Text

Proposed Tech Specs Pages Responding to 980116 RAI Re Improved TS Sections 3.1 & 3.2.Suppl Changes to Improved TS Also Included
	ML20216B769
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Site:	Catawba
Issue date:	03/09/1998
From:	; DUKE POWER CO.
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	NUDOCS 9803130184
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McGuire & Catawba improved TS Review Comments ITS Section 3.1, Reactivity Control Systems 3.1.1, Shutdown Margin 3.1.1-01 Bases JFD 4 Bases discussion of the Applicability for ITS 3.1.1, page B 3.1-4.

The Bases discussion of the Applicability for STS 3.1.1 states that the SDM requirements are applicable to provide sufficient negailve reactivity to meet the assumptions of the safety analyses discussed above. The Bases discussion for corresponding ITS 3.1.1 has replaced

" safety analyses" with " boron dilution accident analysis". Comment: Bases JFD 4 does not provide a specific explanation for this proposed difference. It is not explained why the boron dilution accident analysis has been s'ngled out. The Bases discussion for the Applicable Safety Analyses states that the MSLB is the most limiting accident for SDM requirements.

Provide adequate justificetion for the change or revise the submittal to conform to the STS.

DEC Response:

l l

The submittal has been revised to conform to the STS. 1 i

O l

1 mc3_cr_3.1 3.1-1 March 2,1998 9803130184 980309 '

PDR ADOCK 05000413 p PDR

SDM B 3.1.1

)

BASES

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APPLICABILITY In MODE 2 with k,ff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses l discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, " Boron Concentration."

In MODES 1 and 2 with k f 1.0, SDM is ensured by complyingwithLC03.1.$,r:t" Shutdown Bank Insertion Limits,"

and LCO 3.1.6.

ACTIONS Ad If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.

In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative I_ to raise the boron concentration of the RCS as soon as

\ possible, the boron concentration should be a highly i concentrated solution, such as that normally found in the !

boric acid storage tank, or the refueling water storage l tank. The operator should borate with the best source available for the plant conditions.

In determining the boration flow rate, the time in core life l must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may l approach or exceed 2000 ppm. Using its normal makeup path, the Chemical and Volume Control System (CVCS) is capable of inserting negative reactivity at a rate of approximately 30 pcm/minwhentheRCSboronconcentrationis1000ppmand approximately35pcm/minwhentheRCSboronconcentrationis 100 ppm. If the emergency boration path is used, the CVCS is capable of inserting negative reactivity at the rate of 65pcm/minwhentheRCSboronconcentrationis1000ppmand 75pcm/minwhentheRCSboronconcentrationis100 ppm. ,

Therefore, if SDM had to be increased by 1% Ak/k or 1000 !

pcm, normal makeup path at 1000 ppm could restore SDM in (continued) i Catawba Unit 1 B 3.1-5 Supplement 1 l

i SDM B 3.1.1 O

U BASES APPLICABILITY In MODE 2 with k,ff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses [

discussed above. In MODE 6 the shutdown reactivity requirements are given in LCO 3.9.1, " Boron Concentration."

In MODES 1 and 2 with k complying with LCO 3.1.$ff" aShutdown 1.0, SDMBank is ensured byLimits,"

Insertion and LCO 3.1.6.

ACTIONS Ad If the SDM requirements are not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.

In the determination of the required combination of boration flow rate and boron concentration, there is no unique s requirement that must .be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly concentrated solution, such as that normally found in the boric acid storage tank, or the refueling water storage tank. The operator should borate with the best source available for the plant conditions.

In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may app?oach or exceed 2000 ppm. Using its normal makeup path, the Chemical and Volume Control System (CVCS) is capable of inserting negative reactivity at a rate of approximately 30 pcm/minwhentheRCSboronconcentrationis1000ppmand approximately 35 pcm/ min d.2n the RCS boron concentration is 100 ppm. If the emergency boration path is used, the CVCS is capable of inserting negative reactivity at the rate of 65 pcm/ min when the RCS boron concentration is 1000 ppm and 75 pcm/ min when the RCS boron concentration is 100 ppm.

Therefore, if SDM had to be increased by 1% Ak/k or 1000 pcm, normal makeup path at 1000 ppm could restore SDM in s

O (continued) b Catawba Unit 2 8 3.1-5 Supplement 1 l l .

.. ,. ; : u:.: ?.;.'.'.a ', .-

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D

APPLICABLE coolant temperatures and pressure. The ejection of a rod SAFETY ANALYSES alw produces a time dependent redistrifvaian af : ore power.

(continued) crocn So x /Aef C)

SDM satisfies Criterion 2 of the WIC Po136 star though it is not directly observed frwm the contro'~.room. Even i Y '

t SOM is considered an initial condition process variable because it is periodically monitored to arisure that the unit !

is operating within the bounds of accident analysis

{I h El % assuoptions.

LCO i SDM is a core design condition that can be ensured during ation through control rod positioning (control and banks) and through the sol ele boron ation.

The MStB (Ref. 2) and the boron dilution (Ref. accidents are the most limiting analyses that establish SON value of the LCO. For M5LB accidents if the LCO is violated.

there is a potential to exceed the DWt limit and 4' exce d I 10 CFR 100.

Reactor Site Criteria.* limits (Ref. . For }

the boron dilution accident. if the LCO is viola the minimum required time assumed for operator action to 5~

terminate dilution may no longer be applicable.

\

5 APPLICABILITY In MODE 2 with k < 1.0 and in MXIES 3 the SON T bg requirements are,, applicable to provide,suff ci_ent negative / d. .

g reactivity to meet the assumptions of the a-utemesr ,,,

,,,j i discussed above. J nn resa y, aun As . useroy LW3.1.20 g f g !

pHufuunn gevailnX$DM)/-T /s: 20ly'F.*yIIn nuut e. tne snutar,.r reactivity requirements are given in LCO 3.9.1.

I

BoronConcentration.'$nMODES1and2:SOMisensuredby complyi wi h LCO 3.1 ' Shutdown Bank Insertion Limits.*

andLCO].1 .

U 6

g,g ACTIONS AJ If the SDM requirements are not met. boration must be initiated promptly. A Completion Time of 15 minutes is ;

t adequate for an operator to correctly align and start the l required systems and components. It is assumed that I 4

M B 3.14 Rev 1. 04/07/95 !

c4 A l O

McGuire & Catawba improved TS Review Comments ITS Section 3.1, Reactivity Control Systems 3.1. -02 LA.1, B JFD-4, B JFD-3 CTS 3.1.1.1 ITS 3.1.1 CTS 3.1.1.1 Actions provide instructions on restoring SDM with specific values of boron flow rate and boron concentration. ITS 3.1.1 only requires that boration be initiated to restore SDM. The details of restoring SDM are informative, but the requirement of the specifications is to maintain SDM and the specific requirement may not be appropriate for allinstances.

Therefore, these details are proposed to be moved to plant procedures. These details appear in the STS Bases and not in the ITS Bases. Commec~a Revise the submittal to provide adequate justification for moving details on boration to plant procedures rather than the Bases, as in the STS, or moved the details on boration to the Bases.

DEC Response:

The ITS submittal does include a plant specific boration example in lieu of the generic STS Bases example. The ITS example used values consistent with the normal makeup and emergency boration makeup reflected in the CTS 3.1.1.1 actions. These changes were justified by JFD 3 and JFD 4. No additional action is necessary in the Bases since the ITS Bases are consistent with the intent of the STS Bases to reflect the CTS actions. Discussion of Change LA.1 is revised to indicate that examples of acceptable actions for restoring shutdown margin, including emergency boration, are moved to the Bases for ITS 3.1.1, rather than to procedures. A typographical error is also corrected on Bases insert B 3.1-5.

i mc3_cr_3.1 3.1-2 March 2,1998

01seassien of Changes Section 3.1 - Reactivity Control Systems O TECHNICAL CHANGES - REMOVAL OF DETAIL LA.1 CTS 3.1.1.1 Actions provide insecuctions on restoring SDM with specific values of boron flow rate and boron concentration. ITS 3.1.1 only requires that boration be initiated to restore SDN.

The details of restoring SDM. are informative, but the requirement of the specifications is to maintain SDM and the specific requirement may not be appropriate for all instances. Therefore, examples of appropriate action to restore SIN, including emergency l boration,th= : d t:i h :re moved to the Bases for ITS 3.1.1.pken6 prn:d;rs. Changes to the Basesphnt prn:dars are controlled:=h:t:d by the Bases Control Program in ITS 5.0, addministrative econtrols ;=:r ing prn:dar: :h::;n. These controls ensure that changes to the Basesprn:d;rs are appropriately reviewed. This change is consistent with NUREG-1431.

LA.2 CTS Surveillance Requirements 4.1.1.1.1.d, 4.1.1.1.1.e, 4.1.1.1.2, and 4.1.1.2.b require the following factors to be considered when determining SDM:

1. Reactor Coolant System (RCS) boron concentration,

2. Control Rod Position,

3. RCS average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

l Temperature effect on the fuel is credited by the term Isothermal l Temperature Coefficient (ITC). The fuel and RCS temperatures will L

be changing at the same rate when the reactor is subcritical.

Temperature changes in the fuel creates reactivity changes associated with the fuel doppler coefficient. Depending on the change in temperature, the reactivity addition can be positive or l negative. Thus, ITC is added to the list of factors. All of l these factors are details which nesd to be considered when SDM is

! calculated but unnecessary in the specification and are moved to l

the Bases for ITS 3.1.1. The Bases are subject to the controls described in ITS Chapter 5 " Administrative Controls." Changes to the Bases are evaluated under the 10 CFR 50.59 criteria. Any change using this criteria will be appropriately reviewed. This change is consistent with NUREG-1431.

l lCatawbaUnits1and2 Page LA - 14 Supplement 15/20/07 i

f INSERT Using its normal makeup path, the Chemical and Volume Control System (CVCS) is capable of inserting negative reactivity at a rate of approximately 30 pcm/ min when the RCS boron concentration is 1000 ppm and approximately 35 pcm/ min when the RCS boron concentiation is 100 ppm. If the emergency boration path is used, the CVCS is capable of inserting negative reactivity at the rate of 65 pcm/ min when the RCS boron concentration is 1000 ppm and 75 pcm/ min when the RCS boron concentration is 100 ppm. Therefore, if SDM had to be increased by 1%

Ak/k or 1000 pcm, normal makeup path at 1000 ppm could restore SDM in approximately 33 minutes. At 100 ppm, SDM could be restored in I approximately 29 minutes. In the emergency boration mode at 1000 ppm, the 1% Ak/k could be restored in approximately 15 minutes. With RCS boron concentration at 100 ppm, SDM could be increased by 1000 pcm in approximately 13 minutes using emergency boration. These boration parameters represent typical values and are provided for the purpose of l offering a specific example, m

4 1

1 INSERT B 3.1-5 O l

McGuire & Catawba improved TS Review Comments ITS Section 3.1, Reactivity Control Systems

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3.1.2, Core ReactMty 3.1.2-01 DOC A.5, JFD 3 CTS 4.1.1.1.2 -

ITS SR 3.1.2.1 Note CTS 4.1.1.1.2 states that the predicted reactivity values "shall" be normalized to correspond to the actual core conditions. The Note for corresponding ITS SR 3.1.2.1 states that the predicted reactivity values "may" be normalized. The intent is that if necessary, the reactivity values shall be normalized. JFD 3 deletes the frequency note allowing the first performance of this SR after refueling to be delayed to 60 EFPD (within which time the Bases indicates it must be performed). The remaining note to the SR does not adequately define the SR

{

frequency requirements. Neither DOC A.5 nor JFD 3 adequately address the proposed '

changes. Comment: Revise the submittal to provide the appropriate justification for the proposed changes.

DEC Respones:

Duke Energy agrees that the intent is to normalize the predicted values,"if necessary."

)

- Although the CTS states "shall," this requirement is interpreted to be only "if necessary." That is, if the predicted values are conservative with respect to the measured values, then a ]

O t

change is'not necessary and would not be made. If the values are non-conservative by an amount greater than the margins assumed in the shutdown margin calculations, then an adjustment would be made. Therefore, the use of "may" is an administrative clarification of the current requirements for normalization. DOC A.5 is revised to provide this justification.

The STS frequency note allowance to delay initial performance of the surveillance as long as 60 EFPD is not permitted by the CTS and is not adopted. The CTS frequency of 31 EFPD is retained since the plant staff does not want to delay this meaurement until 60 EFPD. The remaining surveillance note addresses when the predicted values should be normalized to the measured values and does not modify the frequency to actually measure core reactivity.

Therefore, the deletion of the frequency note is independent of the surveillance note. ;

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mc3_cr_3.1 3.1-3 March 2,1998

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Discussi a ef Changes Secticn 3.1 - Reactivity Centrol Systems O

d ADMINISTRATIVE CHANGES i

A.4 CTS LC0 3.1.1.1 specifies the SDM limits for MODES 1-4. CTS 3.1.1.2 specifies the limits for SDM in MODE 5. The two Specifications are combined into ITS LC0 3.1.1, " Shutdown Margin".

The two specifications are identical with the move of the SDM limits to the COLR described in LA.3. This change is administrative and is consistent with NUREG-1431.

A.5 CTS Surveillance 4.1.1.1.2 is included in the LC0 for SDM. The requirement for maintaining core reactivity is retained as ITS LC0 3.1.2, " Core Reactivity" with an appropriate Action and Suueillance in the ITS. The sentence, "The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading" is moved to a Note for the ITS Surveillance 3.1.2.1. Although the CTS note states "shall," this requirement is interpreted to be only if necessary and is changed to "mov." If predicted values are conservative with respect to measured values, a change is not necessary. If values are non-conservative by an amount greater than the margins assumed in shutdown margin 0

calculations, on adjustment would be made. Therefore, the use of -

"may" in the ITS note is an administrative clarification of the current requirements. The referen: to "50 EFP0 after ::ch fuel h: ding" i: O retained for the iTS SP 3.1.2.1 Fr ;tency, :: :n ex::pti:n to the 31 EFPC requirement. The technical requirements are not modified by these changes. These changes are conside,ed administrative and consistent with NUREG-1431. The construction of the ITS LC0 is discussed in a Less Restrictive change later in these discussion of changes.

A.6 " Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading" is moved from the CTS Surveillance 4.1.1.1.1.d to form ITS SR 3.1.2.1 Frequency requirement. This requirement restricted the SDM to be within limits, considering the factors stated in 4.1.1.1.1.e (boron, rods, temperature, burnup, Xenon and Samarium concentrations) with the control banks at the maximum insertion limit. The intent of this CTS SR is to confirm that SDM will be within limits prior to the reactor becoming critical. The CTS SR is combined with CTS 4.1.1.1.2 which compares core reactivity to predicted values. A confirmation of core reactivity will also confirm SDM. This change is necessary since the LCO for SDM is no longer applicable in MODE 1 (see A.2). No technical Catawba Units 1 and 2 Page A - 27 Supplement 15/20/07l

Discussitn of Changes 52ction 3.1 - Reactivity Ccntral Systems ADMINISTRATIVE CHANGES requirements are modified and the change is administrative. The change is consistent with NUREG-1431.

A.7 CTS 3.1.1.3 Action a.1 states withdrawal limits shall be established in addition to insertion limits of Specification 3.1.3.6. LC0 3.0.2 requires that the actions must be met when the LC0 is not met, therefore, the statement in Specification 3.1.1.3 is redundant and therefore eliminated. No technical requirements have been added or deleted, therefore, the proposed change is administrative only. This change is consistent with NUREG-1431.

A.8 Not used.

A.9 CTS Surveillance 4.1.1.3.b requires the MTC be measured under specific conditions. These detail requirements have been transformed into two Notes, which will precede the ITS SR 3.1.3.2.

No technical requirements are modified by this change and it is considered administrative. These changes in ITS 3.1.3 are consistent with NUREG-1431. -

O A.10 CTS 3.1.1.3 Action a.2 requires the control rods to be maintained within the withdrawal limits established in Action a.1 until a subsequent calculation verifies that the MTC has been restored to within limits. This action is deleted because the LC0 actions cannot be exited unless the parameter, system, etc., is restored to within limits or the Mode of Applicability is exited. This requirement is redundant to the requirements of LCO 3.0.2, therefore, this change is administrative. This change is consistent with NUREG-1431.

A.11 CTS 3.1.1.3 Applicability provides a cross reference to the Special Test Exception in CTS 3.10.3 with a footnote. ITS 3.1.3 does not include this reference, consistent with NUREG-1431 philosophy to not include references to other Specifications unless it is specifically needed to specify additional Actions to be taken.

A.12 CTS 3.1.1.3 Action a.1 requires the plant to be in Hot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the control rod limits cannot be established within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the moderator temperature coefficient (MTC) is not within the upper limit. ITS 3.1.3 requires the plant to be

( in Mode 2 with K,,, < 1.0 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months which is below the Catawba Units 1 and 2 Page A - 37 Supplement 15/20/97l

Disc:ssicn of Chtnges S:ctirn 3.1 - Reactivity Control Systems O

V ADMINISTRATIVE CHANGES applicability of the LCO. This is an administrative change because CTS Section 3.0.1 states that the LCO is only applicable during the Modes of Applicability specified in the individual Specifications. Therefore, once the plant reaches Mode 2 with K.,,

< 1.0, the LC0 and its action are no longer applicable. This change is consistent with NUREG-1431.

A.13 CTS 3.1.3.1 Action c.2 states that THERMAL POWER level during subsequent operation with a misaligned rod shall be restricted as specified in Specification 3.1.3.6. This reference does not provide either clarity or additional unique Actions and is therefore deleted. The requirements to maintain rod insertion limits (CTS 3.1.3.6) are not affected by this change and remain applicable. No technical requirements are modified by this change and it is considered administrative. This change is consistent with NUREG-1431.

A.14 Not used.

( A.15 CTS LCO 3.1.3.4 specifies the requirements for rod drop times.

' This requirement is converted to a Surveillance Requirement and l

retained as ITS SR 3.1.4.3. The rod drop time requirement is more appropriate in the rod OPERABILITY under ITS LCO 3.1.4, " Rod Group Alignment Limits" as a Surveillance Requirement. The change maintains the requirement of verifying the rod drop times and is considered to be administrative. This change is consistent with NUREG-1431.

A.16 CTS 3.1.3.1 Action a and c.3 requires that SDM be evaluated to be within limits, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of discovery that a rod (s) is untrippable. ITS LC0 3.1.4, Action A, requires either verifying SDM within limit or, if SDM is not within limit, the initiation of boration to restore SDM is required. The additional action is now included in ITS LCO 3.1.4 because of the modification to ITS LCO t 3.1.1, "SDM", which does not include Modes 1 or 2. These changes ;

are maintain the existing technical requirements and are !

considered administrative. These changes are consistent with NUREG-1431.

A.17 CTS LCOs 3.1.3.1, 3.1.3.5 and 3.1.3.6 Applicabilities reference l I

p the Special Test Exception in CTS 3.10.2 and 3.10.3. The proposed change deletes this reference, consistent with NUREG-1431 I

i Catawba Units 1 and 2 Page A - 47 Supplement 15/20/97l

Discussicn of Changes 5:cticn 3.1 - Reactivity C*ntrol Systems ADMINISTRATIVE CHANGES philosophy to not include references to other Specifications I unless it is needed to specify additional Actions required to be taken. This reference provides information only and is not necessary.

A.18 CTS 3.1.3.5 and 3.1.3.6 actions provide an exception for control and shutdown rods being within required limits during rod freedom of movement testing in CTS SR 4.1.3.1.2. The exception is retained as a Note to the Applicability of ITS LCOs 3.1.5 and 3.1.6 to allow movement of the rods outside the limits during performance of ITS SR 3.1.4.2. No technical requirements are modified by this change and it is considered administrative in nature. This change is consistent with NUREG-1431.

A.19 The Applicability of CTS LCO 3.1.3.5 is stated as Mode 1 and Mode 2 wi th K,,, 2 1.0. Within CTS Surveillance Requirement 4.1.3.5.a, the shutdown rods are required to be within limits prior to withdrawal of any control banks. The most limiting case is prior to withdrawing any control bank. Therefore, the ITS 3.1.5 Applicability for Mode 2 is stated as, "with any control bank not

\ fully inserted." The existing technical requirements are maintained with the proposed change, therefore it is considered administrative. This change is consistent with NUREG-1431.

A.20 Actions have been added to CTS 3.1.3.5 and 3.1.3.6 for shutdown l and control bank insertion limits to verify SDM when insertion limits are not met. CTS 3.1.1.1 on SDM has been revised to delete MODE I and 2 from the applicability (see A.2), therefore, this addition is necessary to maintain the existing requirements in CTS 3.1.1.1 and CTS 3.1.3.1 to verify or restore SDM. These changes are retained in ITS LC0 3.1.5 and 3.1.6 and maintain the existing technical requirements and are considered administrative in nature. These changes are consistent with NUREG-1431.

1 A.21 CTS 3.1.3.5 Actions require within one hour of having a shutdown rod inserted beyond the insertion limit, the rod must be restored within limit or the rod declared inoperable and apply Specification 3.1.3.1. The application of CTS 3.1.3.1 for misaligned rods requires restoration within one hour or the plant must be placed in Mode 3 within the next six hours. The I hour in

/] CTS 3.1.3.5 and the additional hour in CTS 3.1.3.1 are combined V

1 Catawba Units 1 and 2 Page A - 57 Supplement 15/20/07l 1

Disc:ssica cf Ch:ng:s SIctica 3.1 - R: activity C:ntr:1 Systems ADMINISTRATIVE CHANGES and retained in ITS 3.1.5 Action A.2 as 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The requirement to shutdown to MODE 3 is retained as ITS 3.1.5 Action B. The changes are considered administrative and are consistent with NUREG-1431.

A.22 Not used.

A.23 Specific requirements for the control bank sequence and overlap limits have been added to CTS 3.1.3.6 and 4.1.1.1.1.b on control bank insertion limits. ITS 3.1.6 clearly identifies that these parameters are required to be met by including specific actions and surveillance requirements. These parameters have always been a part of the control bank insertion limit as detailed by the figure in the COLR. No technical requirements are modified and the change is considered administrative in nature since it clarifies information already contained within the existing requirements. The change is consistent with NUREG-1431.

A.24 CTS 3.10.3 allows exceptions, to rod alignment and insertion limits during PHYSICS TESTING in MODE 2. With the deletion of CTS 3.1.1.1 SDM is MODE 1 and 2 (see Doc A.2), it is necessary to add appropriate requirements, actions, and surveillances to the test exception LCO. These requirements are retained as ITS LC0 3.1.8.

The change is administrative in nature, and no technical change is made. This change is consistent with NUREG-1431. >

A.25 CTS 3.10.3.b requires the reactor trip setpoints of the intermediate and power range channels to be set at 25% during performance of PHYSICS TESTS. This infcrmation is redundant to the LCO 3.3.1, "RTS Instrumentation" which require.s that the trip j setpoints for the channels be set to 25% for the intermediate and I power ranges in MODE 2. The deletion of redundant requirements is !

administrative and does not represent a te:hnical change. This change is consistent with NUREG-1431.

A.26 The exceptions to SDM provided by CTS 3.10.1, the exceptions for rod insertion and power distribution limits provided by CTS 3.10.2, and the exceptions for rod position indication provided by CTS 3.10.5 are no longer needed and are deleted. SDM will be maintained within the limits specified in the COLR during PHYSICS C TESTS. PHYSICS TESTS will be conducted in MODE 2, thus the MODE 1 exception provided by CTS 3.10.2 is not needed. Rod position Catawba Units 1 and 2 Page A - 6;L Supplement 15/20/07l

Discussicn of Chang 2s S2cticn 3.1 - Reactivity Centrol Systems l

ADMINISTRATIVE CHANGES exception is no longer needed in MODES 3, 4, or 5 and in MODE 2 wi th K.,, < 1.0. The ITS 3.1.8 test exceptions for PHYSICS TESTS in MODE 2 provides an exception to rod alignment requirements.

