.in Technical

[

Specifications 3.1.3,6 and'3.2.1.

Technical Specification 3.2.3

[

provides the nuclear enthalpy rise hot channel-(Fag)-peaking limit.

I The methodology for surveillance for the= core hot channel peaking factors is-described in Reference 7.

In this application of the l-methodology, peaking margin calculations are. performed whenever an incore flux map'is taken for surveillance monitoring.

Specifications 4.2.2 and.4.2.3 have been. written in a form that-l,

_provides this capability, and.the parameters required by this application of core monitoring are provided in the COLR. _The core-operating limits are provided in accordance with'NRC Generic Letter.88-16 and Technical Specification 6.9.1.9.

Table 8-1 lists the Technical Specification changes required for' Cycle 6, and these changes are 1

identical to those submitted in the approved Catawba 1 Cycle 7 reload y

a L

81

. m

subnittal (Reierence 19), except those identifiod by an asterisk in Table B-1.

Table 8-2 lists the changes to the Core Operating Limits Report.

These changes are being submitted to the NRC under separate covers.

Parametero related to monitoring the core power distribution are defined in Reference 7, and are used by the plant computer software.

These parameters'will be supplied for inclusion in the COLR.

Based on the analysis and revisions to the Technical Specifications, COLR and FSAR described in this report, Cycle 6 of Catawba Unit 2 will operate within the 10CFR 50.46 ECCS acceptance criteria and within the thermal design criterit.

The following pages contain the required Technical Specification, COLR and PSAR revisions.

I e

' I 82

-.___.._.m__.,.____.-.

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

-.-______..._-_..-.m___.__.___________.._.

i I

8 1

a I

]

l I

Table 8-1 Technical Specification Changes i

I I

j Foecification Descriotion of Chance i

l 2 l+1 decreased Fag for Mark BW fuel changed CHF correlation reduced RCS minimum flow to 385,000 gpm 2.2.1 decreased Fag for Mark-BW fuel removed power range neutron flux negative rate reactor trip 2

3/4.2.1 deleted baseload operation 3/4.2.2 changed Fn methodology to reflect Duke nomenclature quantified surveillance requirements 3/4.2.3 changed FAH methodology to reflect Duke nomenclature quantified surveillance requirements increased the tilt ratio at which power reduction-is 3/4.2.4

-required incorporated RCS flow as DNB parameter l

3/4.2.5 deleted Figure 8,

RCS Flow vs. R-Four Loops in Operation" from COLR and remove unit specification from Technical Specification Figure 3.2-1.

reduced RCS minimum measured flow to 385,000 gpm 3/4.3.1 removed power range neutron-flux negative rate reactor-trip 3/4.3.2 increased low-steam line pressure setpoint increased feedwater isolation response-time-incrsased steam line isolation response time removed steam line pressure dynamic compensation 3/4.3.3.12 decreased Reactor Makeup Water Fump flowrate limit from 75 gpm to 70 gpm in Mode 5 3/4.4.2.1 increased ~ pressurizer safety valve lift setpoint tolerance.

8,1 -

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

g -

4 i,

i i

}

3/4.4.2.2 increased pressurizer safety valve lift setpoint l

tolerance 1-4 changed steam generator main feodwater to auxiliary 3/4.6.3 feedwater nozzle isolation valve, auxiliary nozzle temper valve, steam generator feedwater containment j

isolation valve, steam generator.feedwater purge valve, i

main steam isolation valve, and main steam isolation I

bypass control valve stroke time from 5 seconds to llot Applicable, t

3/4.7.1.4 increased main steam line isolation valve stroke time Add NRC approved topical DPC-NE-1004A to list of 6.9.1.9 analytical methods used to determine core operating limits f

i i

Notes "* The proposed Technical Specification change was not included

{

in the approved Catawba 1, Cycle 6 (Reference 20) or. Catawba 1, Cycle 7 l

1 reload submittal (Reference 19). - The items without_the *** are l

identical to those approved for Catawba 1, Cycle 6 or Catawba 1, l

Cycle _7.

i f

a i

i-i'

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04

..--,n,---,

.,.~--n,,n,..,

., +., - - -,

~. ~.

~~-+---.--,...---..~..r.-~.--

. ~.

... ~.

l l-1

.i l

Table B-2 Core Operating Limits Report Changes

,I i

l 1

f Soecification Descriotion of Chance 1

i

)

3/4.1.3.5 revised shutdown bank insertion limits to reflect a j

minimum rod withdrawal limit of 222 steps and a maximum i

rod withdrawal limit of 230 steps i

3/4.1.3.6 revised control bank insertion limits to reflect a minimum rod withdrawal limit of 222 steps and a maximum rod withdrawal limit of 230 steps i

3/4.2.1 revised AFD limits for Cycle 6 operation l

3/4.2.2 revised for Cycle 6 operation to reflect a change in the heat flux hot channel factor Fn methodology 4-i

]

3/4.2.3 revised for Cycle 6 operation to reflect a change in the nuclear enthalpy rise hot channel factor FAH j-methodology 3/4.2.5 moved figure 3.2-1 from COLR to Technical Specifications i

1 1

Note: The proposed Core Operating Limits Report changes are' identical j

to those approved for Catawba 1 Cycle 7 reload submittal,' Reference 19.

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B.1 Changes to Technical Specificat. ions i

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-.+,,==-w.

wee mer ew 2

665 7

TOTAL FLOW 385000 GPM Ef4 5 '\\,

?

-\\,%.

455 psia UNACCEPTABLE

~

OPERATION g

w%

"~

2406 pois y

1 640 -

1 635 ;

2200 psia M

sa5 a e

L" 2100 psia

,2 g 620 -

'=

615i 1945 psia 605 0 600 -

596 i 4

ACCEPTABLE 590.

OPERATION

~

Sa5 5a0 0

0.2 0,4 -

0.6 0.8 1

1.2 Fraction of Rated Thermal Power FIGURE 2,1-la.

REACTOR CORE SAFETY LIMITS - FOUR LOOPS IN OPERATION JJHi-T' CATAWBA UNITS 1 & 2 2-2 Amendment No.101.(Unit 1) g,7 Amendment No. 95 (Unit 2)

I Delete.

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np n g..

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2 seconds

> 2 seconds i

4.

Intermediate' Range,

$25% of RTP*

$31% of RIP

l Neutron flux
~ ~ '
i 5.
Source Range, Neutron Flux
$10 cps
$1.4 x 105 cps 5
See Note 1 See Note 2 6.
Overtemperature AT See Note 3 See Note 4 ru 1.
7.
.0verpower AT
' 8.
Pressurizer Pressure-Low
>1945 psig
>1938 psig***
y-
-i
9.. Pressurizer Pressure-High '
$2385 psig
$2399 psig 1
l 10.
Pressurizer Water Level-High
.$92% of instrument
$93.8% of instrument span span i
i yy 11.
Reactor Coolant flow-tow
>90% of loop.
>88.9% of loop minimum measured minimum measured flow **
oo flow **
I E E-ee
-[
o3 f* t'.
f RIP = RATED THERMAL POWER.
[
22
L'oop minimum measured flow = %,^^^ ex (Sit 2), %,250 gpa (5f t I)
P?
' - g
Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead
[
and'1'second for. lag. ' Channel calibration shall ensure that these. time constants are adjusted to these l'
~
!: ' 22'
' values.
j
" 3.
i f* f+.
t
. to e vv 3
i
(,
I s
I i
i.
I
. _ _. _ _...- _ _ _ _ _ __. - _ _. _ _.. _ _ _ _. _.. _ _ _ _ _..._ -. _ - __ _ _ - m 1
i o
.=..n.
I..
.y E
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS m
j.
C i
3 TRIP SETPOINT ALLOWA8LE VALUE
. -4 FUNCTIONAL UN11 i
w-t L
12.
Steam Generator Water n
1 Level. Low-Low
a.. Unit 1 317% of span
>15.3% of span from from 0% to 30%
0% to 30% RTP*
RTP* increasing increasing linearly linearly to-to 138.3% of span e
> 40.0% of 'sgian from 30% to 100% RIP
from 30% to 100%
l RTP*
1-to b.
Unit 2 136.8% of narrow.
>35.1% of narrow cp range span range span l
-g 13.
Undervoltage - Reactor-
>77%. of bus 376% (5016 volts) 4 voltage (5082 j -
Coolant Pumps volts) with a j
0.7s response time gy-
, >56.4 Hz with a
>55.9 Hz eu Underfrequency Reactor-
~
3" 14.
kk Coolant Pumps
.6.2s response time 33-
""- '15.
-Turbine Trip 4
.P P
a..Stop Valve EH '
>550 psig
>500 psig zz
[
-Pressure Low
,g
.J b.
Turbine Stop Valve
>1% open
>1% open l
u,
. Closure-
'gy N.A.
N.A.
" ". - 16.
Safety Injection Input 00-from ESF
)
I
.-1, unAa A
d Op TABLE 2.2-1 (Continued)
I
-4
>5 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 1 '
t c-3 2
'M TRIP SETPOINT Att0WABLE VALUE FUNCTIONAL UNIT us.
e
.' 17.
Reactor Trip System t
Interlocks-f y
a.
Intermediate Range 31 x 10 80 amps 26 x 10 88 amps i
Neutron Flux, P-6 i
b.
Low Power Reactor Trips F
Block, P-7 7
1) 'P-10. input
$10% of RTP* -
$12.2% of RTP*
2) :P-13-input
. $10% RTP* Turbine
$12.2% RTP* Turbine Impulse Pressure Impulse Pressure Equivaient-Equivalent ro
+
t on x
$48% of RTP*
250.2% of RTP*
.c.
Power Range Neutron
{
l Flux,'P-8 d;
Power Range Neutron 169% of RTP*
$70% of RTP*
)-
. Flux, P-9
>10% of RTP*
>7.8% of RTP*
e.
Power Range Neutron 5 ?>
Flux, P-10
<10% of RTP*
<12.2% of RTP*
oo 3 0 --
f.
Power Range Neutron i
+
Flux, Not P-10
<10% RTP* Turbine
<12.2% RTP* Turbine er
'zz g.' Turbine Impulse Chamber Impulse Pressure Impulse Pressure P P' Pressure, P-13' Equivalent Equivalent
$S N. A.. -
N.A.
18.
Reactor Trip Breakers CC' 33 N.A.
N.A.
0 3
19. ' Automatic' Trip and ro s
. Interlock togic vv-l
)
' RIP = RATED THERMAL POWER
(
s
,.~.~._..r, w,
t l
ai; 377 1---
TABLE 2.2-1 (Continued) h TABLE NOTATIONS
)
NOTE 1: OVERTEMPERATURE AT
S) 1 I
b AT II * SI (1 + T 5) < ATo [K, - K I
(1 + T 5) [T (1 + r 5) - T'] + K (P - P') - f (AI)]
3 (1 + TzS) 3 2
3 s
y vs H
Measured AT by Loop Narrow Range RTDs; Where:
AT
=
j p.
i N
1+rS Lead-lag compensator on measured AT; 4
=
1+
5 Time constants utilized in lead-lag compensator for AT In = 12 s,
[
=
T1. T2 T2=3s, 1
Lag compensator on measured AT;
=
y, 3
2
' Time constant utilized in the lag compensator for AT, r3 = 0;
=
13 y
e M
r Indicated AT at RATED THEllMAL POWER; AT-
=
oc -
o 2.
1.1953 1
K
=
2 i
0.03163/*F K
=
2 4
1 1+rS The-function generated by the lead-lag compensator for T,,g E
1 + r35
'[g
=
3 dynamic compensation;-
3 EE
' Time constants utilized in the lead-lag compensator for T,yg,1
= 22 s.
ag 14, 13
=
4 1
er 13 = 4 s; zz 1
oo' i
.T
= -Average temperature. *F; i
e 2$
((
7 y
Tsg Lag compensator on measured T,,9;
=
cc 11 ee I
Time constant utilized in the measured T,,g lag compensator, t = 0;.
j
=
is my 1
vv.
e
~
.- -.. ~
-e,-_--
_m
_ _ _. ~.
._ _ _. -. _ __ _.. _ __. _ _ _ m. _
.__..._____.______m...
i l
1 l
nD
';MIT 1 l
5 TA8LE 2.2-1 (Continued)
TABLE NOTATIONS (Continued)
E NOTE 1: (Continued)
[
Z T'
- 500.8'F (Nominal T allowed by Safety Analysis);
r g
avg 0.001414;
{
K
=
3
-r j.
P
=. Pressurizer pressure, psig; 2235 psig (Nominal RCS operating pressure);
P'
=
Laplace transform operator, s 8; 5
=
I 7
and fm(AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:
. m..
co l
(i)
For q 9
***" ~
~
t. b C
f (aI) = 0, where qt '"d a are percent RATED THERMAL POWER in the top and bottom b
halves of the core respectively, and q
N 8
"E*"*"
h t
b I
RATED THERMAL POWER; k k-'
(ii)
For each percent 'AI that the magnitude of q g is som negaWe Wn -395,W I
t b
oo t
g@
AT Trip Setpoint shall be automatically reduced by 3.910% of AT.;
j.
g 3-and ee
~
yy-(iii)
For each percent AI that the magnitude of qt"9b is a re Positive than +3.0%, the AT Trip
[
l4 Setpoint shall be automatically reduced by 2.316% of AT..
l go-m r
~
CC l
NOTE 2:-
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by l
3" more than 3.0%.
I
[
5 e
uw vv.
t l
1 i
.... ~ -.. _.
-. ~. _ - - -.. -
e i
i
=:T 1 1
j>
TA8LE 2.2-1 (Continued)
~
TABLE NOTATIONS (Continued)
E NOTE 3: OYERPOWER AT "e
AT-(1 + r,5) ( 1
-)
Ib
) 'I I
I I I
) - T"] - f (AI)]
(1 + TzS) (1 + T 5) < ATo {K* - Ks (1 + T 5) (1 + 1.5) T - Ke [T (1 + T 5) 3 2
e..
na As defined in Note 1, Where:
AT
=
I# $
As defined in Note 1,
=
1+I S As defined in Note 1 l-
' r3, In
=
I
= As defined in Note 1, 1+T5 3
t.
7 1
As defined in Note 1,
=
13
4...
' oc As defined in Note 1, AT
=
o K
=' 1.0819' l
4 0.02/*F for increasing average temperature and 0 for decreasing average
'g g Ks
=
j-3g temperature..
g !"
y['$
E The function generated by the rate-lag controller for T,,g dynamic
=
3 7
compensation.
22 OO Time constant utilized in the rate-lag controller for T,,g, t, = 10 s,
=
L I,
$2
- 1 As defined in Note 1, y,
s3 =
mm -
cc 11 i
ee t
te-
=..As defined in Note 1, 3, g vv t
L I
t.
1 t
i a
v
..w.
_,. _, -. _ - ~
..-c
.-e-%v---e.m,
9
-41 NET 1 l
g 62 TABLE 2.2-1 (Continued)
(
TABLE NOTATIONS (Continued)
E NOTE 3:
(Continued)
U 0.001291/*F for T > 590.8*F and K. = 0 for T $ 590.8*F, l
K,
=
T As defined in Note 1,
=
m Indicated T,,g at RATED THERMAL POWER (Calibration temperature for AT T
=
instrumentation, 5 590.8'F),
As defined in Note 1, l
S
=
and 17 (AI) is a function of the indicated differences between top and bottom detectors of thE power range neutron ion chambers; with gains to be selected based on measured y
instrument response during plant startup tests such that:
m E
o (i) for q
~S
***" ~
O; 2 (O
' 9 and qb *#' E'"*"E t
b t
.i RATED THERMAL POWER in the top and bottom halves of the core respectively, and q
+q is total THEPMAL POWER in percent of RATED THERMAL POWER; b
(ii) for each percent al that the magnitude of q
~9 is m re negative than t
b
-35% LI, the AT Trip Setpoint shall be automatically reduced by 7.0% of aT.; and g
(iii) for each percent al that magnitude of g g is a re positive than g
b gg
+35% AI, the aT Trip Setpoint shall be automatically reduced by 7.0% of AT..
NN NOTE 4:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by g$
more than 2.8%.
EE "E
r+ r+
TABLE 2.2.-1 FOR UNIT 2 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Ey FUNCTIONAL UNIT TRIP SETPOINT All0WABLE VALUE 1.
ManualheactorTrip N.A.
N.A.
C5 2.
Power Range Neutron Flux if, a.
High~Setpo t
$109% of k.7*
$110.9% of RTP*
b.
Low Setpoint
$25% of RTP*
127.1% of RTP*
3.
Power Range, Neutron lux, 15% of RTP* with 16.3% of -
with 9
High Positive Rate-a time constant a time onstant m
2 2 seconds 12 _conds 4.
Power Range, Neutron Flux 15% of RTP* with
_6.3% of RTP" with High Negative Rate a time constant a time constant
>2 seconds
>2 seconds
.s 5.
Intermediate Range,
<25% of RTP"
-<31% of RTP*
Neutron Flux i
6.
Source Range, Neutron Flux
$10 s
<1.4 x 105 5
cps u
7.
Overtemperature aT See N e See Note 2 7
3 8.
Overpower AT S
Note 3 See Note 4 9.
Pressurizer Pressure-Low 11945 psig 11938 psig***
10.
Pressurizer Pressure-liigh
$2385 psig
$2399 psig gg 11.
Pressurizer Water Level-High
$92% of instrument (93.8% of instrument span s.n gs aE gg 12.
Reactor Coolant Flow-tow 190% of loop
>88.
cf loop minimum measured minimum asured flow **
"e flow **
gg Ul 9
RIP = RATED TliE L POWER 2),96,250 gpm (Unit 1) g g **** Loop minimum ocasured flow = 96,900 gpm (Unit
Time constads utilized in the lead-la lead
((
and 1 sec(nd for lag.
Channel calibration shall ensure that these time constants are adjusted to value.
vv
\\
UNIT 2
\\
TABLE 2.2-1 (Continued) n h
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5
FUNCTIONAL UhlT TRIP SETPOINT ALLOWABLE VALUE E
\\
Q 13.
Steam Generate Water sn level Low-tow g
a.
Unit 1 317% of span
>I5.3% o span from from 0% to 30%
0% to
. RTP" RTP* increasing in asi g linearly linearly to
>38.3% cf span
> 40.0% of span from 30% to 100% RTP*
from 30% to 100%
RTP*
>36.8% of nar' '
>35.1% of narrow b.
Unit 2 ange span range span o,
14.
Undervoltage - Reactor
>7 f us
>76% (5016 volts)
J Coolant Pumps voit e 5082 5
v 16;) wih a
.7s respons time 15.
Underfrequency Reactor 156.4 Hz with a
>55.9 Hz Coolant Pumps 0.2s response time
.y73 16.
Turbine Trip-a.
Stop Valve EH 1550 psig
_ 00 psig kk gg Pressure Low 55 b.
Turbine Stop Valve
>1% open
>1% opb gg Closure
~ h 17.
Safety Injectio nput N.A.
N.A.
go from ESF E2
L :L
[ [
RTP = R D THERMAL POWER
CIT 2 l
TABLE 2.2-1 (Ccntinued)
.n REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 7.
->y N
N
\\
.E ' FUNCTIONAL UNIT TRIP SETPOINT Alt 0WABLE VALUE h
'18.
Reactor Trip. ystem Interlocks g
a.
Intermediate Range
>l x 10 80 amps 16 x 10 13 amps p..
w Neutron Flux, P-6 b.
Low Power Reactor Trips Block, P-7 l
s
1) P-10 input
$ illE of RTP*
$12.2% of P*
2) P-13 input
<10%
F* Turbine
<12.
RTP* Turbine
' 'T uise Pressure impulse P ssure Equivalent Aquivalent-
~
c.
Power Range Neutron
<48% of RTP*
<50.2% of RTP*
[.
F1,ux, P-8 g
d.
Power Range Neutron
<69% of RTP*
170% of RTP*
oc Flux, P-9 e.
Power Range Neutron -
>10% o TP*
17.
of RTP* ~
Flux, P-10 f.
Power Range Neutron.
_ 0% of RTP*
112.2% o TP*
[f.
g g.
Flux, Not P-10 g.
Turbine Impulse Chamb
<10% RTP* Turbine
<12.2% RTP* Tu "ne hl e r+
Pressure, P-13 Impulse Pressure Impulse Pressure Equivalent Equivalent gg.
g 19. Reactor Trip B ers N.A N.A.
u, -
gg 20.
Automatic ip and N.A.
q,A-11 Interlo Logic ee 1"
RIP =
TED THERMAL' POWER
s UNIT 2 TABLE 2.2.-1 (Continued) n TABLE NOTATIONS NOTE 1:
VERTEMPERATURE AT I
3)
I' I *
$5) [T (1 + TsS) - T'] + K (P - P'l - f (al)]
AT II(1 + 1 5) (1 + T 5)
ATo [K, K
3 2 (1 + r3 2
3 5
Measured'ai by Loop Marrow Range RIDS;
=
'd Where:
AT Lead-lag compensator on' measured AT; I
=
p.
y n
Time constants utilized in lead-lag c nsator for AT, In = 12 s,
=
'1 T2 2,
12=35; I
49 C *Pensator on measured a,
=
1+T53 r3 -
= Time c stant utiliz$d' the lag compensator for AT,13 = 0; Indicated a at R D THERMAL POWER; w
=
AT m
o
..w g
K,
= 1.38; L'
0.02401
=
K2 T
function generated by lead-lag compensator for T'*9 I * S
=
1 * '55 ynamic compensation;
$a
=
Time constants utilized in the le lag compensator for T,,8, r, = 22 s, g 3-I, T3 ts = 4 s; 0N Average temperature, 'F;
=
T
EE 1
Lag compensator on measured T,
=
39 y,
3 Time constant utilized in the measured T,g lag compens or,
t. = 0;
=
1
=
^
i UNIT 2 l
)-
TABLE 2.2-1 (Continued) l n
TABLE NOTATIONS (Continued) r
' g' NOTE 1:
Continued)
T' 1
590.8'F (Nominal T allowed by Safety Analysis);
avg
.e r.
0.001189;
=
g g
r Pressurizer pressure, psig;
/.
=
P' 2235 psig (Nominal RCS operating pressure);
l i
lace transform cnerator, s 8; i
S.
=
i and f,(a!) is a function of the i icated differeace between t ad bottom detectors of the l
I ed on measured instrument power-range neutron ion chambers; wit (r gains to be selected response during plant STARTUP tests s that:
7. i -
(i)
For q q between -22.5% and -6.
j cp g
b f,
f (al) = 0, where q and a are percent ED THERMAL POWER in the top and botton l
u t
b is total THEM NR in percent of
{
. halves of the core respectively, and
+
t 1
e
- RATED THERMAL POWER;'
i i
j (ii) for each percent al that t magnitude of q q i re negative than -22.5%, the l
i b
AT Trip Setpoint shall automatecally reduced by 3.15 of AT.; and g,
l I.
I oo l
ll (iii).
~For each percen al that the magnitude of q q,is more posit' e than -6.5%, the AT Trip l
S5 Setpoint sh be automatically reduced by 2.414% of AT..
i 4
xz l'
??
int by i
$ 9 NOTE 2:
The hannel's maximum Trip setpoint shall not exceed its computed Trip Set re than 1.3%.
l CC l
dd ee l
~0 I
i 1
i
. ~ _.,
-. -.. - ~ -.... _...,., _.
- ~.... - -. - - - -.-. -.
c. _ ~ _.
p I
j i
4 UNIT 2 l
L
- N.
TABLE 2.2-1 (Continued)
s TA8LE NOTATIONS (Continued) f
. n s-h NOTE 3: MERPOWERAT
/
f I
I - T*] - f (
(1 + TsS) T - E [T (I I
AT (1 G) ( 1
).<
(1 + tzS P 1 + T 5) - ATo {K K
1 + TsSF l
5 (1 + r,5) i-3 t
c
-4 l
o' l
1 g
AT-As defined in Note 1, Where:-
=
g.
I N
1 + t'S' As defined in Note 1 o
I
=
1 +-1 5 2
I
,/'
i i
i
= As fined in Note 1
.ts, 12 f
I i
= As defi d in Note 1, 3
1 + TsS l
I As defined in e 1, i
=
13 i
'l' '
As defined in Note t
4
.AT'
=
t o
j-9*
j.
- If K.
= 1.0704 f
l 0.02/*F r increasing average.
rature and 0 for decreasing average K
=
3 temper ure.-
t
,3 kk y['f,3 function generated by the rate-lag troller for T,8 dynamic
=
1 --
i
=3 compensation.
I.@ -
i l
-33 1,
Time constant utilized in the rate-lag control r for T t, = 20 s.
=
i
-avg,.
t zz 1
As defined in Note 1,
(
=
y?
y,
[
. w> o
. us -.
' =
As defined in Note 1,
[
L ' EE Is
~11
?
c+ s
mg e
' vv j.
V c.
d
,w..
..m
4 t
i I
l UNIT 2
- n-TA8LE 2.2-1 (Continued)
TABLE NOTATIONS (Continued) 3-i G
NOT (Continued) e
.i.
0.001707/*F or T > 590.8*F and K. = 0 for T < 590.8'F
)
c.
2 K.
=
+
i
-4
?
sn As defined in Note 1 T
=
[
- r t
at RATLD THElWEL POWER (Cali ation temperature for AT l
e-Indicated T,,g T
=
, u instrumentation, < 590.8'F),
t
{
r
i..
S As defined in Note 1 and j
i.
i or all AI.
j f (AI)
=
2
[
.- t as
. NOTE 4:
The channel's maximum T Setpoint 11 not exceed its competed Trip Setpoint by 7,
more than 2.8%.
r l
O-t i
e t
}
.B i
u 33
.Eo-
l. "
h 33.
?
..N N f
n!'
I l
N__
s U
~
YY
..00-t 3
I i
d
\\
'Y "F
4 4--
y-m.
,_q_,,..
i
2.1 SAFETY LlHITS BASES 2.1.1 REACTOR CORE ff0R-tiffli 1)
The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.
Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surf ace temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excussive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to ONB through the BWCMV correlation.
The BWCMV DNB correlation has been developed to predict the ONB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.
The local DNB heat flux ratio, (DNBR), is defined as the ratio of the heat flux that would cause ONB at a particular core location to the local heat flux, and is indicative of the margin to DNB.
The ONB design basis is as fdllows:
there must be at least a 95%
probability that the minimum DNBR of the limiting rod during Condition I and 11 events is greater than or equal to the ONBR limit of the ONB correlation being used (the BWCMV correlation in this application).
The correlation DNBR limit is established' based on the entire applicable experimental data set such that there is a 95% probability with 95% confidence that ONB will not occur when the minimum DNBR is at the ONBR limit.
In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters, and the BWCMV l
DNB correlation are considered statistically such that there is at least a 95% confidence that the minimum DNBR for the limited rod is greater than or equal to the ONBR limit.
The uncertainties in the above parameters are used to determine the plant DNBR uncertainty.
This ONBR uncertainty is used to establish a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.
The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated ONBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.
CATAWBA - UNITS 1 & 2 B 2-1 8 23 Amendment No.101 (Unit 1) l Amendment No. 95 (Unit 2)
2.1 SAFETY LIMITS BASc5 These curves are based on a nuclear enthalpy rise hot channel factor, N
FaH, of 1.50 and a reference c m ine with a peak of 1.55.
l An allowance is included for an increase in F at reduced power based on the 3g expression:
N F3g = 1.50 [1 + 1/RRH (1-P)]
l Where P is the fraction of RATED THERMAL POWER.
RRH is given in the COLR.
l i
These limiting heat flux conditions are higher than those calculated for l
the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f
(a!) function of the Overtemperature AT trip, when the axial power ikbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.
hl.1 REACTOR CORE (FOR UNIT 2),
it restrictions of this Safety Limit prevent overheating of the fuel and possible ladding perforation which would result in the release of fi3sfon 4
Overheating of the fuel clidding)ing regime products to he reactor coolant.
s prevented by 4 tricting fuel operation to withir the nucleate D
where the heat t sfer coefficient is large and the claddin Pface temperature is slig above the coolant saturation temper ure.
Operai. ion above the aper boundary of the nuclea)( boiling regime could result in excessive claddingKemperatures because pf'the onset of departure from nucleate boiling (DNB) an he resultant spatp reduction in heat transfer i
coefficient.
DNB is not a direct measurabirparameter during operation, and therefore THERMAL POWER and Reactor olant' Temperature and Pressure have been related to ONB through the WRB-1
lation, The WRB-1 DNB' correlation has been developed to predict the DNikfl d the location of DNB for axially i
uniform and nonuniform heat flux tiitributio >
The local DNB heat flux ratio, (DNBR), is defined as the rati f the heat flux hat would cause ONB at a particular core location to e local heat flux, an is indicative of the margin to DNB.
The ONB design sis is as follows:
there must be at ast a 95%
probability that,th,e minimum DNBR of the limiting rod during dition I and II events is edater than or equal to the DNBR limit of the ONB relation stablished based on the entire applicable experimental data (on DNBR e WRB-1 correlation in this application).
The correl t oeing used
_Stt such limit is that ere is a 95% probability with 95% confidence that DNB will not occ%
en the minimum DNBR is at the DNBR limit.
CATAWBA - UNITS 1 & 2 B 2-2 Amendment No.101. (Unit 1)
/
g Amendment No. 95 (Unit 2)
2.1 SAFETY t.!MITS i
BASES
\\ In meeting this cesign basis, uncertainties in plant operating paramet s,
nuel ar and thermal parameters, and fuel f abrication parameters are consip red stati cally such that tnere is at least a 95% confidence that the min 11ium s
DNBR for *ne limiting reo is greater than or equal to tne ONBR limit.
he uncertaint in the above plant parameters are used to determine e plant This ONBR uncertainty, combined with the corr ation DNBR DNBR uncerta -.
limit, establis s a design DNBR value which must be met in p nt safety analyses using va s of input parameters without uncertaint es.
The curves of Figura 2.1-1 show the loci of point of THERMAL POWER, Reactor Coolant System pr'tqure and average tempera re below which the calculated DNBR is no less t. n the design DNBR v ue, or the average enthaley at the vessel exit is less tha the enthalpy o aturated liquid.
y Fh,Thiscurveisbasedonanucle enth py rise hot channel factor, of 1.49 and a reference cosine w a peak of 1.55 for axial power shape.
An allowance is included for an iner de F g at reduced power based on the expression:
N Fg = 1.49 (1 + 0.3
]
1 Where P is the fr tion of RATED THERMAL POWER.
4 These limiti heat flux conditions are higher than the e calculated for the range of al control rods fully withdrawn to the maximum a wable control red insertie assuming the axial power imbalance is within the 1 ts of the f (a!) fu tion of the Overtemperature trip.
When the axial power balance tis not hin the tolerance, the axial power imbalance effect on the r-temce ture AT trips will reduce the Setpoints to provide protection con tent core Safety Limits.
w 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of racionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor vessel, pressurizer, and the Reactor Coolant System piping, valves, and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Safety Limit of 2735 psig-is therefore consistent with the design criteria and associated Code requirements.
The entire Reactor Coolant System is hydrotested at 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation.
e M
CATAWBA - UNITS 1 & 2 B 2-2a Amendment No. 86 (Unit 1) l Amencment No. 80 (Unit 2)
l LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)
The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.
In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity.