This change is consistent with NUREG-1431.

A.27 CTS 3.10.3.a allows exceptions to certain LCOs for the performance of physics tests provided power is limited to 5 5% of Rated Thermal Power (RTP). This statement is redundant because the Applicability for this LCO is MODE 2 (s 5% RTP). ITS LCO 3.1.8

\ retains this same applicability during PHYSICS TESTS. With this

)

deletion, no technical requirements are modified and the change is considered to be administrative in nature. This change is 1 consistent with NUREG-1431.

A.28 CTS 4.1.1.1.1.a and 4.1.1.2.a require verifying SDM when an I

inoperable (imovable or untrippable) control rod is discovered.

) This requirement is already contained in CTS 3.1.3.1 and is retained in ITS LC0 3.1.4 for an untrippable control rod. These requirements already provide adequate assurance that SDM is verified and therefore, the requirements of 4.1.1.1.1.a and

4.1.1.2.a are redundant and eliminated. No technical requirements are deleted by the elin'.ination of this redundant requirement and the change is considered administrative. This change is consistent with NUREG-1431.

A.29 CTS 3.1.3.1 Action c.2 and c.3 contain the phrase, "The rod is declared inoperable" when a rod is not within alignment limit?

This wording does not add any clarity to the actions and is eliminated. The format in the ITS is such that actions are only entered when the LC0 is not met, i.e. the component is inoperable.

Therefore, the additional wording is not necessary for inclusion within ITS 3.1.4. No technical requirements are deleted by the elimination of this wording and the change is considered administrative. This change is consistent with NUREG-1431.

A.30 CTS 3.1.3.6 requires that an out of limit control bank be restored within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or that power be reduced to match the power limit for the existing insertion position. ITS 3.1.6 only requires that the insertion limit be restored. The existing actions are somewhat redundant. Since there are only two ways to restore compliance (i.e., withdraw the control banks or reduce power to Catawba Units 1 and 2 Page A - 77 Supplement 15/20/97l

Discussign of Changes Section 3.1 - RIactivity Ccntrol Systems ADMINISTRATIVE CHANGES match control bank position), the requirement to restore limits is all that is necessary to be specified. This change is considered administrative since there is no change to the technical requirements. This change is consistent with NUREG-1431. l O

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Catawba Units 1 and 2 Page A - 87 Supplement 15/20/97l

\

g McGuire & Catawba Improved TS Review Comments Q ITS Section 3.1, ReactMty Control Systems 3.1.3, Moderator Temperature Coef5cient 3.1.3-01 B JFD 2 Bases discussion of the Applicability for ITS 3.1.3, page B 3.121.

The Bases discussion of the Applicability for STS 3.1.3 states that in Modes 4,5, and 6 the LCO is not applicable because there are no DBAs using tne MTC as an accident analysis assumption are initiated from these modes. The Bases discussion for corresponding ITS 3.1.3 states that no DBAs using the MTC as an analysis assumption are " limiting when" initiated from these modes. Comment: The proposed difference is ambiguous. Revise the Bases to clarify the intent, or conform to the STS.

DEC Response:

MTC may be considered as an input to accident analyses pedarmed in MODES 4,5, and 6, however, accidents initiated from these modes are not exMted to be limiting events.

Therefore, Duke Energy beileves that the proposed cle- 2ication to the Bases is appropriate and necessary to remain consistent with the safety retalysis. JFD 14 is added to the STS markup to reflect this justification.

O O mc3_cr_3.1 3.14 March 2,1998

)

l MrC 1 B 3.1 l BASES LCO (continued) advantage of improvert fuel management and changes in unit operating schedule.

APPLICABILIlY Technical Specifications place both LCO and SR values on MTC. based on the safety analysis assumptions described above.

In MODE 1. the limits on NTC must be maintained to ensure that any accident initiated from TrERMAL POWER operation will not violate the design assumptions of the accident analysis. In NODE 2 with the reactor critical. the upper l limit must also be maintained to ensure that startup and i subcritical accidents (such assembly oc g will notas the f.heuncontrolleds. uni @

assumptions of t withdrawal) violate accident analysis. The lower 6TC limit must De maintained in MODES 2 and 3. in addition to MODE 1 to ensure ass that cooldown accidents will not violate the ions of the accident analysis. In MODES 4. 5. and 6 this is not applicable since no Design Basis Accidents ,

using the NTC as an analysis assumption a nitiated from i these MODES. 3 !

ACTIONS &J (5

d If the BOC NTC limit is violated administrative withdrawal limits NTC withinfor control banks must be established to maintain the its limits. The NTC becomes more negative with control bank insertion and decreased boron concentration.

Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides enough time for A

evaluating the MTC measurement and computing the required bank withdrawal limits.

As cycle burnup is increased. the RCS boron concentration will be reduced.

MTC to become more negative.The reduced boron concentration causes the Using physics calculations.

the time in cycle life at which the calculated NTC will meet the LCO requirement can be determined. At this point in core life Condition A no longer exists. The unit is no longer in the Required Action, so the administrative withdrawal limits are no longer in effect.

(continued)

B 3.1 21 an Rev 1. 04/07/95 P

G

1 Justificcticn f:r Deviatiens S:ction 3.1 - Reactivity Control Systems I l

[ ~\ l U' BASES

9. Reload Physics Testing Program includes Critical Boron (AR0), J Control Rod Vorth (Rod Swap method), and ITC determination. Ref.

UFSAR Chapter 14.3 Physics Test Program.

I

10. This change represents a clarification of the Bases desired by the ,

plant staff. '

11. This level of detail is not needed since the tests are described in the UFSAR.

I?. This change reflects a generic change to NUREG-1431 proposed by the Mdustry owners groups. The justification for this change is cont 31ned in Technical Specification Task Force (TSTF) change number TSTF-89.

13. This change reflects a generic change to NUREG-1431 proposed by the industry owners groups., The justification for this change is j contained in Technical Specification Task Force (TSTF) change number l p TSTF-108. l b 14. The Bases discussion of the Applicability for NUREG LCO 3.1.3 states that in Modes 4, 5, and 6 the LCO is not applicable because there are no DBAs using the MTC as an accident analysis assumption which are initiated from these modes. The Bases discussion for ITS 3.1.3 l states that no DBAs using the MTC as an analysis assumption are l limiting when initiated from these modes. MTC may be considered as an input to accident analyses performed in MODES 4, 5, and 6, however, accidents initiated from these modes are not expected to be limiting events. Therefore, the proposed clarification to the Bases is appropriate and necessary to remain consistent with the safety analysis.

15. The Bases Background discussion for STS 3.1.6 states that the control banks are used for precise reactivity control of the reactor. Additionally, the Bases Background discussion for STS 3.1.6 states that the control banks must be maintained above design insertion limits and are typically near the fully withdrawn position l during normal full power operation. The STS 3.1.6 Bases Background in general describes requirements which are generic to both shutdown and control banks. However, STS 3.1.6 addresses only the requirements for shutdown bank insertion limits. Therefore, the STS (d

Catawba Units 1 and 2 22 Supplement 15/40/97l l

1 McGuire & Catawba improved TS Review Comments q ITS Section 3.1, Reactivity Control Systems 3.1.4, Rod Group Alignment IJmits 3.1.4-01 LA.8 JFD 13 STS SR 3.1.5.2 ITS SR 3.1.4.2 CTS 4.1.3.1.2 STS SR 3.1.5.2 requires verifying rod freedom of movement (trippability) by moving each rod not fully inserted in the core 210 steps in either direction. Corresponding ITS SR 3.1.4.2 requires verifying rod trippability. Comment: This change does not provided clarity and is not a Justifiable plant specific difference. Revise the submittal to conform to the STS.

DEC Response:

The submittal is revised to conform to the STS.

O O

mc3_cr_3.1 3.1-5 March 2,1998

l

Rod Group Alignment Limits 1

3.1.4

\

l SURVEILLANCE REQUIREMENTS (continued)

)

v SURVEILLANCE FREQUENCY l

SR 3.1.4.2 Verify rod freedom of movement 92 days (trippability) by moving each rod not fully inserted in the core 210 steps in either direction.

i l

SR '3.1.4.3 Verify rod drop time of each rod, from the Prior to fully withdrawn position, is s 2.2 seconds reactor from the beginning of decay of stationary criticality

gripper coil voltage to dashpot entry, after each with: removal of the reactor head i

a. T.y, a 551*F; and I

b. All reactor coolant pumps operating.

O l

f !

i O 3.1-10 Supplement 1 l Catawba Unit 1

Red Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS (continued)

O SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify rod freedom of movement 92 days (trippability) by moving each rod not' fully inserted.in the core 2 10 steps in either direction.

SR 3.1.4.3 Verify rod drop time of each rod, from the Prior to fully withdrawn position, is s 2.2 seconds reactor from the beginning of decay of stationary criticality gripper coil voltage to dashpot entry, after each with: removal of the reactor head

a. T ag a 551*F; and

b. All reactor coolant pumps operating.

t O

O 3.1-10 l Catawba Unit 2 Supplement 1

Rod Group Alignrrent Limits B 3.1.4 BASES ACTIONS D.1.1 and 0.1.2 (cc '.inued) required valves and start the boric acid pumps. Boration will continue until the required SDM is restored.

D.1 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. The unit must be brought to a MODE or Condition in which the LC0 requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The allcwed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS i Verification that individual rod positions are within -

\ alignment limits at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. If the rod position deviation monitor is inoperable, a Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months accomplishes the same goal. The specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected.

SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control l rod every 92 days provides increased confidence that all rods continue to be OPERABLE without n

(continued) u)

Catawba Unit 1 B 3.1-29 Supplement 1 l

Rod Grou9 Alignment Limits B 3.1.4 BASES ACTIONS D.1.1 and D.1.2 (continued) required valves and start the boric acid pumps. Boration will continue until-the required SDM is restored.

D.s2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. The unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve'this status, the unit must be brought to at least M0DE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS -

Verification that individual rod positions are within O(./ alignment limits at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides a history that allows the operator to detect a rod that 14 beginning to deviate from its expected position. If the rod position deviation monitor is inoperable, a Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months accomplishes-the same goal. The specified Frequency j

- takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected.

SR 3.1.4.2 i l

Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control l rod every 92 days provides increased confidence that all rods continue to be OPERABLE without l

l (continued)

O Catawba Unit 2 B 3.1-29 Supplement 1 l

1, .. <

bee tCtcman .

s .. . "

,' \ 3,1' ) 4 i .:. .

. ff. .REXCTIVfTY . . ,

CONTdOL SYSTEMS ,

LIMITINGCONSITIONFORUPERATION

[

g J -

s ACTION , ntinued)' --

h,2,y @ wer e4 t.o SR 3.2.f.i a-.J SO.2M Aiiouu on map 37 vos incu .~ v. dic ,

g'y* p dete ante incore]

s and Fn(Z) and1"... _are verified t e within 1hei limits /within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and .> A.\

g,g A#te 2.

Edkth2(HERNAL POWERWW to less than or equal to j 751.of RATFD THERMAL POWFR within the next hourj na he followingfhours the High N 4

,et-

+ educed to 1(ss than or equal ron tiux Trip oin% t is *3 AmnC 85% of RATED T L POWERJ g ,p 4Cr1M D 4withgrethanonairi.ienstgy causes opier than ad essed b n ON ma iut ino rme e.

e %

42ntinvei provided t t :f-- PERAT may /Y).3

'1. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder or 6ne s in tne cankts) wi incperable r s are aligned to wi able rods a12 steps of the in e r-th) 11e maintaining the sequence and inserti limits of pecification 3.1.3.6.

g#3 The THERMAL POWER lev be restr cted pursuant to Speci cation 3.1.3.6 durin shall subseq nt operation, and

2. The operable rods are rest hour to OPERA 8LE statu within 72 )

~

& f SURVEILIANCE REOUIREMENTS M 3,l.4-{ [ teofy vJs.r.dulend O MI ins position N n within yte growlemaD limit ey ven , Av utn rod shil' tusi-ba M termined to @ I

- tne ina' 60nce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Except dusMa t' me inttrva '

aa=I J65 positioWiD Deviation Monitor iy inoperablC tnen vj Swhen the Rod Position *

' wrry tne on-mosioonn at least once-kelk 4 W, opoll.

SR 3.lM.L C1M M ltaprui shensen roa not furysrTnse tEined to WDPERABLE b/ movement of atrienst least once per days. 1_0 stess in any one difectionfat

% L . tl g

\

l F

CATAWBA - UNIT 1 3/4 1-16 Amendment No. 148 n fW Y v

$fec rP Coso -

3. f. Y G EACTIVfTY CONT d SY$TEMS A.I O LIMITING ColeMION FOR OPERAff0N

/ '

r

( ACTION (Cfntinued) JCpe,(.m sa.3.2.f.iM 5U.2..T.0 _

' 6 2.4 5 'n power di ibutton esp tatnea tras tne movenie incore' s .2..f detecto nd F,,fZ) and m are verified to b/within J l their f ttNithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; ana 3,2,.L @THeRMALDER w EgMto less than or equal to a 75% of RATED THERMAL POWER w' thin the next hourjne ipenin 3 41 e followl duend ta stuaeree= A ma,na narm r p-- laint hcars the M1i lh Eggsrun ruum arip onurs ;

is e es8itAf An.,C g .~

't

__ rod m "e~ms inoera w == =

T101t may , m"M bO c ses =ra other +" - 'su-i Y no is a aasve. POWER Oasd addressad *h tinue prov thate /

', g 1 ~.h

[ # 1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , inoperable rod are all reesinder of the sinthebank(s)with to withi t12 steps of the I -

/H,7 able rods whi enintal i the e and insertion limits of s ification 3. 3.6. ' POWER level 11 p1,5 ne restrict ,.rs.nt to s,.cif tien 3.1.3.6 du,:n ,

isubsequent ration, and

2. The in able rods are restsfed to OPERAKE status thlm 72 j hours...

6

$URVEILUUICE REQUIREMENTS Ver,W ad eJ .I 'md3 O 5A 1./.8f,j 1Li e hsition >

6 once per. rs - --

- -- n - ,

- when ".he MoS Position Deviation Monitor is inoperah' e, == -- r- -nu--Mt least once , ,

did b) , ._

)

58 Llleast 4'LMg,ll,'}-{y'"=

once per days.

"#;'NI4";'y y"gQ*[g Le /l $TET~

l c

CATAWBA - UNIT 2 3/4 1-16 Amendment No. 142 f.y. 2 + 7 i

l

Disc:ssitn cf Changes Section 3.1 - Reactivity Control Systems b] TECHNICAL CHANGES - REMOVAL OF DETAIL are evaluated under the 10 CFR 50.59 criteria. Any change using this criteria will be appropriately reviewed. This change is consistent with NUREG-1431.

LA.7 Not used.

LA.8 Not used. CTS '.1.3.1.2 ::nt:in: d t il: :n 50; c;d'tripp:bility i:

veriH :d. Thi: inf;=ti:n i: being rel: =ted te th: S= : for ITS SR 3.1.'.2. Th: int:nt Of the SR i: t: ;;rify 7;d tripp:bility, h;;;;;r th: 1:n;;;;; in the CTS ddre:::: 7:d frc:d:: Of =v:=nt. Int rpr:t:ti= pr:bl = with th: CTS lang;;;; cre n::;nter:d h:n th: r;d: c;=in trippb!: =

required by th: =fety :=1y:i: :nd LCO, but cr n:t ==b1: =

t ted in the Surveill==. Thi: int:rpr;t:ti= in cl: rift:d in the "" REC l'31 S:: = . The S= : cre ::bj=t t th: :=tr:1:

spee+fied in ITS CS:pter 5.0, "f.drini tr:tiv: C=tr:1:", =d ;r enl=t;d in :=:rd== with 10 CPR 50.50. Th r:f re, =y ch= ;=

te-R=: requir:=nt; will bc :ppr:priately r=f:x:d.

C LA.9 The CTS 3.1.1.3 Action a.3 requirement to submit a special report to the NRC in the event that MTC exceeds the upper limit is deleted. The 10 CFR 50.72 and 10 CFR 50.73 requirements clearly identify notifications and reports to be made to the NRC. If a condition were discovered that met the criteria within these regulations, a report would be required. The deleting of the special report in this specification does not change the requirement within the regulations for reporting to the NRC Reports to the NRC are controlled by plant procedures which implement the regulations. This change is consistent with NUREG- '

1431. i i

LA.10 CTS 3.1.3.1 action c allows power operation to continue with one rod not within alignment limits provided the misaligned rod is restored to limits in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (action c.1) or the other rods in the group are moved within the alignment limits (c.2). ITS 3.1.4 only requires that the alignment limit be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The alternatives of how the limit is restored are not necessary for l

I inclusion within u.; TS and are relocated to the Bases for ITS 3.1.4. The Bases are subject to the controlc Specified in ITS i

\

Page LA - 34 Supplement 15/20/97 l Catawba Units 1 and 2

' *? ~,

^

~'_'

. ", ,. s. .'

Rod Group Alignamt 1.12its.

3.1 SLRVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE

{ $16f FREQUENC '

7 SR 3.1 veri 2 rod u,-- or e 92 days di inri.~

he c 0 sMn$$)

~

d SR 3.1 Verify rod drop time of each rod, from the Prior to fu thdrawn position. is reactor seconds from the beginning of decay criticality t ionary gripper coil voltage to after each dashpot entry wi'h: t removal of the a.

reactor head OS T, k : and

b. All reactor coolant pumps operating.'

[

V j

$ 3.1 11 Rev 1. 04/07/95 CAbWh O

l Justificaticn f!r Deviaticns Section 3.1 - Reactivity C:ntrol Systems O TECHNICAL SPECIFICATIONS

10. This change reflects a generic change to NUREG-1431 proposed by the industry owners groups. The justification for this change is contained in Technical Specification Task Force (TSTF) change number TSTF-12.

11. This change reflects a generic change to NUREG-1431 proposed by the industry owners groups. The justification for this change is contained in Technical Specification Task Force (TSTF) change number TSTF-108.

12. This change reflects a generic change to NUREG-1431 proposed by the industry owners groups. The justification for this change is contained in Technical Specification Task Force (TSTF) change number TSTF-14.

13. Not used'_h: pr:p:::d_.,__. :h:ng:

.. climin:te: ::nfu__2i n within

..o._,,.u.. .__ the pl:nt

,__...u

,_.,,____o...,__.u....

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

With the SR :: :urrently a rded, :n interpr:t:ti:n ::uld bc =d:

th:t up:n : f:iiure Of the r:d: t: ::ve, the SR :: f:11:d nd ther:f re, :in:: n: ::ti:n: wer: :::il:ble, :n entry in LCO 3.0.3 would be required. Th: pr:p:::d ;;rding cl:rifit: th: intent Of the SR, :: d:::ribed in th: S::::, t: Only infer tripp:bility Of the r:d:. Currently, thi: i inf- r:d by d;;;n:tr: ting ::::::nt. The pr:p:::d a:rding f : ::: n tripp:bility, r:ther th n ::::::nt which i: ::::i: tent with th: intent :: d:: ribed in the S:::: f:r LCO 3r4 ,4.

\

i l

l Catawba Units 1 and 2 23 Supplement 15/20/97l

' c ' ' * *, -

e.

.,s,*'.' .

~. ,* v."

.g Ja!. *"a

. , - , l' y - . *~

Rod Group Alignment Limit B 3.1 BASES ACTIONS -

(continued) plant systems. conditions in an orderly manner and without challenging the

~-

SURVEILLANCE SR 3.1 1 REQUIREMEN13 1

Verification that individual rod positions are within alignment limits at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position.

If the rod 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months accesplishes the same, goal. position deviation monitor is takes into account other rod position information that isThe specified Fmqu continuously so that durin available to the operator in the control room ,

be detected. g actual md motion deviations can immediately SR 3.1 2 each rod be tripped. Verifying each control rod is OPERABLE would req each control rod would result in radial or axial powerHowever, b in M

' ('T1, -

od

LO gh _ rod tilts, or oscillatinaeJExercising each individual control agj vz days provides increased confidence that all meAse <* rods continue limit even if theytoare benot OPERABLE regularly tripped.ng without the exceedialignment ne t.il AJ control rod by 10 steps will not Moving each ge6, g. c to u+L tilts, or oscillations to occur.cause radial or axial power The 92 day Frequency 4 'g . operator performed in the control room ande SR 3.13.1. wnicn istak more frequentl OPERASILITY of the rods.y and adds to the determination of SR 3.115 Between required movement),2 (determination of control rod rfonnances of

. if a control rod (s) is discovere=to d be!L;"l 4 immovable, but remains trippable and aligned the control rod (s) is considered to be OPERABLE. .

control rod (s) is immovable, a determination of theAt any time, if a trippability made, and app(OPERA 81LITV) m priate actinn taken. of the control rod (s) must be Ime)tL Q '

_ (continued)

STS

[b CA M B 3.1 32 Rev 1. 04/07/95

) '

McGuire & Catawba improved TS Review Comments ITS Section 3.1, ReactMty Control Systems 3.1.4-02 B JFD-10 Bases Background discussion for ITS 3.1.4, page B 3.1-24.

The first sentence of the Bases Background discussion for STS 3.1.5 discusses rod Operability and trippability. The Bases Background discussion for corresponding ITS 3.1.4 has not adopted this material. An insert has been proposed in its place. Comment: While this change seems to add some clarity, it is not a justifiable plant specific difference. Submit a TSTF change request.

DEC Responee:

The submittal has been revised to conform to the STS.

O o _3x3., 3.1.e me,ch2.1eee

Rod Group Alignment Limits B 3.1.4

(]

V B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Rod Group Alignment Limit' BASES BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50 Appendix A, GDC 10. " Reactor Design," 60C 26, " Reactivity Control System Redundancy and Protection" (Ref. 1), and 10 CFR 50.46,

" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2).

Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group.

Control rod inoperability or misalignment may cause increased power peaking, due to the asyneetric reactivity

[ .

distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on control rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved {

by their control rod drive mechanisms (CRDMs). Each CRDM 1 moves its RCCA one step (approximately % inch) at a time, I but at varying rates (steps per minute) depending on the signal output from the Rod Control System.

l l

O (continued)

U Catawba Unit 1 B 3.1-21 Supplement 1 l

Rod Group Alignment Limits B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Rod Group Alignment Limits BASES BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown and control rods is an initial assumption in all safety analyses

, that assume rod insertion upon reactor trip. Maximum rod

/ misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, " Reactor Design," GDC 26, " Reactivity Control System Redundancy and Protection" (Ref. 1), and 10 CFR 50.46,

" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2).

Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. ,

Control rod inoperability or misalignment may cause '

p increased power peaking, due to the asynmetric reactivity Q

distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in ,

design power peaking limits and the core design requirement i of a minimum SDM. j l

Limits on control rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately % inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System.

O (continued) I Catawba Unit 2 B 3.1-21 Supplement 1 l

a

,.,, . .. . s... .

. :. . s . .

' Rod Gtoup Alignment Linits '

B 3.lg p B 3.1 REACTIVITY CONTROL SYSTEMS W G d 8 3.1 Rod Group Alignment Limits l

BASES BACKGROUND The uitunu. te.g., tri m ys or t shutdown control is an initial tion in p/ T that ace

rnd in weti 1julfety lyse esmetne f n f Max 1Rm roG -

- // misalignment is an initial asstaption in the safety analy::is l that directly affects core power distributions and STET assumptions of available SDN.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50. Appendix A.

GDC 10. " Reactor Design," GDC 26,

Reactivity Control S Redundancy and Protection" (Ref.1). and 10 CFR S0.46, ystem

Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2). !

j Nechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group.

Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod .

worth for reactor shutdown. Therefore, control rod alignment design powerand OPERABILITY are related to core operation in of a minism aking limits and the core design requirement -

v Limits on control rod alignment have been established, and all rod positions are monitored and controlled durinq power operation to ensure that the power distribution and reactivity SDM limitslimits defined by the design power peaking and are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately % inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System.

The RCCAs are divided among control banks and shutdown banks.

to provide Each forbank may be further subdivided into two groups of two or more ise reactivity control. A group consists that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups (continued)

M 'B 3.1 24 Rev 1. 04/07/95 Cd b 0

INSERT i j

Not used. l j i

I l

1 l

O ,

1 l

I i

Cedawk INSERT Page B 3.1-24 O

McGuire & Catawba improved TS Review Comments ITS Section 3.1, Reactivity Control Systems 3.1.4-03 B JFD-4 Bases discussion of the Applicable Safety Analyses for ITS 3.1.4, page B 3.1-27.

The last paragraph of the Bases discussion of the Applicable Safety Analyses for STS 3.1.5 addresses why the specification was retained in the Technical Specifications. The Bases discussion of the Applicable Safety Analyses for corresponding ITS 3.1.4 has not adopted this material. An insert has been proposed in its place. The insert states that the alignment requirements in the LCO do not satisfy any 10 CFR 50.36 criteria. Comment: Rod alignment can directly affect power distributions and SDM which are initial conditions assumed in the safety analyses. This is not a Justifiable plant specific difference. Revise the submittal to conform to the STS. Utilize the current reference for criterion 2.

DEC Response:

The submittal has been revised to conform to the STS.

O l

i O mc3_cr_3.1 3.1 -7 March 2,1998

l Rod Group Alignment Limits B 3.1.4 BASES l APPLICABLE rods (Ref. 4). With control banks at their insertion l SAFETY' ANALYSES limits, one type of analysis considers tt,e case when any one

! (continued) rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod-is misaligned from its group by 12 steps. Another type of

. l misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5).

The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local

-linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.

Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (F (X,Y,Z)) and the nuclear enthalpy hot channel factor

( (X,Y)) are verified to be within the'r limits in the C R and the safety analysis is verif%d to remain valid.

When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. .Therefore, the limits may not preserve the design peaking factors, and Fa(X,Y,Z) and F1n(X,Y) must be verified directly by incore mapping. Bases Section 3.2 (Power Distribution Limits) contains more com Fg(X,Y,Z) and F!n(plete discussions X,Y) to the operatingoflimits.

the relation of l Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in the safety analyses.

l Therefore they satisfy Criterion 2 of 10 CFR 50.36 (Ref. 6).

l (continued) l Catawba Unit 1 B 3.1-24 Supplement 1

Rod Group Alignment Limits B 3.1.4 BASES LCO The requirements on rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The limits on shutdown and control rod alignments ensure that the assumptions in the safety analysis will remain valid, and that the RCCAs and banks maintain the correct power distribution and rod alignments.

The requirement to maintain the alignment of any one rod to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 irches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed. Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the p OPERABILITY (i.e., trippability) and alignment of rods have Q

the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are normally bottomed and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, " SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LC0 3.9.1,

" Boron Concentration." for boron concentration requirements during refueling.

n b (continued)

Catawba Unit 1 B 3.1-25 Supplement 1 l

I Rod Group Alignment Limits B 3.1.4 BASES l APPLICABLE rods (Ref. 4)., With control banks at their insertion SAFETY ANALYSES limits, one type of analysis considers the case when any one I (continued) rod is comnietely inserted into the core. The second type of analysis considers the case of a completely withdrawn sind e rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps. Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to detennine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5).

The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.

Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (F (X,Y,Z)) and the nuclear enthalpy hot channel factor O ( ,(X,Y)) are verified to be within their limits in the C R and the safety analysis is verified to remain valid.

I When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and Fo(X,Y,Z) and Flu (X,Y) must be verified directly by incore mapping. Bases Section 3.2 (Power Distribution Limits) i contains more com Fo(X,Y,Z) and F1,(plete discussions X,Y) to the operatingoflimits.

the relation of l Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in the safety analyses.

l Therefore they satisfy Criterion 2 of 10 CFR 50.36 (Ref. 6).

(continued) l Catawba Unit 2 B 3.1-24 Supplement 1 I l

Rod Group Alignment Limits B 3.1.4 BASES LCO The requirements on rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The limits on shutdown and control rod alignments ensure that the assumptions in the safety analysis will remain valid, and that the RCCAs and banks maintain the correct power distribution and rod alignments.

The requirement to maintain the alignment of any one rod to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed. Failure to.

meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs. all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES l in which neutron (or fission) power is generated, and the i OPERABILITY (i.e., trippability) and alignment of rods have O the potential to affect the safety of the planc. In MODES 3, 4, 5, and 6 the alignment limits do not apply because the control rods are normally bottomed and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LC0 3.1.1, " SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1,

" Boron Concentration," for boron concentration requirements during refueling. ,

)

(continued)

Catawba Unit 2 8 3.1-25 Supplement 1 l

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Rod Group Alignment L1::its B 3.1 .

o -

APPLICABLE SAFETY ANALYSES Another type of misalignment occurs if one RCCA fails to (continued) insert upon a reactor trip and remains stuck fully withdrawn.

detemine that the required SON is set with the maximumThis condition worth RCCA also fully withdrawn (Ref. 5).

The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that 1)ERMAL POWER will be adjusted so that excessive local linear heet rates (URs) will not occur, and that the requirements on SOM and ejected rod worth are preserved.

Continued operation of the reactor with a misaligned control rod the and is allowed nuclear enthalpy if the heat hot channel flux hot channel factor (F" factor

)) (Fo I jy verified to be within their limits in the COUt safety analysis is verified to remain valid, leen a contro p>g rod is misaligned, the assumptions that are used to determine the rod insertion limits. AFD limits, and quadrint power tilt limits are not preserved.

Therefore. .the limits mayaust notbepreserve the design peaking factors, and Fog and verified directly by incore mapping, hw.

y 89 ion 3.2 (Power Distribution Limits) contains more {

complete discussions of the relation of Fo )

operating limits. F",, to the

[nuTGGian and w. ~.vl rod

~

d *>E\T directly rela initial to power di if ano all butions and N h.ich en:are I 4 A ;they sati ions asstmedd safety analyses./Therefore Criterian ? er rm enrystatemenO J 54.c I l

/ i gpg,so.% (Re $.(e$h LCO The limits on s control rod ali t

)e awMions in the safety analysis wi s ensure that i Qo Tie requir;._

trip the assumed % activity wi .es onyurtJumILHT ensure wasemain vali upon re and 10 dnser,ted_J t inm n ITY Mr 1 be available and will em==sts aisaremwe that the RCCAs rod alignment. and banks slaintain the correct power distribution and

{g ,y ..,e The or minusrequirement 12 steps isto maintain the@ alignment /to within plus conservative.

assumed in safety analysis is 24 steps (15 inches), and inThe minimum misalignm fully inserted is assumed.some cases a total misalignment from fully withdrawn to ,

(continued)

WO % B 3.1 27 CA b Rev 1. 04/07/95 (O

{

l INSERT 1 I Not Used-o i

l INSERT B 3.1-27 I

O Co/aA N -

McGuire & Catawba Improved TS Review Comments ITS Section 3.1, Reactivity Control Systems 3.1.5, ShteGw i. Bank insertion Limits 3.1.5-01 Bases JFD 2 Bases JFD 4 Bases Background discussion for ITS 3.1.5, page B 3.1-34.

The Bases Background discussion for STS 3.1.6 states that the control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally automatically controlled by the rod control system, but they can also be manually controlled.

They are capable of adding negative reactivity very quickly (compared to borating). This descriptive material has not been adopted in the Bases Background discussion for corresponding ITS 3.1.5. Additionally, the Bases Background discussion for STS 3.1.6 states that the control banks must be maintained above designed insertion limits and are typically near the fully withdrawn position during normal full power operation. The Bases discussion for corresponding ITS 3.1.5 states that the shutdown banks must be maintained above designed shutdown bank insertion limits and are typically near the fully withdrawn position during normal full power operation. Comment: These are not justifiable plant specific differences. Provide adequate justification for these changes or revise the submittal to conform to the STS.

DEC Response:

ITS 3.1.5 specifies the requirements for shutdown bank insertion limits. The Bases Background in general describes requirements which are generic to both shutdown and control banks. The information which has been deleted is specific only to the control banks and is not appropriate to this Bases for shutdown banks. Therefore, the STS control bank information was deleted and/or modified to reflect the bases specific to the shutdown bank insertion limits. The detailed information for control banks is described in the Bases for ITS 3.1.6. This change v.as justified by JFD 2 which indicates a revision to be consistent with the writer's guide, i.e., the Bases discussion should reflect the LCO which it supports. JFD 15 has been added providing the above explanation.

mc3_cr_3.1 3.1 -8 March 2,1998 i

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Sfurtdown Bank Insertion L1:it 8 3.1 ,

B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1 Shutdown Bank Insertion Limits 1@

j BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available ejected rod worth. SON and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50. Appendix A.

E)C 10.

Reactor Design " G)C 26. " Reactivity Control System Redundancy and Protection." G)C 28. " Reactivity . Limits *

(Ref.1), and 10 CFR 50.46. "Acceptana Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design -i power peaking and SON limits are preserved.

The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further st& divided into two groups to provide for ise reactivity control. A group consists of two or more that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered foshion. but always within one step of each other. t= = m >

L D' ants MED four control banks anC K m=--- @down /

5 Danks. See LCO 3.1 3.

Algnment Limits.' for g. -

control and shutdown rod (PERABI M and alignment requirements, and LCO 3.1;B.

Rod Position Indication.* for A

@ position indication requirements.

l I cont the r banks are used fo precise reactivi control o j or. The positions f the control be are norma y automatically rolled by the Rod rol Sys y can also be na but 11y controlled. e k of dding rwaative rea ivity very auickly ( y are aHna)J The wi.swpbanks must be maintained above-redc desi nsertion limits and are t ically near the fully withdr position during normal fu power operations.

j af* ba-A.) %wNQ (continued)

% B 3.1 34 Rev 1. 04/07/95 cM O

Justificatien fer Deviatitns 5:ctica 3.1 - Rrctivity Centrol Systems O mu

9. Reload Physics Testing Program includes Critical _ Boron (ARO),

Control Rod Worth (Rod Swap method), and ITC determination. Ref.

UFSAR Chapter 14.3 Physics Test Program.

10. This change represents a clarification of the Bases desired by the plant staff.

11. This level of detail is not needed since the tests are described in the UFSAR.

12. This change reflects a generic change to NUREG-1431 proposed by the industry owners groups. The justification for this change is contained in Technical Specification Task Force (TSTF) change number TSTF-89.

13. This change reflects a generic change to NUREG-1431 proposed by the industry owners groups. The justification for this change is contained in Technical Specification Task Force (TSTF) change number TSTF-108.

O 14. The Bases discussion of the Applicability for NUREG LC0 3.1.3 states that in Modes 4, 5, and 6 the LCO is not applicable because there are no DBAs using the MTC as an accident analysis assumption which are initiated from these modes. The Bases discussion for ITS 3.1.3 states that no DBAs using the MTC as an analysis assumption are limiting when initiated from these modes. MTC may be considered as an input to accident analyses performed in NODES 4, 5, and 6, however, accidents initiated from these modes are not expected to be limiting events. Therefore, the proposed clarification to the Bases is appropriate and necessary to remain consistent with the safety analysis.

15. The Bases Background discussion for STS 3.1.6 states that the control banks are used for precise reactivity control of the reactor. Additionally, the Bases Background discussion for STS 3.1.6 states that the control banks must h maintained above design insertion limits and are typically near tJe fully withdrawn position during normal full power operation. The STS 3.1.6 Bases Background in general describes requirements which are generic to both shutdown and control banks. However, STS 3.1.6 addresses only the requirements for shutdown bank insertion limits. Therefore, the STS O

Catawba Units 1 and 2 23 Supplement 15/20/07l

Justificction for Devictions SIcticn 3.1 - R:cetivity C ntrc1 Systems O eses control bank information was deleted and/or modified in ITS 3.1.5 to reflect the Bases specific to the shutdown bank insertion limits.

The control bank discussion is not appropriate to this Bases for shutdown banks. The detailed information for control banks is described in the Bases for ITS 3.1.6 consistent with STS 3.1.7.

16. The STS 3.1.7 Bases provides an example figure which illustrates insertion limits. The specific insertion limit figures for unit operation are maintained in the COLR in accordance with the current licensing basis. The inclusion of " example" curves within the Bases of the Technical Specifications is not currently required and creates an unnecessary administrative burden for maintaining examples which remain consistent with any changes in COLR methodology. The inclusion of these examples is not necessary to maintain compliance with the LCO and could lead to inappropriate use of figures which do not reflect the current cycle core design }

limits. Therefore, the ITS 3.1.6 Bases does not contain any example insertion limit figures.

1

\ 17. The STS 3.1.5 Bases for Actions A.1.1 and A.2.2 imply that the i required baration to restore shutdown margin is emergency boration, however, this is not consistent with the actual stated action. The operator will use the appropriate means for restoring SDM based on ;

the amount that it is out of limits. ITS 3.1.4 Bases does not include " emergency." This change is justified in Discussion of Change LA.1.

j l

O Catawba Units 1 and 2 33 Supplement 15/20/07l

McGuire & Catawba improved TS Review Comments ITS Section 3.1, Reactivity Control Systems l l

l l

3.1.6, Control Bank insertion Limits 3.1.6-01 B JFD-3 Bases Background discussion for ITS 3.1.6 STS Bases Figure B 3.1.7-1 The Bases Background discussion for STS 3.1.7 refers to Figure B 3.1.7-1. STS Figure 3.1.7-1 illustrates control bank insertion versus percent RTP. Neither the refe rence nor the figure have been adopted in the Bases for corresponding ITS 3.1.6. Comment: These are not justifiable plant specific differences. Provide adequate justification for these changes or revise the submittal to conform to the STS.

DEC Response:

The STS Bases provides an example of figures which illustrate insertion limits. The specific curves for operation are maintained in the COLR in accordance with the current licensing basis. Duke Energy believes that the inclusion of " example" curves within the Bases of the Technical Specifications creates an unnecessary administrative burden for maintaining examples which remain consistent with any changes in COLR methodology and may also lead to innppropriate use of curves which do not refle.ct the current cycle core design. JFD 16 has been added providing the above explanation.

mC3_cr._3.1 3.1-9 March 2.1998

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Control Bank Insertion L'iaits B 3.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1 Control Bank Insertion Limits BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip.

affect core power and fuel burnup distributions andThe insertion limits d assumptions insertion rate.of available SOM. and initial reactivity 1

The applicabla criteria for these reactivity and power distribution design requirements are 10 CFR 50. Appendix A.

GDC 10.

Reactor Design.* GDC 26.

Reactivity Control System Redundancy and Protection.* GDC 28.

Reactivity Limits

s (Ref.1), and 10 CFR 50.46.

Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2).

Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SOM limits are preserved.

The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks.

subdivided into two groups to provide forEach bank may be further control. A group consists of two or more ise reactivity

(- electrically paralleled to step simultaneously.

that are f RCCAs consists of two groups that are moved in a stA bank of

\ "S fashion. always within one step of each other.

)lan four control banks and at least tw3 s I

2anks.

See LCO 3.1[R.

Rod Group Aligruent Limits 7 mr p, ve, control ano m utoown rod OPERABILITY and alignment reapirements. and LCO 3.1 position indication

Rod Position Indication." for i irements.

( The An ermle controlis bank ppoviinsertion 'imits are specified in the COLR.

of' nformattorvenly iO

\figuri33.1.M1.

lo or aDove the "nsertion limit lines. control banks are required to be at t

(daure moved in Erl.71an overlap alasinme==

pattern. Juverlap Ann /he control banks are a

[ 'aven ieo wv a.ym uy twopshtrol ba The rei (crud.inued) ca B 3.1-39 Rev 1. 04/07/95 O

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C

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(17.231)

,231 (67.231) 200' BANK 8 FULLY WITHORAWN

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m Y 150 A S

C m ?

2 BANK C >

100 m

a ,8A D

o 1 '

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)

1 o

o 50 e' 7

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/ THIS FIGURE FOR {

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FULLY lNSERTED DO NOT USE FOR l 0 (19.cl . /

/

f' OPERATION.

! i o 20 f('%

e

,46 so 80 100 D'< ;

' PERCENT OF RTP l

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Control Bank Insettl0n vs. Percent RTP l

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E STS /

/ 8 3.1-45 Rev 1. 04/07/95 N) k_

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l l

Justificatitn for Deviatiens 1 Sectitn 3.1 - R: activity Control Systems j BASES l

1 control bank information was deleted and/or modified in ITS 3.1.5 to reflect the Bases specific to the shutdown bank insertion limits.

7 The control bank discussion is not appropriate to this Bases for shutdown banks. The detailed information for control banks is described in the Bases for ITS 3.1.6 consistent with STS 3.1.7.

16. The STS 3.1.7 Bases provides an example figure which illustrates insertion limits. The specific insertion limit figures for unit operation are maintained in the COLR in accordance with the current l licensing basis. The inclusion of " example" curves within the Bases of the Technical Specifications is not currently required and creates an unnecessary administrative burden for maintaining examples which remain consistent with any changes in COLR methodology. The inclusion of these examples is not necessary to maintain compliance with the LCO and could lead to inappropriate use of figures which do not reflect the current cycle core design limits. Therefore, the ITS 3.1.6 Bases does not contain any example insertion Iimit figures.

17. The STS 3.1.5 Bases for Acttons A.1.1 and A.2.2 imply that the required baration to restore shutdown margin is emergency boration, however, this is not consistent with the actual stated action. The operator will use the appropriate means for restoring SDM based on the amount that it is out of limits. ITS 3.1.4 Bases does not include " emergency." This change is justified in Discussion of Change LA.1.

I I

I O

Catawba Units 1 and 2 3B Supplement 15/20/07l l

I

f McGuire & Catawba improved TS Review Comments ITS Section 3.1, Reactivity Control Systems l

1 i

3.1.7, Rod Position Indication l 3.1.7 No comments 3.1.8, Physics Test Exceptions - Mode 2 3.1.8 No comments

)

O

,,e e g

O ENCLOSURE 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION SECTION 3.2 I

i O l l

McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Dietribution Limits 3.2.1, Heat Flux Hot Channel Factor (Fo(x,y,z,))

3.2.1-01 JFD 6 l ITS 3.2.1 Required Action B.1 CTS 4.2.2.2.C.2(a)

Required Action B.1 of STS 3.2.1 requires reducing the AFD limits 21% for each 1% F o

( exceeds the limit. Required Action B.1 of ITS 3.2.1 has inserted the phrase "from COLR i

limits" after 21%. Ct,mment: This is not a Justifiable plant specific difference. The phrase "from COLR limits" changes the meaning in the proposed location, and is incorrect. Revise the submittal to conform to the STS.

DEC Response:

CTS 4.2.2.2.c.2.a.1 and associated footnote 3 clearly indicate that the revised AfD limits is a reduction in the established COLR limits based on the change in margin to the operational transient limit. The S TS model does not employ the NRC approved methodology utilized by Duke Energy for the Catawba and McGuire Nuclear Stations. The changes proposed to the STS are necessary and appropriate to maintain the current licensing basis reflected in the CTS.

3.2.1-02 JFD 6 -

ITS 3.2.1 Required Action C.1 Required Action C.1 of ITS 3.2.1 states reduce the OTAT trip setpoint from COLR limit by M

KSLOPE for each 1% Fa (x,y,z) exceeds limit. This is neither clear nor consistent with the format of STS 3.2.1 Required Action B.1. Comment: Revise the submittal to delete the phrase "from COLR limit" from Required Action C.1 of ITS 3.2.1.

DEC Response:

M CTS 4.2.2.2.c.3 requires reducing the OTAT K1 value by KSLOPE for each 1% Fo (x,y,z) exceeds limit. CTS footnote 3 indicates that KSLOPE is in the COLR and footnote 4 indicates that the K1 value is in CTS Table 2.2-1. A review of CTS Table 2.21 shows that the K1 value is in the COLR also. The Bases for Action C provides a detailed explanation of the relationship of K1 to OTAT. The STS model does not employ the NRC approved methodology utilized by Duke Energy for the Catawba and McGuire Nuclear Stations. The changes proposed to the STS are necessary and appropriate to maintain the current licensing basis reflected in the CTS.

mc3_cr_3.2 3.2-1 March 2,1998

McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits 3.2.1-03 JFD 6 ITS Required Action C.2 Bases discussion for ITS 3.2.1 Required Action C.2, insett page R 3.2-15 (insert 6).

It has been pre. posed to add the requirement to adjust Fa '(x,y,z)RPS by the equivalent reduction in OTaT trip setpoint to CTS 4.2.2.2.c.3 as corresponding ITS 3.2.1 Required Action C.2. Nn justification has been provided for this proposed change to the CTS. Comment: ITS M

3.2.4 Required Action C.2 is initiated by Condition C which states that Fo (x,y,z) >

Fa'(x,y,z)RPS. It is not Clear Why exceeding the limit should result in an adjustment to the limit and why this is an appropriate compensatory action. Is the adjustment only applied to subsequent sury;illances? Revise the submittal to provide the technical justification for this proposed charga.

DEC Response:

^

The Fa'(x,y,z)RPS includes a margin factor to the center line fuel melt temperature. If that limit is exceeded, then compensatory action to adjust the OTAT trip setpoint is required in order to trip the unit sooner and prevent conter line fuel melt following a transient. Once the OTAT trip setpoint is reduced, the available margin is increased. An adjustment is then -

O necessary in the limit, using the increased margin, in order to restore compliance with the LCO and exit the condition. These adjustments maintain a constant margin and ensure that centerline fuel melt does not occur. Without a corresponding adjustment in the limit, the condition would always be applicable and the surveillance would not be met even though the appropriate compensatory actions (reduction of OTAT trip setpoint) were already taken. In accordance with SR 3.0.1, the surveillance would no longer be required to be performed since SRs are not required for variables not within limits. Clearly this is not the intent of the existing requirements but would be permissible if the proposed action were not required. This adjustment is currently made as part of the actions taken in CTS 4.2.2.2.c.3. JFD 16 is added to justify this change in the STS and a Bases clarification was added.

The adjustment must be sufficient to ensure that tha surveillance is expected to be met at the next surveillance interval. This is done by ecapolating previous measurements to 31 EFPD ;

into the future. If the survoillance is not expected to be met at the 31 EFPD point, an j additional adjustment of 2% is made or the surveillance is performed at a shorter frequency consistent with when the parameter would be expected to be not met.

mc3_cr_3.2 3.2-2 March 2,1998

Fa(X,Y,Z)

B 3.2.1 BASES O ACTIONS C.1 and C.2 (continued)

If the RPS margin is less than zero, then F5(X,Y,Z) is greater than Fh(X,Y,Z)RPS and there exists a potential for F5(X,Y,Z) to exceed peak clad temperature limits during certain Condition 2 transients. The Overtemperature AT K1 value is required to be reduced as follows:

Kimusno = K1 - l KSLOPE * % RPS Margin l where Kimusno is the reduced Overtemperature AT K1 value KSLOPE is a penalty factor used to reduce K1 and is defined in the COLR

% RPS Margin is the most negative margin determined above.

Reducing the Overtemperature AT trip setpoint from the COLR limit is a conservative Ntion for protection against the consequences of transients since this adjustment limits the peak transient power level which can be achieved during un anticipated operational occurrence. Once the OTAT trip setpoint is reduced, the available mar An

. adjustment is then necessary in the Fh(gin is increased.