The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the accident analysis to enhance the overall reliability of the Reactor Trip System.
The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is d
initiated.
This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary 4
actuation of the Engineered Safety Features Actuation System.
Manual Reactor Trio The Reactor Trip System includes manual Reactor trip capability.
Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting.
The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.
g g
Power Rance, Neutron Flux, 44eh 9 te:-
The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents.
--The Pcwer Range-Hegat4ve-Rate-t+49--peev4 des-peeteet4er, for conttc4-cod-
-drop-ace 44ents,--At-M9h-power -: red 4eop-aco44ent-cou14-saus: 10:01 flux--
-peak 4ng-wh4ch-4ou14-caut: :n-unconser-v4tive-local-ONSR te exist.
The Power--Renge
-Negat4ve-Rate-t+4s u!" prevent-thi+-f+om-occer4ng-dy-t+4pping the reactor.
4ie-eeedi4.-+s--taken-f+c-eperat4en-of-the-Pcw: r-Range-Negat4ve-Rate-ttip-f 0 r
-ttese-control rod-deep-see44ent4-for which GNBR: wil' be greatee-than-the app 44eable-des 4<jn '"it ONBR vale f e r-each-Jue4-type,----
The P:wer Rsnge-Negat4*e-Rat: 'ip 5:0 been-deleted f er--Urit: L 8 26 CATAWBA - UNITS 1 & 2 B 2-4 Amenoment No.101 (Unit 1)
Amencment No. 95 (Unit 2)
i l
-w:T 1-l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX OIFFERENCE (AFO)
LIMITING CONDITION FOR 0.0ERATION 3.2.1 The indicated AXIAL FLUX O!FFERENCE (AFO) shall be maintained within the acceptable limits specified in the CORE OPERATING LIMITS REPORT (COLR).
"~
APPLICABILITY:
MODE 1, above 50% of RATED THERMAL POWER.* (Uni'.1)
ACTION:
i a.
For operation with the indicated AFD outside of the limits specified l
in the COLR, 5
1.
Either restore the indicated AFD to within the COLR limits within 15 minutes, or 2.
Reduce THERMAL POWER to less then 50% of RATED THERMAL POWER 1
within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 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.
THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR.
d 4
See Special Test Exceptions Specification 3.10.2.
CATAWBA - UNITS 1 & 2 3/4 A2-1 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)
_ =
-_ =
-UN f4-4--
l POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:
a.
Monitoring the indicated AFD for each OPERAf;2 excore channel:
1)
At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2)
At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status, b.
Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the _first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoper-able.
The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging, c.
The provisions of Specification 4.0.4 are not applicable.
4.2.1.2 The indicated AFD shall be considered outside of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the limits.
9 8-28 l
CATAWBA - UNITS 1 & 2 3/4 A2-2 AmendmentNo.$b((Unit 1)
Amendment No.
Unit 2)
vHii-t-l' POWER OISTRIBUTION LPITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (X,Y,Z) l LIMITING CONDITION FOR OPERATION 3.2.2 F (X,Y,Z) shall be limited by imposing the following relationships:-
1 q
MA Fq (X,Y,Z) i F0 K(Z) for P > 0.5 P
MA(X,Y,Z) 1 RTP F
F q
RTP = W (Z) for P < 0.5 0
K
~
Where:
Fg q Limit at RATED THERMAL POWER (RTP) the F specified in the CORE OPERATING LIMITS REPORT (COLR),
MA Fq (X,Y,Z) = the measured heat flux hot channel factor F9 (X,Y,Z),
M witn adjustments as specified in-4.2.2.3, p
THERMAL POWERRATED THERMAL PO E, and
~
K(Z) = the normalized F (X,Y,Z) limit specified in the q
COLR for the appropriate fuel types.
APPLICABILITY:
MODE 1.-("r.it 1)
ACTION:
With F (X,Y,Z) exceeding its limit:
9 NA a.
Reduce THERMAL POWER-at least 1% for each 1% Fq (X,Y,Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and b.
Control the AFD to within new AFD limits which are' determined by reducing the allowable power at'each' point along the AFD limit lines 1
M of Specification 3.2.1 at least M for each 1% Fq (X,Y,Z) exceeds the limit-within 15 minutes and reset the AFD' alarm setpoints to the modified limits within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and POWER OPERATION may proceed for_up to a total off72 hours; subsequent c.
POWER OPERATION may proceed provided-the Overpower AT
_(value of K,) have been reduced at least 1% (in aT spa. Trip Setpoints n) for each 1%
MA(X,Y,Z) oxceeds the limit,-and Fq d.
Identify and correct-the cause of the out of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a.,-
above; THERMAL-POWER may then be increased provided F (X,Y,Z) is-
[
9
_ demonstrated through>incore mapping to be withir. its limit.-
CATAWBA - UNITS 1 & 2
-3/4 A2 Amendment No. 86 (Unit 1)
Amendment No. 80 (Unit 2)-
.-=._...-- - -.. - -. - -
t a
f
. UNIT.lh I
i_
POWER DISTRIBUTION LIMITS
'j SURVEILLANCEREhUIREMENTS 4.2.2.1 The_ provi'sions of' Specification 4.0.4 are not applicable.
4.2.2.2 F N(XYZhl}hallbeevaluatedtodeterminewhetherF(X,Y,Z) iswithin9tslimitby:
9
~
a.
Using the movable incore detectors to obtain a power distribution
!=
map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.-
N I
b.
Measuring-Fq (X,Y,Z) at tha earliest of:
1.
At least once per :31 Effective Full Power Days, or-i 2.
Upon reaching equilibrium conditions after exceeding by 10% or
[
more of RATED THERMAL POWER, the THERMAL POWER at which a
M l
q (X,Y,Z)1was last determined (2), or F
3.
At each time the QUADRANT. POWER TILT RATIO indicated by the excore-j detectors is normalized using-incore detector measurements.
{
c.
Performing the following. calculations:
1.
For each location, calculate the % margin to the maximum j'
allowable design as_follows:
M
% Operational Margin = (1 -
F(X,Y,Z)
) x 100%
n
[F((X,Y,1)]oP M
F(X,Y,Z)
) x 100%-
l
% RPS Margin =_(1 -
n
[
[Fh(X,Y,Z)]RPS-L OP
-L-RPS where (F (X,Y,Z)]
and(F(X,Y,Z)]
are the Operational and q
q l
RPS design peaking limits defined in the COLR.
i-2.
Find the minimum Operational Margin of all-locations ~ examined in
'4.2.2;2.c.1 above. :If any margin is-less than zero, then.either;
' of:the. following actions shall be taken:
D o additional uncertainties are required in-the following equations-for N
M Fq (X,Y,Z), be'cause' the limits include uncertainties.
( )During power escalation'at the beginning of each cycle, THERMAL. POWER may-
-be = increased until a. power level for extended operation hasl been achieved and a power distribution map;obtained.
CATAWBA.- UNITS 1 & 2 3/4'A2-4 '30 Amendment No. 86I(Unit 1)
[
8 Amendment No'. 80- (Unit 2)
.a.
= ~.
.~
i
-uni-T 1 POWER DISTRIBUTION' LIMITS l
SURVEILLANCE REQUIREMENTS (Continued) v (a) Within.15 minutes:
1 l
(-
(1) Control the AFD to within new AFD limits that are j
determined by:
reduced COLR(a).
(AFD' Limit) negative (AFD Limit) negative I
i min
+ (Margin 0P] absolute value 4
reduced COLR(3)
(AFD Limit) positive = (AFD Limit)p,3$ggy, i
min
- (Margin 0P] abs lute value min where Margin is the minimum margin from-4.2.2.2.c1,Nd 3
l j
(2) -Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to the modified limits of 4.2.2.2.c.2.a. or b
(b) Comply with the ACTION requirements of-Specification 3.2.2, j
treating the margin violation in 4.2.2.2.c.1 above as the NA j
amount by which F is exceeding its limit.
l q
3.
Find the minimum RPS Margin of all locations examined in l.
4.2.2.2.c.1 above.
If any margin is less than zero, then the following action shall be taken:
i i
- Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . - reduce the K1 value for OTAT by:
i t adjusted = K I )
-[KSLOPE( ) x Margin 3] absolute value
-K i
q min s
where MARGIN is the minimum margin from 4.2.2.2.c.1.
RPS
0) Defined and specified in the COLR per Specification 6.9.1.9.
U)K value from Table 2.2-1.
t CATAWBA - UNITS 1 & 2 3/4 A2-5 8-31 Amendment No.101(Unit '1).
Amendment No. -95(Unit-2) i
-_ __ = -
-uni T i _
l POWER OfSTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) d.
Extrapolating (5),g,) east two measurements to 31 Effective Full j
Power Days beyond the most recent measurement and if:
l M
L OP (F (X,Y,Z)) (extrapolated) 1 (F (X,Y,Z))
(extrapolated), and q
9 N
N (F (X,Y,Z))
(extrapolated)
(F (X,Y,Z))
(Fh(X,Y,Z))0P-(extrapolated)
(Fh(X,Y,Z))0P or i
M L
RPS l
(F (X,Y Z)) (extrapolated) 1 (F (X,Y,Z))
(extrapolated), and g
g i
M (F (X,Y,Z))
(extrapolated)
(F (X,Y,Z))
(F (X,Y,Z))RPS (extrapolated)
(Fh(X,Y,Z))RPS i
either of the following;. actions shall be taken:
M
]
- 1.
_F (X,Y,Z) shall be' increased by 2 percent over that specified in 0
j 4.2.2.2.a. and the calculations of 4.2.2.2.c repeated, or s
2.
A movable incore detector. power districution map shall be obtained, and the calculations of 4.2.2.2.c.1 shall be performed-no later than the time at which the margin in 4.2.2.2.c.1 is i
extrapolated to be equal to zero.
l e.
The limits in Specifications-4.2.2.2.c and 4.2.2.2.d are not applicable in the following core plano regions as measured in percent of core j
height from the bottom of the fuel:
1.
Lower core region-from 0 to 15%, inclusive.
I:
2.
Upper core region from 85 to 100%, i_nclusive.
l 4.2.2.3 When a full core power distribution map is taken for reasons other -
than meeting the requirements of Specification 4.2.2.2, an'overall F "(X,Y,Z) q i
shall be. determined, then increased by 3% to account for manufacturing
)
tolerances, further increased by 5% to account for measurement uncertainty, I
and further increased by the radial-local peaking factor to obtain-a maximum local peak.
This value shall be compared to the limit in Specification 3.2.2.
(5)ExtrapolationofF)fortheinitialfluxmaptakenafter-reachingequili-~
brium conditions _is not required since the initial flux map extablishes 4
the baseline _ measurement for future trending.
Also, extrapolation of F limits are not valid for core locations that were previously redded, or '
for core locations that were previously within $2% of the core height about c
j the-demand position ~of the rod tip.
CATAWBA - UNITS 1 & 2 3/4 A2-6 Amendment No.101 (Unit 1) 8 32 Amendment No. -95 (Unit 2)
E w
---t-,c y
y-v-
w-,
-UNIT 1 POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR FAH(X,Y)
LIMITING CONDITION FOR OPERATION 3g(X,Y)shall be limited by imposing the following relationship:
3.2.3 F
FAH"(X,Y) 5, (FAH'(X,Y]L 0 Where:
FAH"(X,Y) = the measured radial peak.
[FAH'(X,Y)]LCO
= the maximum allowable radial peak as defined in the CORE OPERATING L! HITS REPORT (COLR).
APPLICABILITY:
MODE 1. (UNIT 1) i ACTION:
With FAH(X,Y) exceeding its limit:
a.
Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce the allowable THERMAL POWER from RATED THERMAL POWER at least RRH%( ) for each 1% that FAH"(X,Y) exceeds the limit, and l
b.
Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:
1 1.
Restore FAH"(X,Y) to within the limit of Specification 3.2.3 for l
RATED THERMAL POWER, or 2.
Reduce the Power Range Neutron Flux-High Tri: Setpoint in Table 2.2-1 at least RRH% for each 1% that FA[(X,f) exceeds that limit, i
and c.
Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the lir.it of Specification 3.2.3, either:
1.
Restore FAH"(X,'r) to within the limit of Spe:ification 3.2.3 for RATED THERMAL POWER, or 2.
Perform the following actions:
)
term in Table 2.2-1 oy at least TRH (a)
Reduce the OTAT Kt N
for each 1% that FAH (X,Y) exceeds the limit, and l
(b) Verify through incore mapping that FabX,Y) is restored to within the limit for the reduced THERMAL POWER allowed by ACTION a, 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 .
0)RRH is the amoung(of THERMAL POWER reduction required to compens each 1% that FaH X,Y) exceeds the limits of Specification 3.2.3 provided in the COLR per Specification 6.9.1.9.
0)TRHistheamoung(ofOTATK setpoint reduction recuired to compensate for 1
each 1% that FaH X,Y) exceeds the limit of Specif':ation 3.2.3, provided in the COLR per Specification 6.9.1.9.
CATAWBA - UNITS 1 & 2 3/4 A2-7 Amendment No.101 (Unit 1)
'mendment No. 95 (Unit 2)
-UNIT-1 POWER DISTRIBUTION LIMITS i
LIMITING CONDITION FOR OPERATION i
ACTION (Continued) d.
Identify and correct the cause of the out-of-limit condition prior i
to increasing THERMAL POWER above the reduced THERMAL POWER limit--
required by ACTION a. and/or c.2., above; subsequent POWER OPERA-N TION may proceed provided that FAH (X,Y) is demonstrated, through 1
i incore flux mapping, to be within the limit specified in the COLR prior to exceeding the following THERMAL POWER levels:
1) 50% of RATED THERMAL POWER, 2) 75% of RATED THERMAL POWER, and l
3)
Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95%
of RATED THERMAL POWER.
]
SURVEILLANCE REQUIREMENTS l1 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.
N 4.2.3.2 FAH (X,Y) shall be evaluat'ed to determine whether FAH(X,Y) is within l
its limit by:
l a.
Using the movable incore' detectors to obtain a power distribution j
map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
N l
b.
Measuring FAH (X,Y) according to the following schedule:
i 1.
Upon reaching equilibrium conditions after exceeding by 10%
orgoreofRATEDTHERMALPOWEg3)theTHERMAL'POWERatwhich FAH (X,Y) was last determined
, or 2.
At least once per 31 Effective Full Power Days, or
[
3.
At each time the QUADRANT POWER TILT RATIO indicated by the excore detectors is normalized using incore detector measurements, c.
Performing the following calculations:
1.
For each location, calculate the % margin to the maximum allowable design as_follows:
N FAH (X,Y) x 100%
%F Margin =
1-g
[FAH'(X,Y)]b[
L V
Where (FAH (X,Y)]
is the design peaking limit defined in the i
COLR.
No additional uncertainties are required for FAH"(X,Y),
because (FAH (X,Y)]SURV, includes uncertainties.
l (3) During power escalation at the beginning of each cycle, THERMAL POWER may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
4 CATAWBA - UNITS 1 & 2 3/4 A2-8 8-34 Amendment No.101(Unit 1)
Amendment No.
95(Unit 2)
MNIT -1 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR-OPERATION ACTION (Continued) 2.
Find.the minimum margin of all locations examined in 4.2.3.2.c.1 above.
If any margin is less than zero, comply with the ACTION requirementsofSpecification3.2.3asif(Fj(X,Y))SE i s the sameas(Fj(X,Y)]LCO d.
Extra'solating(4) at least two measurements to 31 Effective Full Power Days seyend the most recent measurement and if:
l Fh(X,Y) (extrapolated) > (Fj(X,Y))SURY (extrapolated), and Fh(X,Y)
(extrapolated)
F (X,Y)
(Fh(X,Y))SURV(extrapolated)-
(FhX,Y))SURV either of the following actions shall be taken 1.
Fb(X,Y)shallbeincreasedby2percentoverthatspecified itt"4.2.3.2.a. and the calculations of 4.2.3.2.c repeated, or 2.
A movable incere detector power distribution map shall be obtained, and the calculations of 4.2.3.2.c shall be performed no later than the time at which the margin in i
l 4.2.3.2.c is extrapolated to be equal to 'zero.
4 7
a 4
i N
Extrapolation of F for the initial flux map taken after reaching (4) equilibrium conditNns is not required since the initial flux map establishes the baseline measurement for future trending.
CATAWBA - UNITS 1 & 2 3/4 A2-9 Amendment No.101 (Unit 1) 8-35 Amendment No. 95 '(Unit 2) w y
y y
e'Yr
w W
-WIT 1 l
4 POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.
APPLICABILITY:
MODE 1 above 50% of RATED THERMAL POWER-(UM t 1).*,**
l ACTION:
a.
With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but
]-
less than or equal to 1.09:
a 1.
Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
I a)
The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)
THERMAL POWER is reduced to less than 50% of RATED THERMAL 4
POWER.
2.
Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:
a)
Reduce the QUADRANT POWER TILT RATIO to within its
~
limit, or b)
Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron l
Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
3.
Verify that the CUADRANT POWER TILT RATIO 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 50% 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 reduce the Power Range Neutron Flux-High Trip 3
Setooints to less than or equal to 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 4.
Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the
~
QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%
or greater RATED THERMAL POWER.
See Special Test Exceptions Specification 3.10.2.
Not applicable until calibration of the excore detectors is completed subse-quent to refueling.
CATAWBA - UNITS 1 & 2 3/4 A2-10 -36 8
Amendment No. b ((Unit 1) l Amendment No.
Unit 2)
-tNIT 1 I
POWER OISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) b, With the QMADR NT POWER TILT RATIO determined to exceed 1.09 due to i
misalignment of either a shutdown'or control rod:
1.
Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a)
The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)
THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
~
2.
Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each L% of indicated QUADRANT POWER TILT RATIO in excess of i
1.02, within 30 minutes; I
3.
Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL 1
POWER to less than 50% 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 reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 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 4
Identify and correct the cause af the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%
or greater RATED THERMAL POWER.
"~
c.
With the QUADRANT POWER TILT RATIO determined to exceed l.09 due to causes other than the misalignment of either a shutdown or control rod:
1.
Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a)
The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)
THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
CATAWBA - UNITS 1 & 2 3/4 A2-118-37 Amendment No. 86 (Unit 1) l
~
Amendment No. 80 (Unit 2)
-UNIT 1 l
POWER DISTRIBUTION LIMITS L,IMITING CONDITION FOR OPERATION ACTION (Continued) i 2.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER 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 less than or equal to 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 3.
Identify and correct the cause of the out-of-limit condi ion prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.
j d.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:
I a.
Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and
.b.
Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alarm is inoperable.
c.
The provisions of Specification 4.0.4 are not applicable.
4.2.4.2 The QUADRANT POWER TILT RATIO shall be determine'd to be within the limit when above 75%' of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm-that the normalized symmetric power distribution, obtained from two sets of four symmetric-thimble 1
locations or full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
8-38 CATAWBA - UNITS 1 & 2 3/4 A2-12 Amendment No. 86 (Unit 1)
(
Amendmant Nn. 80 (Unit 21
k
--UN W 1 POWER DISTRIBUTION ~ LIMITS 3/4.2.5 DNB PARAMETERS d
i-5 i
LIMITING CONDITION FOR OPERATION 1
l
)
3.2.5 The following DNB related parameters shall be maintained within the j
limits shown on Table 3.2-1:
Reactor Coolant System T,yg, a.
)
{
b.
Pressurizer Pressure, i:
c.
Reactor Coolant System Total Flow Rate.
l I
l APPLICABILITY:
MODE 1. (' nit 1)
J i
l ACTION:
1-i a.
With either of the parameters identified in 3.2.5a. and b. above l
exceeding its limit, restore the parameter to within its limit i
within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce
THERMAL-POWER to less than 5% of RATED THERMAL POWER within the'next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
i 6
b.
With the combination of Reactor Coolant System total flow rate and THERMAL POWER within the region of restricted operation specified on i
Figure 3.2-1, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reduce the Power Range Neutron Flux-High Trip Setpoint to below the nominal setpoint by the same amount
(% RTP) as the power reduction required by Figure 3.2-1.
l c.
With the combination of Reactor Coolant System total flow rate and THERMAL POWER within the region of prohibited operation specified i
on Figure 3.2-1:
s 1.
Within-2 hours either:
a)
Restore the combination of-Reactor Coolant System total flow rate and THERMAL' POWER-to within the-region of i
permissible operation,-or b)
Restore the combination of Reactor-Coolant-System total flow rate and-THERMAL POWER to within the region of restricted operation and comply with action b. above,.or c)
Reduce l THERMAL POWER to-less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux _- High Trip Setpoint to less than or equala to 55% of RATED
-THERMAL POWER within the next.4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
CATAWBA - UNITS 1 & 2 3/4 A2-13 -39 Amendment No.101 '(Un'it 1)-
8 p
Amendment No. 95 (Unit 2)
-UNIT 1-l POWER OISTRIBUTION LIMITS
~
i 3/4.2.5 DNB PARAMETIRS LIMITING CONDITION FOR OPERATION 2.
Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being within the re,gion of prohibited operation specified on figure 3.2-1, verify that the combination of THERMAL POWER 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 .
i SURVEILLANCE REOUIREMENTS i
t 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4.2.5.2 The Reactor Coolant System total flow rate indicators shall be sub-jected to a CHANNEL CALIBRATION at least once per 18 months.
The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement.
1 4.2.5.3 The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months.
4 a
E CATAWBA. UNITS 1 & 2 3/4 A2-14 gao Amendment No.
86 (U' it 1) n
}
Amendment No. go,(Unit 2)
vNIT 1 l
TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS Four Loops in Operation Average Temoerature Meter Average
- 4 channels:
< 592 F
- 3 channels:
i592F Comouter Average
- 4 channels:
< 593'F
- 3 channels:
i593*F Pressurizer Pressure Meter Average
- 4 channels:
> 2227 psig*
- 3 channels:
[2230psig*
Computer Average
- 4 channels:
> 2222 psiga
- 3 channels:
> 2224 psig*
Reactor Coolant System Total Flow Rate Figure 3.2-1 4
i
Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per min'ute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.
CATAWBA - UNITS 1 & 2 3/4 A2-15841 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)
j 1
M l
l i
388850 A penalty or o.1% for unoelected fe.o.atw venturi room, and a Permissible rneasurement ecwtanty or 2.t% for
. Operation-t
flow are ncuoes n true figure, Region -
4 iss.sssocc) 385000 F
i i
Restricted
'I (ss,3stuoi l
l E 381150; Operattori Region
~
O i
i C
f 4
I 1
(94.3 773001 S 377300-
-.I L
l l
4 E
j 3
l.
a 1
C/3 (92.37345e)
I 1
5373450 i
s o
4 0
}
[
- Prohlblted i
o Operation o
m (9c.389400)
Region C
- c::: 369600-3 1-4 L
.i t
c 365750 1
4 4
1 361900 86 88
=90 92 94 96' 98' 10 0 10 2 Fraction'of Rated Thermal Power; i
Figure 3.2-1 Reactor Coolant System Total Flow Rate Versus Rated Thermal Power
-Four Loops in.0peration
("ait 1) i; 8-42 CATAWBA - UNITS 1 & 2 3/4 A2-16
' Amendment No. 86 (Unit -1)
I
-Amendment No. 80--(Unit 2).-
LONIT 2 l
1 POWER OISTRIBUTION !.IMITS
'3/4.2 i
l 3 4.2.1 AXIAL FLUX DIFFERENCE (AFD) l
\\
LIMITN!,G CONDITION FOR OPERATION 1
c
)
l 3.2.1 The 'ndicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained thin:
i a.
the llowed operational space as specified in the CORE _OP RATING LIMIT REPORT (COLR) for RAOC operation, or b.
within t target band specified in the COLR about t target flux difference during baseload operation.
APPLICABILITY:
MODE 1, above 50% of RATED THERMAL-_ POWER
nit 2)
{
ACTION:
i a.
For RAOC operation ith the indicated AFD utside of'the limits i
specified in the COL i
l 1.
Either restore the ndicated AFD o within the COLR limits within
-15 minutes, or 2.-
Reduce THERMAL POWER to less than 50%.of RATED THERMAL-POWER within 30 minutes and re the Power Range Neutron Flux-High j
Trip setpoints to less tha or equal to 55% of RATED THERMAL l
POWER within the next 4 ou b.
For Base. Load operation'a ve APLND
with t'he. indicated AXIAL FLUX l
DIFFERENCE outside of.th applicable rget-. band about the target j
flux difference:
1.
Either restore he-indicated AFD to w thin the COLR specified target band 1'mits within.15 minutes, i
Ni 2.
Reduce THE. L POWER to less than APL o RATED THERMAL POWER-l and disc tinue Base Load operation within__
minutes.
j c.
THERMAL PO R shall not be increased above 50%-of TED~ THERMAL PGWER unless th indicated AFD zis within the limits specif d in the COLR.
3 i
i^
V 1
i.
See decialTestExceptionsSpecification3.10.2..
j.
4 0
A is the' minimum allowable (nuclear design) power level for base load peration and'is specified in the CORE OPERATING LIMITS REPORT per Speci fi cati on-6. 9.1'. 9.
CATAWBA _ UNITS'1 &.2 3/4 82 ga3 Amendinent No. 86< (Unit 1)J
' Amendment No. 80--(Unit 2)'
,e
-n.--,
r---
-_n~..,
m.
-, --m,-www-~m-,
1 i\\
i UN!T 2 l
Y l-
\\
i s
l POWER DISTRIBUTION LIMITS l
j LIMIT NG CONDITION FOR OPERATION s
SURVEILLhCERE0VIREMENTS
/
j s
\\
f l
4.2.1.1 The ndicated.AFO shall be determined to be within 'its limits uring POWER-OPERATIO above 50% of RATED THERMAL POWER by:
I a.
Monitor g the indicated AFD for each OPERABLE excore e nnel:-
1)
At le t once per 7 days when the AFD Monitor A rm is OPERABLE, t
and 2)
At least-o ce per hour for the first 24 hou s after restoring j
the AFD Mon or Alarm to OPERABLE status.
l-b.
Monitoring and logg g the indicated 'AFD for each OPERABLE excore i
ch'annel at'least once er hour for the-fir 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes th eafter, when the F0 Monitor Alarm is inoper-l able.
The logged values of the indicat d AFD shall be assumed _to j-exist during the interval receding e h logging.:
a l_
c.
The provisions of Specificat n 4.
.4 are not applicable.
i 4.2.1.2 The indicated AFD shall be consi red-outside of its-limits when-at.
t least two OPERABLE-excore channels.are i i
ting the AFD to be outside the j
limits.
i
=
i 4.2.1.3 When in Base Load operatio, the target axial ~ flux difference of eacn OPERABLE excore channel shall e determined y measurement at least once per 92 Effective Full Power Days.
The provisions o Specification 4.0.4 are l-not applicable, i
4.2.1.4 When in Base Load o eration, the target flux d ference shall be i
updated at least once per. Effective Full Power Days by either determining l
the target flux differenc in conjunction with the surveil nce requirements of L
Specification 3/4.2.2 o/by-linear interpolation between the est recently mea-i sured values and the a'lculated value at the end of cycle-lif The provisions l
of Specification 4.0 are not applicable, f
I i
1-j a
i 844 CATAWBA - UNITS 1.& 2-3/4 B2-2
' Amendment No. 86 (Unit 1)
\\.
Amendment.No. 80' (Unit-2)
F
4 l
~
10 NIT 2
,s
'N POWER DISTRIBUTION LIMITS
\\
I
3/4.2.2 HEAT FLUX HOT-CHANNEL FACTOR - F,(2).
p w
i l
LIMI NG CONDITION FOR OPERATION
\\.
I j
3.2.2 F
) shalllbe limited by the following relationships:
q i
i RTP _
i (Z) 1. F l
RTP F (Z) 1 F K(Z) for P 1 0.5 l_
9 Where:
F.
= the F Limit at RATED T RMAL _- POWER '(RTP) q specified in the CORE PERATING LIMITS REPORT.
j O LR),-
P _ THERMAL OWER
, and r'
RATED THE L POWER K(Z) ='the norma i ed F (Z) for a given core height 9
i specified the COLR.
APPLICABILITY:
~
l 1
-MODE 1. (Unit 2)
ACTION:
1.
j With F (Z) exceeding its mit:
q h
a.
' Reduce THERMAL P0 R.at 1 east 1% for each 1.
F (Z) exceeds the._ limit q
j within 15 minut s and similarly reduce the Po r' Range Heutron
, Flux-High Tri Setpoints'within-the next 4 hou ;-POWER OPERATION 4
may proceed or_up'to a total.'of'72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subse ent POWER OPERATION-may procee provided the Overpower aT: Trip Setpoin s-(value of _K ) have 4
j
.been red ed at least 1% (in AT. span) for leach 1%' F Z) exceeds the-9 j-
limit, nd:
i b.
'Ide fy and correct the cause.of the out-of-limit cond ion prior to nereasing THERMAL POWER above the reduced limit 1requi dLby i-ION a., above;-THERMAL-POWER may then be increased prov1 ed q(Z) is.-demonstrated.through incore mapping'to be within-it limit.
[
L
-p
. b
$~
CATAWBAJ, UNITS 1 & 2 3/4'B2-3 Amendmart No. So.(Unit 1);
845_
- Amendment-No. 80 f(Unit -2) l
-,r-
~ch~--
4 m
u
.,4
..,.,,,-[',,
,..,-r..w,w,me...'.y
.,,...,,w,..
,...,...s.,,
,w, m,mg.,.-.#y.mu.m_.
UNii 2_
[
OWER OISTRIBUTION'LIAITS
\\
SUR ILLANCE REOUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not. applicable.
4.2.2.2 F
RAOC operation, F (I) shall be evaluated to-determine if F
.)
is within.it limit by:
9 a._
Usin the movable incere detectors to obtain a power d'istri ution map at any THERMAL POWER greater than 5% of RATED THERMAL OWER.
b.
Increasin the measured F (z) component-of the power d tribution-q map by 3% t account for manufacturing tolerances an further in-creasing the alue by 5%.to account.for measuremen uncertainties.
Verify the req frements of Specification 3.2.2 ar satisfied, c.
Satisfying the f lowing. relationship:
N Fq (I) 1 x K(z) for P > 0.5 Px (z) l RTP M
F j
Fq (7) i 9
) for P < 0.5 W(z) x 0.
where F (z) is the measured F (.
increased by the allowances for q
l manufacturing tolerances and asumentuncertainty;FhTPis the F limit, K(z)-is the norma zed F (I as a function of core height, q
q l'
P is the relative THERMAL OWER, and W(
is the cycle dependent i
function that accounts r power distribu ion transients encountereo during normal-operati FfP K(z), and W ) are specified in the l'
CORE OPERATING LIMI 5 REPORT per Specificatio 6.9.1.9.
1 d.
Measuring F (z) ccording to the-following sched e:
1.-
Upon'ac eving equilibrium conditions after ex eding by 10% or i
more o RATED THERMAL POWER, the THERMAL POWER a which F (z)
{
was 1 st determined,* or 9
i 2.
A east once per 31 Effective Ful1~ Power Days, which ver occurs
rst, f
During p f escalati' n.at the beginning of each-cycle, power leve1Lmay b o
i
-increase until a power level-'for-extended operation has been achieved and l
power d tribution map obtained.
l l-CATAWBA'- UNITS 1 & 2 3/4 B2-4 8-46 Amendment No. 86 _(Unit 1)-
Amendment No. 80 (Unit 2)
.s
.,. - ~. _.. -.. --
u.----
1
. UNIT 2 l.
DOWER DISTRIBUTION LIMITS YURVEILLANCEREOUIREMENTS(Continudd) i e.
With measurements indicating Fh(z) maximum over z-A(z)
[
ha increased since the previous determination of F (z) ei her of l
the ollowing actions shall be taken:
f 1)
F(
shall be increased by 2*4 over that specif ed in j
Speci cation 4.2.2.2c., or f
2)
F (z) shal be measured at least once pe 7 Effective Full Nq, Power Days u il two successive maps i dicate that maximum M (z) is not i reasing.
i over I K( '
}
f.
With the relationships sp ified in pecification-4.2.2.2c. above j
not being satisfied:
l 1)
-Calculate the percent F
) exceeds its limit by the following 3
-expression:
q l'
4 i
maximum F (z)
W(z) 4 x 100 for P > 0.5
]
p
~
over z l
x K(z) J Y
a i-(
"M(z) x-W(z) maximum F
1 x 100 fo P < 0.5 l
/ over I RTP p
i I
O x K(z) y L
0. s-
]
2) h One f the following actions shall be taken:
a)
-Within'15' minutes, control the AFD to within w AFD limits i
'which are determined by rLducing the AFD limits f
Specification 3.2.1 by 1% AFD for each percent F z) exceeds q
j:
its-limits as determined'in' Specification'4.2.2.2f.
Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset-the AFD alarm setpoints-to._the mod-ified limits, or b)
Comply with the requirements of Specification 3.2.2;for-j.
F (z) exceeding its limit by the percent calculated above, or 9
c)
Verify that the requirements of Specification 4.2.2.3 for -
Base Load >peration;are satisfied and enter Base Load operatio.
CATAWBA - UNITS 1:& 2' 3/4 B2-5 8-47 Amendment No. 86. (Unit _1)~
h
. Amendment No. 80L (Unit 2) i1
i UNIT 2.
I OVER DISTRIBUTION LIMITS 5 VEILLANCE REQUIREMENTS (Continued)
The limits specified in Specifications 4.2.2.2c., 4.2.2.2e., and 4.2.2.2f., above are not applicable in the following core plane l
regions:
l Lower core region from 0 to 15%, inclusive 2.
Upper core region from 85 to 100%, inclusive.
4.2.2.3 Base ad operation is permitted at powers above APLND* if he following 4
conditions are s tisfied:
a.
Prior to ntering Base Load' operation, maintain THE POWER'above l
ND APL and ss than or equal to that allowed by Spe fication 4.2.2.2 j
for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
Maintain Base oad operation surveillance
..FD within the target band about t target flux differ-ence of Specifi tion 3.2.1) during this time p riod. : Base load operation is the permitted providing THERMAL OWER is maintained ND OL between APL and.A L or between APLND: an 100%-(whichever is most j
limiting) and FQ surs illance.is maintain pursuant to Specification El 4.2.2.4.
APL is def ed as:
I RTP l
APLOL = "I"I""*
II)
] x-100%
over Z F (Z)
W(Z)BL Ff(z)isthemeasured z) increased by the allowances for where:
FhTP is the manufacturing tolerances and a rement uncertainty.
F limit, K(z) is the norma zed "q Z) as a function of core height, j
q j
W(z)BL is the cycle depen ent functio that accounts for limited power distribution transients encountered dur' g Base Load operation, j
F
, K(z),-and W(Z are specified in e CORE OPERATING LIMITS L
REPORT per Specifi tion-6.9.1.9.
i b.
During Base Load operation, if the THERMAL PO is decreased below ND APL then th conditions of 4.2.2.3a shall be s isfied before re-entering se Load operation.
4.2.2.4 During Base oad Operation F (Z) shall be evaluated t determine if q
F (Z) is within it limit by:
l-q a.
Using e movable incore detectors to obtain a power dis ribution-ND map t any THERMAL POWER above APL j
b.
I reasing the measured F (Z) component of the power distrib tion map q
.y 3% to account for. manufacturing tolerances-and further inct sing t
the value by 5% to account for measurement uncertainties.
Veri the i
requirements -of Specification 3.2.2 are satisfied.
l
A/LND is the minimum allowable (nuclear design) power level for Base Lead l
operation in Specification 3.2.1.
8'48 CATAWBA - UNITS 1 & 2 3/4 B2-6 Amendment No. 86 (Unit 1)
Amendment No. 80 (Unit 2)
UN!T 2 l
l 1-OWER DISTRIBUTION LIMITS'
\\
i SURVEILLANCE REOUIREMENTS (Continued) y
?
4 c.
Satisfying the following relationship:
_i j
RTP ND F' Z) 1
[
for P > APL q
M where:
F (Z) is the measured F (Z).
FfistheFq JImit..
q j
K(Z) is t e normalized F (Z) as a function o c;e n ight.
P is the q
relative T RMAL POWER.
W(Z) is the_ cycle depe ent function that j
accounts for imitadpowerdibributiontransien*
encountered during j
Base load opera ion.
F
, K(Z), and W(Z)g are specified in the CORE OPERATING LI ITS REPORT per Specificati n 6.9.1 9.
i d.
Measuring F (Z) in c junction.with targ flux difference deter-j mination ac ording to he following sch ule:
1.
Prior to entering Ba e load ope tion after_ satisfying-surveil-lance 4.2.2.3 unless full co e flux map-has been _taken in the i
previous 31 EFPD with t e re tive thermal. power having been ND j
maintained above APL fo he 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and 2.
At_least once per 31 ef-cti l full _ power days-e.
With measurements indicat g M
maximum F
7)
(I) 3 over I hasincreasedsinfethepreviousdeterminatic F (Z) either of the following actio shall be taken:
l 1.
F (Z) s 11 be increased by 2 percent over t t specified in 4.2.2 c, or f
M 2.
F ( ) shall be measured at least once per.7 EFPO til 2 9
Wccessive maps indicate that
I"""
FM (z) is not increasing.
[ K(I) 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 N
CATAWBA - UNITS 1 & 2 3/4 B2-7 8 Amendment No. ' 86. (Unit' 1)
- Amendment No. 80 (Unit 2) a
. -. ~. -- -
-~~ -
l UNIT 2 1
-PO R DISTRIBUTION LIMITS I
' SURVEIL NCE REQUIREMENTS (Continued) j Comply with the requirements of Specification 3.2.2_for-F (Z) exceeding.its limit by the. percent calculated with l
9
[-
the following expression:
F$(Z) x W(Z)BL ) ) -1 ) x 100 f r P > A ND
((m, over z of (
~
FhTP x'K(Z)
P g.
The-limits spec ied in_4.2.2.4c., 4.2.2.4e., a 4.2.2.4f.-
above are not app icable in the following. core plan. regions:-
1.
Lower core regt n 0 to 15 percent, inc usive.-
2. -
Upper core region 5 to 100 percen inclusive.
4.2.2.5 When F (I) is measured for easons et r than meeting the requirements 9
of Specification 4.2.2.2-an _overall m sured q(z)'shall'be obtained from;a power distribution map and increased by 3% to c unt'for manufacturing-tolerances and further increased by 5% to account-fo measurement uncertainty.
\\
~
' CATAWBA - UNITS 1 & 2 3/4 ~B2-8 8-50 Amendment No. 86 '(Unit 1)
Amendment No. 80 (Unit 2)'
UN!T 2 POWER DISTRIBUTION LIMITS
/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMINNG CONDITION FOR OPERATION 3.2.3 e combination of indicated Peactor Coolant System total flow ra and R shall bh maintained within the region of permissible operation specif ed in N
the CORE OPFtATING LIMITS REPORT (COLR) for four loop operation.
Where:
N" a.
R=
F 1.0 + mfg (1.0 - P))
4 THER AL POWER b.
p RATEDTHER(1ALPOWER c.
F H = Measured vat s of F obtained by usin the movable incere g
i detectors to o tain a power distribut n map.
The measured l
valuesofFhsh11beusedtocalc ate R since the figure l
specified in the C0 includes p nalties for undetected feed-
[
water venturi fouling f 0.1% nd for measurement uncertainties ncoremeasurementofFh, of 2.1% for flow and 4%
r j
Fh limit at RAT d.
F
= The THE L POWER (RTP) specified in the COLR, and e.
mfg = The power factor multiplier specifie in the COLR.
APPLICABILITY:
MODE 1 (UNIT.
).
4 ACTION:
a.
With the combinati of Reactor Coolant System total ow rate and R within the region of res ricted operation within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months redue the Power Range Neutron Flux-Hi Trip Setpoint to below the nominal setp int by the same amount (% RTP) as the power reduction required by the figu specified in i
the COLR.
b.
With the mbination of Reactor Coolant System total flow rate nd R within the regi n of prohibited operation specified-in the COLR:
1.
ithin 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 rate'and R to within the region of permissible operation, or b)
Restore the combination of Reactor Coolant System total flow rate and R to within the region of rentricted operation and comply with action a. above, or CATAWBA - UNITS 1 & 2 3/4 B2-9 8-51 Amendment No. 86
(' Unit 1)
Amendment No. 80 (Unit 2)
\\
l-l UNIT 2
~
i l
P erb 1STRIBUTIONLIMITS f
3/4)t.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHAL'PY-RISE HOT CHANNEL FACTOR LIMITIN. CONDITION FOR OPERATION
\\
/
i ACTION (Conti ued) c)
R uce THERMAL POWER to less than 50% of RATED THER L POWER an reduce the Power Range Neutron Flux - High Tri Setpoint.
i to 1 s than or equal to 55% of RATED THERMAL P0 R withtm 2.
Within 24 ho of itially being within the reg n of prohibited-i operation spec ied in the COLR, verify through ncore flux mapping l
and Reactor Cool nt System total flow rate cum rison that the com-bination of R and eactor Coolant System tot flow rate are restored.
j to within the regio s of restricted or perm sible operation, or-reduce THERMAL POWER o less than 5% of R D THERMAL POWER within j-the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
1 i
i I
L i
l t
{
8-52 CATAWBA - UNITS 1 & 2
-3/4 B2-10 Amendment No. 86 (Unit :1)-
a..namont un. An (nnit 31
. ~..
4 4
-UNIT 2
.l OVER DISTRIBUTION LIMITS-
?
i LIM ING CONDITION PR OPERATION e
x ACTIONhContinued) 3.
dentify and correct the cause of the out-of-limit condition rior t increasing THERMAL POWER above the reduced THERMAL POWER imit re ired by ACTION b.1.c) and/or b.2., above;-subsequent P WER OPERA-~
TIO.ay proceed provided that the combination'of R and dicated React Coolant System total-flow rate are demonstrate through_
l incore ux mapping and Reactor Coolant System total ow rate
i comparis q, to be within the regions of restricted o permissible operation i ecified in the COLR prior to exceeding he following THERMAL POW 1evels:
4
[
a)
A nominal 0% of RATED THERMAL POWER,
[
b)
A nominal 75 of RATED THERMAL POWER, d
i c)
Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months f attaining greate than or equal to 95%
l of RATED THERMAL OWER.
SURVEILLANCE REOUIREMENTS 4
j 4.2.3.1 The provisions of Specificathn 4.d4 are not applicable.
4.2.3.2 The combination of indicated Re tor Coolant System total flow rate j
determined by process computer reading or igital voltmeter measurement and R i
shall be determined to be within the gions of restricted or permissible l
operation specified in the COLR:
l a.
Prior to operation abov 75% of RATED ERMAL POWER after each fuel j
loading, and b.
At~1 east once per 3. Effective Full Power 0 s.
4.2.3.3 The indicated Re tor Coolant System total flow rate shall-be verified i
to be within the' regions f restricted or permissible ope tion specified in the COLR at least once er 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently btained value of R, obtained per. Specific ion 4.2.3.'2, is assumed to exist.
4.2.3.4 Th'e Reacto Coolant System total flow rate -indicators hall.be subjected to a CHANNEL CALI TION at least once per 18 months.
The-measu ent-.
instrumentation all be calibrated within 7 days prior'to the pe 'orman'e of-c the calorimetr flow measurement.
4.2.3.5 The eactor Coolant System total flow rate shall be determin by f
precision h at balance measurement at least once per_18 months; 4
\\
CATAWBA --UNITS 1 & 2 3/4 B2-11 -53
- Amendment No. 86 (Unit 1)-
\\-
8 Amendment No. 80-(Unit 2)
j UNIT 2-
'l j
[
P lER DISTRIBUTION LIMITS-3/4 4 QUADRANT-POWER TILT RATIO LIMITI ONDITION FOR OPERATION i'
3.2.4 The QUA' ANT POWER TILT RATIO shall not_ exceed 1.02.
-APPLICABILITY:
E 1, *above 50% of RATED THERMAL POWER (Unit 2) i ACTION:
l a.
With the QUAD NT POWER TILT RATIO determined to exce d11.02 bat j
less than or e al to 1.09:
j 1.
Calculate th QUADRANT POWER TILT RATIO at ast once per hour' until either:
i l
a)
The QUADRAN OWER TILT _ RATIO is r duced to within its limit, or
~
i i
b)
THERMAL POWER is educed to'1 ss than 50% of RATED THERMAL
. POWER.
l 2.
Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:
a)
Reduce the QUADRANT P
-TILT RATIO to within its
-limit, or i
b)
Reduce THERMAL P ER at leas 3% from RATED THERMAL POWER l-for each 1% _ of - dicated QUAD NT POWER TILT RATIO in excess of 1 a similarly redue the Power Range Neutron Flux-High Tr'p Setpoints within t e next 4. hours.
4 i
3.
Verify-that t.
QUADRANT POWER TILT RATI is within its. limit within 24 ho s after exceeding the limite reduce THERMAL POWER to 19 s-than-50% of-RATED THERMAL POW within the next
]
2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s-a#d reduce the Power Range Neutron F1
-High Trip i
Setpoin to less than or equal to 55% of RATE THERNAL POWER withi the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 4.
Ide tify and correct the cause of the out-of-limit ndition p ior to increasing THERMAL ' POWER; subsequent POWER ERATION bove 50% of RATED THERMAL. POWER may proceed provided at the QUADRANT POWER TILT RATIO is verified within.its limit at least:
once. per hour for 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s-or> until verified. acceptable a 95%.
i or greater RATED THERMAL POWER.
4 4
See pecial Test Exceptions Specification 3.10.2.
i i
8 ~
CATAWBA - UNITS 1 & 2 3/4 B2-12 Amendment No. 86 (Unit.1)-
Amendment No. 80_ (Unit 2) -
. - -... =
i UN!T 2-l-
i OWER DISTRIBUTION LIMITS l
y
LIM TING CONDITION FOR OPERATION ACTION ontinued)-
}-
l l
b.
Wi the QUADRANT, POWER TILT RATIO determined to exceed 09 due ton misa ignment:-of. either a shutdown or control rod:
L i
1.
Ca culate the QUADRANT POWER TILT RATIO at leas once per hour-j unt 1 either:
f a):
T QUADRANT POWER TILT RATIO _is-reduc d to wi_ thin
-its limit, or.
b)
THE POWER is reduced to less an 50% of RATED THERMAL _
i POWER.
I 2.
Reduce THERMAL WER at least 3% f om RATED THERMAL POWER for j
each 1% of indica ed QUADRANT PO R TILT RATIO in ex' cess of-1, within 30 minut i
3.
Verify that the QUAD T POW TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after e eed'ng the limit or reduce THERMAL POWER to less than 50% o TED-THERMAL POWER within the next1 j
E2 hours'and reduce the P r Range Neutron Flux-High Trip.
Setpoints to less than r e ual to 55% of RATED THERMAL; POWER i
within the next 4 hou ; and j
l 4.
Identify and corre o the cause the out-of-limit' condition j
prior to increa jsi g THERMAL POWE subsequent POWER OPERATION-j:
above 50% of RA}ED THERMAL. POWER m
_ proceed provided that-the-i.
QUADRANT-POWE ILT RATIO-is-verifie within.its : limit aileast once per hou for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until v ified acceptableiat 95%
j
-or greater TED-THERMAL POWER.
l-c.
With the QUAD ANT! POWER TILT. RATIO determined t exceed 1.09 due to causes othe than the misalignment _of either a s tdown-or control.'
l-rod:-
t I
1.-
Ca culate-the QUADRANT POWER TILT RATIO.at_least once per hour til either:
l
_ a)= -.The QUADRANT POWER TILT RATIO-iL reduced to wit in -
~
its limit, or 2
i.
- b).
THERMAL-POWER is reduced to less than 50% of ' RATED ERMAL-F POWER.
t 8-55 CATAWBA - UNITS'I & 2-13/4 B2-13 Amendment No. 86?(Unit 1)
Amendment No. 80:(Unit:2)
L
/
UNIT 2 i
POW OISTRIBUTION LIMITS LIMITIN CONDITION FOR OPERATION ACTION (Co nued) 2.
qduce THERMAL POWER to less than 50% of RATED TH L POWER wiYhin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutro Flux-High Trik Setpoints to less than or equal to 55% of TED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 3.
Identify and correct the cause of the out-o -limit condition prior to creasing THERMAL POWER; subseg nt POWER OPERATION above 50% o RATED THERMAL POWER may pro eed provided that the QUADRANT POW TILT RATIO is. verified y thin its limit at least i
once per hour r 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL P ER.
d.
The provisions of Speci cation 3.0.4 re not applicable.
SURVEILLANCE REQUIREMENTS
\\
4.2.4.1 The QUADRANT POWER TILT RATIO hal be determined to be within the limit above 50% of RATED THERMAL POWE by:
a.
Calculating the ratio at east once pe 7 days when the alarm is OPERABLE, and b.
Calculating the rati at least once per 12 ours during steady-state operation when the arm is inoperable.
c.
The provisions o Specification 4.0.4 are not a licable.
4.2.4.2 The QUADRANT P WER TILT RATIO shall be determined t be wi. thin the limit when above 75%
RATED THERMAL POWER with one Power Ran e channel inoperable by using e movable incore detectors to confirm tha the normalized symmetric puer di ibution, obtained from two sets of four symm tric thimble locations or full ore flux map, is consistent with the indicated ADRANT POWER TILT RATIO at le t once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
4 8-56 CATAWBA - UNITS 1 & 2 3/4 B2-14 Amendment No. N (, Unit 1)
...a___.....
POWER DISTRIBUTION LIMITS i
3/ 2.5 DNB PARAMETERS LIMIT ONDITION FOR OPERATION j
3.2.5 The fo lowing DNB related parameters shall be maihtained withi the limits shown o > Table 3.2-1:
a.
Reactor colant System T,yg, and b.
Pressurize Pressure.
APPLICABILITY:
MODE 1, nit 2 only.
4 ACTION:
With any of the above paramet rs exceeding its limit restore the parameter to within its limit within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL P ER to less than 5% of RATED THERMAL POWER within the n xt 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
i SURVEILLANCE REOUIREMENTS 4.2,5 Each of the parameters of Ta
.2-1 s all be verified to be within their limits at least once per 12 'ours.
i 4
i 4
N CATAWBA - UNITS 1 & 2 3/4 B2-15 Amendment No. 86 (Unit 1)
Amendment No. 80 (Unit 2)
UNIT 2 TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS Four Loops in Operation Average Temoerature Meter Ave
- 4 channels:
< 592*F
- 3 channels:
5592'F Computer Avera
- 4 channels:
< 593'F
- 3 channels:
{593 Pressurizea Pressur channels:
,,2227 psig*
Meter Average
-3 hannels:
1 2230 psig*
Computer Average
- 4 ch qnals:
) 2222 psig*
3 chan 1s:
> 2224 psig*
N
Lin>[tnotapplicableduringeitheraTHERMALPOWERrampinexcessof5%o.
RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of ATED THERMAL POWER, 4
\\
/
g.sg CATAWBA - UNITS 1 & 2 3/4 B2 Amendment No.86 (Unit 1)
Amendment No,80 (Unit 2)
m. _... _.... _. _. m. _ -. _. _ _. _ _
1ABLE 3.3-1 : = =II 1 l
[
t
+
'Y REACIOR TRIP SYSTEM INSTRUMENTATION l
5 I
MINI M i-TOTAL NO.
CHANNELS CHANNELS APPLICABLE l
1 j'
E FUNCTIONAL UNIT OF CHANNELS TO TRIP-OPERABLE MODES ACTION a
w _ 1.
Manual Reactor Trip.
2 1
2 1 2 1
I 5
2 1
2 3, 4*, 5*
10 I
N I
t 2I Power Range,. Neutron Flux
[
a. _High Setpoint.
4 2
3 1, 2 2
i b.
. Low Setpoint-4 2
3 IfM, 2 2
[
~3.
Power' Range, Neutron Flux.
4 2
-3 1, 2 2
f
'!!igh Positive Rate t
t
)
' 4.
Intermediate Range, Neutron Flux 2
l' 2
1888, 2 3
w p
r
>-. S'.
Source Range, Neutron Flux a.
Startup..
2 1
2 2M 4
[
I w
b..-Shutdown 2
1 2
3*, 4*, 5*
10
[
i.
a l
6.
Overtemperature AT
?Four' Loop Operation.
4 2
3 1, 2 6
I 4
e 7.
0verpower AT EE Four Loop Operation 4
2 3
1, 2 6
i wu 33 i
.RE 8..
Pressurizer; Pressure-Low 4
2 3
1 6**
3 33 g
i r* M -
4' 22'
.4 A
.an i
,cc-
"n "A s
1 i
] _ y _ c* f
{
NW' 4
vv
't
. _ -. -. _ _ -, ~ _,. _.. - _. _. _.
n-TABLE 3.3-1 (Continued)
Wii "
l
'E REACTOR TRIP SYSTEM INSTRUMENTATION g
MINIMUM
[-
TOTAL NO.
CHANNELS CHANNELS APPLICABLE 5
fUNC110NAL UNIT OF CHANNELS TO TRIP' OPERA 8LE MODES ACTION 9.
Pressurizer. Pressure-High
'4 2
3 1, 2 6**
I g
C 10.
Pressurizer. Water level-High 3
2 2
1 6
l
- 11.
Reactor Coolant flow-Low a.
Single Loop.(Above P-8) 3/ loop 2/ loop in 2/ loop in 1
6
[
any oper-each oper-ating loop ating loop 3
a b... 'Two Loops (Above P-7 and.
3/ loop 2/ loop-in-2/ loop 1
6 l
R below.P-8) two oper-each oper-ating loops ating loop y
12.
Steam Generator Water 4/sta 2/sta gen 3/sta gen 1, 2 6**
l Level--Low-Low gen in any each' operating operat 1 sta gen sta gen
13. Undervoltage-Reactor Coolant 4-1/ bus 2
3 1
6 I
.g,
' Pumps (Above P-7)
-gg
14. Underfrequency-Reactor Coolant 4-1/ bus 2
3 1
6 l
-Q $
y g.
Pumps (Above P-7) 55 Turbine Trip I
[ [
15.
a.
Stop.Valwe EH Pressure 4
2 3
1####
6
- Low
$9
'b.
-Turbine Stop Valve Closure 4
4 1
I####
11 1
22 am. '16. : Safety Injection input-37 from ESF-2 1
2 1, 2 9
l 00
TABLE 3.3-1 (Continued)
="!
l 2
j.
REACTOR TRIP SYSTEM INSTRUMENTATION N
MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE
3. FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8LE MODES ACTION I
d a
17.. Reactor Trip 5ystem Interlocks
y g~
a.-
Intermediate Range Neutron Flux, P-6 2
1 2
2N 8
b.
Low Power Reactor Trips Block,.P-7 P-10 Input 4
2 3
1 8'
or P-13 Input 2
1 2
1 8
g c.
Power Range Neutron
' Flux', P-8 4
2 3
1 8
2-y d.
Power Range Neutron 4
2 3
1 8
y Flux, P-9 a-e.
Power Range Neutron Flux, P-10
.4 2
3 1
8 f.
Power Range Neutron
((
' Flux, Not P-10 4
3 4
1, 2 8
oo
'RR g.
Turbine Impulse Chamber
~$$
. Pressure, P-13 2
1 2
1 8
. ee ydL18.. Reactor Trip Breakers 1
2 1; 24*, 5*
I
$5.
4 '19.
Automatic Trip and Interlock 2
1 2
1 2 9
[
I
'((~
Logic 2
1 2
3, 4*, S*
10
20.. Reactor. Irip Bypass Breakers N.A.
N.A.
N.A.
1,2,3*,4*,5*
13 l
TABLE 3.3-1 (Continued)
-UNI T-t-l TABLE NOTATIONS
$0nly if the Reactor Trip System breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.
Comply with the provisions of Specification 3.3.2, for any portion of the channel required to be OPERABLE by Specification 3.3.2.
Below the P 6 (Intermediate Range Neutron Flux Interlock) Setpoint.
Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
Above the P 9 (Reactor Trip on Turbine Trip Interlock) Setpoint, i
ACTION STATEMENTS l
ACTION 1 - With the number of OPERABLE channels one less than the Minimum j
Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT j
STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
4 ACTION 2 - With the number of OPERABLE channels one less than the Total l
Number of Channels,' STARTUP and/or POWER OPERATION may proceed i
provided the following conditions are satisfied:
1 a.
The inoperable" channel is placed in the tripped condition i
within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j
b.
The Minimum Channels OPERABLE requirement'is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification I
4.3.1.1, and i
i c.
Either, THERMAL POWER is restricted to'less than or equal i
to 75% of RATED THERMAL POWER and-the Power Range Neutron L
Flux trip setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ;-or, the-i QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.
1 i
ACTION 3 - With the number of channels OPERABLE one less than the Minimum l
Channels OPERABLE requirement and with the-THERMAL POWER level:
1 a.
Below the P-6 (Intermediate Range Neutron Flux Interlock)
-Setpoint,' restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER-above the P-6' i
Setpoint; or b.
Above the P-6 (Intermediate Range Neutron Flux Interlock) l Setpoint but belos 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to-increasing THERMAL POWER above 10% of RATED THERMAL POWER.
N,2
-CATAWBA - UNITS 1 &'2 3/4 A 3-5 Amendment _No.
(Unit'.1)10* l Amendment No. -(Unit 2) 9:
- - ~.
-MrnN TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued)
ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
ACTION 5 - Delete ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels. STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , and b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.
ACTION 7 - Delete ACTION 8 - With less than the.Hinimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by, observation of the associated permissive' status light (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 ours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.
ACTION 10 - With the number of OPERABLE-channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor trip breakers within the next hour.
ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
ACTION 12 - With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 9.
-The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for perform-ing maintenance to restore the breaker to OPERABLE status.
With the breaker bypassed, apply ACTION 9 ACTION 13 - With any reactor trip bypass breaker inoperable, restore the bypass breaker to OPERABLE status prior to placing it in 3
service, CATAWBA - UNITS 1 & 2 3/4 A 3-6 Amendment No.101(Unit 1) l Amendment No. 95(Unit 2)
TABLE 3.3-2-f0R UN!! 1 9>
52 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES 6
RESPONSE TIME h
FUNCTIONAL UNIT a
N.A.
g; 1.
Manual Reactor Trip ff 2.
Power Range, Neutron Flux 5 0.5 second*
3.
Power Range, Neutron Flux, N.A.
High Positive Rate N.A.
4.
Intermediate Range, Neutron Flux N.A.
5.
Source Range, Neutron Flux
^~
R:
6.
Overtemperature AT
$ 8 seconds"
$ 8 seconds *
'7.
Overpower AT ye 8.
Pressurizer Pressure-Low i 2 seconds
~
t 5 2 seconds 9.
Pressurizer Pressure-High N.A.
10.
Pressurizer Water Level-High NN
se zz
??
Neutron detectors are exempt from response time testing.
Response time of the neutron ficx signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
g 22 E. 3.
se
T ABLE 3.3-2 T0" 1;;;;T I (Contin'ued) l n
C.
'E REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES S?
RESPONSE TIME FUNCTIONAL UNIT c
d
11. - Low' Reactor-Coolant Flow
.h
' Single' Loop (Above"P-8)
$ 1-second a.
b.
Two Loops (Above P-7 and below P-8)
$ 1 second 12.
Steam Generator Water Level-Low-tow
< 3.5 seconds
-a.
Unit 1 b.
Unit 2 1 2.0 seconds 13.
Undervoltage-Reactor Coolant Pumps
.-C 1 1.5 seconds l
~14.
Underfrequency-Reactor Coolant Pumps 5 0.6 second s-I Y
.IS'.
Turbine Trip l
N. A.-
'Stop Valve EH Pressure-Low g
a.
N.A.
b.
Turbine Stop Valve Closure M.A.
16s : Safety Injection Input from ESF.
I
!k. 17. : Reactor; Trip System Interlocks l
N.A.
3o hk.:'18.
Reactor' Trip Breakers N.A.
l
'55 n
N.A.
19. ~ Automatic Trip and Interlock Logic-z2
'??
l 22-'
Ak 26 c
n-TABLE 4.3-1 TOR 0;ui 1 l
C REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
>5 TRIP I
ANALOG ACTUATING M00ES FOR CHANNEL DEVICE WHICH 5
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE
' d FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED
.g
- Manual Reactor Trip N.A.
N.A.
N.A.
R(14)
N.A.
1, 2, 3*, 4*, 5*.
1.
2.
Power Range, Neutron Flux a.
High Setpoint S
O(2, 4),
M N.A.
N.A.
1, 2 M(3, 4),
Q(4, 6),
R(4, 5) b.
Low Setpoint
'S R(4)
-M, N.A.
N.A.
1###, 2
'R 3.
-Power Range,-Neutron Flux, N.A.
R(4)
M N.A.
N.A.
1, 2 High Positive Rate 4.
Intermediate Range, 5-R(4,5) 5/U(1),M N.A.
N.A.
1###, 2
]
Neutron Flux 5.
Source Range, Neutron Flux 5
R(.4, 5)
S/U(1),M(9)
N.A.
N.A.
2##, 3, 4, 5 l
. s N.
6.
Overtemperature AT S
R M
N.A.
N.A.
1, 2
[
3 5
'R M-N.A.
N.A.
1, 2 l
k k. /.
. Overpower AT-
!R gg 8.
Pressurizer. Pressure-Low 5-R M
N.A.
N.A.
1 l
[ [
9.
Pressurizer Pressure-High 5
R M
'N.A.
N.A.
1, 2 l
M 10.-
Pressurizer Water Level-High 5
R M
N.A.
N.A.
1 l
31 11.
Reactor Coolant Flow-Low 5
R M
N.A.
N.A.
1 l
22
~
~. -.
t TABLE 4.3-1 i = = IT 1 (Continued) l n.
-4 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS g.
4 TRIP h
ANALOG ACTUATING MODES FOR 4
Z CliANNEL DEVICE WHICH
.h CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECT CALIBRATION TEST TEST LOGIC TEST IS REQUIRED y
C
12. ' Steam Generator Water Level-S R(13)
M N.A.
N.A.
1, 2 l
Low-Low
13. 'Undervoltage - Reactor Coolant N.A.
R N.A.
M N.A.
.1 g
Pumps-14.
Underfrequency - Reactor N.A.
R N.A.
M N.A.
1 l
1
' Coolant Pumps w.
N
~
15.
Turbine Trip a.
Stop Valve Eli Pressure N.A.
R N.A.
S/U(1, 10)
N.A.
1#
}-
- Low
,o a
ca:6 b.
Turbine Stop. Valve Closure N.A.
R N.A.
S/U(1, 10)
N.A.
18 4
w
~ 16.
Safety Injection Input from N.A.
N.A.
N.A.
R**
N.A.
1, 2 l
ESF j jy 17.
Reactor Trip System Interlocks eegg a.
Intermediate Range Wg Neutron Flux, P-6 N.A.
R(4)
M.
-F.A.
N.A.
2##
l oo b.
Low Power Reactor
' gg Trips; Block, P-7 N.A.
R(4)
M(8)
N.A.
N.A.
1
)
c.
. Power Range Neutron
$ 9, Flux, P-8 N.A.
R(4)
M(8)
'N.A.
N.A.
1 4-gg d.
Low Power Range Neutron Flux, P-9 N.A.
R(4)
M(8)'
N.A.
N.A.
1 o=
3 ee b
14 V
This surveillance need not be performed until prior to entering STARIUP following the Unit 1 first
~
refueling.
4 i
4 G
r TABLE 4.3-1 T0" '" "-1 (Continued) l J
es.
t D
35 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
om
p TRIP ANALOG ACTUATING MODES FOR I
C:
CHANNEL DEVICE WHICH 21 d
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS' REQUIRED so E
l 17.
Reactor Trip System Interlocks (Continued) i e.
Power Range Neutron Flux, P-10 N.A.
R(4)
M(8)
N.A.
N.A.
T f.
Power Range Neutron Flux, Not P-10 M(8)
N.A.
N.A.
1, 2 M.A.
R(4) u, N
l 2
g.
Turbine Impulse Chamber Pressure, P-13 N.A.
R M(8)'
N.A.
N.A.
1 i
o, 0F
~
9R 18.
Reactor Trip Breaker N.A.
N.A.
N. A.
M(7, 11)
N.A.
1, 2, 3*,
4*, 5*
19.
Automatic Trip and Interlock N.A.
~
- N.A.
N.A.
N.A.
M(7) 1, 2, 3*, 4 *, 5*
Logic 20.
Reactor Trip Bypass M.A.
M.A.
N.A.
-M(7.15)R(16) N.A 1, 2, 3*, 4*, 5*
t l[ff Breakers
=o 99 zz
.O.O
[
i
D,_.
o C C ~
bh
. 0?:
e
_q_y.,
TABLE 4. 3-14-OR-UNI-T-1-(Continued)
TABLE NOTATIONS Only if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.
4 Above P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint.
Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
l (1) If not performed in previous 7 days.
(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.
Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(3) Single point comparison of incere to excore axial flux difference above 15% of RATED THERMAL POWER.
Recalibrate if the absolute difference is greater than or equal to 3%.
The' provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5) Detector plateau curves shall be obtained, evaluated and compared to manufacturer's data.
For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into H0DE 2 or 1.
(6)
Incore - Excore Calibration, above 75% of RATED THERMAL POWER.
The provisions of Specification 4.0.4 are not applicable for entry into MODE,2 or 1.
(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(8) With power greater than or equal to the interlock setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive status light.
(9) Honthly surveillance in H0 DES 3*, 4*, and 5* shall also include verifi-cation that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive status light.
(10) Setpoint verification is not applicable.
(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verifi-cation of the operability of the Undervoltage and Shunt trips.
(12) Deleted (13) for Unit 1, CHANNEL CALIBRATION shall ensure that the filter time constant associated with Steam Generator Water Level Low-Low is adjusted to a value less than or equal to 1.5 seconds.
(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip circuits for the Manual j
Reactor Trip Function.
The test shall also verify the CPERABILITY of the Bypass Breaker trip circuit (s).
CATAWBA - UNITS 1 & 2 3/4 A 3 m Amendment No.101 (Unit 1) l Amendment No. 95 (Unit 2)
TABLE 4. 3-1 f0R-4JHI-T-1-(Continued)
TABLE NOTATIONS (15) A local manual shunt trip on the bypass breakers shall be performed prior to placing breaker in service.
(16) The automatic undervoltage trip capability shall be verified OPERABLE.
CATAWBA - UNITS 1 & 2 3/4 A 3-12w)
Amendment No.101 -(Unit 1)
Amendment No. 95 (Unit 2)_
j
.___._..____..____.______._______._m TABLE 3.3-1 FOR UNIT 2 l
n 3
E t
REACTOR TRIP SYSTEM INSTRUMENTATION y
i I
MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE l
j5 d FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION k
1.
Manual Reactor Tri 2
1 2
1, ?
1 5*
10 2
1 2
3*
l 2.
Power Range, Neutron Flux liigh Setpoint 4
2 3
1, 2 2
a.
b.
Low Setpoint 4
2 l###, 2
~ 2 l
t 3.
Power Range, Neutron Flux 4
2 3
1, 2 2
f
[
High Positive Rate i
R 4.
Power Range,' Neutron Flux, 4
3 1, 2 2
High Negative Rate m
Y l
r 5.
Intermediate Range, Neutron Flux 2
1 2
188#, 2 3
i 2
t 6.
Source Range, Neutron Flux 1
2 2##
4 l'
a.
Startup t
b.
Shutdown 2
1 2
3*,
4*, 5*
10 2- [
7.
Overtemperature AT g
Four Loop Operation 4
'2 3
1, 2 6
' ss-kk a=
8.
Overpower AT
((~
Four Loop Operation 4
2 3
1, 6
j t
OO~ i 9..
. Pressurizer Pr sure-Low 4
2 3
1 6**
i I
l
. $9 22 j
o_ o_
h
{
UU I
I
\\
~
.w-,
n w
IABLE'3.3-1 FOR UNIT 2 (Continued) l n
2'; ~
r 4
E MININUM g
TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL T
OF CHANNELS TO TRIP OPERABLE MODES ACTION i
i d
10.
Pressurizer sure-High 4
2 3
1, -
6**
j h ' 11.
Pressurizer Water
-High 3
2 2
1 6
12. Reactor Coolant Flow-to N Single Loop (Above P 4 3/ loop 2/ loop in 2/1 in 1
6 a.
any oper-each oper-
~
ating loop ating loop 4 -
b.
Two' Loops (Above P-7 and 3/ loop 2/ loop ~
2/ loop 1
6 below P-8) wo oper-each r- -
a ~
loops a 'ng loop-R a
13.
Steam Generator Water 4/sta 2/sta gen 3/sta gen 1, 2 6**
y Level -Low-tow gen-in any each operating operating y
gen sta gen
j
.14.
Undervoltage-Reactor Coolant 4-1/ bus 2
3 1
6 I
s Pumps (Above P-7)
[ [
15.
Underfrequency-Reactor Coolant 4-1/ bus 2
3 1
6
.ss Pumps (Above P-7) kI' t