O X , Y , Z)RPS using the increased margin, in orde' to restore compliance with the LCO and exit the condition. These adjustments limit, maintain a constant margin and ensure that centerline fuel melt does not occur. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the small likelihood of a limiting transient in this time period. Adjusting the transient RPS limit by the equivalent change in OTAT trip setpoint establishes the appropriate revised surveillance limit.

L.1 If Required Actions A.1 through A.4, B.1, or C.1 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This allowed Completion Time is reasonable cased on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.

(continued) l Catawba Unit 1 B 3.2-8 Supplement 1 )

Fo(X,Y ,Z)

B 3.2.1 BASES A

b SURVEILLANCE SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.1.3 are modified by a REQUIREMENTS Note. The Note applies during the first power ascension after a refueling. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained. This allowance is modified, however, by one of the Frequency conditions that requires verification that Fy(X,Y,Z) is within the specified limits after a power rise of 210% RTP over the THERMAL POWER at which it was last verified to be within specified limits. BecauseFV(X,Y,Z)couldnothave previously been measured in this reload core, power may be increased to RTP prior to an equilibrium verification of FV(X,Y,Z) provided nonequilibrium measurements of F5(X,Y,Z) physics are performed testing. This ensures at various power that some levels duringofstartug(X,Y,Z) determination Fo is made at a lower power level at which adequate margin is available before going to 100% RTP. The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which Fo was last measured.

SR 3.2.1.1 Verification that FV(X,Y,Z) is within its specified steady state limits involves either increasing F5(X,Y,Z) to allow for manufacturing tolerance, K(BU), and measurement uncertainties for the case where these factors are not included in the Fo limit. For the case where these factors X,Y,Z) to the F g are limit included, can be performed. a directSpecifically, comparison of FyFa(X,Y,Z) is the measured value of Fo(X,Y,Z) obtained from incore flux map results. Values for the manufacturing toler9nce, K(BU), and measurement uncertainty are specified in the COLR.

The limit with which F4(X,Y,Z) is compared varies inversely with power above 50% RTP and directly with functions called K(Z) and K(BU) provided in the COLR.

O) y (continued)

Catawba Unit 1 B 3.2-9 Supplement 1 l l

L _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _

Fa(X ,Y ,Z)

B 3.2.1 BASES O ACTIONS C.1 and C.2 (contineed)

If the RPS margin is less than zero, then F5(X,Y,Z) is greater than Fh(X,Y,Z)RPs and there exists a potential for F5(X,Y,Z) to exceed peak clad temperature limits during certain Condition 2 transients. The Overtemperature AT K1 value is required to be reduced as follows:

Klamusito = K1 - l KSLOPE * % RPS Margin l where K1Amustro is the reduced Overtemperature AT K1 value KSLOPE is a penalty factor used to reduce K1 and is defined in the COLR

% RPS Margin is the most negative margin determined above.

Reducing the Overtemperature AT trip setpoint from the COLR limit is a conservative action for protection against the consequences of transients since this adjustment limits the peak transient power level which can be achieved during an anticipated operational occurrence. Once the OTAT trip setpoint is reduced, the available mar An

[],

adjustment is then necessary in the Fh(gin X ,Y ,Z)is increased.

RPS limit, using the increased margin, in order to restore compliance with the LC0 and exit the condition. These adjustments maintain a constant margin and ensure that centerline fuel melt does not occur. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the small likelihood of a limiting transient in this time period. Adjusting the transient RPS limit by the equivalent change in OTAT trip setpoint establishes the appropriate revised surveillance limit.

El If Required Actions A.1 through A.4, B.1, or C.1 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not appP. cable. This is done by placing the plant in at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.

(continued) l Catawba Unit 2 B 3.2-8 Supplement 1 I

Fa(X,Y,Z)

B 3.2.1 BASES

. SURVEILLANCE SR 3.2.1.1, SR 3.2.1.2, and SR 3.2.1.3 are modified by a REQUIREMENTS Note. The Note applies during the first power ascension after a refueling. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained. This allowance is modified, however, by one of the Frequency conditions that requires verification that F5(X,Y,Z) is within the specified limits after a power rise of 210% RTP over the THERMAL POWER at which it was last verified to be within specified limits. Because FV(X,Y,Z) could not have previously been measured in this reload core, power may be increased to RTP prior to an equilibrium verification of F5(X,Y,Z) provided nonequilibrium measurements of F5(X,Y,Z) physics are performed testing. at various This ensures powerdetermination that some levels duringofstartup(X,Y,Z)

Fg is made at a lower power level at which adequate margin is available before going to 100% RTP. The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which Fo was last measured.

SR 3.2.1.1 .

Verification that F5(X,Y,Z) is within its specified steady state limits involves either increasing F5(X,Y,Z) to allow for manufacturing tolerance, K(BU), and measurement uncertainties for the case where these factors are not included in the F0 limit. For the case where these factors X,Y,Z) to the Fn are limit included, a directSpecifically, can be performed. comparison of FyFa(X,Y,Z) is the measured value of Fn(X,Y,Z) obtained from incore flux map results. Values for the manufacturing tolerance, K(BU), and measurement uncertainty are specified in the COLR.

The limit with which F5(X,Y,Z) is compared varies inversely with power above 50% RTP and directly with functions called K(Z) and K(BU) provided in the COLR.

(continued)

Catawba Unit 2 B 3.2-9 Supplement 1 l

Y,Y,1 FaCZ$RFoMet*1ool 3.2.

(-'. ETIONS (continued)

CONDITION REQUIRED ETION COMPLETION TIME B. F*o(Z) rpt withGn B.1 R its 11mit r hours

_J h1 each f[(X,Y,Z)7@(V,f) ag (g

a .

g {(x,v,t}

P g '

@ 4 Required Action and N.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

ryo l B.'2. A a f* I L4 MW jf

( in AFD. )

Q X,Y,7) > C.I Re1,u Oc oT6T 2 bowes F,'(y, y,37* 5 T" P SeWat 4" .

COLR 1; 14 5 f

\

kuort beus i%

FE &J,O cx*O '#1 AND i

7'2 h*dd C.1 Adjud I*' (v#E}RPS y se niv.W+ ttW" \

,oTar We sew ^

v i

WOpd5 3.2 5 Rev 1. 04/07/95 C.t).S l

Justification f r Deviations Sectitn 3.2 - Power Distributica Limits TECHNICAL SPECIFICATIONS

14. This change is consistent with generic change TSTF-95 submitted to the NRC by the industry owners groups, except that it is also made applicable to the trip setpoint reduction action of LC0 3.2.4 which was deleted from the NUREG but retained in the plant specific ITS.

)

Additionally, this change is not implemented for LC0 3.2.2 since the Duke Power Company methodology in the CTS specifies different actions from the NUREG and the change is not applicable.

15. This change is consistent with generic change TSTF-99 submitted to the NRC by the industry owners groups.

l

16. ITS 3.2.1 Required Action C.2 requires an increase in the Fq (x,y,z)RPS limit by the equivalent reduction in OTAT trip setpoint made in L

Required Action C.1. The Fq (x,y,z)RPS limit includes a margin factor to the center line fuel melt temperature. If that Iimit is exceeded, then compensatory action to adjust the OTAT trip setpoint is required in order to trip the unit sooner and prevent center line fuel melt O following a transient. Once the OTAT trip setpoint is reduced, the O avatlable margin is increased. An adjustment is then necessary in the limit, using the increased margin, in order to restore compliance with the LCO and exit the Condition. These adjustments maintain a constant margin and ensure that centerline fuel melt does not occur. Without a corresponding adjustment in the lirait, the Condition would always be applicable and the Surveillance would.not be met even though the appropriate compensatory actions (reduction of OTAT trip setpoint) were already taken. This adjustment is currently made as part of the actions taken in CTS 4.2.2.2.c.3. The adjustment is made to ensure that the Surveillance is expected to be met at the next surveillance interval.

17. Required Action A.5 of STS 3.2.4 requires calibrating the excore detectors to show zero QPTR. QPTR, by definition is a ratio or fraction and a *zero* QPTR is not possible. ITS 3.2.4 and the associated Bases have been corrected to reflect the plant design and require a zero "QPT.* The NRC has acknowledged the necessity for a similar clartftcation in the Safety Evaluation Report for Vogtle '

Eletric Generation Station for License Amendment 96 (Unit 1) and 74 (Unit 2) in support of the conversion to improved technical specifications.

(

32 Supplement 15/20/07 lCatawbaUnits1and2

l

)

INSERT 6 C.I and C.2 The margin contained'within the reactor protection system (RPS) Overtemperature AT setpoints f during transient operations is based on the relationship between Fo*(X,Y,Z) and the RPS limit, Fo(X,Y,Z)"", as follows:

% RPS Margin = l Fo"(X*Y*Z) '

100%

( Fo'(X, Y,Z)"8 j If the RPS margin is less than zero, then Fo"(X,Y,Z) is greater than Fo'(X,Y,Z)"" and there exists a potential for Fo(X,Y,Z) to exceed peak clad temperature limits during certain Condition 2 transients. The Overtemperature AT K1 value is required to be reduced as follows:

Kl.4, a = K1 - l KSLOPE * % RPS Margin l where Klaj..a is the reduced Overtemperature AT K1 value KSLOPE is a penalty factor used to reduce K1 and is defined in the COLR

% RPS Margin is the most negative margin determined above.

Reducing ti:e Overtemperature AT trip setpoint from the COLR limit is a conservative action for protection against the consequences of transients since this adjustment limits the peak transient

' power level which can be achieved during an anticipated operational occurrence. Once the OTAT trip serpoint is reduced, the available margin is increased. An adjustment is then necessary in the FgL(x,y, FPS limit, using the increased margin, in order to restore compliance with the LCO and exit the condition. These adjustments maintain a constant margin and ensure that centerlinefuel melt does not occur. -The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is sufficient considering the smalllikelihood of a limiting transient in this time period. Adjusting the transient RPS limit by the equivalent change in OTAT trip setpoint establishes the appropriate revised surveillance limit. j i

l i

INSERT Page B 3.2-15 Catawba

I McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distrhution Limits 3.2.1-04 JFD 6 ITS SR 3.2.1.1 Frequency Bases discussion for ITS 3.2.1 Surveillance Requirements, page B 3.2-16.

' Bases JFD 7 Bases discussion for ITS SR 3.2.2.1, insert page B 3.2-26.

The Frequency for STS 3.2.1.1 requires that Fo"(x,y,z) be verified within the steady ' state limit once after each refueling prior to Thermal Power exceeding 75% RTP. This Frequency has not been adopted in corresponding ITS SR 3.2.1.1. The Surveillance Note for corresponding ITS SR 3.2.2.1 states that during power escalation at the beginning of each fuel cycle, Thermal Power may be increased until an equilibrium power level has been achieved, at which time a power distribution map is obtained. The ITS does not provide a specific requirement to verify that Fo"(x,y,z) is within limit prior to reaching RTP. The Bases insert for ITS SR 3.2.2.1 states that because F"a(x,y,z) could not have previously been measured in the reload core, power may be increased to RTP prior to equilibrium verification of Fo(x,y,z) provided nonequilibrium measurements of Fo(x,y,z) are performed at various power levels during startup pifysics testing. Neither the CTS nor the ITS appear to specify requirements for the nonequilibrium measurements referred to in the Bases insert. Comment: The measurement requirements should be explicitly specified in the Technical Specifications.

Revise the submittal to either specify when the first steady state measurement is required, or when the nonequilibrium measurements are required. The justification for these requirements O- should also be provided.. Also see comment 3.2.2-02.

DEC Respones:

Duke Energy is in compliance with the current technical specifications, the ANSI 19.61 1985 standard, and the physics testing program approved by the NRC. An equilibrium flux map at 75% power is not required by the current program, but is required under the STS. Duko Energy is not proposing to add any new more restrictive surveliMee requirements in this area and believes that the approved program is acceptable. The nonequilibrium measurements are not contained in the current TS and are not proposed for addition within ths ITS.

mc3_cr._3.2 3.2-3 March 2,1998

McGuire & Catawba improved TS Review Comments A ITS Section 3.2, Power Distr &Jtion Limits 3.2.1-05 DOC A.3 CTS 4.2.2.2.c.2(b)

CTS 4.2.2.2.c.2(b) requires complying with the Action requirements of Specification 3.2.2, M

treating margin violation in 4.2.2.2.c.1 as the amount by which Fo (x,y,z) is exceeding its limit. This requirement has not been retained in corresponding ITS 3.2.1. DOC A.3 states that this action could penalize the unit by requiring a power reduction; steady state lirnits being applied to steady state and transient conditions. This proposed change has been categorized as administrative. Comment: This proposed change is less restrictive.

DEC Response:

CTS 4.2.2.2.c.2.b is an alternative to the CTS 4.2.2.2.c.2.a requirement which allows a I reduction in AFD limits consistent with the reduction in margin to the Fo "(x,y,z) operational tranalent limit. CTS 4.2.2.2.c.2.a is the preferred option because AFD limits would be appropriately compensated for the reduction in margin to the transient limit and the unit would remain at RTP. Under the CTS 4.2.2.2.c.2.b option, an unnecessary power reduction would be required consistent with the actions taken if the steady state limit were not met. The removal of a non mandatory alternative action which is not used in practice 'is considered admin 1strative as described in DOC A.3.

)

l mc3_cr_3.2 3.2-4 March 2,1998

I McGuits & Catawba improved TS Review Comments ;

ITS Section 3.2, Power Distribution Limits 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (Fm(x,y))

3.2.2-01 JFDs 5 and 6 Completion Times associated with Required Actions A.2.1 and A.2.2 of ITS 3.2.2 Bases discussion of Required Actions A.2.1 and A.2.2 for ITS 3.2.2, page B 3.2-24.

CTS 3/4.2.3 Actions b.1 and b.2 Six hours is provided to complete Actions b.1 and b.2 of CTS 3/4.2.3. The Completion Tirae associated with corresponding Required Actions A.2.1 and A.2.2 of ITS 3.2.2 is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The CTS markup does not identify a justification for these proposed changes. Comment: Revise the submittal to conform to the CTS. This change is beyond the scope of the conversion.

DEC Response: l An additionalless restrictive Discussion of Change (L.11) is added to justify the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months completion time. Duke Energy does not believe this change to be beyond the scope Of the ITS conversion since the STS actions already allow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . i I

O -

4 mc3_cr_3.2 3.2-5 March 2.1998

3Pted=&e- 3,m ,

POW Q DISTRIBUTION LIMITS 30.2 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR FaHfX.Y) dIMITING C0dITION FORAPERATiOE)

W 3.2 Fay (X,Y) shallM11mited by i sing tne following eletionsM M L Fm (X,Y) s F s (X,T]LCO Where: F, X,Y) = the me ured radial peak.

(F (X,Y)]LCO = the ximum allowable r tal peak as defin in e CORE OPERATING MITS REPORf (COLR ,

efPLICABILITY: MODE j:/[ 7d'Yf d.hJ At ACTION:

t ww w A.- MY f/ . ** W .

M With Fm(X,Y) exceeding itsilfr "A.th Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , red POWER fnns RATED THERMAL POWER at i s efor the eachallowable 1% that FTq(X,Y) g exceeds the limit, and L.t h Withi rs e ther:

A. t.1Restore F M(X,Y) to within the limit of Specification 3.2.3 for RATEDTHEkLPOWER,or Q

"A.2 Z

$ Reduce the Power Range Neutron Fi g-Migh Trip Setpoint in Table 2.2 3 at least RRH4 for each 1% that Fg (X,Y) exceeds that limit, ane

@ Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the limit of Specification 3.2.3, either:

ALL @ Restore F M RATED THEk(X,Y) to within the limit of Specifica. ion 3.2.3 for POWER, or g

( Pe gorm th M ollowin d cti_ons3 A 10 I LTabled.z-Dby at least TRHW for

@ Reduce the OTAT each 1% that Fm (X,Y) exceeds the limit, and h3 RRH is the nt of THERMAL reduction requi to compensa foD each 1% th F M(X,Y) exceed the limits of Sp fication 3.2.

provided n th[COLR per Spe fication 6.9.1.9.

m TRH is he amount of OTAT , setpoint reducti required to c ensate OOp-for pr ch 1% that F M(X ded in the C pe

> exceeds the limi of Specificati pecification 6.9. 9.

3.2.3 1

/

CATAWBA - UNIT 1 3/4 2-7 Amendment No. 148 O

v

, , ,. ~

WA.fim 3. Z. 2.

POWER OfSTRIBUTION LIMITS 3 6 2 b LEAR ENTHALPY RISE HOT CHANNEL FACTOR FoHfX.Y)

(LINITIWCONDITION FOR OPERAT10sD LCO 3.2. Fm(X,Y) shall,by/IImited by impo ngthefollowingre'ationshiM M

Fm (X,Y) s F n'(X,f]Lco Where: F (X,Y) gp .h

= the meas ed radial peak.

[ g (X,Y)]'C0 l

= the ximum allowable radi peak as defi d Q t e CORE OPERATING LIH S REPORT (COL .

APPLICABILITY: M;JE 1.

g- NeTE. gd A+m AAI g4 4 EIlQFe j ~M **r C~J.m

+.A u$8 <~

C ff c .s .

, ,J .i Q C M A With Fm(X,Y) exceeding itsllimit:} -

Al @ Within POWER iathours, I as redge'the for each 1%allowable that Fm THERy(X,Y) exceeds POWER from RATED THERPG1

., and the limit

@ Within urs her: f A2.1 @ Restore F "(X,Y) to within the Ilmit of Specification 3.2.3 for RATEDTHENLPOWER,or

?

A.22 @ Reduce the Power Range Neutron Flug-High Trip Setpoint in TabV2.2-3 at least RRH% for each 1% that Fa (X,Y) exceeds that limit, and

@ Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the Itait of Specification O

v A.S.I 3.2.3, either:

hRestoreF u(X,Y) to within the limit of Specification 3.2.3 for RATEDTHEblPOWER,or (2. Perfard the follodne actind A , 3. 2.1 (a) Reduce the OTATgn TJbte 27-f by at least TRH W for each 1% that F g M(X,Y) exceads the limit, and b RRH is the unt of THERMAL reduction requir tocompensatefor) each 1% th F M(X,Y) exceeds he limits of Specif cation 3.2.3 provided th[COLRperSpecifcation6.9.1.9.

L,A - W TRH is amountofjTATK etpoint reduction r utred to compensate for ea 1% that F ceeds provi in the C0$ X p(er,Y)

S g the ification limit of pecification 3.2.3, 6.9.1.9.

CATAWBA - UNIT 2 3/4 2 7 Amendment No. 142 m~ 3 af ,

i J LRYu

i Discussitn of Ch:ngis Ssction 3.2 - Power Distribution Limits TECHNICAL CHANGES - LESS RESTRICTIVE associated potential for a plant transient. The increase in acceptable, considering the small likelihood of a severe transient during this period and the requirements of other actions to reduce )

THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and verify peaking factors within 24 l hours. This change is consistent with NUREG-1431.

L.11 CTS 3/4.2.3 Actions b.1 and b.2 allow 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to either restore F"u(X,Y) to within limits or reduce the Power Range Neutron Flux !

trip setpoint. ITS 3.2.2 Required Actions A.2.1 and A.2.2 allow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to perform these actions. The additional 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allows time to perform a second flux map to confirm results, or determine that the condition was temporary, without implementing on unnecessary trip setpoint change and associated potential for a plant transient. The increase is acceptable, considering the small likelihood of a severe transient during this additional time period and the requirements of other actions to reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This change is consistent with similar actions in NUREG-1431 which allow a completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to

, perform trip setpoint reductions.

l l

Catawba 011ts 1 and 2 Page L - 55 Supplement 15/20/97l

N) Significtnt Hazards C:nsid:raticn S;ctica 3.2 - Power Distributica Linits

!v LESS RESTRICTIVE CHANGE L.11 The Catawba Nuclear Station is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1431, " Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive.

Below is the description of this less restrictive change and the No Significant Hazards Consideration for conversion to NUREG-1431.

CTS 3/4.2.3 Actions b.1 and b.2 allow 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to either restore F%(X,Y) to within 'imits or reduce the Power Range Neutron Flux trip setpoint. ITh 3.2.2 Required Actions A.2.1 and A.2.2 allow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to perform these actions. The additional 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allowe time to perform a second flux map to confirm results, or de' .uine that the condition was t ?mporary, without implementing an unnecessary trip setpoint change and associated potential for a plant transient. The increase is acceptable, considering the small likelihood of a severe transient during this additional time period and the requirements of other actions to reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This change is consistent with similar

{ octions in NUREG-1431 which allow a completion time of B hours to L}/ perform trip setpoint reductions.

In accordance with the criteria set forth in 10 CFR 50.92, the Catawba Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change increases the Completion Time to reduce the Power Range Neutron Flux - High trip setpoint from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The probability of an accident is not affected by this change. The time allowed to reduce trip setpoints is not an initiator of any analyzed events. The consequence of an accident is not affected by increasing the time to reduce the power range neutron flux - high trip setpoint since accident mitigating equipment continue to perform the intended safety function. This change does not alter assumptions relative to the mitigation of an accident. Other actions, such as reducing THERMAL POWER, along with other Specifications for heat flux hot channel factor, axial y flux difference, and quadrant power tilt ratto ensures the power Catawba Units 1 and 2 Page 24N of 2BN Supplement 15/20/07l

No Significar.t Nazards Consider;tica l Secticn 3.2 - Power Distributien Licits O distribution limits are not further degreded. Therefore, the proposed change does not involve ~ a significant increase in the probability or consequence of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from arty accident previously evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or different type of equipment will be installed) or changes in the manner in which the plant is nperated. Monitoring of other required parameters will ensure operation within safety analyses assumptions. Therefore, the proposed change wtIl not create the possibtIity of a new or different kind of accident than any previously evaluated.

3. Does this change involve a significant reduction in a m rgin of safety?

The proposed change increases the Completion Time to reduce the Power Range Neutron Flux - High trip setpoint from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Other actions to reduce THERMAL POWER during this period along with other power distribution limit specifications ensure the power distribution limits are not further degraded. The safety analysis assumpttons wtIl still be maintained, therefore, the proposed change does not involve a significant reduction in a i margin of safety. I b

V lCatawbaUnits1and2 Page 25 2 of 28 9 Supplement 15l20l97

McGuire & Catawba improved TS Review Comments O ITS Section 3.2, Power Distribution Limits 3.2.2-02 JFD 6 ITS SR 3.2.2.1 Frequency Bases JFD 7 Bases discussion for ITS SR 3.2.2.1, insert page B 3.2 26.

The Frequency for STS 3.2.2.1 requires that F,g be verified within limits once after each refueling prior to Thermal Power exceeding 75% RTP. This Frequency has not been adopted in corresponding ITS SR 3.2.2.1. The Surveillance Note for corresponding ITS SR 3.2.2.1 states that during power escalation at the beginning of each fuel cycle, Thermal Power may be increased until an equilibrium power level has been achieved, at which time a power j distribution map is obtained. The iTS does not provide a specific requirement to verify that F,g is within limit prior to reaching RTP. The Bases insert for ITS SR 3.2.2.1 states that M

because F ,g(x,y) could not have previously been measured in the reload core, power may be increased to RTP prior to equilibrium verification of F,g(x,y) provided nonequilibrium measurements of F,g(x,y) are performed at various power levels during startup physics testing. Neither the CTS nor the ITS appear to specify requirements for the nonequilibrium measurements referred to in the Bases insert. Comment: The measurement requirements.

should be explicitly specified in the Technical Specifications. Revise the submittal to either specify when the first steady state measurement is required, or when the nonequilibrium measurements are required. The justification for these requirements should also be provided.

O Also, see comment 3.2.1-04.

DEC Responee:

See response to Comment 3.2.1-04.

mc3_cr_3.2 3.2-6 March 2,1998

McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits O

3.2.2-03 Bases Background discussion for ITS 3.2.2, page B 3.2-21.