16. Turbine Trip a.
Stop Valve EH P ssure 4
2 3
INN 6
l
((,
.ww Low
}
.b.
Turbine S Valve Closure 4
4 1
1 11
. u, -.
Q 17. ' Safety I ~ ction Input
=s-from
.2 1
2 1, 2 9
ee i
i d
i 1
m
i TABLE 3.3-1 FOR UNIT 2 (Continued) l n
h-MINIMUM 1
TOTAL-12.
CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERA 8tE MODES ACTION i'-
c-FUNCTIONAL'UN r
1 g
18.
Reactor Trip'S Interlocks a.
Intermediat nge
.g 2
1 2
2M 8
C-Neutron Flux,
. b.
Low Power Reactor-
Trips Block, P-7
.P-10 Input 4
2 3
1
.8 j
or i
P-13 Input 2
1 2
1 8
5 i.
c.
Power Range Neutron 1 Q-
' Flux,.P-8 4
2 3
1 8
=-
- d.
Power Range' Neutron.
4 2
3 1
8-
y Flux, P-9 m
bf e.
' Power Range' Neutron Flux, P-10 4'
2.-
3 1
8 f
Power Range Neutron Flux,.Kbt P-10 4
.3 4
1, 2 8
! I-
-=g g.
Turbine' Impulse' Chamber h3 Pressure, P 2 1-2 1
8 aa. 19.
Reactor. Trip Breaker 2
1 2
,2 9, 12 22 oo~
2 1
2 3*,
, 5*
10 l
L
$ 9. 20.
Automatic Tr' and Interlock 2
1 2
1, 2 9-
. Logic:
2 1
2 3*, 4*, 5*
10
-l gg ' '
-11 N.A.
M.A.
N.A.
1, 2, 3*, 4*, 5*
13 ee 21.
Rea or-Trip Bypass Breakers 00
4 l
l TABLE 3.3-1 FOR UNIT 2 (Continued) l 1
TABLE NOTATIONS 1
i "Only if the Reactor Trip System breakers happen to be in the closed po tion I
and t Control Rod Drive System is capable of rod withdrawal.
i
"" Comply ith the provisions of Specification 3.3.2, for any portion the
nannel quired to be OPERABLE by Specification 3.3.2.
l
Below the
-6 (Intermediate Range Neutron Flux Interlock) Setp nt.
Below the P 10 (Low Setpoint Power Range Neutron Flux Interlo k) Setpoint.
Above the P-Reactor Trip on Turbine Trip Interlock) Setp nt.
1 ACTION STATEMENTS i
ACTION 1 - With the mber of OPERABLE channels one i ss than the Minimum Channels OPQtABLE reautrement, restore t inoperable channel j
to OPERABLE t stus within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or b in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
ACTION 2 - With the number o OPERABLE channe one less than the Total Number of Channel TARTUP and/o POWER OPERATION may proceed i -
provided the follow conditio are satisfied:
The inoperable $ ch nel i placed in the tripped condition a.
{
within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
b.
The Minimum Channel 0 RABLE requirement-is met; however, i
the inoperable,ch nel be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months -
for surveillance esting other channels per Specification J
4.3.1.1, and c.
Either, THE L POWER is rest eted to less than or equal to 75% of TED THERMAL POWER a d the Power Range Neutron Flux tri setpoint'is reduced to ess than or equal to l
85% of TED THERMAL POWER within hours; or,-the QUADR POWER TILT RATIO is monito d at least once per l
12 h rs per Specification-4.2.4.2.
ACT10N 3 - With t number of channels OPERABLE one less than the Minimum Chan als OPERABLE requirement and with the THE L POWER level:-
i a.
Below the P-6 (Intermediata Range Neutron F1 Interlock)-
Setpoint, restore the inoperable channel-to O RABLE status prior to increasing THERMAL-POWER above eP6-Setpoint;.or b-Above the P-6 (Intermediate Range Neutron Flux Inte ock)
Setpoint but below 10% of RATED THERMAL POWER, restor the-inoperable channel to OPERABLE status prior to increas g THERMAL POWER above 10% of RATED THERMAL POWER.
t 4
CATAWBA - UNITS 1 & 2
- 3/4 B 3 Amendment No.101(Unit 1) l 8 74 Amendment No. 95(Unit 2)'
(
,,,-...-_ _~__,,
_. - - - - _ _. _.- - -. _ - _ - _ _ - -... _. ~,
= _ _..
4 1
j TABLE 3.3 1 FOR UNIT 2 (Continued) l j
ACTION STATEMENTS (Continued) 3 1
ACTIO 4 With the number of OPERABLE channels one less than the M imum Channels OPERABLE requirement, suspend all operations i olving i
i positive reactivity changes.
i ACTION $ - elete ACTION 6 - W h the number of OPERABLE channels one less tha the Total Num r of Channels, STAATUP and/or POWER OPERAT may proceed prov ed the following conditions are satisfie.
i a.
Th inoperable channel is placed in the ripped condition wit n 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and b.
The M imum ChanncIs OPERABLE requirement is met; however, i
the ino rable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surve 11ance testing of othe channels per j
Specificat. n 4.3.1.1.
i ACTION 7 - Delete 1
ACTION 8 - With less than theiN gimum Numb r of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine bytobservati of the associated permissive status light (s) that th)sint riock is in its required state-j for the existing plant so' tion, or apply Specification 3.0.3.
ACTION 9 - With the number of OPERA E hannels one less than the Minimum Channels OPERABLE requ ement be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; howey
, one c nnel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveill nee testing er Specification 4.3.1.1, l
provided the other hannel is OPE BLE.
i ACTION 10 - With the number f OPERABLE channels ne less than-the Minimum Channels OPERAfLE requirement, restore the inoperable channel to OPERABLE ptatus within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or o n the Reactor trip breakers wi nin the next hour.
ACTION 11 - With the number of OPERABLE channels less th the Total Number of Cha els, operation may continue provided t e inoperable chan els are placed in the tripped condition wi in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
ACTION 12 - Wif.h one of the diverse trip features (Undervoltag or shunt i
iip. attachment) inoperable, restore it to OPERABLE tatus-within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply CTION 9.
The breaker shall not be bypassed while one of the div sc trip features is inoperable except for the time reouired for erform-ing maintenance to_ restore the breaker to OPERABLE status.
With the breaker bypassed,-' apply ACTION 9.
e A ION 13 With any reactor trip bypass breaker inoperable, restore the bypass breaker to OPERABLE status prior to placing it in service.
r
't CATAWBA - UNITS 1 & 2' 3/483-675 Amendment No. 101(Unit 1) l 8
}
Amendment No.~
95(Unit 2)
~_.
i l
h n
TABLE 3.3-2 FOR UNIT 2 I
I 2t j
E-REACTOR TRIP SYSTEM INSTRUNENTATION RESPONSE TIMES i-c l
5 fUNC110NAL UNII RESPONSE iIME i
d
. Manual Reactor
-p N.A.
I
)
g
- 1.
I e-m 2.
Power Range, Neutron lux
< 0.5 second*
t 3.-
Power Range, Neutron flux, l
High Positive Rate N.
[
4.
Power Range, Neutron flux, f
i High Negative Rate f 0.5 second*
l t
o R
5.
Intermediate Range, Neutron Flux N.A.
{
l 8'
6.
Source Range, Neutron Flux N.A.
oc T y "
7.
Overtemperature AT-
< 8 seconds
g i
8.
Overpower AT
< 8 seconds *
[
9.
Pressurizer Pressure-Low
-< 2 seconds
[
I' h '.10.
Pressurizer Pressure-High seconds sa
$ k 11.
Pressurizer Water Leve igh M.A.
ss Me zz
..o.o
+
a i
$9 22 i
as
/.
Response time of the neutron flux signa { portion
Neutrordetectors are exempt from response time testing.
j of he channel shall be measured from detector output or input of first electronic component in chhanel.
t m,
l 1'
'/
u rg.
.wm.
e m
+
m u
---i
A-e Y
IABLE 3.3-2 FOR UNii 2 (Continued)
$.f.
REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES h~'FUNCTIONALUNIl
~
RE5PONSE TIME t' Flow y
_12.g Low Reactor Coo C'
a.
Single Loop (Abowg P-8)
$ 3 second b.
Two Loops (Above and below P-8)
$ I second
- 13. -Steam Generator Water Level-Low
.5 seconds a.
Unit 1 b..
Unit 2
[2.0 seconds w-14.'
Undervoltage-Reactor Coolant Pumps -
1 1.5 seconds
.= 15.. Underfrequency-Reactor Coolant Pumps 1 0.6 second w
. 4 ' 16.
Turbine Trip
~
-4 N.A.
' a.
Stop Valve EH Pressure-Low
~ b.-
-Turbine Stop Valve' Closure M.A.
.A.
. 17.
Safety. Injection Input from EST
.$$'18
~
N.A.
Reactor Trip System Interlocks 3i-E M.A.
19; Reactor Trip Breakers gg
- E E 20. LAutomatic Trip'.and erlock Logic' N.A.
. g{
La 33.
00 mM
N g
n D
TABLE 4.3-1 FOR UNIT 2 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP c5 ANALOG ACTUATING MODES FOR M
CilANNEL DEVICE WilICH CilANNEL CHANNEL OPERATIONAL OPERATIONAL AG UAIION SURVEILLANCE IEST OGIC TEST IS REQUIRED
~
C FUNCTIONAL UNIT CHECK CALIBRATION TEST 1.
Manual Reactor Trip N.A.
N.A.
N.A.
R
)
N.A.
1, 2, 3*,
4*, S' 2.
Power Range, Neutron Flux x
a.
liigh Setpoint S
0(2, 4),
M N.A.
N.A.
I, 2 M(3, 4),
Q(4, 6),
R(4, 5) b.
Low Setpoint 5
(4)
M H.A.
N.A.
1###, 2
{
3.
Power Range, Heutron Flux, H.A.
R(4 M
N.A.
N.A.
1, 2 Y
High Positive Rate
/
w N
4.
Power Range, Neutron Flux, N.A.
R ]
M N.A.
N.A.
1, 2 High Negative Rate M
N.A.
N.A.
1###, 2 5.
Intermediate Range, S
R(4. 5)
S/U(2),\\
Neutron Flux gg R g 6.
Source Range, Neutron Flux 5
R(4, 5)
S/U(1),H(9)
N.A.
N.A.
2##, 3, 4, 5 2.
55 7.
Overtemperature AT S
R M
4 N.A.
1, 2 zz
??
8.
Overpower al 5
R H
N.A.
N.A.
1, 2 9.
Pressurizer P 'ssure-Low R
H N.A.
N.A.
I 22 11 10.
Pressuri r Pressure-High 5
R M
H.A.
N.A.
1, 2
<* e OC 11.
Pr urizer Water Level-High 5
R H
N.A.
N.A.
I\\
12 Reactor Coolant flow-low 5
R H
N.A.
N.A.
I m
.i' N,
l 1
n N TABLE 4_3-1 FOR UNII 2 (Continued) k.K.
M-REACIGR TRIP SYSTEM INSTRilMENTATION StIRVEIti ANCE REQulREMEN15
'7 TRIP ANA10G ACTUAi!NG MODE 5 ION 3
CHANNEL DEVICE milch g
CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEltLANCE
{ FUNCTIONAL UNIT CIIECK CALIBRATION TEST TEST LOGIC IEST 15 REQtilRED y.
13.
Steam Generator Water (evel-S R(13)
M
.A.
N.A.
I, 2 Lo& Low 14.
Undervoltage - Reactor Coolan M.A.
R N.A.
M N.A.
1 Pumps
15..Underfrequency - Reactor N.A.
R
.5
.A.
M N.A.
I
?
_yg Coolant Pumps
,. 16.
Turbine Trip 4
a.
Stop Valve EH Pressure.
N.A.
R N.A..
S/U(1, 10)
N.A.
la o
- Low-x d$
b.
Turbine Stop Valve Closure M.A.
R N. A.
S/U(1,10)
N.A.
1#
17.
Safety. Injection Input from.
N.A.
N.A.
N>.
R**
N.A.
1, 2 ESF-5 l" 18.
Reactor Trip System Interlock g
(l-
.a.
Intermediate Range Neutron Flux, P-6 N.A.
R(4)
M N.A.
M.A.
2##
gg
' z z..
b'.'
Low Power Re or.
Trips Block P-7 N.A..
R(4)'
M(8)
N.A.
N.A.
I e E>
c.
Power Range Neutron
,XX Flux
-8 N.A.
R(4)
M(8)
N.A.
N.A.
1
' EE d.
.L
- ower Range Neutron ux, P-9 N.A.
R(4)
M(8)
N.A.
M..
1
.00 This surveillance need not be performed until prior to entering STARTUP following the Unit I first refueling.
-i:
l k-TABLE 4.3-1 FOR UNIT 2 (Continued) i p
y REACTOR TRIP SYSTEN INSTRUMENTATION SURVEILLANCE REQUIRENENTS TRIP e
ANALOG ACTUATING MODES FOR
[
5 CHANNEL DEVICE WHICH CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUAi10N SURVEILLANCE l
y, C FUNCIl0NAL' UNIT-CllECK CALIBRATION TEST TEST
/10GIC TEST 15 REQUIRED 18.
Reactor Trip System Inte locks (Continued) i e.
Power Range Neutron j-Flux, P.A.
R(4)
N(8)
N.A.
N.A.
I i
3 j
f.
Power Range Neutron Flux, Not P-10
}
N. A.
R(4) y M(8)
N.A.
M.A.
1, 2 w
i
?
ar -
g.
Turbine Impulse Chamber J.
Pressure, P-13 N.A.
R M(8)
N.A.
N. A. -
1
[
w
'fe "-
s 19.
Reactor Trip Breaker N.A.
.A.
'N.A.
M(7, 11)
N.A.
1, 2, 3*, 4*, S*
l 3
20.
Automatic Trip and Interlock N.A.
N.A.
M.A.
M(7) 1, 2, 3*, 4*
5*
t
. Logic k k. 21.
Reactor Trip Bypass
'. A.
N.A.
N.A M(7,15)R(16)
N.A 1, 2, 3*, 4*, 5*
'33 Breakers r
ff i
t ge 55 l
d 22.
11 e <+
j' UU W? - - --
1
\\\\'
TABLE 4.3 1 FOR UNIT 2 (Continued)
TABLE NOTATIONS Only if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.
ove P 9 (Reactor Trip on Turbine Trip Interlock) Setpoint.
Be w P 6 (Intermediate Range Neutron Flux Interlock) Setpoint.
Belo P-10 (Low Setpoint Power Range Neutron Flux Interlock) Satpoi (1)
If not rformed in previous 7 days.
(2)
Comparts of calorimetric to excore power indication above 15% f RATED i
THERMAL P0 R.
Adjust excore channel gains consistent with c torimetric 1
power-if absb ute difference is greater than 2%.
The provis ns of j
Specification
.0.4 are not applicable for entry into MODE or 1.
)
(3)
Single point com rison of incore to excore axial flux o forence above i
15% of RATED THERM (L POWER. Recalibrate if the absolut difference is l
greater than or equbl to 3%.
The provisions of Speci cation 4.0.4 are
]
not applicable for en ry into MODE'2 or 1.
l (4)
Neetron detectors may b excluded from CHANNEL C BRATION.
l (5) Detector plateau curves s s11 be obtained, eval e ted and compared to i
manufacturer's data.
For t e Intermediate Ra e and Power Range Neutron Flux channels the provisions f Specificatio 4.0.4 are not applicable j
for entry into MODE 2 or 1. ",
(6)
Incore - Excore Calibration, abo e 75% o RATED THERMAL POWER.
The provisions of Specification 4.0.4 re t applicable for entry into i
j MODE 2 or 1.
(7)
Each train shall be tested at less e ry 62 days on a STAGGERED TEST BASIS.
2 (8) With power greater than or equa to the terlock setpoint the required j'
ANALOG CHANNEL OPERATIONAL TE shall cons it of verifying that the interlock is-in the required state by obseryl g the permissive status j
light.
(9) _ Monthly surveillance in DES 3*, 4*, and 5* sh 11 also include verifi-i cation'that permissive P-6 and P-10 are in their equired state for j
existing plant condit ons by observation of the pe issive status light.
j (10) Setpoint verificat n-is not applicable.
(11) The TRIP ACTUATI DEVICE OPERATIONAL TEST shall include independent verifi-cation of the dervoltage and Shunt. trips.
(12) Deleted j
(13) For Unit 1 CHANNEL CALIBRATION shall ensure that the filter me constant associat with Steam Generator Water Level Low-Low is adjuste to a value less t n or. equal to 1.5 seconds.
(14)The/RIPACTUATINGDEVICEOPERATIONALT.!STshallindependentlyveri the-OPfRABILITY of the.unoervoltage and shunt trip circuits for the Manua factor Trip Function, The test sha11'also verify the OPERA 8ILITY of e
j ypass Breaker trip l circuit (s).
CATAWBA --UNITS 1 1 2
'3/4 B 3-12
- Amenoment-No.101 (Unit-1)
Amenoment No. 95_(Unit 2)
. g.g g _
/1
J\\
TABLE 4,3-1 FOR UNIT 2 (Continued)
TABLE NOTATIONS (15) A local manual shunt trip on the bypass breakers shall be performed prior to lacing breaxer in service.
(16) The automatic undervoltage trip capability shall be verified OPERABL i
4 h
1 4
1 4
CATAWBA - UNITS 1 & 2 3/4 B 3-12a Amendment No.101 (Unit 1) l g.82 Amendment No. 95 (Unit 2)
TABLE 3.3-4 T;r l'-it I nij ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS l FUNCTIONAL UNIT' TPIP SETPOINT ALLOWA8LE VALUE
=
~
d'
1. Safety Injection (Reactor Trip, Phase "A"' Isolation, Feedwater C
Isolation, Control Room Area
-Ventilation Operation, Auxiliary Feedwater-Motor-Driven Pump,.
Purge & Exhaust Isolation, Annulus
~
ventilation Operation, Auxiliary Building Filtered Exhaust Operation, Emergency
. Diesel Generator Operation, Component Cooling Water.
w-)
Turbine Trip, and Nuclear
Service Water Operation)
'a..
Manual IniLiation N.A.
N.A.
de b.
Automatic Actuation Logic N.A.
M.A.
and Actuation Relays c.
Containment Pressure-High
$ 1.2 psig
$ 1.4 psig 22 d.
Pressurizer. Pressure-tow
> 1845 psig
> 1839 psig
2, 2, RA W!
e.
. Steam Line Pressure-Low 3 775 psig 3 744 psig
. s s,
2. Containment. Spray.
zz
??
a.
Manual Initiation.
N.A.
N.A.
22 b.
~ Automatic Actuation Logic M.A.
N.A.
and.' Actuation Relays Containment Pressure-High-High' $ 3 psig i 3.2 psig c.
TABLE 3.3-4 for fr.it 1 (Continued) 9 ENGINEERED SAFETY FEATURES A TUATION SYSTEM INSTRUNENTATION TRIP SETPOINTS
.a7 E
TRIP SETPOINT ALLOWABLE VALUE FUNCTIONAL UNIT C5
3. Containment Isolation
,g Phase "A" Isolation a.
w E
1) Manual Initiation N.A.
N.A.
2)' Automatic Actuation Logic N.A.
N.A.
ar.d Actuation-Re1ays See Item 1. above for all Safety Injection Setpoints and Allowable Values.
3) Safety Injection L-b.
Phase "B" Isolation (Nuclear p
Service Water Operation) m
1) Manual Initiation M.AJ N.A.
y A
2), Automatic Actuation M.A.
N.A.
Eg Logic and Actuation Relays
,3) Containment Pressure-
$ 3 psig 5 3.2 psig High-High
-22 Purge and Exhaust Isolation.
c.
'g g-
1) Manual Initiation N.A.
N.A.
gg
,E,E ~
2) Automatic Actuation M.A.
N.A.
Logic and Actuation
-go Relays See Item 1. above for all Safety Injection Setpoints and Allowable Values.
22 3)' Safety injection gg.
eao U
=
_______...__..m_-_._
TABLE 3.3-4 for Ur.it 1 (Continued) nD ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
-6 I
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE J
.g
4. Steam'Line Isolation a.
Manual Initiation N.A.
M.A.
l j
3 g
~b.
Automatic Actuation Logic N.A.
N.A.
1 and' Actuation Relays j
N Containment Pressure-High-High $ 3 psig i 3.2 psig j
c..
r d.
Steam Line Pressure - Low 1 775 psig 3 744 psig e.
Steam Line Pressure-
$ 100 psi
$ 122.8 psi **
3 Negative Rate - High
'5. leedwater Isolation
(
}
a.
Automatic Actuation Logic-N.A.
M.A.
w Actuation Relays-w b.
Steam Generator Water
}
E L'evel-High-High (P-14) 60 g
1.
Unit 1 5 82.4% of
$ 84.2% of narrow narrow range range instrument.
j l
instrument span span t
2.
Unit 2
-< 77.1% of
-< 78.5% of narrow l
t a3 narrow range range instrument ee instrument span
'yy l
span ee n3
-> 561*F T
> 564*F c.
T g Low av
.zz L
d.
Doghouse Water Level-High 11 inches 12 inches above 577' above 577'
,g' floor level floor level ui -
Sa'fety Injection See Item 1..above for.a11' Safety Injection Setpoints and Allowable Values _
e.
t de to 6
l vv L
TABLE 3.3-4 far-Unit 1 (Continued) nD E
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SET._4
.a FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE 2>
SE
6. Turbine Trip.
a.
Manual Initiation N.A.
N.A.
hh
$2 b.
Automatic Actuation N.A.
N.A.
Logic and Actuation Relays
^'
c.
Steam Generator Water Level-fligh-liigh (P-14) 1.
Unit I
< 82.4% of
< 84.2% of narrow narrow range range instrument instrument span span 5 77.1% of 5 78.9% of narrow h$
2.
Unit t narrow range range instrument 2
Instrument span w
span g
d.
Trip of All Main N.A.
N.A.
Feedwater Pumps e.
Reactor Trip (P-4)
N.A.
N.A.
f.
Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable
,, 3, Values.
33
's '$
((((
7. Containment Pressure Control gg System a.
Start Permissive 5 0.4 psid 5 0.45 psid r+ e z2??
b.
Termination
> 0.3 psid
> 0.25 psid
$h
8. Auxiliary Feedwater IEIE a.
Manual Initiation N.A H.A.
oo EIN b.
Automatic Actuation Logic N.A.
N.A.
and Actuat*on Relays N >"
vv
_ _ _ _ _.. _.. _ _ _. _. _. _...... ~.... _ _. _ _ _ _
T ABLE 3. 3-4 for l'r.it 1-- (Centinued)
L n-ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Y,
E FUNCTIONAL-UNIT TRIP SETPOINT ALLOWA8LE VALUE
' 8.
Auxiliary Feedwater (Conthw o.,
C5 c.
Steam Generator Watt-d Level - Low-Low 5
1) 11 nit I
> 17% of span
> 15.3% of t
from 0% to span from 0% to 30% RTP
.30% RTP increasing increasing linearly to linearly to 1 38.3% of. span
> 40.0% of.
from 30% to 100%
span from 30%
RTP to 100% RIP
)
'.2)
Unit 2
> 36.8% of
> 35.1% of narrow w-narrow range range instrument t
-span span See Item 1. above for all Safety. Injection Setpoints and Allowable Values.
Y d.
Safety Injection b
Loss-of-Offsite Power 1 3500 V 3 3200 V e.
f.
Trip of All Main Feedwater Pumps H.A.
N.A.
L g.
Auxiliary feedwater Suction Pressure-Low'
.M y
.1)
CAPS 5220, 5221, 5222 3 10.5 psig-3 9.5 psig aa aa"5
2) CAPS 5230, 5231, 5232 3 6.'2 psig 3 5.2 psig j
35 a.
Unit 1
'3 6.2 psig 3 5.2 psig b.
Unit'2 3 6.0 psig.
.3 5.0 psig 22 OO
~ ~ 9.
Containment-Sump Recirculation
$b Automatic Actuation Logic N.A.
N.A.
a 22 and Actuation Relays oa Refueling Water Storage" 1 177.15 inches 1 162.4 inches
$3 b.
i Tank Level-Low mg
. Coincident'With Safety See item 1. above for all Safety. Injection Setpoints and Allowable Values.
' Injection i
TABLE 3.3-4 For "r.~t I (Continued)
^
n i.
--4
- ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS -
7
.m
> ' FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE'VALUE r
c-jz!
10;. loss of, Power
- -4 m
^
a.
4 kV Bus Undervoltage-Loss 3 3500 V 3 3200 V
.. gg of Voltage b.
4 kV Bus Undervoltage-1 3685 V 3 3611 V Grid-Degraded Voltage.
4
{
.11.
Control Room Area Ventilation
Operation a.
. Automatic Actuation Logic
..y and. Actuation Relays
.- N. A.
N.A.
b
. Loss-of-Of fsite Power 1 3500 V 1 3200 V c.
Safety Injection
.See Item L above for all Safety Injection Setpoints and Allowable -
E Values.
12.
Containment Air Return and l
Hydrogen Skimmer Operation sa a.
-Manual.. Initiation
N.A.
N.A.
>> 2,
.. uu oo
- i ss-b.
-Automatic-Actuation Logic.
'N.A.
N. A..
an wo and Actuation Relays as 4
o r+
oo c.
' Containment Pressure-
'5 3 psig; 5 3.2 psig l
zz
~]
High-High'
.-e ma
.o=
q fu H..
vv e
Y e
. -. -__.. ~. _ _ _ _ _ _ _ _ _ _.. _ _ _ _ _ _ -....
TABLE 3.3-4 Fcr Unit 1 (Continued) n
^
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
~
lg TRIP SETPOINT ALLOWABLE VALUE FUNCTIONAL UNIT c:
25
13. Annulus Ventilation Operation Manual Initiation N.A.
N.A.
a.
r s,
R$
b.
Automatic Actuation Logic N.A.
N.A.
and Actuation Relays See Item 1. above for all Safety Injection Setpoints and A1,10wable Values.
c.
Safety Injection
14. Nuclear Service Water Operation a.
Manual Initiation' N.A.
N. A-u, ha b.
' Automatic Actuation Logic H.A.
N.A.
3,
~
.and Actuation Relays u,
J.
l e'
.c.
Loss-of-Of fsite Power 1 3500 V 2 3200 V I
See Item 2.. above for all Containment Spray Setpoints and Allowable Values.
d.
Containment Spray t
See Item 3.b. above for all Phase "B" Isolation Setpoints and Allowable e.
Phase "B" Isolation' Values.
E $I See Item 1. above for a11' Safety. Injection Setpoints and Allowable' Values.
kk f.
Safety Injection 3U Suction Transfer-Low Pit Level
>E1. 554.4 f t.. >E1.T552.9 ft..
a llll g.
2z i
PP 15.
Emergency' Diesel Gene itor I
Operation (Diesel.Butading ES S3 Ventilation Operation, Nuclear 22 ;5 Service Water Operation) oa IIII Manual Initiation N.A.
N.A.
a.
SU
.... _.. _. _ _ _..... _.. _. _.. _ _. _ _ -. _ _. _ _ _. - _ _ ~. _ _ _ _ -.. _ -.
i TABLE 3.3-4For'Jnit1-(Continutd).
'g
-E-ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
-4 4
'3E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE
!k'15.
Emergency Diesel Generator
[
' D!.
Operation (Diesel. Building
. Ventilation Operation, Nuclear g
Rl Service. Water Operation) (Continued) i.
b.
-Automatic Actuation Logic N.A.
N.A.
a'nd Actuation Relays c.
Loss-of-Of fsite Power 3 3500 V 3 3200 V' d.
Safety Injection See Item. I above for a11' Safety. Injection Setpoints and Allowable Values.
h
16. Auxiliary Building Filtered Exhaust Operation 3,
i J,
a.
Manual Initiation N.A.
N.A.
w s
b.
Automatic Actuation Logic N.A.
N.A.
h
\\
.and Actuation' Relays See Item 1. above 'for all Safety ' Injection Setpoints and Allowable Values.
l c.
Safety Injection i
2, p 8g 11 Diesel Building Ventilation
'Eg Operation 55 llll a.
' Manual-Initiation N.A.
N.A.
?'
.o.o b.
Automatic Actuation Logic N.A.
N.A.
zz i
-and Actuation Relays j
u$ 5?
See Item 15. above for all Emergency Diesel Generator Operation Emergency Diesel Generator l
q;;;
c.
EL EL -
Operation Setpoints-and Allowable Values.
i n re h) H
~ - - -
i 5
[
i i-
I TABLE 3.3-4 is: 'J:-it 1 (Continued)
I-
'a n.
.4.g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM udSTRUMENTATION TRIP SETPOINTS
{
' s-TRIP SETPOINT ALLOWA8LE VALUE-
' FUNCTIONAL UNIT e.
25 18.~ Engineered' Safety. Features
'd
' Actuation System Interlocks f
e.
7
~C a.
Pressurizer Pressure,'P-11 1955 psig
>1944 psig b'.
Pressurizer Pressure, not P-11.1955 psig
$1966'psig
. l c.
Low-Low Tavg.-P-12
->553*F
>550*F'
. i d.
Reactor: Trip, P.-41 N.A.
'N.A.
e.
Steam Generator Level, P-14 See Item 5. above for all Steam Generator Water Level Trip Setpoints w) and Allowable Values.
l W
eW
'j J"
2 6
1 4
'~
'33...
t 33-
& W.:
g 22
- o.o
- un _s Ah CC 33 a.
e.
y hY i
.Np~
vv
+
a.
i.
1-t e
~.
TABLE 3. 3-4 -FeHJrt+t (Conti nued)
TABLE NOTATIONS
The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconds.
Channel calibration shall ensure that this time constant is adjusted to this value.
I a
i l
4 i
CATAWBA - UNITS 1 & 2 3/4 A 333k.
Amendment No.101 (Unit 1)
Amendment No. 95 (Unit 2)
m_._,__,____
\\s s
, TABLE 3.3-4 For Unit 2 r,
I 21!
NGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
'f
. 3> :
~. FUNCTIONAL T
TRIP SETPOINT ALLOWABLE VALUE
.e.
. c:
z :-
Li
11. Safety Injectio (Reactor Tr.ip,
. Phase'"A" Isolati Feedwater
r,
' R; Isolation.. Control Area Ventilation. Operation, Auxiliary Feedwater-Notor-Driven P
' Purge'& Exhaust Isolation, nulus Ventilation Operation, Auxiliary Building Filtered Exhaust Operation, Emergency 4
Diesel Generator. Operation,
..-5
-Cocponent. Cooling Water,.
j u,
3r Turbine Trip, and' Nuclear.
Service Water. Operation) a3 us N.A.
.A.
4, a.
' Manual Initiation f
~a 4:
.b.
Automatic Actuation Logic N.A.
N.
i Oc and: Actuation Relays l
o
/ Containment Pressure-High
.2 psig
$l1.4 psig c.
g g
-d.
. Pressurizer. Press ~re-Lowf
>-1845:psig
> 1839 psig 22, 4
u 2
t roo
'$ g-e.
Steam Line Pressure-Low it 725 psig
> 694 psig*
CL CL sa-r+ (+
.'2. Containment Spray 22
.O.O "
N.A.
N.A.
a.
Manual Initi ion j
(_
u,.
2373L b.
Autnma
Actuation Logic N.A.
N. A. -
j E' 5L.
-and-tuation Relays c+ c+
l ontainment Pressure-liigh-liigh 5 3 psig.
5 3.2 psig 2323 c.
i.
4 l
1-t 3:
i N.
. -.. _.. ~,. _. - _ _.. -..... _.... _ _.... _ _.
TABLE 3.3-4 For Unit 2 (Centinued) nn
~
E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS r
co>
s FUNCTIO L UNIT.
TRIP SETPOINT -AtLOWABLE VALUE l
3. Contain Isolation 4
[
W
'a.
Phase "A" 'I olation
- l5
1) Manual Ini ~ tion N.A.
N.A.
2) Automatic Actuat' n Logic H.A.
N.A.
and Actuation Rela
'3)
Safety Injection-See Item 1. above for all Safe Injection Setpoints and Allowable Values.
b.
Phase "B" Isolation (Nuclear w
'S -
Service' Water Operation) co
1) ' Manual Initiation H.A.
N.
w to
2) Automatic Actuation
'N.A.
N.A.
[
.y Logic and Actuation Relays
3) Containment Pressure.
psig
< 3.2 ps' High-High l
Purge and Exhaust Isolatio
.S E' c.
3 R$
'SS 1). Manual Initiatio N.A.
N.A.
-zz Automatic A ation N.A.
N.A.
i
??
2)._ Logic and ctuation 4
IE 2 Relays 22-
. 3. 3.
3).5 ety-Injection See. Item 1. above for all Safety Injection Setpoints d Allowable Values.
r+ r j
.' ((
m d.
i I
IABLE 3.3-4 For Unit 2 (Continued) n De ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS i
FUNCTIONAL U II TRIP SETPOINT ALLOWA8LE VALUE
.E 4.SteamLineIsD1 tion 3
a.
Manual Initi ion N.A.
-N.A.
g-b.
Automatic Actuat' n Logic N.A.
N.A.
N and Actuation Rela c.
. Containment Pressure-h-High 5 3 psig 5 3.2 psig
d., Steam Line Pressure - Low 1 725 psig 1 694 psig*
e.
Steam Line Pressure--
< 100 psi
-< 122.8 psi **
Negative Rate High
'S. feedwater Isolation Q
a.
Automatic Actuation Logic N.A.
N.A.
, Actuation Relays-y b.
Steam Generator Water g
Level-High-High (P-14) 1.
Unit'l 5 82.4%
$ 8 $ of narrow i
narr ange range strument inst ument span i
an
. 2.
Unit 2 5 77.1% of 5 78.9% of narr
-g,
narrow range range instrument l
g,g g
instrument
' span span 3
,((
T,yg-Low 1 564'F 1 561"F c.
i
~PP d.
Doghouse Wa Level-High 11 inches 12 inches i
e3 above 577' above 577'
Z floor' level floor level i
j- $$
' e.
.Sa y-Injection See Item 1. above for all Safety Injection Setpoints an Allowable Values.
.j
. ' < + < +
NN l
vv j
1
,t
, i;
!<[
j!
!i
I;!
!I tlIe! -
l i.; ;, iIi e
~
l baw o
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s t
n i
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te S
n S
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S d d N
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f CS a b A a b U
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.,ll I
ciI *
f
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4 1i
TABLE 3.3-4 For Unit 2 (Continued)
n..
ENGINEERED SAFETY' FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 2g
'kFUNCTIONAL'IT-TRIP SETPOINT ALLOWABLE VALUE' l
' 8.
-Auxiliary e6 water (Continued)'
.c5-c.
Steam Ge ator Water
.[
j.
Level'- Low-p
1) Unit I-3 17% of span 3 15.3% of 1
from 0% to span from 0% to.
j N
30% RTP 30% RTP increasing i
increasing linearly.to linearly to "1 38.3% of, span 40.0% of from 30% to 100%.
l spa from 30%
RTP-l 10.10 RTP-
2) -Unit 2 1 36.8% o 1 35.1% of trow l
narrow range range.in unent m
.}
span span w
,d.. Safety Injection See. Item 1.'above r.all Safety. Injection Setpoints and Allowable Values.
.l o,
e e.
Loss-of-Of fsite Power
-> 3560 V
> 3200 r
w e
l!
f.' Trip of All Main Feedwater.
l Pumps N.A.
N.A.
g.
Auxiliary feedwater Suction-Pressure-Low' NN gg
1) CAPS 5220, 5221, 522 3;10.5 psig 2 9.5'psig gy
-2) CAPS 5230, 5231, 32' 1 6.2 psig
> 5.2 psig.
2a a.
Unit 1 3'6.2.psig 1 5.2 psig-gg
.zz
? 5.0 psig
'b.
Unit 2-1 6.0 psig
?P,9.
. Containment S ecirculation N3'
.a.
Automat Actuation _. Logic H.A.
'N.A.
and uation-Relays gg b.
eling' Water Storage 3 177.15-inches 1.162.4, inches
[-h-l Tank Level-Low
' vv Coincident With~ Safety es.
-Injection See item 1. above for all Safety Injection'Setpoints ared Allowable Va I
-s,N TABLE 3.3-4 For Unit 2 (Continued) n D
E-NGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS m
TRIP SETPOINT ALLOWABLE VALUE FUNCTIONAL UNIT i
g*
10.
Loss of Powe Y
4 kV Bus Un e voltage-Loss
~> 3500 V
-> 3200 V s.
a.
~
U of Voltage b.
4 kV Bus Undervoltag -
. > 3685 V 3 3611.V.
Grid Degraded Voltage 11.
Control Room Area Ventilation Operation Automatic Actuation Logic e -
a.
)
and Actuation Relays N.A.
N.A.
w b.
. Loss-of-Of fsite Power
> 3500 V
_ 3200 V m
wO M
c.
Safety Injection See Item. abo for all Safety Injection Setpoints and Allowable Values.
N-12.
Containment Alr Return and liydrogen Skimmer Operation-N.A.
H.A.
a.
Manual Initiation-I 2R R S5 22 b.
Automatic Actuation gic-N.A.
N.A.
55 and Actuation Rei s
?P
. c.
Containment. essure-
'$ 3 psig
.$ 3.2 psig zz High-High g
a,--
22
.la r+ <+
,. ~.. - ~.
-.. - - - ~... -
.. - ~ - -.... - - - -.. -.. -...
. - -.. ~. -. -. -...
N TABLE 3.3-4 For Unit 2 (Continued) g NGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS g
FUNCTIO NIT TRIP SETPOINT ALLOWABLE VALUE Eg
.13. Annulus' Vent ation Operation m
a.
Manual Initi ion H.A.
N.A.
g s
to b.
Automatic Actuati Logic N.A.
N.A.
-and Actuation Relays c.
Safety Injection See Item 1. above for all Safety I ection Setpoints and Allowable Values.
14. Nuclear Service Water Operation w
a.
Manual Initiation N.A.
N. A.-
t N.
N. A.,
N.
i cn b.
Automatic Actuation Logic and Actuation Relays.
w
-> 3500 V
- 200 V c.
Loss-of-Of fsite Power 3
a2.aboveforal(ContainmentSpraySetpointsandAllowableValues.
d.
Containment Spray See I Ji e Item 3.b. above,.for albhase "B" Isolation Setpoints and Allowable e.
Phase "B" Isolation
/ Values.
28 EE f.
Safety Injection See Item 1. above for all Safety Inj tion Setpoints and Allowable Values.
l 35 5S g.
Suction Transfer-L Pit Level
>El. 554.4 ft.
>El. 552.9 ft.
zz
??
15.
Emergency Diesel G erator m6 Operation.(Dies Building ventilation.0 ration, Nuclear 22 Service Wa r Operation)
.55
~ N.A.
nual Initiation M.A.
a.
3g t
/
1-.
s TABLE 3.3-4 For Unit 2 (Centinued) gy 3E ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
.-4 m
TRIP SETPOINT Att0WABLE VALUE FUNCTI0fALAUNIT C
33' 15.
Emergency Die (el Generator Building d
Operation (Diesa{ tion, Nuclear Ventilation Opera s
E3 Service Water Operation) (Continued) b.
Automatic Actuation gic N.A.
N.A.
a'nd Actuation Relays c.
Loss-of-Of fsite Power 1 3500 V 3 3200 V ee Item. I above for all fety Injection Setpoints and Allowable Values.
d.
Safety Injection w
2:
16. Auxiliary Building filtered Exhaust Operation m
Manual Initiation N.A.
N.A.
w0 a.
P si b.
Autcmatic Actuation Logic H.A.
H.
and Actuation Relays
~
e Item 1. above for all fety Injection Setpoints and Allowable Values.
Safety Injection c.
s%
II. Diesel Building Ventilation ER Operation
?$
Manual Initiation N.A.
N.A.
5!5!
a.
PP b.
Automatic Act ion Logic N.A.
N.A.
zz and Actuati Relays
,g See Item 15. above for all Emergency Diesel Generato Operation w-72 72 c.
Emerg cy Diesel Generator Setpoints and Allowable Values.
EL EL Ope tion en
TABLE 3.3-4 For Unit 2 (Continued) n D
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS g
2 TRIP SETPOINT ALLOWABLE VALUE FUNCTION A UNIT C5
18. Engineere afety Features d
Actuation Syh m Interlocks Pressurizer Preqsure, P-11 1955 psig
>1944 psig h
a.
b.
Pressurizer Pressu not P-11 1955 psig
$1966 psig Low-Low T,y, P-12 3553*F 3550 F c.
d
. Reactor Trip, P-4' N.A.
H.A.
Steam Generator-Level, P-14 Se Item 5.aboveforp Steam Generator Water Level Trip Setpoints e.
an llowable Val es.
w
}
co 5
~
a*
8R u s.+
e s.e O 22 hh Oc
TABLE 3.3-4 For Unit 2 (Os tinued) l TABLE NOTATIONS
Time constants utilized in the lead-lag controller for-Steam t.ine Pressur -
ow are t i > 50 seconds-and is < 5 seconds.
Channel calibration shall-e sure that these time constanti are adjusted to these values.
The ime constant utilized in the rate-lag controller for Steam Li Press e-Negative Rate-High is greater than or equal to 50 secon Channel calibration shall ensure that this time constant-js adj ted to this lue.
f-L
[.
?
1 l
. CATAWBA - UNITS 1&2 3/4 B 3-36 Ame'ndment No.101: (Unit 1) l 8-102 Amendment-No. ' 95 (Unit 2)
~
v v
w..,.--,33
...,rw
or_,
..e.3 a
.me
,y
.,w.
-,.+,
j 4
~
)
TABLE 3.3-5 FOR UNIT 1 ENGINEERED SAFETY FEATURES RESPONSE TIMES l
INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
]
1.
Manual Initiation a.
Safety Injection (ECCS)
N.A.
b.
Containment Spray H.A.
I c.
Phase "A" Isolation N.A.
d.
Phase "B'? Isolation N.A.
l e.
Purge and Exhaust Isolation N.A.
]
f.
Steam Line Isolation N.A.
g.
Diesel Building Ventilation Operation N.A.
b.
NuclearServiceWaterOpefation N.A.
i.
Turbine Trip N.A.
l j.
Component Cooling Water N.A.
]
k.
Annulus Ventilation Operation N.A.
1.
Auxiliary Building Filtered H.A.
Exhaust Operation m.
Reactor Trip N.A.
j n.
Emergency Diesel Generator Operation N.A.
o.
Containment Air Return and Hydrogen i-Skimmer Operation N.A.
i p.
Auxiliary Feedwater N.A.
l 2.
Containment Pressure-High a.
Safety Injection (ECCS) i 27(1)/12(3) f
1) Reactor _ Trip-12
2) Feedwater Isolation
< 12 l
l
3) Phase "A" Isolation (2)
-1 18(3)/28(4)
4) Purge and Exhaust Isolation
<6 l
5) Auxiliary Feedwater(5)
{A.
'6) Nuclear Service Water Operation i65(3)/76(4)
7) Turbine Trip N.A.
8) Component Cooling Water 1 65(3)/76(4)'
9) Emergency Diesel. Generator Operation i 11
10) Control Room Area Ventilation Operation-N.A.
4 l-CATAWBA - UNITS 1&2
- 3/4 A 3.F})03 Amendment No.101'(Unit 1).
[
Amendment No. - 95 (Unit 2)-
1 4
TABLE 3. 3-5'-FOR-(MI-T--l- (Continued) l-ENGINEERED SAFETY FEATURES RESPONSE TIMES i
j INITIATING SIGNAL AND FUNdTION RESPONSE TIME IN SECONOS, 2.
Containment Pressure-Nigh (Continued).
11) Annulus Ventilation Operation 1 23 i
12) Auxiliary Building filtered
-N.A.
Exhaust Operation
13) Containment Sump Recirculation N.A.
i 3.
Pressurizer Pressure-Low a.
Safety Injection (ECCS) 1 27(1)/12(3) i 1)
Reactor Trip i2
}
2)
Feedwater Isolation.'
< 12 l
3)
Phase "A" Isolation (2).
}18(3)/28(4) 4)
Purge and Exhaust Isolation 16 j
5)
/.uxiliary Feedwater(5)
N.A.
)
6)
Nuclear Service Water Operation i 65(3)/76(4) 7)
Turbine Trip N.A.
I-8)
Component Cooling Water 1-65(3)/76(4) 9)
Emergency! Diesel Generator Operation 1 11.-
'10)
Control Room Area Ventilation N. A.
Operation 11)
Annulus Ventilation Operation 1 23 l
12)
Auxiliary Building Filtered N.A.
Exhaust Operation i
13)
Containment Sump Recirculation N.A.
i 4.
Steam Line Pressure-Low a.
Safety Injection (ECCS)
.1 12(3)/22(4) f h-
, 1)
Reactor Trip 12 i
2)
Feedwater-Isolation
< 12 1
3)
Phase "A" Isolation (2) k18(3)/28(4) 4)
Purge and Exhaust-Isolation-
<6
5) -Auxiliary Feedwater(5) h60 6)
Nuclear' Service Water Opera' tion 1 65(3)/76(4) 7)
Turbine Trip N.A.
I 8)
Component Cooling Water 1 65(3)/76(4) 9)
Emergency Diesel Generator Operation 5 11
' CATAWBA - UNITS.1&2 3/4 A 3 3 P Amendment No.101 (Unit 1)
}
3
~
. Amendment No. 95(Unit 2)
s TABLE 3.3-5 TOR UNIT 1-(Continued)-
-l ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS 4.
-Steam Line Pressure-Low (Continued) 10)
Control Room Area Ventilation N.A.
~
Operation 11)
Annulus Ventilation Operation 5 23 12)
Auxiliary Building Filtered Exhaust Isolation N. A.
13)
Containment Sump Recirculation N.A.
i 10 l
b.
Steam Line Isolation S.
Containment Pressure-High-High a.
Containment Spray
< 45 b.
Phase "B" Isolation 1 65(3)/76(4)
Nuclear Service Water Operation N.A.
c.
Steam Line Isolation 1 10 l
I d.
Containment Air Return and Hydrogen
< 600 Skimmer Operation i
6.
Steam Line Pressure - Negative Rate-High Steam Line Isolation 1 10 l
7.
Steam Generator Water Level-High-High a.
Turbine Trip i3 b.
Feedwater Isolation i 12 l
8. ' Tfyg-Low Feedwater Isolation N.A.
9.
Doghouse Water level-High Feedwater Isolation N. A.
10.
Start Permissive Containment Pressure Control System N.A.
'll.
Termination l
Containment Pressure Control System N.A.
t b
l CATAWBA..-: UNITS 1&2-3/4 A 3 M5
- Amendment No.101 (Unit 1)
-l 8
Amendment No. -95 (Unit 2) 11
~~
\\
TABLE 3. 3-5 f0R-UNIT-1-(Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 12.
Steam Generator Water Level-Low-Low a.
Motor-0 riven Auxiliary Feedwater Pumps 1 60 b.
Turbine-Oriven Auxiliary Feedwater Pump 1 60 13.
Loss-of-Offsite Power a.
Motor-Driven Auxiliary Feedwater Pumps 1 60 b.
Turbine-Oriven Auxiliary Feedwater Pumps 1 60 Control Room Area Ventilalian Operation N.A.
c.
d.
Emergency Diesel Generator Operation 1 11
1) Diesel Buildiag Ventilation Operation N.A.
2) Nuclear Service Water Operation 1 65(3)/76(4) 14.
Trip of All Main Feedwater Pumps a.
Motor-Driven Auxiliary Feedwater Pumps 1 60 b.
Turbine Trip N.A.
15.
Auxiliary Feedwater Suction Pressure-Low Auxiliary Feedwater (Suction Supply 1 16(0)
Automatic Realignment) 16.
Refueling Water Storage Tank Level-Low Coincident with Safety Injection Signal (Automatic Switchover to Containment Sump)
$ 60 17.
Loss of Power a.
4 kV Bus Undervoltage -
< 8.5 Loss of Voltage b.
4 kV Bus Undervoltage-
-< 600 Grid Degraded Voltage 18.
Suction Transfer-Low Pit level Nuclear Service Water Operation N.A.
CATAWBA - UNITS 1&2 3/4 A 3 4@6 Amendment No.101 (Unit 1) l 8
Amendment No. 95 (Unit 2)
TABLE 3.3-5 FOR UNIT 2 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATION JGNAL AND FUNCTION RESPONSE TIME IN ECONDS 1.
ManualIh ation Containm(njection(ECCS)
Safety N.A.
a.
t Spray N. A. -
b.
c.
Phase "A" I olation N.A.
d.
Phase "B" Is lation N.A.
e.
Purge and Exha t Isolation N.' A f.
Steam Line Isola ion N
g.
Diesel Building Ve tilation Operation
.A.
h.
Nuclear Service Wate Ope /ation N.A.
1.
Turbine Trip N.A.
j.
Component Cooling Water N.A.
- k.
Annulus Ventilation Operat on N.A.
1.
Auxiliary Building Filtered N.A.
Exhaust Operation-m.
Reactor Trip N.A.
n.
Emergency Ofesel Generator trat on N.A.
o.
Containment Air Return an Hydrogen Skimmer Operation N.A.
p.
Auxiliary Feedwater N.A.
2-,
Containment Pressure-High a.
Safety Injection ( CCS)
<27(1)/1253)
1) Reactor Trip i2
-2) Feedwater solation
3) Phase "A' IsolationI3) 18 )/28(4)
4) Purge y d Exhaust Isolation 16
5) Auxi ary Feedwater(5)
N.A.
6) Nuflear Service Water Operation 1-65(3)/76 4) 7)
trbine Trip N.A.
8) Component Cooling Water 1 65(3)/76(4)
' Emergency Diesel Generator Operation i 11-
0) Control Room Area Ventilation Operation N.A.
CATAWBA : UNITS 1&2 3/4 B 3-37 Amendment No.101 (Unit 1) 1 8-107 Amendment No. 95'(Unit 2)
..,,..I -
\\ :
I
TABLE 3.3-5 FOR UNIT 2 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES fNITIATING IGNAL AND FUNCTION RESPONSE TIME IN SECON05 2.
Containm t Pressure-High (Continued)
11) A lus Ventilation Operation 1 23
12) Auxi ary Building Filtered N.A.
Exhau t Operation
13) Containment Sump Recirculation N.A.
Pressurizer Pressu\\
r -Low 3.
a.
Safety-Injection (ECCS) 1 27(1)/12(3) l 1)
Reactor Trip 12 2)
Feedwater Isolation 3)
Phase "A" Isolat n(2) 18(3)/28(4) 4)
Purge and Exhaust I olation
<6 4
5)
Auxiliary Feedwater N.A.
1 65(3)/76(4) 6)
Nuclear Service Water eration 7)
Turbine Trip N.A.
8)
Component Cooling Water 1 65(3)/76(4) 9)
Emergency Diesel Ge'nerator Op ration i 11 10)
Control Room Area Venti tion N.A.
Operation 11)
Annulus Ventilation 0peration 1 23 Auxiliary Buildin 'pFiltered N.A.
12)
Exhaust Operati 13)
Containment Sump Recirculation
.A.
4.
Steam Line Pressure-L a.
Safety Injection (ECCS).
1 12(3)/22(4) 1)
Reacto rip 52 2)
Feedvater Isolation
<7 3)
Phase "A" Isolation (2)
}gg(3)/2(4) p 4) urge and Exhaust Isolation 16 5
Auxiliary Feedwater(5)
< 60 Nuclear Service Water Operation 65(33/76(4) 7)
Turbine Trip H.A.
8)
Component Cooling Water 1 65(3)/76(4) 9)
Emergency Diesel Generator Operation i 11 CATAWBA - UNITS 1&2 3/4 8 3-38 Amendment No.101 (Unit 1) 810g Amendment.No. 95 (Unit 2)
n d
5 j
TABLE 3.3-5 FOR UNIT 2 (Continued)
[.
b ENGINEERED SAFETY FEATURES-RESPONSE TIMES INITIA 4NG SIGNAL AND FUNCTION RESPONSE TIME IN SEC0 %S 4.
Stea Line Pressure-Low (Continued) j 10 Control Roon Area Ventilation N.A.
Operation l
11)-
nulus Ventilation Operation 1 23 i
12)
Aux (liary Building Filtered 2
Exhaust Isolation N.A.
13)
Contal ent Sump Recirculation N.A i
l b.
St,eam Line Iso tion
_7 i
5.
Containment Pressure-H h-High l
a.
Containment Spray 1 45 5 65(3)/76(4) i b.
Phase "B"' Isolation 1
Nuclear Service Water op ration-N.A.
3 c.
Steam Line Isolation 17 1
d.
Containment Air Return and H r gen
< 600
~
i Skimmer Operation
^
6.
Steam Line Pressure - Negative te-H h i
Steam Line Isolation-
<7 1
7.
Steam Generator Water Leve -High-High a.
Turbine Trip
- <3
]
b.
Feedwater Isolat n 17
-T,yg-Low 8.
Feedwater I olation N.A.
4 9.
Doghouse Water Level-High
./
l.
Feedwater Isolation
N.A.
10.
Start P reissive ontainment Pressure Control-System N.A.
11.
T reination-Containment Pressure Control System H.A.
+
1
./
J CATAWBA - UNITS 1&2 3/4 B 3 Amendment No.101 (Unit 1) 8-109 Amendment-No. ' 95 ' (Unit 2) c
..r m.
v..
,,..e
~-
t l
/
TABLE 3.3-5 FOR UNIT 2 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATI SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS
\\
12.
Steam Ghnerator Water Level-Low-Low a.
Mot Driven Auxiliary Feedw ter Pumps 5 60 7
b.
Turbine riven Auxiliary i
Feedwater Pump i 60 13.
Loss-of-Offsite (wer i
a.
Motor-0 riven A xiliary Feedwater Pumps 1 60 b.
Turbine-Oriven A xiliary (eedwater Pumps 1 60
]
c.
Control Room _ Area entilatioh Operation N.A.
d.
Emergency Diesel Gen rator Operation i 11
1). Diesel Building Ve tilation Oper tion N.A.
2) Nuclear Service Wate Operati 1 65(3)/76(4)
Trip of All Main Feedwater Pum\\
14.
ps a.
Motor-Driven Auxiliary Feedwo er Pumps 1 60-
b. _
Turbine Trip N.A.
15.
Auxil'iary Feedwater Suction essure-L Auxiliary Feedwater (5 ction Supply
< 16(6) l Automatic Realignm t) j 16.
Refueling Water Stora Tank Level-Low Coincident wit,h> Safety Injection Signal (Automatic Sw tchover-to Containment Sump)
< 60 17.
Loss of Power a.
4 kV B.us Undervoltage -
1 8.
Loss of Voltage 4 k! Bus Undervoltage-
< 600 b.
G jd Degraded Voltage 18.
Suc on Transfer-Low Pit Level Nuclear Service Water Operation H.A.
~
i CATAWBA - UNITS 1&2 3/4 8 3-40 Amendment No.101 (Unit 1 l-8-110 Amendment No. 95 (Unit 2) l
4 1
INSTRUMENTATION BORON DILUTION MITIGATION SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.11 As a minimum, two trains of the Boron Dilution Mitigation System l
shall be OPERABLE and operating with Shutdown Margin Alarm ratios set at less than or equal to 4 times the steady-state count rate.
)
APPLICABILITY:
MODES 3, 4, AND 5 ACTION:
(a) With one train of the Boron Dilution Mitigation System inoperable or 4B hours, or, restore the inoperable train to OPERABLE status within not operating 1
(1) suspend all operations involving positive reactivity changes and verify that valve NV-230 is closed and secured within the next hour, or (2) verify two Source Range Neutron Flux Monitors are OPERABLE with Alarm Setpoints less than or equal to one-half deca # (square root of 10) above the steady-state count rate and v wify that thecombinedflowratefrombothReactorMakeup)WaterPumpsis less than or equal to 150 gpm (Mode 3 or 4) or F gpm (Mode 5) within the next hour.
3 (b) With both trains of the Boron Dilution Mitigation System inoperable or not operating, restore the inoperable trains to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , or (1) suspend all operations involving positive reactivity changes and verify that valve NV-230 is closed and secured within the next hour, or (2) verify two Source Ran AlarmSetpointslessf.eNeutronFluxMonitorsareOPERABLEwith han or equal to one-half decade (square root of 10) above the steady-state count rate and verify that the combined flow rate from both Reactor Makeup Water Pumps is lessthanorequalto150gpm(Mode 3or4)orf6gpm(Mode 5) within the next hour.
SURVEILLANCE REQUIREMENTS 4.3.3.11.1 Each train of the Boron Dilution Mitigation System shall be demon-l strated OPERABLE by performance of:
(a) A CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , CATAWBA - UNITS 1 & 2 3/4 3-85 Amendment No.10XUnit 1) l Amendment No. 9XUnit 2)
INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued)
(b) An ANALOG CHANNEL OPERATIONAL TEST at least once per 31 days, and (c) At least once per 18 months the BDMS shall be demonstrated OPERABLE by:
(1) Verifying that each automatic valve actuated by the BDMS moves to its correct position upon receipt of a trip signal, and (2) Verifying each reactor makeup water pump stops, as designed, upon receipt of a trip signal.
4.3.3.11.2 If using the Source Range Neutron Flux Monitors to meet the require-ments of Technical Specification 3.3.3.11, (a) The monthly surveillance requirements of Table 4.3-1 for the Source Range Neutron Flux Monitors shall include verification that the Alarm Setpoint is less than or equal to one-half decade (square root of 10) above the steady-state count rate.
(b) The combined flow rate from both Reactor Makeup Water Pumps shg11 be verified as less than or equal to 150 gpm (Mode 3 or 4) or)5'gpm (Mode 5) at least once per 31 days.
l l
l CATAWBA - UNITS 1 & 2 3/4 3-86 Amendment No.103 (UNIT 1)_
Amendment No. 97 (UNIT 2) i
- - =.
1 t'
REACTOR COOLANT SYSTEM 3/4.4.2 -SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of-one pressurizer Code safety valve shall be OPERABLE with-4 a lift setting of'2485 psig + 3%, -2% fer UMt 1 and 1Y fer U M t 2.*
l 1
l APPLICABILITY:
MODES 4 and 5.
}
j ACTION:
With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE resi-dual heat removal loop into operation in the shutdown cooling mode.
i t
4 SURVEILLANCE REQUTREMENTS J
4.4.2.1 No additional-requirements oth'er than those required by 5pecification 4.0.5, 2
4-4
The lift setting pressure shall correspond to ambient conditions of the valve at nominal. operating temperature' and ' pressure, i
4 CATAWBA - UNITS 1 & 2 3/4 4-7 Amendment No.101
(. Unit 1) 8-113 Amendment No. 95 (Unit 2)
REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig + 3%, -2% 4eHM4; -1 :nd : 1% f:r Un M-2.*
l.
APPLICABILITY:
HODES 1, 2, and 3.
ACTION:
With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE001REMENTS 4.4.2.2 No additional requirements other than those required by Specification 4.0.5.
The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
CATAW8A - UNITS 1 & 2 3/4 4-8 g.t:4 Amendment No.101 (Unit 1)
Amendment No. 95 (Unit 2)
TABLE 3.6-2a 2
p 5;' :
UNIT 1 CONTAINMENT ISOLATION VALVES J:
2 MAXIMUM FUNCTION ISOLATION TIME (s)
VALVE NUMBER f
m Phase "A" Isolation 1.
.~
l 88-578#.
Steam Generator IA Blowdown Containment Outside Isolation
$10 88-218#
Steam Generator IB Blowdown Containment Outside Isolation
<10 m
{
BB-618#
Steam Generator IC. Blowdown Containment Outside Isolation
{10
[
BB-10B#
-Steam Generator ID Blowdown Containment Outside Isola? ion
$10 t
1 88-56A#
Steam Generator.IA Blowdown Containment Inside Isolation
<10 t
BB-19A#
Steam Generator 18 Blowdown Containment Inside Isolation 210 I
BB-60A#
Steam Generator IC Blowdown Containment Inside Isolation 210 i
'BB-BA#'
Steam Generator ID Blowdown Containment Inside Isolation 710 f
88-1488#.
Steam Generator IA Blowdown Containment Isolation Bypass 310
[
}
y ~
B8-1509#
Steam Generator IB Blowdown Containment Isolation Bypass
$10 i
}
B8-1498#
Steam Generator IC Blowdown Cositainment Isolation Bypass
$10 t
a BB-1478#
Steam Generator ID Blowdown Containment Isolation Bypass
<10 m
n Steam Generator IA Main feedwater to Auxiliary feedwater Nozzle Isolation g #4 NA
.cA-149#
=
Steam Gewrator IB Main feedwater to Auxiliary Feedwater Nozzle Isolation 9J I
CA-150#.
CA-151#
Steam Generator IC Main Feedwater to Auxiliary Feedwater Nozzle Isolation ~ NA l
I
-CA-152#
Steam Generator ID Main feedwater to Auxiliary Feedwater idozzle Isolation
4 '
l CA-1BS#
Auxiliary Nozzle Temper SGIA frNA l
A J'
CA-186#
Auxiliary Nozzle Temper SGIB.
NA i
CA-187#
Auxiliary Nozzle Temper SGIC j;
j CA-18Bf Auxi1iary Nozzle Temper SGID f5 #4 l
CF-60#'
Steam Generator 10 feedwater Containment Isolation
A l
l CF-51#
Steam Generator IC Feedwater Containment Isolation
A CF-42#
Steam Generator 18 Feedwater Containment Isolation fJNA l
CF-33#
Steam Generator IA Feedwater Containment Isolation g5"NA t
CF-90#
Steam Generator IA Feedwater Purge Valve iXN4 CFrB9#
Steam Generator IB Feedwater Purge Valve fB~NA t
CF-BB#.
Steam Generator IC Feedwater Purge Valve g#A 3
CF-B/#
Steam Generator ID Feedwater Purge Valve.
,g3rgA i.
+-
6 i
7 O
f Y
l
TABLE 3.6-2a (Continued) 9 UNTI 1 CONTAINMENT ISOLATION VALVES g
a,i MAXIMUM 2
ISOLATION TIME (s)
FUNCTION g
VALVE NUMBER R
2.
Phase "B" Isolation (Continued)
<60 Supply to NC Pumps and LCVU Supply Outside Containment Isolation 760 Return from NC Pumps and LCVU Return Inside Containment Isolation RN-437B 760 Return from NC Pumps and LCVU Return Outside Containment Isolation 210 m
RN-484A RN-4878 Supply to Upper Containment Supply Ventilation Units Containment RN-4048 Isolation (Outside)
Return from Upper Containment Ventilation Units Containment Isolation
-<10 RN-429A (Ins de) i Return from Upper Containment Ventilation Units Containment Isolation
-<10 RN-4328 (Outside)
Mu
<10 Instrument Air Containment Outside Isolation h
VI-778 g VA Main Steam ID Isolation g#A SM-1 #
Main Steam IC Isolation f6 N A SM-3 #
Main Steam IB Isolation JS'#4 SM-5 #
Main Steam 1A Isolation Main Steam ID Isolation Bypass Ctri.
NA SM-7 #
A SM-9 #
Main Steam IC Isolation Bypass Ctrl.
NA SM-10 #
Main Steam IB Isolation Bypass Ctr1.
gNA SM-11 #
Main Steam 1A Isolation Bypass Ctri.
SM-12 #
<5 SV-19 #
Main Steam 1A PORV 25 SV-13 #
Main Steam IB PORV 75 SV-7 #
Main Steam IC PORV E5 Main Steam ID PORV 710 SV-1 #
Containment Vent Unit Drains Inside Containment Isolation 210 WL-867A**
Containment Vent Unit Drains Outside Containment Isolation WL-8698**
p.
TABLE 3.6-2b
-9
'Y:
UNIT 2 CONTAINMENT ISOLATION VALVES 6'
MAXIMUM S
VALVE NUMBER.
FUNCTION ISOLATION TIME (s)
R 1.
Phase "A" Isolation w
'8B-578#-
Steam Generator 2A Blowdown Containment Outside Isolation
<10
'88-218#
Steam Generator 2B Blowdown Containment Outside Isolatior.
210 BB-618#
Steam Generator 2C Blowdown Containment Outside Isolation 210 BB-10B#
Steam Generator 2D Blowdown Containment Outside Isolation 210 BB-56A#'
Steam Generator 2A Blowdown Containment Inside Isolation 210 BB-19A#'
Steam Generator 28 Blowdown Containment Inside Isolation ilo BB-60A#-
Steam Generator 2C. Blowdown Co'ntainment Inside Isolation 210 BB-BA#
Steam Generator 2D Blowdown Containment Inside Isolation 210 BB-1488#
Steam Generator 2A Blowdown Containment Isolation Bypass 210 w
1 _
BB-150B#
Steam Generator 28 Blowdown Containment Isolation Bypass 310 88-1498#
- Steam Generator 2C Blowdown Containment Isolation Bypass
$10 m
88-1478#
Steam Generator 2D Blowdown Containment Isolation Bypass
$10 Steam Generator 2A Main Feedwater to Auxiliary Feedwater Nozzle Isolation f['#A
A
P CA'-149#
Steam Generator 28 Main Feedwater to Auxiliary Feedwater Nozzle Isolation f5 C-
'CA-150#.
INA Steam Generator 2C Main Feedwater to Auxiliary Feedwater Nozzle Isolation J'5'NA CA-151#