Bases discussion of the Applicable Safety Analyses for ITS 3.2.2, page B 3.2-22.

The Bases discussions for STS 3.2.2 describe limiting the minimum DNBR below a specific value. The Bases discussions for corresponding ITS 3.2.2 contain a general reference to a design limit value and associated critical heat flux (CHF) correlation, without providing the value or CHF correlation. Comment: This is not a justifiable plant specific differenec.

Revise the Bases discussions to provide a more specific description of the design limit value and associated CHF correlation for the minimum DNBR.

DEC Response: i The critical heat flux correlations and corresponding design limit values of DNBR are .

described in approved methodologies. The approved methodologies for DNBR are ik din -

lTS 5.6.5. Different methodologies are applicable for different vendor fuel types. Si < there 1 is no single correlat!on or design limit applicable to all fuel types, a discussion in the tsases would be quite lengthy and potentially confusing to the operating staff. The identification of

)

J the various correlations and specific design limit values in the Bases would not provide any I useful information to the operators. The exclusion of the detailed discussion within the Bases .

does not preclude the requirement to utilize approved methodologies. JFD 9 has been added d to the STS Bases markup to justify this difference.

3.2.2-04 Bases JFD 4 Bases Background discussion for ITS 3.2.2, page B 3.2 22.

The Bases Background discussion for STS 3.2.2 states that the DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant. The Bases Background discussion for corresponding ITS 3.2.2 has not adopted this information, and has not substituted any information in its place. Comment: Revise the Bases Background discussion to provide a description of the DNB design basis objective.

DEC Response:

The STS Baces are not consistent with the current licensing basis. There are three events which allow some fuel failures, i.e., locked rotor, rod ejection, and single rod withdrawal.

These ca.%s are acceptable since they do not exceed the 10 CFR 100 dose limits. The proposed ITS Bases already indicate that the DNB design basis precludes DNB for operational transients and events of moderate frequency in the paragraph preceding the subject STS paragraph. Ne other substitutions are necessary to address the DNB design basis. JFD 10 has been e.Jded to the STS Bases markup to justify this difference.

mc3_cr_3.2 3.2-7 March 2,1998

B 3.2 8 3.2 POWER DISTRIBUTION LIMITS g

B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F",,,

FL CA Psdx,A BASES BACKGROUW The purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. The design limits on local (pellet) and integrated fuel red peak power density are expressed in tems of hot channel factors. Control of the core power distribution with respect to these factors ensures that local conditions in the fuel rods and coolant channels do not challenge core integrity at arty location during either normal operation.or a postulated accident analyzed in the safety analyses, k is defined as the ratio of the integral of the linear power along the fuel rod with the highest integrated ,

to the average integrated fuel rod power. Therefore, is a measure of the maximum total power produced in a fuel rod.

kissensitive 1 loading patterns, bank insertion, and fuel burnup. typically increases with control bank insertion and typi ly decreases with fuel burnup.

disnotdirect1 distribution map bseasurable but is inferred from a power O system.

power di determine ined with the novable incore detector fically, the results of the three dimensional ion map are analyzed by a computer to This factor is calculated at least every 31 EFPD. However, during operation, the global power distribution is monitored LCO 3.2.3 *AX1AL FLUX DIFFERENCE (AFD).* and LCO .2.4. " QUADRANT POWER TILT RATIO (QPTR)

which address directly and continuously measured process variables.

~

The COLA design basis value provides of the departurepeaking from nucleatefactor (DW) is met for normal operation, operational transients, boiling limits that ensu 9tb and any transient condition arisino from awmts of noderate tremica+s J

eg g V g frequency. The D e design basisAprecluots'D W 4 limiti the minimim local DW heat flux ratio to tesM F Nds I

' i 8" g using correlation. All ) 74 de sgn limited ransient events are assumed to begin with an value that satisfies'i,'g ,4 w ch,$g.%(tw] the LCO requirements.

(continued) y4fS B 3.2 21 Rev 1. 04/07/95 i

I e/J $

s O

B 3.2.2 BASES BACKGROUND (continued)

Operation outside the LCO limits may produce une-table consequences if a DNR limitino event occurs, fThe 05 dati w sis res at there is overheati of the fuelfthat O i.

resul in po ible claddi rforation th the release of '

cfiss cts to the or coolantg 1

APPLICABLE Limitsondprecludecorepowerdistributionsthatexceed SAFETY ANALYSES the following fuel design limits:

a. [ There st W at ast 958 111ty at O h; confi hot level the 95/95 fuel in the criterion) does not the a DE itinn* ,

b. bring a large break loss of coolant accident (LOCA).

adding temperature (PCT) must not exceed

c. During an e.)ected rod accident the ition to the fuel must not exceed 280 cal /gu f, and

d. Fuel design limits required by GDC 26 (Ref. 2) the condition when control rods must be capable of shutting down the reactor with a minimas required SON with the hi withdrawn. ghest worth control rod stuck fully For transients t may be 08 limited. the Reactor Coolant System flow and are the parameters of most importance. The limits on ensure that the De design basis is met for norinal operation, operational transients, and any transients arising from events of moderate frequency. The De desian basis is met by lim ting the inum DER to theM0lgrern;erson er WJirusing Ellel_Q ll4,7

% aR a

\

'yg c

4 CHF correlation. This value provides a high degf1Mt of Qak j 9 experience athat rance DWthe

, hottest fuel rod in the core does not The allowable l@,, limit increases with decreasing power level. This functionality in FE is included in the analyses that vide the Reactor Core Safety Limits (SLs) of SL 2.1.1. fore, any D e events in which the calculation of the core limits is modeled implicitly use (continued)

WOG4f$ B 3.2 22

/ Rev 1. 04/07/95 Cf6 O

L -

I

\

Justificatign f:r Deviatiens t Section 3.2 - Power Distribution Limits am i Westinghouse Reactors. This methodology .is contained in an NRC approved Topical Report, DPC-NE-2011PA. As a result, NUREG Bases 3.2.3 has been modified to reflect the Duke Power Company methodology. The DPC methodology is similar to the relaxed axial offset control (RA0C) methodology. Details have been added to the Bases to describe the DPC methodology.

9. The Bases Background and Applicable Safety Analysis discussions for STS I 3.2.2 describe limiting the minimum DNBR below a specific value. The Bases for ITS 3.2.2 contain a general reference to a design limit value and associated critical heat flux (CHF) correlation, without providing the value or CHF correlation. The critical heat flux correlations and corresponding design limit values of DNBR are described in approved methodologies listed in ITS 5.6.5. Different methodologies are applicable for different vendor fuel types. Since there is no single correlation or design limit applicable to all fuel types, a discussion in the Bases would be quite lengthy and potentially confusing to the operating staff and would not provide any useful information to the operators. The exclusion of the detailed discussion within the Bases:

does not preclude the requirement to utilize approved methodologies.

10. The Bases Background discussion for STS 3.2.2 states that the DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant. The STS Bases are not consistent with the current licensing basis and are deleted. There are three events which allow some fuel failures, i.e., locked rotor, rod ejection, and single rod withdrawal. These cases are acceptable since they do not exceed the 10 CFR 100 dose limits. The proposed ITS Bases indicate that the DNB design basis precludes DNB for operational transients and events of moderate frequency.

11. STS 3.2.2 Action A.1.1 requires that power be reduced to less than 50%

RTP. The Bases discussion of STS 3.2.2 Action A.1.1 refers to the Condition A Note and describes the requirements to detennine Fa prior to exceeding 50% RTP and 75% RTP, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching or exceeding 95% RTP. CTS 3.2.3 oction a only requires that power be reduced by RRH as defined in the COLR. This power level would be much higher than 50%. The STS bases are predicated on the presumption that power is reduced to less than 50% and that Fg wtll be determined at O

33 Supplement 15/20/97 l Catawba Units 1 and 2

McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits 3.2.2-05 Bases JFD 7 Bases discussion of the Applicable Safety Analysis for ITS 3.2.2, page B 3.2 23.

The Bases discussion of the Applicable Safety Analysis for STS 3.2.2 states that all transients N

that may be DNB limited are assumed to begin with an initial F aH as a function of power level defined by the COLR limit equation. The Bases discussion of the Applicable Safety Ant!ysis for ITS 3.2.2 has not adopted this matsrlal, and has not substituted any material in its place.

Comment: Justify the deletion of this information or revise the Bases discussion of the Applicable Safety Analysis.

DEC Response:

The Bases statement in the Applicable Safety Analysis for STS 3.2.2 does not accurately reflect Duke Energy Company's licensing basis for Catawba and McGuire Nuclear Stations. In N

the evaluation of limiting DNB transients, an initial F aH value that satisfies the LCO requirements as delineated in the COLR limit equation is assumed. In the Duke Energy methodology, the margin to the DNB limit is monitored through routine flux mapping of the reactor core. If the margin to DNB is exceeded, then compensatory actions as described in the TS are performed. The Duke Energy methodology is described in the NRC approved Q

Topical DPC NE-2011PA, ' Core Operating Limits for Westinghouse Reactors." This change is already justified by JFD 7 in the ITS submittal.

l s

i mc3_cr_3.2 3.2-8 March 2,1998 I

1 4

l McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits l 3.2.2-06 Bases JFD 5 Bases discussion of ITS 3.2.2 Action A.1, page B 3.2-24.

Tne Bases discussion of STS 3.2.2 Action A.1.1 refers to the Condition A Note and describes the requirements to determine Fa prior to exceeding 50% RTP and 75% RTP, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching or exceeding 95% RTP. This material has not been adopted in the Bases discussion for corresponding ITS 3.2.2. Comment: Justify excluding this information or revise the Bases to conform to the STS.

DEC Response:

STS 3.2.2 Action A.1.1 requires that power be reduced to less than 50% RTP. CTS 3.2.3 action a only requires that power be reduced by RRH as defined in the COLR. The STS bases are predicated on the presumption that power is reduced to less than 50% and that Fa will be determined at this power level during subsequent operation. This discussion is not appropriate for Catawba or McGuire based on differences in methodology and is not retained in the Bases for ITS 3.2.2 Action A.1. Additionally, the discussion of when additional measurements would be made is already discussed in the Bases for STS Action A.3 (ITS A.4) and is unnecessarily repeated in the Bases for STS Action A.1.1. JFD 11 has been added to the STS Bases markup to justify this difference.

O l

I mc3_cr_3.2 3.2 9 March 2,1998

F*a )

B 3.2.2 BASES

[FMX,dY)

I LCO thermal feedback and greater cont 1! insertion at low (continued) power levels. The limiting value o is allowed to increase .3% for every it RTP reduction in T}ERMAL POWER.,7

. Appe w M 9 _

g intrean h APPLICABILITY The M. limits must be maintained in MODE 1 to preclude core power distributions from exceed 1 the fuel design limits ne guyf' for OleR and PCT, Applicability n other modes is not 43 4 4 g required because thert is either insufficient stored energy kbed o~a+

in the fuel or insufficient bei transferred to the aC t.ca. ,,,,,<.i coolant to ire a limit on dist bution of core power. Speci 11y, the design bases events that "gg gg,.

sensitive to in other modes (MODES 2 through 5) M-t__

significant in to DDR. and.therefore, there is no need we c

-J )

8

U" to restrict

' inthesemodes.'f N U k 49" Y 4 Le J ACTIONS _ A /1/h ' ,, g A 6 A' G a.aa. w .I O i m er ,($th FZ.,Ed!seding Jfs lir ib unit is allowed @ hours to

{ it. wi rm,to witnin its ,ist -

irns reir/mation may. mr e 1e, involve reali ng any saligned rods or 3suc1re to bring within itse-- .

1 Init.

i When the limit is exceeded, the 015R limit is not 11xely .

W violated in steady state ation, because events that could significantly O -

10 A.11 o.J A1.sJ theIT,value(e control red misalignment) are considered in analyses. However, the DISR limit may be violated if a 018 1 miting event occurs. Thus, the allowed Comp 1

, static safety on Time of hours provides an acceptable time to restore to within itsy11mits without allowing the plant to remain 'n an RT -- unacceptable condition for an extended period of time.

Condition A is andified a Note that reautres that %p eru h wA 1y Reau' red Actio 4 and be completta wheneverh g,, eg o { 5 wnd tion A is entered. . i f -.

~

peca th w usarirw me:ian n ~ 1 wi ---

the A D 4,c)#

a ,

tt4=. _

1aA1 Requ1 red Ac; ion

_ igless requires i

) another measurement and calculati of Pm' ar1mn a "

in accordance with SR 3.2.2.1.

\ "However, A.3.2.M d power recucea low bus ni naquired Action .3 requi that a deterni ion of F"4,.

i I pe prior exceedi

RTP. pri to exceedi (continued) ;

g B 3.2 24 Rev 1. 04/07/95 CtJ S s

O

B 3.2.2 BASES O ACTIONS W continued)

)

0II 75t TP, and wi RTP. In a n 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months fter reach tion. Recu or I Action .2 is per if I r ascensi is dela w naet 24 J AJO2.1andAd.2.2 f b*

IMMT. - - !

pg*" .If the value of 8 is, not. ruti,.M to within its saaaified limitAnn. _ ov an;-aine a mman-I U -'*- ar or . ---- un F

f"j_f'*_Y fD rpRRH'!* p mgM.

~

i

- 2r n reauca sne rdwer Range usutrtin flux-Highdo SrMiN n c ='++- wi':h Reauired Action VY 29 4xtm i 4 no ly use h" "* ad " C+ '

st -

an )

The reGuction on trio Wweints ensures Inst continu ng operation reasigsrat _an weie low powerT 4 level with adequate D e R serginL_p na an - wuan

'nus or = nour or c. . r,,ggg at t with those al ion A.I.2.1 1 consistent 3W", g, .

for in red Action A. 1 and provi t

an acceptabi' ime to reach level f g full power wation wi required p.

allowing the apt to rwaai irf Me +,us i.J P'W" h r.M hI W an unac >1e condition The an iod of t1 .

w ei A t.<.- k ion Times of hours for Actions .1.1

_and A. .1 are not ,tive., ---

i .ekt,,,.J 4,J,3

' , , y,mo,A.

O The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to reset the trip setpoints per Required Action Af)2.2 recognizes that, once power is reduced, the safety ane ysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that any inadvertently trip the Reactor Protection System.

b g g ,n A ,e. 4 T Tr;e k N pr once thew 1m has been reduced &DGHFlffp)per Required Actions.v2.n an incore flux map (SR 3.2.2.1) must De ootained and the seasured value of % verifieq not A.tti to exceed the all limit at the lower power level.TThe unit is providad g 6 over and above the additional hours to perforunnism hours allowed by either Action A.(Cl' or

{ Action N . The letion Time of @ hours is f!)

i acceptable because of the increase in the De margin, wRfch j

AM (continued)

B 3.2 25 Rev 1, 04/07/95 C 4/5 O

Justificatirn for Devictiens Section 3.2 - Power Distribution Limits o

BASES Westinghouse Reactors. This methodology is contained in an NRC approved Topical Report, DPC-NE-2011PA. As a result, NUREG Bases 3.2.3 has been modified to reflect the Duke Power Company methodology. The DPC methodology is similar to the relaxed axial offset control (RA0C) methodology. Details have been added to the Bases to describe the DPC methodology.

9. The Bases Background and Applicable Safety Analysis discussions for STS 3.2.2 describe limiting the minimum DNBR below a specific value. The Bases for ITS 3.2.2 contcin a general reference to a design limit value and associated critical heat flux (CHF) correlation, without providing the value or CHF correlation. The critical heat flux correlations and corresponding design Iimit values of DNBR are described in approved methodologies listed in ITS 5.6.5. Different methodologies are applicable for different vendor fuel types. Since there is no single correlation or design limit applicable to all fuel types, a discussion in the Bases would be quite lengthy and potentially confusing to the operating staff and tould not provide any useful information to the

(% operators. The exclusion of the detailed discussion within the Bases does not preclude the requirement to utilize approved thodologies.

i

10. The Bases Background discussion for STS 3.2.2 states that the DNB design basis ensures that there is no overheating of the fuel that ;

results in possible cladding perforation with the release of fission i products to the reactor coolant. The STS Bases are not consistent with the current licensing basis and are deleted. There are three events which allow some fuel failures, i.e., locked rotor, rod ejection, and single rod withdrawal. These cases are acceptable since they do not exceed the 10 CFR 100 dose limits. The proposed ITS Bases indicate that the DNB design basis precludes DNB for operational transients and events of moderate frequency.

11. STS 3.2.2 Action A.1.1 requires that power be reduced to less than 50%

RTP. The Bases discussion of STS 3.2.2 Action A.1.1 refers to the Condition A Note and describes the requirements to determine Fg prior to exceeding 50% RTP and 75% RTP, and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching or exceeding 95% RTP. CTS 3.2.3 action a only requires that power be l reduced by RRH as defined in the COLR. This power level would be much l higher than 50%. The STS bases are predicated on the presumption that power is reduced to less than 50% and that Fg will be determined at i

1 l

lCatawbaUnits1and2 33 Supplement 15/20/97 l l

Justificatirn f r Deviatiens Section 3.2 - Power Distribution Limits BASES this power level during subsequent operation. The STS discussion is not appropriate based on differences in methodology and is not retained in the Bases for ITS 3.2.2 Action A.1. Additionally, the discussion of when additional measurements would be made is already discussed in the Bases for STS Action A.3 (ITS A.4) and is unnecessortly repeated in the Bases for STS Action A.l.1.

12. The STS 3.2.3 Bases provides an example of figures which illustrate AFD limits. The specific AFD limits for operation are maintained in the COLR in accordance with the current licensing basis. The inclusion of

" example" figures within the Bases of the Technical Specifications is not currently required and creates an unnecessary administrative burden for maintaining examples which remain consistent with any changes in AFD methodology. The inclusion of these examples is not necessary to j maintain compliance with the LCO and could lead to inappropriate use of j figures which do not reflect the current cycle core design limits.

Therefore, the ITS 3.2.3 Bases does not contain any example AFD limit figures.

i O

Catawba Units 1 and 2 42 Supplement 15/20/07 l

McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits 3.2.3, Axial Flux Difference 3.2.3-01 Bases JFD 4 Bases discussion of the Applicable Safety Analyses for ITS 3.2.3, i page B 3.2-39 l The Bases discussion of the Applicable Safety Analyses for STS 3.2.3 identifies the most important Condition 2,3, and 4 events. This material has not been adopted in the Bases for corresponding ITS 3.2.3. Comment: Justify not including this information or revise the Bases to identify these events for McGuire and Catawba.

DEC Response:

l The ITS Bases are revised to provide the appropriate events for Catawba and McGuire. l l

l O

mc3_cr_3.2 3.2-10 March 2,1998

AFD 8 3.2.3 BASES APPLICABLE within respective limits for these postulated accidents.

SAFETY ANALYSES The most important Condition 3 and 4 event is the I (continued) LOCA. The most important Condition 2 events include loss of flow.-uncontrolled bank withdrawal, and boration or dilution accidents. Condition 2 and 3 accidents simulated to begin from within the AFD limits are used to confirm the adequacy of the Overpower AT and Overtemperature AT trip setpoints.

The limits on the AFD satisfy Criterion 2 of ID CFR 50.36 (Ref.2).

LCO The shape of the power profile ~ in the axial (i.e., the vertical) direction is largely under the control of the operator through the manual operation of the control banks '

or automatic motion of control banks. The automatic motion of the control banks is in response to temperature deviations resulting from manual operation of the Chemical and Volume Control System to change boron concentration or from power level changes.

Signals are available to the operator from the Nuclear !

O . Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in l

l l

normalized flux signals between the top and bottom excore l detectors in each detector well. For convenience, this flux l difference is converted to provide flux difference units l expressed as a percentage and labeled as %A flux or %AI. l The AFD limits are provided in the COLR. The AFD limits do not depend on the target flux difference. However, the target flux difference may be used to mir.imize changes in the axial power distribution.

Violating this LCO on the AFD could produce unacceptable consequences if a Condition 2, 3, or 4 event occurs while the AFD is outside its specified limits.

i (continued) l Catawba Unit 1 B 3.2-26 Supplement 1

AFD B 3.2.3

[ BASES APPLICABLE within respective limits for these postulated accidents.

SAFETY ANALYSES The most important Condition 3 and 4 event is the (continued) LOCA. The most important Condition 2 events include loss of flow, uncontrolled bank withdrawal, and boration or dilution accidents. Cor.dition 2 and 3 accidents simulated to begin from within the AFD limits are used to confirm the adequacy of the Overpower AT and Overtemperature AT trip setpoints.

The limits on the AFD satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).

LC0 The shape of the power profile in the axial (i.e., the vertical) direction is largely under the control of the operator through the manual operation of the control banks or automatic motion of control banks. The automatic motion of the control banks is in response to temperature deviations resulting from manual operation of the Chemical and Volume Control System to change boron concentration or from power level changes.

g. Signals are available to the operator from the Nuclear t' Instrumentation System (NIS) excore neutron detecton3 (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as %A flux or %AI. 4 The AFD limits are provided in the COLR. The AFD limits do not depend on the target flux difference. However, the ,

target flux difference may be used to minimize changes in i the axial power distribution. !

Violating this LCO on the AFD could produce unacceptable consequences if a Condition 2, 3, or 4 event occurs while the AFD is outside its specified limits.

(continued) l Catawba Unit 2 B 3.2-26 Supplement 1

App gmenanaino B 3.2.

BASES (continued)

APPLICABLE SAFELY ANALYSES The AFD is a measure of the axial power distribution skewing to either the top or bottom half of the core. The AFD is sensitive to man bank positions, y core related parameters such as control core power level, axial burnip, axial xenon distribution, and, to a lesser extant, reactor coolant temperature and boron concentration.

The allowed range of the l is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions that meet safety analysis requimments.

p f The RAOC arthodol distribution 11 (Ref. 2) establi with tentativel aW de AFD limits.

dimensional ax power distributi calculations are Mr performed to trate that operation power are accept e for the LOCA oss of flow acci . and for init conditions of ipoted transients. The tentat limits are as necessary to the

_,saf analysis requi s.

CO %f The

o>'

' c*-limits on the AFD

"* ensure that the Heat Flux Hot Channel w.'6iwi or in the event of menon m' distribution following power changes.

The limits on the AFD also restrict the range of power distributions that are used as initial conditions in the analyses of Condition 2. 3. or 4 events. h This ensures that the fuel cladding integrity is maintainoa Aa"WN-O _

for these postulated accidents. ; me

-- . is w= . =-~^ .f.,- 6 6 gg IM axrtant ition 2 r =- O led banit MQop

g qr ynnaraurTrane horarianrar mTF*ia . l acc16,as T1mulated to begin from within the AFD limits areid=a+ M Conditiof 2 Q l=25 0

% used to confirm the adequacy of the Overpower AT and Overtemperature AT trip setpoints.

The limits on the AFD satisfy Criterion 2 of 6ema, roncs QFRfD.% (Ref.2 LCO The shape of the power profile in the axial (i.e., the vertical) direction is largely under the control of the operator through the asnual operation of the control banks or automatic motion of control banks. The automatic motion

                                                                                                          *(continued)

B 3.2 39 Rev 1. 04/07/95 GL O

I McGuire & Catawba improved TS Review Comments l ITS Section 3.2, Power Distribution Limits I 3.2.3-02 Bases JFD 3 ) STS Bases Figure B 3.2.3B-1, page B 3.2-42. STS Bases Figure B 3.2.3B 1 illustrates the AFD limits. This figure has not been adopted in the Bases for corresponding ITS 3.2.3. Comment: This is not a justifiable plant specific difference. Revise the submittal to conform to the STS, or provide a figure that better I illustrates the AFD limits for McGuire and Catawba. DEC Response: The STS Bases provides an example of figures which illustrate AFD limits. The specific curves for operation are maintained in the COLR in accordance with the current licensing basis. Duke Energy believes that the inclusion of " example" curves within the Bases of the 3 Technical Specifications creates an unnecessary administrative burden for maintaining examples which remain consistent with any changes in AFD methodology and may also lead to inappropriate use of curves which do not reflect the current cycle core design. JFD 12 has been added to the STS Bases markup to justify this difference. 3.2.3-03 Bases JFD 4 Bases discussion for ITS LCO 3.2.3 page B 3.2-40. The Bases discussion for STS LCO 3.2.3 refers to Figure B 3.2.38-1. This has not been adopted in the Bases for corresponding ITS 3.2.3. Comment: This is not a justifiable plant specific difference. Revise the submittal to conform to the STS. DEC Response: The STS Bases provides an example of figures which illustrate AFD limits. The specific curves for operation are maintained in the COLR in accordance with the current licensing basis. Duke Eriergy believes that the inclusion of " example" curves within the Bases of the Technical Specifications creates an unnecessary administrative burden for maintaining examples which remain consistent with any changes in AFD methodology and may also lead ; to inappropriate use of curves which do not reflect the current cycle core design. JFD 12 has been added to the STS Bases markup to justify this difference. mc3_cr_3.2 3.2 11 March 2,1998

AFD (RAOCAGWnserrna

B 3.2. BASES LCO of the control banks is in response to temperature (continued) deviations resulting from manual operation of the Chemical and Volume Control System to change boron concentration or from power level changes. Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are tak4ni free the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as 24 flux or 2AI. The AFD limits are provided in the COLR. - a:== umen --- -n nauto The AFD 11 ts 2 not depend on the target f' et difference. However, the target flux difference any be used to minisize changes in the axial power distribution. Violating this LCO on the AFD could produce unacceptable conseguences if a Condition 2. 3. or 4 event occurs while the AFD is outside its specified limits. APPLICABILITY The AFD requirements are applicable in MODE 1 or equal to 502 R1P when tne combination of ter than -

                                                    '"' ' '- i s . ' "' '                                                             P0bER (J                                                 analys                                                             -
                                                                                                                  '!_ ('f,,',',',','3 ,,,gj,]

For AFD limits developed usingMaethodology, the value of the AFD does not affect the limiting accident _ consequences with TERMAL poler < SOS RTP and for lower operating power MODES. ACTIONS L1 As an alternative to restoring the AFD to within its specified limits. Required Action A.1 requires a THERMAL POWER reduction to < 502 RTP. This places the core in a ; condition for which the value of the AFD is not important in i the applicable safety analyses. A Coopletion Time of i e (continued) B 3.2 40 Rev 1. 04/07/95 cil6 O

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

Justification f;r Deviaticns Section 3.2 - Power Distribution Limits BASES this power level during subsequent operation. The STS discussion is not appropriate based on differences in methodology and is not retained in the Bases for ITS 3.2.2 Action A.J. Additionally, the discussion of when additional measurements would be made is already discussed in the Bases for STS Action A.3 (ITS A.4) and is unnecessarily repeated in the Bases for STS Action A.1.1.

12. The STS 3.2.3 Bases provides an example of figures which illustrate AFD limits. ~e specific AFD limits for operation are maintained in the COLR in accordance with the current licensing basis. The inclusion of

             " example" figures within the Bases of the Technical Specifications is not currently required and creates an unnecessary administrative burden for maintaining examples which remain consistent with any changes in AFD methodology. The inclusion of these examples is not necessary to maintain compliance with the LCO and could lead to inappropriato use of figures which do not reflect the current cycle core design limits.

Therefor =, the ITS 3.2.3 Bases does not contain any example AFD limit figures. l O Catawba Units 1 and 2 42 Supplement 15/20/07 l

AFD 6Acrargt l

             'N r

(-15,100) (6,ID0) { UNACCEPTABLE

                                                                                 .l UNACCEPTABLE OPERATION OPERATION I  80                       /                     k x

N s

    @                    )         ACCEPTABLE g                   I          OPERATION

40 ' -31,50) (20,50)

O 20

           ~ilISFIGUREISFOR T

ILLUSTRATION 04.Y. 00 NOT USE FOR OPERATION. 0 '

         -50           -30            -10           10
               -40                                              30          50
                               -20             0           20         40 AXIAL FLUX DIFFERENCE (4)
                                                                                   )

Figure B 3.2.38-1 (page 1 of 1) AXIAL FLUX DIFFERENCE Acceptable Operation Limits as a Function of RATED THERMAL POWER 8 3.2 42 Rev 1, 04/07/95 cv5 1 o !

McGuire & Catawba imp'.oved TS Review Comments ITS Section 3.2, Power Distribution Limits 3.2.4, Quadrant Power Tilt Ratio 3.2.4-01 L.9 JFD 12 ITS 3.2.4 Required Action A.6 Bases discussion for Required Actions A.6 and A.7 for ITS 3.2.4, pages B 3.2-46 and 47. Required Action A.5 of STS 3.2.4 requires calibrating the excore detectors to show zero QPTR. Required Action A.6 of corresponding ITS 3.2.4 requires normalizing excore detectors to eliminate tilt. JFD 12 references TSTF 25 ae che justification for this oifference. Comment: TSTF-25 has been rejected. This is not a justifiable plant specific difference. Revise the submittal to conform to the STS. Revise the Bases discussions also. DEC Response: Duke Energy acknowledges that TSTF-25 was rejected, however, OPTR, by definition is a ratio or fraction. A "zero" OPTR, therefore, is not possible to acheive. The ITS and associated Bases have been corrected to reflect the plant design and show a zero "QPT." This proposed change is a slight departure from the STS format, however, it corrects the m error. The NRC has acknowledged the necessity for a similar clarification in the Safety Evaluatiort Report for Vogtle Electric Generation Station for License Amendment 96 (Unit 1) and 74 (Unit 2) in support of the convercion to improved technical specihcations. JFD 12 has been replaced by JFD 17 to provide the jt stification for this difference. 1 I mc3_.cr 3.2 3.2-12 March 2,1998

McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits 3.2.4-02 JFD 12 ITS 3.2.4 Required Action A.7 Note The Note for Required Action A.6 of STS 3.2.4 requires performing Required /.ction A.6 only after Required Action A.5 is completed. The Note for Required Action A.7 of corresponding ITS 3.2.4 requires that Required Action A.7 must be completed when Reqeired Action A.6 is implemented. JFD 12 references TSTF 25 as the justification for this difference. Comment: TSTF 25 has been rejected. This is not a justifiable plant specific difference. Revise the submittal to conform to the STS. DEC Response: Duke Energy acknowledges that TSTF-25 was rejected, however, the STS Actions are not consistent with the STS Bases discussion. The Bases indicate that once the tilt is zerood out, an added check of the hot channel factors is required. LCO 3.0.2 states that once the LCO is met, Required Actions are no lenger required, unless ctherwise stated. Therefore, once the OPT has been zeroed in STS Action A.5, all subsequeni actions (STS Action A.6) no longer must be met. The STS Note to Action A.6 relates to the sequence of action performance, i.e., A. 6 cannot be performed before A.5, however, it does not mandate that A.6 must be performed once the tilt has been normalized. The NRC has acknowledged the O necessity for this clarification in the Safety Evaluation Report for Vogtle Electric Generation blution for License Amendment 96 (Unit 1) and 74 (Ur,:t 2) in support of the conversion to improved technical specifications. JFD 12 has been replaced by JFD 18 to provide the justification for this difference. 3.2.4-03 JFD 12 ITS 3.2.4 Required Action A.7 Completion Time Bases discussion for Required Action A.7 for ITS 3.2.4, page B 3.2-46. The Completion Time for Required Action A.6 of STS 3 2.4 states within 24 hours after reaching RTP or within 48 hours after increasing Thermal Power above the limit of Required ;

Action A.1. The Completion Time for Required Action A.7 of corresponding ITS 3.2.4 states i within 24 hours after achieving equilibrium conditions with Thermal Power increased above the limit of Required Action A.1. JFD 12 references TSTF 25 as the justification for this' ,

difference. Comment: TSTF-25 has been rejected. This is not a justifiable plant specific : difference. Revise the submittal to conform to the STS. Revise the Bases discussion also. Note, the additional difference that has been proposed for the Completion Time based on JFD 11 is acceptable and should be incorporated into the response. DEC Response: The submittal is revised to conform to the STS. mc3 cr_3.2 3.2-13 March 2,1998

QPTR 3.2.4 O actioas CONDITION REQUIRED ACTION COMPLETION TIME l A. (continued) A.4 Reduce Power Range 72 hours Neutron Flux - High Trip Setpoint 2 3% for each 1% of QPTR > 1.02. AIEl A.5 Reevaluate safety Prior to analyses and confirm increasing results remain valid THERMAL POWER for duration of above the more operation under this restrictive condition. limit of Required Action A.1 or A.2 Atul A.6 --------NOTE--------- Perform Required >O Action A.6 only after Required Action A.5 is completed. Calibrate excore Prior to detectors to show increasing zero QPT. THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2 A!H1 (continued) i O l Catawba Unit 1 3.2-12 Supplement 1

QPTR 3.2.4 () ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.7 --------NOTE--------- Required Action A.7 must be completed when Required Action A.6 is completed. l Perform SR 3.2.1.1 Within 24 hours and SR 3.2.2.1. after reaching RTP DE Within 48 hours after increasing THERMAL POWER above the more restrictive limit of O - Required Action A.1 or A.2 B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to s 50% RTP. Time not met, s i O Catawba Unit 1 3.2-13 Supplement 1 l

QPTR 3.2.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l A. (continued) A.4 Reduce Power Range 72 hours Neutron Flux - High Trip Setpoint a 3% for each 1% of QPTR > 1.02. M A.5 Reevaluate safety Prior to analyses and confirm increasing results remain valid THERMAL POWER for duration of above the more operation under this restrictive condition. limit of l Required l Action A.1 or A.2 M A.6 --------NOTE--------- Perform Required Action A.6 only after b Required Action A.5 is completed. l Calibra excore Prior to detectccs to show increasing l zero QPT. THERMAL POWER I above the more restrictive limit of Required Action A.1 or A.2 M (continued) l l l Catawba Unit 2 3.2-12 Supplement 1

QPTR 3.2.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.7 --------NOTE--------- Required Action A.7 must be completed when Required Action A.6 is completed. l Perform SR 3.2.1.1 Within 24 hours and SR 3.2.2.1. after reaching RTP QE within 48 hours after increasing THERMAL POWER above the more restrictive limit of O-Required Action A.1 or A.2 B. Required Action and B.1 Reduce THERMAL POWER 4 hours associated Completion to s 50% RTP. Time not met. O Catawba Unit 2 3.2-13 Supplement 1 l

QPTR B 3.2.4 BASES O ACTIONS M If QPTR exceeds a value of 1.02, the Power Range Neutron Flux-High trip setpoint is reduced by 3% for each 1% QPTR exceeds 1.02. Lowering this setpoint maintains the same margin to trip by limiting the transient response of the l core. The 72 hour Completion Time is sufficient for this activity to be performed and is acceptable based on the low probability of a transient occurring in this time frame. M

Although F3 g(X,Y) and Fa(X,Y,Z) are of primary importance as l initial conditions in the safety analyses, other changes in ) the power distribution may occur as the QPTR limit is ! exceeded and may have an. impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is O accomplished by examining the incore power distribution. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure that, before increasing THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2, the reactor core conditions are ; consistent with the assumptions in the safety analyses. M If the QPTR has exceeded the 1.02 limit and a re-evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are recalibrated to show a zero QPT prior to increasing THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2. This is done to detect any subsequent significant changes in QPTR. (continued) l Catawba Unit 1 B 3.2-32 Supplement 1

                                                                           -QPTR B 3.2.4 BASES ACTIONS        AJi (continued)

Required Action A.6 is modified by a Note that states that the QPT is not-zeroed out until after the re-evaluation of l the safety analysis has determined that core conditions at RTP are within the safety analysis assumptions (i.e., RequiredActionA.5). This Note is intended to prevent any ambiguity about the required sequence of actions. LZ Once the flux tilt is zeroed out (i.e., Required Action A.6 l 1s performed), it is acceptable to return to full power operation. However, as an added check that the core power distribution at RTP is consistent with the safety analysis assumptions, Required Actior. A.7 requires verification that Fa(X,Y,Z) andiF g(X,Y) are within their specified limits within 24 hours of reaching RTP. As an added precaution, if the core power does not reach RTP within 24 hours, but is increased slowly, then the peaking factor surveillances must be perfonned within 48 hours of the time when the more restrictive of the power level limit determined by Required Action A.1 or A.2 is exceeded. These Completion Times are O intended to allow adequate time to increase. THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2, while not permitting the core to remain with j unconfirmed power distributions for extended periods of time. Required Action A.7 is modified by a Note that states that the peaking factor surveillances must be done after the excore detectors have been calibrated to show zero tilt l (i .e., Required Action A.6) . The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are calibrated to show zero tilt and the core returned to power. - L.1 If Required Actions A.1 through A.7 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to s 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience (continued) O Catawba Unit 1 B 3.2-33 Supplement 1 l I

QPTR B 3.2.4 BASES ( ACTIONS M If QPTR exceeds a value of 1.02, the Power Range Neutron Flux-High trip setpoint is reduced by 3% for each 1% QPTR exceeds 1.02. Lowering this setooint maintains the same margin to trip by limiting tha transient response of the l core. The 72 hour Completion Time is sufficient for this activity to be performed and is acceptable based on the low probability of a transient occurring in this time frame. Lli Although F3 g(X,Y) and Fo(X,Y,Z) are of primary importance as , initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety , concern exists. It does mean that there is an indication of I a change in the gross radial power distribution that requires an investigation and evaluation that is / accomplished by examining the incore power distribution.

                  'Specifically, the core peaking factors and the quadrant tilt     l must be evaluated because they are the factors that best        {

characterize the core power distribution. This 1 re-evaluation is required to ensure that, before increasing ) THERMAL POWER to above the more restrictive limit of ) Required Action A.1 or A.2, the reactor core conditions are consistent with the assumptions in the safety analyses. M If the QPTR has exceeded the 1.02 limit and a re-evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are recalibrated to show a zero QPT prior to increasing THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2. This is done to detect any subsequent significant changes in QPTR. l I (continued) l Catawba Unit 2 B 3.2-32 Supplement 1

1 QPTR B 3.2.4 I BASES . j ACTIONS L 1 (continued) Required Action A.6 is modified by a Note that states that the QPT is not zeroed out until after the re-evaluation of l { the safety analysis has detemined that core conditions at i RTP are within the safety analysis assumptions (i.e., Required Action A.5). This Note is intended to prevent any ambiguity about the required sequence of actions. L.Z. Once the flux tilt is zeroed out (i.e., Required Action A.6 l 1s performed), it is acceptable to return to full power operation. However, as an added check that the core power distribution at RTP is consistent with the safety analysis l assumptions, Required Action A.7 requires verification that i Fo(X,Y,Z) and Fay (X,Y) are within their specified limits l within 24 hours of reaching RTP. As an added precaution, if the Core Power does not reach RTP within 24 hours, but is increased slowly, then the peaking factor surveillances must be perfonned within 48 hours of the time when the more restrictive of the power level limit determined by Required Action A.1 or A.2 is exceeded. These Completion Times are O intended to allow adequate time to increase THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2, while not permitting the core to remain with unconfirmed power distributions for extended periods of time. Required Action A.7 is modified by a Note that states that ; . the peaking factor surveillances must be done after the ! excore detectors have been calibrated to show zero tilt l (i .e., Required Action A.6) . The intent of this Note is to have the peaking factor surveillances perfonned at operating power levels, which can only be accomplished after the excore detectors are calibrated to show zero tilt and the core returned to power. l L1 If Required Actions A.1 through A.7 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to s 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience (continued) O Catawba Unit 2 B 3.2-33 Supplement 1 l

1 Specification 3.2.4 INSERT 1 (continued Q L.8 L.9 A. (continued) A.6 -----N OTE--------- { Perform Required l Action A.6 only after Required Action A.5 is completed. Calibrate excore detectors Prior to increasing ! to show zero QPT. THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2 ANE A.7 ----NOTE--------- Required Action A.7 must be completed when Required Action A.6is completed.- O . Perform SR 3.2.1.1 and Within 24 hours after SR 3.2.2.1. reaching RTP OR Within 48 hours after increasing THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2 B. Required Action B.1 Reduce THERMAL 4 hours and associated POWER to s 50% RTP. Completion Time not met. O Catawba l Page3 of h l

Specification 3.2.4 INSERT 1 (continued) L.P L.9 A. (continued) A.6 ----------NOTE--------- Perform Required Action A.6 only after Required Action A.5 is completed. Calibrate excore detectors Prior to increasing to show zero QPT. THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2 i AND i A.7 -------NOTE--------- Required Action A.7 must be completed when Required Action A.6is completed. . b Perform SR 3.2.1.1 and Within 24 hours after SR 3.2.2.1. reaching RTP OR 8 l Within 48 hours after increasing THERMAL . POWER above the more j restrictive limit of j Required Action A.1 or A.2

B. Required Action B.1 Reduce THERMAL 4 hours and associated POWER to s 50% RTP. Completion Time not met. O - Catawba ~2 Page 3 of (

QPTR 3.2.4 ACTIONS V CONDITION REQUIRED ACTION CONPLETION TINE A. (continued) A - NOTE - -- Perform Required Action A;5 only after RequiredActionA.Q is completed. Op aimra detectors to@jgid Prior to increasing w we ut. 1 TERNAL POWER y g%t,g above of thet11mit Required k rtdr O h Action A.1 Q ) hg an

A. M ........N0tt......... 6 ? w'f U" m+,djE Required Action A @ " \

                                         !'.Q.0'f,4 )                                                       g Perform SR 3.2.1.1           Within 24 hours and SR 3.2.2.1.              after                        /

at

                                                                                          +wt, ew m3 CD'                                                                                        '~'     ""

ques ad a m6' l 3/2if ithin hours J^ > after / i sing i Ye Action A.1 p j g B. Required Action and 8.1 associated Completion Reduce THERMAL POER 4 hours to :s 50% RTP. Time not met. Wo r dS 3.2 19 Rev 1, 04/07/95 cas N O

l Justification fzr Deviations Section 3.2 - Power Distribution Li;its O TECHNICAL SPECIFICATIONS _

14. This change is consistent with generic change TSTF-95 submitted to the q NRC by the industry owners groups, except that it is also made applicable to the trip setpoint reduction action of LC0 3.2.4 which was deleted from the NUREG but retained in the plant specific ITS.

Additionally, this change is not implemented for LCO 3.2.2 since the Duke Power Company methodology in the CTS specifies different actions from the NUREG and the change is not applicable.

15. This change is consistent with generic change TSTF-99 submitted to the NRC by the industry owners groups.

L

16. ITS 3.2.1 Required Action C.2 requires an increase in the Fq (x,y,z)RPS limit by the equivalent reduction in OTAT trip setpoint made in Required Action C.1. The Fq l (x,y,z)RPS limit includes a margin factor to the center line fuel melt temperature. If that limit is exceeded, then compensatory action to adjust the OTAT trip setpoint is required in order to trip the unit sooner and prevent center line fuel melt following a transient. Once the OTAT trip setpoint is reduced, the N avatlable margin is increased. An adjustment is then necessary in the limit, using the increased margin, in order to restore compliance with the LCO and exit the Condition. These adjustments maintain a constant margin and ensure that centerline fuel melt does not occur. Without a corresponding adjustment in the limit, the Condition would always be applicable and the Surveillance would not be met even though the apprornate compensatory actions (reduction of OTAT trip setpoint) were already taken. This adjustment is currently made as part of the actions taken in CTS 4.2.2.2.c.3. The adjustment is made to ensure that the Survetilance is expected to be met at the next survetilance interval.

17. Required Action A.5 of STS 3.2.4 requires calibrating the excore detectors to show zero QPTR. QPTR, by definition is a ratto or fraction and a *zero' QPTR is not possible. ITS 3.2.4 and the associated Bases have been corrected to reflect the plant design and require a zero "QPT.* The NRC has acknowledged the necessity for a similar clartftcotton in the Safety Evaluattor, Report for Vogtle Eletric Generation Station for License Amendment 96 (Unit 1) and 74 (Unit 2) in support of the conversion to improved technical specifications.

b l Catawba Units 1 and 2 32 Supplement 15/20/97

Justificati n f;r Deviaticns Section 3.2 - Power Distribution Limits TECHNICAL SPECIFICATIONS

18. The Note for Required Action A.6 of STS 3.2.4 requires performing Required Action A.6 only after Required Action A.5 is completed. The Note for Required Action A.7 of corresponding ITS 3.2.4 requires that Required Action A.7 must be completed when Required Action A.6 is implemented. The Bases indicates that once the tilt is zeroed out, an added check of the hot channel factors is required. LCO 3.0.2 states that once the LCO is met, Required Actions are no longer required, unless otherwise stated. Therefore, once the QPT has been zeroed in STS Action A.5, all subsequent actions (STS Action A.6) no longer must be met. The STS Note to Action A.6 relates to the sequence of action performance, i.e., A.6 cannot be performed before A.5, however, it does not mandate that A.6 must be performed once the tilt has been nonnalized. The NRC has acknowledged the necessity for this clarification in the Safety Evaluation Report for Vogtle Eletric Generation Station for License Amendnent 96 (Unit 1)' and 74 (Unit 2) in support of the conversion to improved technical specifications.

O Catawba Units 1 and 2 43 Supplement 15/20/07l

QPTR B 3.2.4 BASES ACTIONS

                                         .A. / (continued) and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as onnk worths and           ing factors for rod malfunction accidents, lesen the exceeds its limit, it does not necessarily mean a safety               concern exists.

It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is examining the incore power distribution. lished by ifically, the because power they are the factors that best characterize th distribution. This re evaluation.is required to ensure limit of Required Actionincreasing.T)ElWAL that, before A.V t poler to above thet.ic*s dam are consistent with the asy.he reactor enre condit4ons ions in the safety analyses. '" f \

6 If the of thesafety OPTR has exceeded analysis is completedthe and1.02showslimit that and safetya/%. re evaluati W SfE tmtiert w iraments are us, the excore detectors are verm4tmumt td- f gNn M. unecs - - aprior to increasing TERMAL POER to- th:4 + above theylimit any W of~ Required Action A.E This is done to ia g g a i T ^_ significant C1._ in urIK. bn n.1- M - @ ylred Action A I unri y. .= s modified by a Note that states that i .4e in ,4 - =

until after the re evaluation of .

RTP are within the safety analysis assunptions (i.e.. ti Required Action A antiguity about the required sequence of actions.. This Note is inten i n

        -exto Jt4tc es g                                    gf                                                          '

o't " d'd i ys t l'~, ds

Once the s um m it W / i

             ~*e d     'l 1     is performed) operation.             it is acceptable to return to full powerw r e * (i.e. Req       I
                    <     i distribution at RTP is consistent with the safety analysis asstasptions. Required Action A Fo 24      and FL are within their specified limits withinrequires verification that                   I s(lp                      an    - -     z_ L. er tre w
                     ~

Q - Sii!I (continued) psfS B 3.2 46 g Rev 1. 04/0 s e c,t%tv; 3 %d %d% p ,7/95_

vM THetML W. m. mQ&

uc.n,3 ok

                                                                                             '.1
                                                                                                   *C Rm o Ach.. A. \ c A.3.                        2)

(

QPTR 8 3.2.4 - BASES ACTIONS A f ,(continued) 6gr ulW Em 3,Q .t A m l

                                                   -: rm rea        RTP W1       24 increased owly t            he peaki     factor                            *3      M be perf
        @                             m - -m within      hours of
                                                              =rweero,i n       ti= '

1 Gr-e ' g k 4,9 allow adequate time ~to increase Tl O NAL . ,. tended to limit of Required Action A. w soove e d- f

to remain with unconfi While not persitting the wg periods of time. distributions for extendsf l i (9 # excore '"' Required'"'"' "'* ' '""" Action AIis modified " by a Note @" " state

                                                                                              *"8'             ~

( detectors have boar & = "--- - m r= - W il l have the peaking factor surveillances(i.e.. Required Action A491 Y at operating 2=! power levels wfiich can only be ; ' 1shed after the t excore detectors are calibrated to aere tilt and the ) core returned to power. ' Ed If Required Actions A.1 through A art not coupleted within to a N00E or condition in which the requirements do no apply.