Steac. Generator 2D Main feedwater to Auxiliary feedwater Nozzle Isolation f CA-152#
2 ' NA 5
CA-185#'
Auxiliary Nozzle Temper SG2A CA-186#.
Auxiliary Nozzle Temper SG2B f5I NA CA-1873 Auxiliary Nozzle Temper SG2C 25'pA 6
CA-188#
Auxiliary Nozzle Temper SG20 f 'go CF-60#'
Steam Generator 20 feedwater Containment Isolation 35NA CF-Sl#
Steam Generator 2C Feedwater Cor.tainment Isolation f5'#A CF-42#
Steam Generator'2B Feedwater Containment Isolation g5 #4 CF-33#
Steam Generatcr 2A Feedwater Containment Isolation f5'#A CF-90#
Steam Generator 2A Feedwater Purge Valve f5'#4 CF-89#-
Steam Generator 28 Feedwater Purge Valve 56 NA CF-88#
Steam Generator 2C Feedwater Purge Valve JS #A CF-87#
Steam Generator 2D Feedwater Purge Valve
{F#A r
m
_.m. _ _... _. _
v---
i
]
TA8LE 3.6-2b (Continued) t n
r UNTI 2 CONTAllMENT ISOLATION VALVES E
MAXIMUM g
l
.e FUNCTION ISOLATION TIME (s) i g
VALVE NUMBER i
-e 2.
Phase "B" Isolation (Continued)
<60 l
o.
RN-4378-Supply to NC Pumps and LCVU Supply Outside Containment Isolation Return from NC Pumps and LCVU Return Inside Containment Isolation 760 l
E60 l
RN-484A~
Return from NC Pumps'and LCVU Return Outside Containment Isolation RN-4878 710
[
RN-404B Supply to Upper Containment Supply Ventilation Units Containment l
Isolation (Outside)
'RN-429A Return from Upper Containment Ventilation Units Containment Isolation
<10 l
(Inside)
' Return from Upper Containment Ventilation Units Containment Isolation
-<10 RN-4328 (Outside) i A-
- Instrument Air Containment Outside Isolation
<10 i
VI-778 g;-
i' w.
' Main Steam 2D Isolation g NA
[
=
m SM-1 #
j Main Steam 2C Isolation jS NA f
j
--SM-3 #
JrNA l
..SM-5 #
. Main Steam 28 Isolation SM-7 #
Main Steam 2A isolation y NA r
.SM-9 #
Main Steam 2D Isolation Bypass Ctrl..
p5r NA
'SM-10 #
Main Steam'2C Isolation Bypass Ctr1.
g yA I
Main Steam 2B Isolation Bypass Ctr1.
gINA i ?
SM-12 #
Main Steam 2A Isolation Bypass Ctr1.
ffM/*
f SM-11 #
t
<S i
i Main Steam 2A PORV 25 1'
'SV-19 #
'SV-13 #
Main Steam 28 PORY-l fSV-7 #
Main Steam 2C PORY_
35
[
i
<5 710 f
SV-1 #
' Main Steam 2D PORV Containment-Vent Unit Drains Inside Containment Isolation WL-867A**
' Containment vent Unit Drains Outside Containment Isolation
{10 l
WL-869B*
l l
[
b
.w
,,..s
t
.1
,I PLANT SYSTEMS
[
{
MAIN STEAM LINE ISOLATION VALVES i
i i
j L8MITING CONDITION FOR OPERATION l
1
)
l 3.7.1.4 Each main steam line isolation valve (MSLIV) shall be OPERABLE.
I APPLICABILITY:
MODES 1, 2, and 3.
4 f
ACTION:
1 MODE 1:
With one MSLIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within j
4' hours; otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUT 00WN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
MODES 2 and 3:-
With one MSLIV inoperable, s'ubsequent overation in MODE 2 or 3 may proceed i
provided the isolation valve is main'ii n @ 4c The provisions of j
Specification 3.0.4 are not applicabim 0 a. e.e,
'S in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT Shfha t t'9o E. following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
L SURVEILLANCE REQUIREMENTS i
i 4.7.1.4 Each MSLIV shall be demonstrated OPERABLE by verifying full closure within 8 seconds 4er Urit 1 :nd 5 :::: d: 'Or-Unu-2 when tested pursuant to Specification 4.0.5.- The provisions of Specification 4.0.4 are not applicable i
for entry into MODE 3.
s 1
i i
r L
L CATAWBA - UNITS 1 & 2 3/4 7-8 g.ggg Amendment No.101(Unit 1)
- Amendment No. 95(Unit 2)
..g><=,
-4
~ + _ -
p
-,-,r w,,e,g.aenn,
--amp.,ms-w,.
g
.-..~ewwr,..
1 3/4.2 POWER OISTRIBUTION LIMITS (Unit-t)-
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 calculated DNBR in the core greater than or equal to design limit DNBR during normal operation and in short-term transients, and (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 the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria are not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
F (X,Y,Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat 9
flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; FAH(X,Y) 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.
X(z) is defined as the normalized F (X,Y,Z) limit for a given core height.
9 3/4.2.1 AXIAL FLUX DIFFERENCE-Ornt-i-The limits on AXIAL FLUX DIFFERENCE (AFO) specified in the CORE OPERATING LIMITS REPORT (COLR) ensure that the F (X,Y,Z) and the F6H(X,Y) limits are n
not exceeded during either normal operation or in the event of xenon redistrib-ution following power changes. _The AFD envelope specified in the COLR has been adjusted for measurement uncertainty.
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, AND NUCLEAR ENTHALPY RISE liOT CHANNEL FACTOR (Unit 1)
The limits on heat flux hot channel factor, and nuclear enthalpy" rise hot channel factor 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 ECCS ac-ceptance criteria are not exceeded.
The peaking limits are specified in the CORE OPERATING LIMITS REPORT (COLR) per Specification 6.9.1.9.
The heat flux hot channel factor and nuclear enthalpy rise hot channel factor are each 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:
CATAWBA - UNITS 1 & 2 B 3/4 2-1 Amendment No.101(Unit 1) 8 120 Amendment No. 95(Unit 2)
3/4.2 POWER DISTRIBUTION LIMITS (Unit-i-)--
BASES i
HEAT FLUX HOT CHANNEL FACTOR, AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR
-(4)ni t - 1) = ( Conti nueo) 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 overlappir.g groups as described in Specification 3.1.3.6; c.
The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
FAH(X,Y) will be maintained within its limits provided Conditions a, through d. above are maintained.
The limits on the nuclear enthalpy rise hot channel factor, FAH(X Y, are specified in the COLR as Maximum Allowable Radial Peaking limits, obtaIne)d by
~
dividing the Maximum Allowable Total Peaking (MAP) limit by the axial peak (AXIAL (X,Y)] for location (X,Y).
By definition, the Maximum Allowable Radial Peaking limits will, for Hark-BW fuel, result in a DNBR for the limiting tran-sient that is equivalent to the ONBR calculated with a design FAH(X,Y) value of 1.50 and a limited reference axial power shape.
For transition cores, MAP i
limits may be applied to both Mark-BW and optimized fuel types provided allowances for differences in DNBR are accounted for in the generation of MAP limits.
The MAP limits specified in the COLR include allowances for mixed core DNBR effects.
The relaxation of FAH(X,Y), as a function of THERMAL POWER allows changes in the radial power for all permissible control bank insertion limits.
This relaxation is implemented by the application of the following factors:
k = (1 + (1/RRH) (1 - P)]
.~
where k = power factor multiplier applied to the MAP limits P = THERMAL POWER / RATED THERMAL POWER RRH is given in the COLR N
N FQ (X,Y,Z) and FAH (X,Y) are measured periodically, and comparisons to i
the allowable limit are made to provide reasonable assurance that the core is I
operating as designed and that the limiting criteria will not be exceeded for operation within the Technical Specification limits of Sections 2.2 (Limiting Safety Systems Settings), 3.1.3 (Movable Control Assemblies), 3.2.1 (Axial Flux Difference), and 3.2.4 (Quadrant Power Tilt Ratio).
A peaking margin calculation is performed to provide a basis for decreasing the width of the AFD and f(AI) limits and for reducing THERMAL POWER.
CATAWBA - UNITS 1 & 2 8 3/4 2 h;!
Amendment No.101(Unit 1) 121 Amendment No. 95(Unit 2)
1 4
1 1
POWER DISTRIBUTION LIMITS i
I BASES l
HEAT FLUX HOT CHANNEL FACTOR, AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR
-(hn-+)-(Continued) j When an FQ (X,Y,Z) measurement is obtained from a full core map in accord-l 1
ance with the surveillance requirements of Specification 4.2.2, no uncertainties are applied to the measured peak since the required uncertainties are included j
M j
in the peaking limit.
When FQ (X,Y,Z) is measured for reasons other than meet-ing the requirements of Specification 4.2.2, the measured peak is increased by the radial-local peaking factor and allowances of 5% for measurement uncertainty and 3% for manufacturing tolerances.
1 N
When an FAH (X,Y) measurement is obtained from a full core map regardless of the reason, no uncertainties are applied to the measured peak since the f
3 required uncertainties are included in the peaking limit, j
3/4.2.4 00ADRANT POWER TILT RATIO-(Unit 1)-
1 The QUADRANT POWER TILT RATIOLlimit assures that the radial power distribu-1 tion satisfies the design values u5ed 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 ONB and linear heat generation rate protection with x-y plane power tilts.
A i
peaking increase that reflects a QUADRANT POWER TILT RATIO of 1.02 is included in the generation of the AFD limits.
i' The 2-hour time allowance for operation with a tilt condition greater j
than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.
In the event such action does not
.}
correct the tilt, the margin for uncertainty on F (X,Y Z) is reinstated by 0
reducing the maximum allowed power by 3% for each percent of tilt in excess of 2%.
4 l
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 dictribution is consistent with the QUADRANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.
l 3/4.2.5 DNB PARAMETERS (UN:T 1)
The limits on the ONB-related parameters assure that each of the parameters l
are maintained within the normal' steady-state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a 3
4 d
4 CATAWBA - UNITS 1 & 2 B 3/4 2-3 Amendment No.101(Unit 1) 8 122 Amendment No. 95(Unit 2) i
.=
^ /ER O!STRIBUTION LIMITS 1
BAS'ES 3/4.2.5 DNB PARAMETERS-fUNi4-it (Continued) design limit DNBR throughout each analyzed transient.
As noted on Figure 3.2-1, Reactor Coolant System flow rate and THERMAL POWER may be " traded off" against one another (i.e., a low measured Reactor Coolant System flow rate is acceptable if-the THERMAL POWER is also low) to ensure that the calculated DNBR will not be below the design DNBR value.
The relationship defined on Figure 3.2-1 remains valid as long as the limits placed on the nuclear enthalpy rise hot channel f ac-tor, FAH(X,Y) in Specification 3.2.3 are maintained.
The indicated T,y value l
i and the indicated pressurizer pressure value correspond to analytical 1 mits of 594.8'F and 2205.3 psig respectively, with allowance for measurement uncer-tainty. When Reactor Coolant System flow rate is measured, no additional 4
allowances are necessary prior to comparison with the limits of Figure 3.2-1 since a measurement error of 2.1% for Reactor Coolant System total flow rate has been allowed for in determination of the design DNBR value, l
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 manner.
Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-1.
Any fouling which might bias the heactor Coolant System flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters.
If detected, action shall be taken before performing 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 these parameters through instrunnent readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
Indica-tion instrumentation measurement uncertainties are accounted for % the limits provided in Table 3.2-1.
l Amendment No.101 Unit 1)
Ainendment No. 95 ((Unit 2)
CATAWBA - UNITS 1 & 2 B3/42gj2 '
l
\\
3/4.2 POWER 0!$TRIBUT10N LIMITS (Unit 2)
[
BAS 5 T
pecifications of this section provide assurance of fuel integr.y during Cond tion 1 (Normal Operation) and II (Incidents of Moderate Fr vency) 1 events by:
) maintaining the calculated ONdR in the core greater t an or equal to design lim \\ ONBR during normal operation and in short-term tran ents, and (2)limitingthAfissiongasrelease,fuelpellettemperature,andcladding mechanical properties to within assumed design criteria.
In add ion, limiting thepeaklinearpohrdensityduringConuitionIevantsprovide assurarice that the initial conditions assumed for the LOCA analyses are met d the ECCS acceptance criteria 11 it of 2200'F is not exceeded.
The definitions of rtain hot channel and peaking-f ctors.as used in these specifications are a follows:
F (Z)
Heat Flux Hot Chann 1 Factor, is defined a the maximum local heat 9
flux on the surface a fuel red at cor elevation Z divided by the average fuel rod heat lux, allowing to manufacturing tolerances-on fuel pellets and rods; Fh Nuclear Enthalpy Rise Hot annel. F ctor, is defined as the ratio of the integral of linear power lon the rod with-the highest integrated power to the average rod power 3/4.2.1 AXIAL FLUX OIFFERENCE (Unit l
The limits on AXIAL FLUX DIFF.EHCE (AFD) a sure'that'the F (Z) upper g
Fflimi specified in the ORE OPERATING-LIMITS REPORT bound envelope of the i
(COLR) times the normalized a al ceaking. factor is t exceeded during either l
normal operation or in the-e ent of xenon redistributi
-following power changes.
4 l
Target flux differen e is determined at equilibrium non conditions.
The full-length rods ma be positioned within the core in cordance with their respective inser on limits and should be inserted nea their normal position for-steady y ate operation at high power levels.
The value of the i
target flux differepce obtained under these conditions divided the fraction ofRATEDTHERMALPAWERiisthetargetfluxdifferenceatRATEDTH L POWER' for-the associated core burnup conditions. Target flux differences or other~
THERMAL POWER }iate fractional THERMAL POWER level,;4vels-are obtained by mul WER value:
by the appropf The periodic upda ing.of l-
~the target iux_ difference value is necessary to reflect core burnup-l considera ens.
i f
i i
CATAWBA.m UNITS 1 & 2 B 3/4 F524 Amendment Noi 86 (Unit.1)
L Amendment No. 80 (Unit 2)
.n
...,..w..
OWER OISTRIBUTION LIMITS l
i BASE x j;
ND At wer levels below APL
, the limits on AFD are defined in the C R,
i.e., that defined by the RAOC operating procedure and limits.
These mits 1
were calcul ed in a manner such that expected operational transients e.g.,
i load follow rations, would not result in the AFD deviating outsi of those
limits, Howev
, in the event such a deviation occurs, the short riod of time allowed outside the limits at reduced power levels will not r uit in signi-ficant xenon redi.ribution such that the envelope of peaking f tors wou,1d l
ND py,7 j
change sufficiently o prevent operation in the vicinity of t e APL level.
j At power levels gre ter than APL
, two modes of op ation are permis-ND sible; 1) RAOC, the AFD 11 its of which are defined in e COLR, and 2) Base i
Load operation, which is de ined as the maintenance o he AFD within a COLR
l specified band about a targe value.
The RAOC opera ng procedure above ND ND APL is the same as that defin d for operation b ow APL However it is j
i possible when following extended load following neuvers that the AFD limits j.
may result in restrictions in the aximum allo d power or AFD in order to
~
j guarantee operation with F (z) less than its miting value.
To allow operation q
l at the maximum permissible-value,'the ase oad operating procedure restricts i
i i
L i
i l
4 j.
l-i i
i 8 125
~ CATAWBAS-UNITS 1-& 2 8 3/4 2-6
' Amendment No. 86.(Unit 1)
Amonhant Nn. An-Uln1t y
_, ~.,. _ _ _. -, -.... _,. _,. - - _.., _ _. _,
,,.......,_..~_.m._
4 I
i 1
POWER 0!$TRIBUT!0N LIMITS 1
i B SES the in icated AFD to relatively small target band and power swings (AF rget
}
band as ecified in the COLR, APLND < power < APLOL or 100% Rated T rmal Power, i
whichever s lower).
ForBaseLoadoperation7itisexpectedthat he Units will j
operate wit.in the target band.
Operation outside of the target nd for the j
short time pegiod allowed will not result in significant xenon r distribution j
such that the bqvelope of peaking factors would change suffici tly to p.rchibit continued operat n in the power region defined above.
To as ure there is no residual xenon re stribution impact f rom past operation on he Base load f
operation, a 24 hou waiting period at a power level abov APL and allowed ND j
by RAOC is necessary.
During this time period load cha es and rod motion are restricted to that all ed by the Base Load procedure.
After the waiting i
period extended Base Loa operation is permissible.
The computer determine the one minute avera of each of the OPERABLE excore detector outputs and evides an alarm me sage immediately if the AFD l
for at least 2 of 4 or 2 of 3 ERABLE excore annels are:
1) outside the
]
allowed al power operating spac (for RAOC op ration), or 2) outside the-allowed al target band (for Base ad opera on).- These alarms are active when
)
power is greater than:
1) 50% of TED TH Al POWER (for RAOC operation), or ND (for Base Load operation),
2) APL en ty deviation minutes for Base Load
- operation are not accumulated based on he short period of. time during which operation outside of the target band s 11 owed, i
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT HANNEL CTOR, and REACTOR COOLANT SYSTEM j
FLOW RATE AND NUCLEAR ENTHALPY RME HOT CHANNEL FACTOR
' Unit 2)
I j
The limits on heat flux I channel factor, coolant flow rate, and nuclear i
enthalpy rise hot channel fa or ensure that:
(
the casign limits on peak local power density and min um DNBR are not excee q(d and (2) in the i
i a LOCA the peak fuel clad emperature will not excee the 2200*F ECCS acceptance criteria limit.
These 1 its are specified in the COM OPERATING LIMITS REPORT per Specification 6.9.
9.
i Each of these i measurable but will normally only be determined i
periodically as sp ified in Specifications 4.2.2 and 4,2.3.
This periodic j
surveillance is fficient to insure that the limits are main ained provided:
l a.
Con 1 rods in a single group move together with no dividual rod i
rtion differing by more-than_i 12 steps, indicated, rom the-foup demand position;
- b.
Control rod groups are sequenced with overlapping groups as escribed in Specification 3.1.3.6; i
.j i
8-126-CATdWBA-UNITSli2 B 3/4 2-7 AmendmentNo.8(-(U..ni.t1) l j
. o r
m
-rms,
,.r.
w..-s t
r ow*v-'--<re---
-*w-----
r-r ve+e-
-w-"
-+,---e
+w
-e vw+
ev-v-+
--e-a
POWER OISTRIBUTION LIMITS i
BAS S l
HEAT F HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND N EAR ENTHALPY NISE M0T CHANNEL FACTOR (Unit 2) (Continuco) l c.
control rod insertion limits of Specifications 3.1.3. /and 3.1.
6 are maintained; and d.
The ax 41 power distribution, expressed in terms of A. Al FLUX.
O!FFERERRE, is maintained within the limits.
~
Fhwillbema ainedwithinitslimitsprovidedConftions4. through d.
above are maintained.
s noted on the figure specified h the CORE OPERATING LIMITS REPORT (COLR), Rea tor Coolant System flow rate and F may be " traded g
of f" against one another (i.
, a low measured Reac r Coolant System flow rate is acceptable if the measured is also low) to nsure that the calculated g
DNBR will not be below the desig DNBR value, e relaxation of F as a H
function of THERMAL POWER allows e nges in se radial power shape for all permissible red insertion limits.
R as calculated in Specification 37 3 and used in the figure specified in the COLR, accounts for F less th h or equal to the F limit specified g
in the COLR.
This value is used i the var us accident analyses where Fhinfluencesparametersother an DNBR, e.g.
peak clad temperature, and thus is the maximum "as measured" v ue allowed.
The od bow penalty as a function of burnupapplied-forFhiscalculatedwiththemethosdescribedinWCAP-B691, i
Revision 1, " Fuel Rod Bow valuation," July 1979, and he maximum rod bow penalty is 2.7% DNBR.
Since the afety analysis is performed
'th plant-specific safety DNBR limits compared t the design DNBR limits, there is ufficient thermal margin available to o set the rod bow penalty of 2.7% ON The hot channe-factor F (z) is measured periodically an increased by a cycle and heigh ependent power factor appropriate to either R C or Base Lead operation,1W(z or W(I)BL, to provide assurance that the limit o the hot channel fact r, F (z), is met.. W(z) accounts for the effects of no mal oper-l q
ation tra ients and was determined from expected power control maneutgrs over the full ange of burnup conditions in the core.
W(z)BL accounts for more restri tive operating limits allowed by Base Load operation which result less evere transient values The W(z) function for normal operation-and e
W function for Base Load Operation are specified in the CORE OPERATING
.()MITSREPORTperSpecification6.9.1.9.
gg 4
CATAWBA - UNITS 1 & 2 B3/4I-[
Amendment No. b ((Unit 1) q Amendment No, Unit 2)
POWER OISTRIBUTION LIMITS ASES HEA LUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCL.E ENTHALU RISE HOT CHANNEL FACTOR (Unit 2) (Continued)
Reactor Coolant System flow rate and Fh are measured, no ad) tional Wh allowances re necessary prior to comparison with the limits of the igure specified in he COLR.
Measurement errors of 2.1% for Reactor Cool nt System total flow rate and 4% for Fh have been allowed for in determin ion of..the design DNBR value.
ThemeasuremenNerrorforReactorCoolantSystemtotal ow rate is based upon performing a pre ision heat balance and using the res t to calibrate the Reactor Coolant System low rate indicators.
Potential ling of the feedwater venturi which might not e detected could bias the resu from the precision heat balance in a noncons vative manner.
Therefore, penalty of 0.1% for undetected fouling of the dwater venturi is inclup d in the figure specified in the COLA.
Any fouling wh might bias the Rea t'or Coolant System flow rate measurement greater than 0.1%
n be detected by nitoring and trending various plant performance parameters, detected, acti n shall be taken before per-forming subsequent precision hea alance meas
ements, i.e., either the effect of the fouling shall be quantified nd compe ated for in the Reactor Coolant System flow rate measurement or the Nenturi. hall be cleaned to eliminate the fouling.
The 12-hour periodic surveillance indicated Reactor Coolant System flow is sufficient to detect only flow de adation which could lead to opera-tion outside the acceptable region o opera ion specified on the figure spec-ified in the COLR.
3/4.2.4 00AORANT POWER TILT RATI (Unit 2) l The QUADRANT POWER TILT ATIO limit assures t t the radial power distribu-tion satisfies the design v lues used in the power c ability analysis.
Radial power distribution easurements are made durin STARTUP testing and periodically curing powe operation.
The limit of 1.0, at which corrective action is req red, provides ONB and linear heat geneyation rate protection with x y plane p wer tilts.
A limit of 1.02 was ydlected to provide an allowance for the u ertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt conditi greater than 1.02 but ess than 1.09 is provided to allow identification d correction of a dropped or misaligned control rod.
In the event such action d s not correct ths/ ilt, the margin for uncertainty on F is reinstated by ducing t
A the maxi./um allowed power by 3% for each percent of tilt in excess of 1 or purposes.of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm th th4 normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore CATAWBA'- UNITS 1 & 2 B 3/4 f- >
Amendment No. 86 (Unit 1) s Amendment No. 80 (Unit 2) 4
0WER DISTRIBUTION LIMITS
-l m
BASE OUADRANh\\ POWER TILT RATIO (Unit 2) (Contin'ued).
flux map or two sets of four symmetric thimbles.
The two sets of fou symmetric thimbles is a unique set of eight detector locations.
The normal 1 ations are C 8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.
Alternate locations ar availaole ifanyofthenorgallocationsareunavailable.
3/4.2.5 DNB PARAMETERS (Unit 2)
The limits on the NB-related parameters assure that e ch of the parameters l
are maintained witnin th normal steady-state envelope of peration assumed in the transient and acciden analyses.
The limits are co istent with the-i j
initial FSAR assumptions an have been analytically de nstrated adequate to _
)
{
maintain a design limit ONBR hroughout each analyze transient.
The indicated 1
]
T,y value and the indicated p essurizer pressure y ue correspond to analytical i
lim ts of 594.8'F and 2205.3 psi respectively, w h allowance for measurement j
uncertainty.
l The 12-hour periodic surveillanc of th e parameters through instrument l
readout is sufficient to ensure that t pa ameters are restored within their i
limits following load changes and other acted transient operation.
Indica-l tion instrumentation measurement uncerta ies are accounted for in the limits provided in Table.3.2-1.
i i
l l
l l
4 lg 6
L I
i b
. CATAWBA - UNITS 1 & 2 B 3/4 2-Id29 Amendment No. 85 (Unit 1)
. Amendment No. 80 (Unit 2)
_ _,_,.__ _. _ _ _,_.- _. _.., _, ~. _._... _,_,,
ADMINISTRATIVE CONTROLS-
. CORE OPERATING LIMITS REPORT (Continued) 9.
OPC-NE-3000P-A, Rev. 1 " Thermal-Nydraulic Transient Analysis Methodology,'! November 1991.
(Modeling used in the system thermal-hydraulic analyses)
The core operating limits shall be determined so that all appitcable Ilmits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits,ident ECCS limits nuclear limits such as shutdown margin, and transient and acc 7
analysIslimits)ofthesafetyanalysisaremet.
1
]
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 in accordance with 10 CFR 50.4.
\\
i
/0, bPC - AIE-/004Al Wulear he r.'y me%dalo3y Ain3 Chsmo-3/SuvatArg-y" Alo v exber /992
( Aedhedology )or S ec.lfo'eds'e n '3.1.l 3 M o derkar Tempenivre, S
f Coeffs'c.o'ent.
9 s
I4. CATAWBA - UNITS 1.& 2 6-19b
' Amendment No.101 (Unit-1)
. g,g Amendment No. 95 (Unit 2)
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ML20126F014

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