           \m f/         *,,                    To achieve this status. TIERNAL P(MR aust be .

reduced toA)S0t RTP within 4 hours. The allowed Coupletion ilme of 4 hours is reasonable, based on operating experience O regarding the amount of time siquired to roads the reduced power level without cha,11enging plant systems. SURVElli.ANCE SR 3.2.4.1 REQUIRDENTS SR 3.2.4.1 is modified by m N7 g% i ' Note 1 _ lated with three power range. channels T umy< ng4sr ="theisinput from one Power Range i i i raux cnanne' i performance of SR 3.2.4.2 n able. Note 2 allows Lsg nput rom Power)thnge Neutro dv. d._. 1s 4.z 4.1 1 tg (() o s I a l 5F This Surveillance verifies that the WIR as indicated by the Nuclear Instrumentation System (NIS) excore channels. is (continued) M 8 3.2 47 CN5 Rev 1. 04/07/95

O

McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits 3.2.4-04 DOC A.23 JFD 9 ITS SR 3.2.4.1 Note 1 CTS Table 3.3-1 Action 2 (c) CTS 4.2.4.2 Bases discussion for ITS SR 3.2.4.1 Note 1 to STS SR 3.2.4.1 addresses the Condition of the input from one Power Range Neutron Flux channelinoperable and Thermal Power < 75% RTP. Note 1 to corresponding ITS SR 3.2.4.1 addresses the Condition of the instrument channel being inoperable, but is silent regarding the Thermal Power level. The STS Note is consistent with the requirements ' of CTS Table 3.3-1 Action 2 (c) and CTS 4.2.4.2. Comment: This is not a justifiable plant specific difference. Revise the submittal to conform to the STS. Revise the Bases discussion also.

DEC Response: ITS SR 3.2.4.1 is revised to be consistent with the STS. O l O -ac,a2 - 22-'4 -c"2.'ae

1 ( McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits O 3.2.4-05 JFD 2 l ITS SR 3.2.4.2 Note An editorial difference with the Note for STS SR 3.2.4.2 has been proposed in the Note for ITS SR 3.2.4.2. Comment: This is not a justifiable plant specific difference. Revise the j submittal to conform to the STS. l DEC Response: f ITS SR 3.2.4.2 is revised to be consistent with the STS. l 1 l O - I 1 mc3_cr_3.2 3.2-15 March 2,1998 l

McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits O, 3.2.4-09 Bases JFD 5 Bases discussion for ITS SR 3.2.4, page B 3.2-47. An editorial difference with the Bases discussion for STS SR 3.2.4 regarding Note 2 has been proposed in the Bases discussion for corresponding ITS SR 3.2.4. Comment: The change deleting the requirement related to 75% RTP is incorrect and is not a justifiable plant specific difference. Revise the submittal to conform to the STS. DEC Roeponse: The editorial change is removed and the STS wording is adopted. The 3hange deleting the

  ~iS% RTP requirement is deleted and the STS wording is adopted. See also response to comment 3.2.4-04.

O O l l l [ mc3_cr_3.2 3.2-19 March 2,1998

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 -------_----_-----NOTES--_--_---_---_-----

1. With input from one Power Range ,

Neutron Flux channel inoperable and THERMAL POWER <75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance.

3. This SR is not required to be performed until 12 hours after exceeding 50% RTP.

7 days Verify QPTR is within limit by calculation. AliQ ; Once within 12 hours and O every 12 hours thereafter with the QPTR alarm inoperable SR 3.2.4.2 -------------------NOTE------------------- Only required to be performed if input from one or more Power Range Neutron Flux l channels are inoperable with THERMAL POWER 2 75% RTP. Verify QPTR is within limit using the 12 hours movable incore detectors. O 3.2-14 Supplement 1 j l Catawba Unit 1

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS _ SURVEILLANCE FREQUENCY SR 3.2.4.1 -----.--- --------NOTES---.---_-----------

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER <75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance.

3. This SR is not recuired to be performed until 12 hours after exceeding 50% RTP.

7 days Verify QPTR is within limit by calculation. AlH1 Once within 12 hours and every 12 hours O thereafter with the QPTR alarm inoperable SR 3.2.4.2 -------------------NOTE------------------- Only required to be performed if input from one or more Power Range Neutron Flux l channels are inoperable with THERMAL POWER 2 75% RTP. Verify QPTR is within limit using the 12 hours movable incore detec+. ors.

is.L'i 1

O l Catawba Unit 2 3.2-14 Supplement 1

QPTR B 3.2.4 BASES ACTIONS ILj, (continued) regarding the amount of time required to reach the reduced power level without challenging plant systems. SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by three Notes. Note 1 allows QPTR to be calculated with tt,ree power range channels if THERMAL POWER is <75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1 if more than one input from Power Range Neutron Flux channels are inoperable. Note 3 states the SR is not required to be performed until 12 hours after exceeding 50% RTP. This is necessary to establish core conditions necessary to provide a meaningful calculation. This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. The Frequency of 7 days when the QPTR alarm is OPERABLE is a'cceptable because of the low O' probability that this alarm can remain inoperable without detection. When the QPTR alarm is inoperable, the Frequency is in:reased to 12 hours. This Frequency is adequate to detect any relatively slow changes in QPTR, because for those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt. SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is required only when the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is 2 75% RTP. l With an NIS power range channel inoperable, tilt monitoring ) for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. Performing SR 3.2.4.2 at a (continued) l Catawba Unit 1 B 3.2-34 Supplement 1

QPTR B 3.2.4 BASES O ACTIONS R d (continued) regarding the amount of time rcquired to reach the reduced power level without challenging plant systems. SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by three Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is <75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows perfonnance of SR 3.2.4.2 in lieu of SR 3.2.4.1 if more than one input from

Power Range Neutron Flux channe~:; are inoperable. Note 3 states the SR is not required to be performed until 12 hours after exceeding 50% RTP. This is necessary to establish core conditions necessary to provide a meaningful calculation. This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. The Frequency of 7 days when the QPTR alarm is OPERABLE is acceptable because of the low probiility that this alarm can remain inoperable without -

-                  detection.

When the QPTR alarm is inoperable, the Frequency is increased to 12 hours. This Frequency is adequate to detect any relatively slow changes in QPTR, because for those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt. SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is required only when the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL i POWER is a 75% RTP. With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. Performing SR 3.2.4.2 at a p (continued) O l Catawba Unit 2 8 3.2-34 Supplement 1

Specification 3.2.4 INSERT 2

                                                             .L
 .________..____________________ NOTES-..-_---_-------                           ..-___

l. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER < 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance.

( l I Catawba (A l Pageb of b

Specifict. tion 3.2.4 INSERT 2 4,g

                                      .__. NOTES-.---          -.-- - . '-         --.~.

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER < 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance.

O I 1 I Catawba [(2 Page b of (, i

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 .. - NOTES - -

1. With input from one Power Range ron F m-_ lux channel w ino,perableg~ .t l remaining three power range channels 1 can be used for calculating QPTR. !

2. SR 3.2.4.2 may be perfonned in lieu of this Surveillancein tel v i.v. lux lj o I

t LEr. j Verify QPTR is within limit by 7 days calculation.

I SIE r . 4 ,nd b k pol ** h ;t l[Ees See exocMag M M-

Once within U

                                                                         ,y, a

rs thereafter with the QPTR alare inoperable

/-

l SR 3.2.4.2 ( - - - - . NOTE - - Only required to be performed if input from one or more Power Range Neutron Flux channels (E inoperable with 114ERMAL POWER 2: 752 R )

                    ..........4         .
                                          ....p....................                            ' I Verify QPTR is within limit using the               (Oncew       in' movable incore detectors.                           (12       s i

12 hours _ G!!ergem 1 l 3.2 20 Rev 1, 04/07/9S CMS : 1 l f ,

Justification f r Deviaticns Section 3.2 - Power Distribution Limits TECHNICAL SPECIFICATIONS has been modified to reflect the Duke Power Company methodology. The DPC methodology is similar to the relaxed axial offset control (RA0C) methodology.

8. The current TS 3.2.4 requires power adjustments on the amount the tilt exceeds 1.02 which is included in the Duke Power Company methodology.

This methodology is approved in DPC-NE-2011PA.

9. Not used.The ":t: t: """iC SR 3.2.4.1 indi=t= th:t thr= ps;r r=;;

ch:=:S := b; =:d t: =lecht: 0"TR =ly ah= bel = 75t ps=. The

=r=t TS definiti= Of QPTR d=: =t pl=: thi: retricti= = th:

     =huhtin Of OPTR. Thh retricti= h= t= deleted fra p=p= d

pcifin ti = 3.2.4.1. At p n = l =: h gr= ter th= 75t, the inn r det= t = r: = d t: :=ff=:==: indicti =:.

10. This change is consistent with generic change TSTF-109 submitted to the NRC by the industry owners groups.

O 11. The frequency in LC0 3.2.4 Require'd Action A.5, A.6, and A.7 has been ( j modi fied. The existing frequency in the NUREG does not account for additional power reductions mandated by Required Action A.3 which could be more restrictive than those in Required Action A.I.

12. This change is consistent with generic change TSTF-25 submitted to the NRC by the industry owners groups.

13. The current TS allow a the 25% extension to apply to this surveillance and is retained in the proposed ITS SR 3.2.4.2. Surveillance testing of power range channels is required quarterly and results in taking a channel out of service. The surveillance takes approximately 8 - 10 hours to complete. If there is a problem with the channel, only two hours would be left to complete the flux map required by SR 3.2.4.2. i

     'hese maps generally take 4 hours to complete. Without the 25%

allowance of SR 3.0.2, the surveillance could not be completed and a missed surveillance would be declared. The current frequency of 12 hours, therefore, is retained in the proposed ITS. ( Catawba Units 1 and 2 23 Supplement 15/20/07l

QPTR B 3.2.4 BASES O ACTIONS [(continued) mn nyg,,4 .h * '

                                    @    increased :. not rea    RTP wit       24 hour h                      be perfo     owly. then he peaki within      hours of factor ti t is g
                                                                                                                *[

g,g 4,9

                                                                                                                           ,,,W '

eer Ms MEh ~

se lonpletion 11 N3* {l allow adequate time ~ to increase TERNAL @ phtended to i d limit of Required Action A. to aoove c to remain with unconft while not permitting the "gg ' periods of time. u .c d stributions for extended l \ [ Required Action A1is modified by a Note the peaking factor surveillances aF2WUhbe done after t states that ( excore detectors have (i.e.. Required Action A emm rm snar umM L 6j[ M The intent of this Note is to Y'. ^* power levels, which can only be accomplished d. ,I after

                                                                                                                          ;       thehl excore core        detectors returned to power.are   calibrated to show zero tilt and the M                                                                                                     !

i If Required Actions A.1 through A are not completed within to a N00E or condition in which the requirements do not apply. fg W To achieve this status. TERNAL PGER sust be _ reduced toJD50% RTP within 4 hours. The allowed Completion sime of 4 hours is reasonable, based on operating experience {g~} regarding the amount of time required to reach the reduced v power level without challenging plant systems, SURVEILLANCE SR 3.2.4.1 REQUIRENENT3

                                                              + <ee                           f SS '

f SR 3.2.4.1 is modified by gT .tes' ' Note I allows.QPTR lated with three power range. channels 6f TE to w n T s s /M A IP andithe input from one Power Range raux cnannel is i rable. Note 2 allows (s h. performance of SR 3.2.4.2 n te 4.2.4.1 i (( fe,. 04 % 5 U 9W l m inputfrom Power)thnge Neut inoperabre.i4 N ux channels re th)

This Surveillance verifies that the QPIR as indicated by the Nuclear Instrumentation System (NIS) excore channels is (continued) M B 3.2 47 cns Rev 1. 04/07/95

r 1

I McGuire & Catawba improved TS Review Comments ITS Section 3.2, Power Distribution Limits 3.2.4-06 DOC A.1 CTS 4.2.4.2 ITS SR 3.2.4.2 CTS 4.2.4.2 provides requirements for monitoring the OPTR with one Power Range channel' inoperable. The Note for corresponding ITS SR 3.2.4.2 addresses the Condition of one or more Power Range Neutron Flux channels inoperable. This proposed change has been categorized as administrative. Comment: This proposed change is less restrictive. Is the phrase "one or more" even applicable; if two Power Range channel are inoperable thn plant should trip? Revise the submittal to provide the appropriate justification for the proposed change. DEC Response: CTS 4.2.4.1 requires a determination of OPTR once per 7 days. Should a power range channel input be inoperable above 75% RTP, then CTS 4.2.4.2 is applicable and requires OPTR be determined using the incore detectors. The CTS does not explicitly require the incore detectors to be used if more than one power range input to OPTR is inoperable. However, it also does not preclude the use of incore detectors to determine OPTR in this case. This method of determining QPTR would be utilized since the LCO remains applicable O d and is how the CTS is interpreted. Since this method of determining QPTR with more than one power range input inoperable is not precluded, the ITS conversion considers this a format only change and is neither more nor less restrictive. A plant trip would not occur if only the inputs to OPTR from the power range channels were Inoperable. If the entire power range channel were inoperable, a plant trip would occur. l mc3_cr_.3.2 3.2-16 March 2,1998

McGuire & Catawba improved TS Review Comments l ITS Section 3.2, Power Distribution Limits [ 3.2.4-07 Bases JFD 2 Bases Background discussion for ITS 3.2.4, page B 3.2-43. An editorial difference with the Bases Background discussion for STS 3.2.4 has been proposed in the Bases Background discussion for corresponding ITS 3.2.4. Comment: This is not a justifiable plant specific difference that does not enhance understanding. Revise the submittal to conform to the STS. DEC Response: The submittalis revised to conform to the STS. I O V g 9 h

QPTR B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR) BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation. The power drensity at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, " AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, and LCO 3.1.G., " Control Rod Insertion Limits," provide limits on process variables that characterize and control the three l dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses. O APPLICABLE This LC0 precludes core power distributions that violate d SAFETY ANALYSES the following fuel design criteria:

a. During a large break loss of coolant accident, the peak cladding temperature must not exceed 2200"F (Ref. 1);

b. The DNBR calculated for the hottest fuel rod in the core must be above the approved DNBR limit. (TheLC0 alone is not sufficient 'o preclude DNB criteria violations for certain accidents, i.e., accidents in

! which the event itself changes the core power distribution. For these events,, additional checks are l made in the core reload design process against the ! permissible statepoint power distributions.) i

c. During an ejected rod accident, the energy deposition I to the fuel must not exceed 280 cal /gm (Ref. 2); and

d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).

(continued) b

   ' Catawba Unit 1                       B 3.2-29                       Supplement 1 l

QPTR B 3.2.4 B 3.2 -POWER DISTRIBUTION LIMITS O ' B 3.2.4- QUADRANT POWER TILT RATIO'(QPTR) : BASES i BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution 4 measurements are made during startup testing, after ) refueling, and periodically during power operation. The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, " AXIAL FLUX DIFFF'lENCE (AFD)," LCO 3.2.4, and LCO 3.1.6, " Control Rod Ir. ertion Limits " provide limits on process variables that characterize and control the three l dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bourds used in the safety analyses. APPLICABLE This LCO precludes core power distributions that violate O SAFETY ANALYSES the following fuel design criteria:

a. During a large break loss of coolant accident, the :

peak cladding temperature must not exceed 2200"F (Ref. 1);

b. The DNBR eCoulated for the hotter *uel rod in the core must be above the approved DL

jimit. (The LC0 alone is not sufficient to preclude UNB criteria l violations for certain accidents, i.e., accidents in which the event itself changes the core power distribution. For these events, additional checks are made in the core reload design process against the-permissible statepoint power distributions.);

c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal /gm (Ref. 2); and

d. The control rods must be capable of shutting down the reactor with a ninimum required SDM with the highest worth control r :d stuck fully withdrawn (Ref. 3).

l (continued) Catawba Unit 2 B 3.2-29 Supplement 1 l

QPTR 8 3.2.4 8 3.2 POWER DISTRIBUTION LIMITS V B 3.2.4 QUADRANT POWER TILT RATIO (QPTR) BASES BACKGROUND The OPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling and periodically during power operation. The power density at any point in the core must be limited so tnat the fuel design criteria are maintained. Together. LCO LCO 3 3.2.3. ' AXIAL FLUX OlFFERENCE (AFD).* LCO 3.2.4. and h prvces.17 " Control Rod Insertion Limits.* provide limits orL571T 5-s variables that characterize and canarW4me sneee ' dimensional power distribution of the reactor core. Control g, t of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses. APPLICABLE SAFETY ANALYSES This LCO precludes core power distributions that violate the following fuel design criteria:

a. During a large break loss of coolant accident. the

,4 peak cladding temperature must not exceed 2200'F (Ref.1); . 5

'                      b.     ,Ouring a lo of forced rea   r coolant f1     acciden there must        at least     probability at y                                                                             95t k@-               confiden level (the 95 mili      (DNB) criteri departure f     nucleate that the hot f -l rod in theJ I

ore s not experi , e a DMI conditi -

c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal /gm (Ref. 2); and d. The control rods must be capable of shutting down the ! reactor with a minimum required SDN with the highest worth control rod stuck fully withdrawn (Ref. 3). The LCO limits on the AFD the QPTR the Heat Flux Hot Channel Factor (Fog)). the Nuclear Enthalpy Rise Hot - 1 l

(continued) WOG STS

     '                                        B 3.2 43                       Rev 1. 04/07/95 CIn

McGuire & Catawba improved TS Review Comments

ITS Section 3.2, Power Distribution Limits 3.2.4-08 L.6 Bases JFD 5 ,

Bases discussion for Action A.1 for ITS 3.2.4, page B 3.2-44. f I An " editorial" difference with the Bases discussion for Action A.1 of STS 3.2.4 has been I proposed in the Bases discussion for Action A.1 of corresponding ITS 3.2.4. Comment: This is neither an editorial change nor a Justifiable plant specific difference. The addition of the word "from" changes the meaning, making it wrong. Revise the submittal to conform to the i STS. DEC Response: { l Duke Energy disagrees that the change is not editorial. STS Action A.1 clearly states to f reduce THERMAL POWER z 3% from RTP...The ITS Bases are only revised to match the associated action statement. i O mc3_cr_3.2 3.2-18 March 2,1998

McGuire & Catawba improved TS Review Comments l iTS Section 3.2, Power Distrbution Limits 3.2.4 10 (Catawba only) DOC M.7 CTS 4.2.4.1.c ITS SR 4.2.4.1 Note 3 CTS 4.2.4.1.c states that the provisions of Specification 4.0.4 are not applicable. Note 3 of corresponding ITS SR 4.2.4.1 states that the SR is not required to be performed until 12 hours after exceeding 50 % RTP. DOC M.7 does not provide a specific justification for the 12 hour allowance. The change is not more restrictive since the note does not place any restriction at all on plant operation. How is the proposed change consistent with the STS? Comment: Revise the submittal to provide the justification for the proposed change or revise the submittal to conform with the STS. DEC Response: The Mode of Applicability of CTS 3.2.4 is Mode 1 above 50% RTP. The 4.0.4 exception in the CTS would allow power to be increased above the mode of applicability prior to performing CTS 4.2.4.1. The CTS would allow power to be increased to RTP and the surveillance would not have to be performed until the first scheduled frequency, i.e.,7 days. The proposed note in ITS 3.2.4.1 maintains the current 4.0.4 alloviance to increase power above the mode of applicability without performing the SR, however, it places a limit on how long the unit may operate before the SR must be met (12 hours). This requirement is much more restrictive than the current non specific 4.0.4 allowance. The time limit of 12 hours was O selected based on the time required for radial power distributions to dampen out after 50% power is reached. Radial power distributions at 50% RTP are considerably different than radial power at 100% RTP. A minimum of 8 hours is required at typicalloading rates (10% per hour between 50% and 90% RTP and less than 3% per hour above 90% RTP) to increase power from 50% to 100% RTP, leaving 4 hours to perform the required surveillance. DOC M.7 has been revised to provide a justification for the 12 hours and to remove the statement that the change is consistent with the NUREG. l t mc3_cr_3.2 3.2 20 March 2,1998

1 Discussien of Changes Section 3.2 - Power Distribution Limits i TECHNICAL CHANGES - MORE RESTRICTIVE H.1 If the actions of CTS 3.2.2 could not be completed in the required time, on entry into LC0 3.0.3 would be required and a shutdown to q MODE 2 in 7 hours would become applicable. ITS 3.7. 1 provides a specific action for this condition which requires placing the unit in MODE 2 in 6 hours. This addition is more restrictive because it requires the plant to be placed in MODE 2 within 6 hours, whereas CTS LC0 3.0.3 would allow an additional hour before l starting to shutdown to MODE 2 within 6 hours. This change is consistent with NUREG-1431. M.2 CTS 4.2.2.2.b.2 and 4.2.3.2.b.1 require verification of hot channel factors after reaching equilibrium conditions following power changes greater than 10% from when these factors were last determined. ITS SR 3.2.1.1, 3.2.1.2, 3.2.1.3, 3.2.2.1, and 3.2.2.2 require this verification within 12 hours after reaching equilibrium conditions. This addition is more restrictive since the CTS does not currently address how long the unit may operate at equilibrium conditions before the SR is completed. This change is consistent with NUREG-1431. V M.3 Not used. M.4 Not used. M.5 The QPTR limit of 1.09 in CTS 3.2.4 action a has been deleted. ITS 3.2.4 bases all required actions on QPTR in excess of 1.02, rather providing increased actions at increasing intervals of QPTP. Therefore, this change is consioered slightly more restrictive. This change is consistent with NUREG-1431. Other less restrictive changes are discussed in these Discussion of Changes. M.6 Not used. M.7 The CTS 4.2.4.1.c exception to Specification 4.0.4 has been replaced. ITS SR 3.2.4.1 includes a note which allows up to 12 hours to perform the surveillance after exceeding 50% RTP. This change is considered more restrictive since the CTS would allow up to 7 days prior to performing the SR. Thi; ch:ng: i: con:iztent

          =ith "J"EC 1431. The proposed note in ITS 3.2.4.1 maintains the O          current 4.0.4 allowance to increase power above the mode of

\ opplicability without performing the SR, however, it places a l Catawba Units 1 and 2 Page M - la Supplement 15/20/97

[ 7 l 1 Discussicn of Changes , l- I i Section 3.2 - Power Distribution Limits A b TECHNICAL CHANGES - MORE RESTRICTIVE limit on how long the unit may operate before the SR must be met (12 hours). The time limit of 12 hours was selected based on the time required for radial power distributions to dampen out after 50% power is reached. Radial power distributions ct 50% RTP are considerably different than radial power at 100% RTP. A minimum of 8 hours is required at typical loading rates (10% per hour between 50% and 90% RTP and less than 3% per hour above 90% RTP) to increase power from 50% to 100% RTP, leaving 4 hours to perform the required survetllance. M.8 The CTS 4.2.4.1.b frequency of once per 12 hours has been revised to once within 12 hours and every 12 hours thereafter. This change, retained in ITS SR 3.2.4.1, is slightly more restrictive since the 25% frequency extension does not apply to "once within" ; frequencies whereas the extension would apply to the CTS. This change is consistent with NUREG-1431. 'G 0 . l l I l }O I Catawba Units 1 and 2 Page M - 23 Svoplement 15/20/07l

McGuire & Catawba Improved TS Review Comments ITS Section 3.2, Power Distribution Limits Editorial Comments < 3.2-01 ITS 3.2.4 Bases insert page B 3.2-45 ITS SRs 3.2.2.2 and 3.2.1.3 Bases insert page 8 3.2-19 Bases discussion of Required Action A.4 of ITS 3.2.4, insert page B 3.2-45. The end of the second sentence refers to "during unanticipated transients". This seems awkward and extraneous. Suggest that it be deleted. Bases discussion of ITS SRs 3.2.2.2 and 3.2.1.3, insert page B 3.2-19. On the next to last line, "with

2%" should be changed to "within
2%".

DEC Response: ! The suggested changes have been incorporated into the ITS submittal. l mc3_cr_3.2 3.2-21 March 2,1998 l

Fo(X,Y , Z) B 3.2.1 BASES O V SURVEILLANCE SR 3.2.1.2 and 3.2.1.3 (continued) REQUIREMENTS is not required to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending. Also, extrapolation of F$(X,Y,Z) limits are not valid for core locations that were previously l rodded, or for core locations that were previously within

                     *2% of the core height about the demand position of the rod tip.

Fo(X,Y,Z) is verified at power levels :t 10% RTP above the THERMAL POWER of its last verification,12 hours after achieving equilibrium conditions to ensure that Fa(X,Y,Z) is within its limit at higher power levels. The Surveillance Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup. The Surveillance may be done more frequently if required by the results of Fn(X,Y,Z) evaluations. The Frequency of 31 EFPD is adequate to monitor the change ; of power distribution because such a change is sufficiently s'ow, when the plant is operated in accordance with the TS, (' to preclude adverse peaking factors between 31 day surveillances. l REFERENCES 1. 10 CFR 50.46, 1974.

2. UFSAR Section 15.4.8.

3. 10 CFR 50, Appendix A, GDC 26.

4. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

5. DPC-NE-2011PA " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors", March 1990. i 1

l t

%J l Catawba Unit 1                 7 8 3.2-12                         Supplement 1    >

QPTR B 3.2.4 BASES O ACTIONS M If QPTR exceeds a value of 1.02, the Power Range Neutron Flux-High trip setpoint is reduced by 3% for each 1% QPTR exceeds 1.02. Lowering this setpoint maintains the same margin to trip by limiting the transient response of the l core. The 72 hour Completion Time is sufficient for this activity to be performed and is acceptable based on the low probability of a transient occurring in this time frame. M Although Fi g(X,Y) and Fo(X,Y,Z) are of primary importance as initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is (' , accomplished by examining the incore power distribution. s Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure that, before increasing THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2, the reactor core conditions are consistent with the assumptions in the safety analyses. M If the QPTR has exceeded the 1.02 limit and a re-evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are recalibrated to show a zero QPT prior to increasing THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2. This is done to detect any subsequent significant changes in QPTR. (continued) U l Catawba Unit 1 B 3.2-32 Supplement 1

Fo(X,Y,Z) BASES O SURVEILLANCE SR 3.2.1.2 and 3.2.1.3 (continued) REQUIREMENTS is not required to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending. Also, extrapolation of F5(X,Y,Z) limits are not valid for core locations that were previously l rodded, or for core locations that were previously within 12% of the core height about the demand position of the rod tip. Fo(X,Y,Z) is verified at power levels :t 10% RTP above the THERMAL POWER of its last verification,12 hours after achieving equilibrium conditions to ensure that Fo(X,Y,Z) is within its limit at higher power levels. The Surveillance Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup. The Surveillance may be done more frequently if required by the results of Fo(X,Y,Z) evaluations. The Frequency of 31 EFPD is adequate to monitor the change of power distribution because such a change is sufficiently slow, when the plant is operated in accordance with the TS, to preclude adverse peaking factors between 31 day ( surveillances. REFERENCES 1. 10 CFR 50.46, 1974.

2. UFSAR Section 15.4.8.

3. 10 CFR 50, Appendix A, GDC 26.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

5. DPC-NE-2011PA " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors", March 1990.

O ! l Catawba Unit 2 B 3.2-12 Supplement 1

QPTR B 3.2.4

BASES ACTIONS M $ If QPTR exceeds a value of 1.02, the Power Range Neutron Flux-High trip setpoint is reduced by 3% for each 1% QPTR exceeds 1.02. Lowering this setpoint maintains the same margin to trip by limiting the transient response of the - l core. The 72 hour Completion Time is sufficient for this activity to be performed and is ceptable based on the low probability of a transient occuri ing in this time frame. M Although F3 s(X,Y) and Fo(X,Y,Z) are of primary importance as initial conditions in the safety analyses, other changes in l the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution ccn affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is

  ;                  accomplished by examining the incore power distribution.
>                    Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure that, before increasing THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2, the reactor core conditions are consistent with the assumptions in the safety analyses.

M If the QPTR has exceeded the 1.02 limit and a re-evaluation i of the safety analysis is completed and shows that safety requirements are met, tne excore detectors are recalibrated to show a zero QPT prior to increasing THERMAL POWER to i above the more restrictive limit of Required Action A.1 or A.2. This is done to detect any subsequent significant changes in QPTR. (continued) p(_) l Catawba Unit 2 B 3.2-32 Supplement I l 1

INSERT 8

     $(X,Y,Z) is not required to be extrapolat:d for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending. Also, extrapolation of R(X,Y,Z) limits are not valid for core locations that were l previously rodded, or for core locations that were previously within 2% of the core height about the demand position of the rod tip.

1 0 - l l l l INSERT Page B 3.2-19 O u Catawba

1 f-INSERT A.4 If QPTR exceeds a value of 1.02, the Power Range Neutron Flux-High trip setpoint is reduced by 3% for each 1% QETR exceeds 1.02. Lowering this setpoint maintains the same margin to trip l by limiting the transient response of the core 6.::;; :: &ip:::d *.:=i::::. The 72 hour Completion Time is sufficient for this activity to be performed and is acceptable based on the low probability of a transient occurring in this time frame. INSERT Page B 3.2-45 O Catawba

O l ENCLOSURE 3 SUPPLEMENTAL CHANGES TO ITS SUBMITTAL O l l 1 O

Issue Number l 3] Affected Section 13.4.15 RCS Leokoge Detection l ANected Units CNS: MNS: l Yes) Affected Pogos ITS: ITS: ITS Boses: B 3 4-84 ITS Bones: B 3.445 CTS: CTS: DOCS: DOCS: NRG: NRG: NRG Bones: B 3.446 NRG Bones: B 3.4-86

                       #D:                                                 #D:

NSHC: NSHC: Desenpr.cn ITS 3.4.15 Bases Background, Porograph 4, line 3: Change ' Level would indicate..

to 'An abnormd level increase would indicate .' to occurately represent the concern. Also, for Cotowbct correct tre ronge of tre porticulate monitor.

l 'N l i i l 4 I

1 RCS Leakage Detection Instrumentation l B 3.4.15 J B 3.4. REACTOR COOLANT SYSTEM (RCS) B 3.4.15 RCS Leakage Detection Instrumentation BASES BACKGROUND. GDC 30 of Appendix A to 10 CFR 50 (Ref. 1) requires means for detecting and, .to the extent practical, idertifying the l location of the source of RCS LEAKAGE. Regul:nory l Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

Leakage detection systems must have the capability to detect ! significant reactor coolant pressure boundary (RCPB) i degradation as soon after occurrence as practical to minimize the potential for propastion to a gross failure. ) j Thus, an early indication or warning signal is necessary to j permit proper evaluation of all unidentified LEAKAGE. j The primary method of detecting leakage into the Containment is measurement of the Containment floor and equipment sump level. There are small sumps located on either side of the containment outside the crane wall. Any leakage would. fall to the floor inside the crane wall and run by a sump. drain O - line to one of the two sumps. Any leakage outside the crane wall would fall.to the floor and gravity drain to these sumps. The sump level rate of change, as calculated by the plant computer, would indicate the leakage rate. This method of detection would indicate in the Control Room a water leak from either the Reactor Coolant system or the , Main Steam and Feedwater Systems. A 1 gpm leak (cumulative I in both sump A and B) is detectable in 1 hour. The containment ventilation unit condensate drain tank level , change offers another means of detecting leakage into the l l containment. An abnormal level increase would indicate ' removal of moisture from the containment by the containment air coolers. The plant computer calculates the rate of change in level to detect a leak of 1 gpm. The reactor coolant contains radioactivity that, when I released to the containment, can be detected by radiation monitoring instrumentation. Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects. (continued) l Catawba Unit 1 B 3.4-84 Supplement 1

RCS Leakage Detection Instrumentation B 3.4.15 BASES BACKGROUND Instrument sensitivities of 10-10 l (continued) particulate monitoring and of 10 pCi/ccCi/cc radioactivity radioactivityfor for gaseous monitoring are practical for these leakage detection systems. Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS LEAKAGE. An increase in humidity of the containment atmosphere would indicate release of water vapor to the containment. Dew point temperature measurements can thus be used to monitor humidity levels of the containment atmosphere as an indicator of potential RCS LEAKAGE. A 1*F increase in dew point is well within the sensitivity range of available instruments. Since the humidity level is influenced by several factors, a quantitative evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in liquid level into the containment floor and equipment sump and condensate level from air coolers. Humidity level monitoring is considered most useful as an indirect alarm or indication to alert the operator to a 3 potential problem. Humidity monitors are not required by this LCO. Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to the containment. Containment temperature and pressure fluctuate slightly i during plant operation, but a rise above the normally indicated range of values may indicate RCS leakage into the containment. The relevance of temperature and pressure , measurements are affected by containment free volume and, for temperature, detector location. Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment. Temperature and pressure monitors are not required by this LCO. 1 APPLICABLE The need to evaluate the severity of an alarm or an SAFETY ANALYSES indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary. The system response times and sensitivities are described in the UFSAR (Ref. 3). Multiple instrument I locations are utilized, if needed, to ensure that the i transport delay time of the leakage from its source to an i (continued) p%.) ~ Catawba Unit 1 B 3.4-85 Supplement 1 l

RCS Leakage D2tection Instrumentation - B 3.4.15 8 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.15 RCS Leakage Detection Instrumentation BASES BACKGROUND GDC 30 of Appendix A to 10 CFR 50 (Ref.1) requires means for detecting and, to the extent practical, identifying the location of the source of RCS LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems. Leakage detection systems must have the capability to detect significant reactor coolant pressure boundary (RCPB). degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure. Thus, an early indication or warning signal is necessary to permit proper evaluation of all unidentified LEAKAGE. The primary method of detecting leakage into the Containment i is measurement of the Containment floor and equipment sump level. There are small sumps located on either side of the containment outside the crane wall. Any leakage would fall , to the floor inside the crane wall and run by a sump drain line to one of the two sumps. Any leakage outside the crane O. wall would fall to the floor and gravity drain to these sumps. The sump level rate of change, as calculated by the plant comouter, would indicate the leakage rate. This method of detection would indicate in the Control Room a water leak from either the Reactor Coolant system or the Main Steam and Feedwater Systems. A 1 gpm leak (cumulative in both sump A and B)'is detectable in I hour. The containment ventilation unit condensate drain -tank level change offers another means of detecting leakage into the l containment. An abnormal level increase would indicate ) removal of moisture from the containment by the containment air coolers. -The plant computer calculates the rate of change in level to detect a leak of 1 gpm. The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation. Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been. formed and fission products appear from fuel eiement cladding contamination or cladding defects. (continued) l Catawba Unit 2 8 3.4-84 Supplement 1

l. ___-__-_ _ _ _ _ _ _ _ _ _ _ _ .

RCS Leakage Detection Instrumentaticn B 3.4.15 BASES BACKGROUND Instrument sensitivities of 1040 l (continued) particulate monitoring and of 10 pCi/ccpCi/cc radioactivity radioactivity for for gaseous monitoring are practical for these leakage detection systems. Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS LEAKAGE. An increase in humidity of the containment atmosphere would indicate release of water vapor to the containment. Dew point temperature measuraments can thus be used to monitor humidity levels of the containment atmosphere as an indicator of potential RCS LEAKAGE. A l'F increase in dew point is well within the sensitivity range of available instruments. Since the humidity level is influenced by several factors, a quantitative evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in liquid level into the containment floor and equipment sump and condensate level from air coolers. Humidity level monitoring is considered most useful as an indirect alarm or indication to alert the operator to a potential problem. Humidity monitors are not required by O this LCO. Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to the containment. Containment temperature and pressu5e fluctuate slightly during plant operation, but a rist above the normally indicated range of values may ind'.c.te RCS leakage into the containment. The relevance of temperature and pressure measurements are affected by containment free volume and, for temperature, detector location. Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment. Temperature and pressure monitors are not required by this LCO. APPLICABLE The need to evaluate the severity of an alarm or an SAFETY ANALYSES indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary. The system response times and sensitivities are described in the UFSAR (Ref. 3). Multiple instrument locations are utilized, if needed, to ensure that the transport delay time of the leakage from its source to an (continued) l' Catawba Unit 2 B 3.4-85 Supplement 1 l

RCS Leakage Detection Instrumentation B 3.4.15 B 3.4 REACTOR C00UWT SYSTEN (RCS) B 3.4.15 RCS Leakage Detection Instrumentation 8ASES BACIGOUW GDC 30 of Appendix A to 10 CFR 50 (Ref.1) requires m location of the source of IICS LEAbE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for ' selecting leakage detection systems. . significant reactor coolant pressure boundary (RCPS) Lea degradation as soon after occurrence as practical to minimize the potential for. przpagetion to a gross failurt. Thus. / early indication or warning signal is necessary to , permit proper evaluation of all unidentified tJAKAE. Industry practi to 1.0 gpa can shown that water flow readily detected in ofD Mk sonitoring operating f collect in water level. in y of a . The ined Muses by ' rate, or in the Deent dentified 1si( ,)'and t*? sed i for flow rate moaf . ases of 0.5 to 1.0 gpa[a instrusierited to al the noriaal flow r sensitivity is acceptabt or detecting . identified IIAKAE. 1 in J

  "                                         The reactor coolant contains radioactivity that, when                                                     '

released to monitoring the containment, can be detected by radiation instrumentatica. Reactor coolant radioactivity levels will be low during initial reactor starty and for a few weeks thereafter, until activated ccerosion products have been formed and fission products appear frus' fuel

    -                                   element cladding contaminationar claddina defer +=.

Instrument sensitivities of low'gc1/cc radioactivity form 7 particulate monitoring and of Ing,1/cc racioactivity for gaseous monitoring are practical for these leakage detection systems. g/9 Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to llCS LEAKAE. An increase in hueidity of the containment atmosphere would indicate release of water vapor to the containment. Dew point temperature seasurements can thus be used to monitor nueidity levels of the containment atmosphere as an (continued) m 8 3.4 86 Rev 1, 04/07/95 s NO 1

q INSERT 15 [ The primary method of detecting leakage into the Containment is measurement of the Containment floor and equipment sump level. There are small sumps located on either side of the containment outside the cran.e wall. Any leakage would fall to the floor inside the crane wall and run by a sump drain line to one of the two sumps. Any leakage outside the crane wall would fall to the floor and gravity drain to these sumps. The sump level rate of change, as calculated by the plant computer, would indicate the leakage rate. This method of detection would indicate in the Control Room a water leak from either the Reactor Coolant System or the Main Steam and Feedwater Systems. A 1 gpm leak (cumulative in both sump A and B) is detectable in 1 hour. The containment ventilation unit condensate drain tank level change offers another means of detecting leakage into the containment. An abnormal level increase would indicate removal of moisture from the containment by the containment air coolers. The plant computer calculates the rate of change in level to detect a leak of 1 gpm. O INSERT Page B 3.4-86 )

losue Number l 6l. Affected Section 13.0.6 ] Affected Units CNS: l Yes) MNS: l Yes! Affected Pogos ITS: 3.0-2 ITS: 3.0-2 ITS Boeos: ITS Boeos: CTS: CTS: DOCS: DOCS: NRG: NRG: NRG Bones: NRG Boess: JFD: jpp: NSHC: NSHC: Descripflon LCO 3.0.6. patograph 2, change 'Wh en' to "When'. Add misang line between LCO 3.0.6 ond 3.0.7. I

LCO Applicability 3.0

  <-    3.0 LCO APPLICABILITY LCO 3.0.4        Specification shall not prevent changes in MODES or other (continued)   specified conditions in the Applicability that are required to comply with ACTIONS.                                        j Exceptions to this Specification are stated in the individual Specifications.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the required testing to demonstrate OPERABILITY.

LC0 3.0.6 When a supported system LCO is not met solely due to a support system LC0 not being met, the Conditions and Required Actions associated with this supported system are

  /                      not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LC0 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.5.15, " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered.

l When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered 4 in accordance with LC0 3.0.2. I LC0 3.0.7 Test Exception LCOs 3.1.8 and 3.4.17 allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements ramain I (continued) V l Catawba Unit 1 3.0-2 Supplement 1

LCO Applicability 3.0 m 3.0 LCO APPLICABILITY V LCO 3.0.4 Specification shall not prevent changes in MODES or other (continued) specified conditions in the Applicability that are required to comply with ACTIONS. Exceptions to this Specification are stated in the individual Specifications. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative cor. trol solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LC0 3.0.2 for the system returned to service under administrative control to perform the required testing to demonstrate OPERABILITY. , LCO 3.0.6 When a supported system LCO is not met solely due to a support system LC0 not being met, the Conditions and Required Actions associated with this supported system are not required to be ent'ered. Only the support system LC0 O , ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and lin'itations may be required in accordance with Specification 5.5.15, " Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss , of safety function exists are required to be entered. l l When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2. I LC0 3.0.7 Test Exception LCOs 3.1.8 and 3.4.17 allow specified Technical Specification (TS) requirements to be changed to l permit performance of special tests and operations. Unless ; otherwise specified, all other TS requirements remain (continued) l Catawba Unit 2 3.0-2 Supplement 1

issue Number l 8} ANected Section l5.2.1 Onsrte and Offsite Organizations 1 A#ected Units CNS: l Yesl MNS: l Yesj ANected Pogos ITS: 5.0-2 ITS: 5.0-2 ITS Bases: ITS Bases: ) CTS: 61 CTS: 6-1 DOCS: DOCS: NRG: 502 NRG: 5.0-2 l NRG Boeos: NRG Bases: l JFD: JFD: ! NSHC: NSHC: Descripilon The correct tme in 5.2.1.d is ' Executive Vice President Nuclear Generation Department' rcrher than Senior Vice President. i 1

l l i

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2_ Organization 5.2.1 Onsite and Offsite Oraanizations Onsite and offsite organizations shall be. established for unit operation and corporate management, respectively. The onsite and affsite organizations shall include the positions for activities eifecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and consnunication shall be defined and established throughout highest manage:nent levels, intennediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departn' ental responsibilitiet and relationships, and job descriptions.for key personnel positions, or in equivalent forms of documer.tation. These requirements-shall be documented in the UFSAR;

b. The Station Manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant; - -

c. The Vice President of Catawba Nuclear Site shall have corporate responsibility for overall plant nuclear. safety and shall take any_ measures needed to ensure acceptable :

performance of the staff in operating, maintaining, and I providing technical support to the plant to ensure nuclear safety; I d. The Executive Vice President Nuclear Generation Department will be the Senior Nuclear Executive and have corporate responsibility for overall nuclear safety; and

e. The individuals who train the operating staff, carry out radiation protection, or perform quality assurance fo:tions may report to the appropriate onsite manager; howev9r, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

(continued) O 5.0-2 Supplement 1

 -l Catawba _ Unit 1 I

I

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2~ Organization 5.2.1- Onsite and Offsite Orannizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels. intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the UFSAR;

b. The Station Manager shall be responsible for overall safe operation of the plant and shall have control over those onsite activities necessary for safe operation and maintenance of the plant;

c. The Vice President of Catawba Nuclear Site shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety; l d. The Executive Vice President Nuclear Generation Department will be the Senior Nuclear Executive and have corporate responsibility for overall nuclear safety; and

e. The individuals who train the operating staff, carry out radiation protection, or perform quality assurance functions may report to the appropriate onsite manager; however, these ;

individuals shall have sufficient organizational freedom to ensure their independence from operating pressures. 1 l (continued) l Catawba Unit 2 5.0-2 Supplement 1 l

Spe6kt.k 5 5. S.O ADMINISTRATIVE CONTROLS OA.I $7.1 RESPONSIBILITY . f'f.1.1 The Station Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6#.1.2 The Shift Supervisor der a6on- --a.or t ram ttv enn% +aa _-g aren onueo umi v i nua a shall be responsible for the control room connand func-

  .Dh        u un.o mana g/(CatawbaNuc1              4 .   . . . m mu      eu , signea  o sne nce rr n oe Sitesha[mbereissued[eo all Catawb        uclear Sit    ersonnel un an annua       ast s.r f A.
          $~4. 2 ORGANIZATION
         $ A.2.1 TQF[SITQ AND T  QNS,1T,D ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a .- Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the fom of organization charts, functional descrip-tions of departinental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent foms of documentation. These requirements shall be documented in the SAR.

b. The Station Manager shall be responsible for overall unit safe O operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c. The Vice President of Catawba Nuclear Site shall have responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintain-ing, and providing technical support to the plant to ensure nuclear safety. -

d.

      @..              The Senior Vice President Nuclear Generation Department will be the clear Executive and have corporate responsibility for over-   /

all nuclear safety.

e. The individuals who train the operating staff and those who carry out radiation protection and quality assurance functions may report to the appropriate onsite manac r; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

CATAWBA - UNIT 1 6-1 Amendment No. 148 O

                                                                               ,,au

Spe.f% % .s.O

r. 0 STRATIVE CONTROLS g,) 4.1 RESPONSIBILITY IA.1.1 The Station Manager shall be re$ponsible for overall unit operation and I

shall delegate in writing the succession to this responsibility during his absence. A1 jf.1.2 The Shift Sumrvisordor durghis absenc(from the conefoi room, aJ N (estshateda nt vidua 3 shall be responsible for the control room comand func-

  < f *y    - Nr b-t / tion. AfA           ag=- uirective o tnis e Tect, signe oy tne v1 e Presiaent r CataWDa          ear    e shall b reissued to all Cat      a Nuclea   ite personnel

( on an a al ba s.

                 #'t.2 ORGANIZATION                                                            \    L,A , l OA'l                          '

54'.2.1 (dfiSi!Yh*AND h0RGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and comunication shall be established and defined for the highest management levels through '

intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descrip-tions of departmental responsiwilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of A,l documentat' ion. Theserequ'irementsshallbedocumentedinthagSAR.

b. The Station Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c. The Vice President of Catawba Nuclear Site shall have responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintain-ing, and providing technical support to the plant to ensure nuclear safety.

d. The Vice President Nuclear Generation Department will be the !

Senior Nuclear Executive and have corporate responsibility for over-all nuclear safety.

e. The individuals who train the operating staff and those who carry out radiation protection and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient orynizational freedom to ensure their independence from operating pressures.

CATAWBA - UNIT 2 6-1 Amendment No. 142 O fye- t :

g Organizat fy ($,kly 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and offsite Oraanizations Onsite and offsite organizations shall be established for unit operation and corporate management. respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be defined and established thrsughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requiramante tha11 he documented in TI I (t y .w.., m . 3, 9

b. The plant warW3shall be safe operation of the plst and shall ible for overall ve control over @

those onsite activities necessary for safe operation and maintenanea af the n1aa+ Tvece pum,,J. .c c./.wie w c ked fek)

c. The((a spectr1uso w.-sur escutivepsitionDshall -have corporate respons1D111ty for overati nant. nuclear safety @

and shall take an/ measures needed to ensure acceptable O'S performance of the staff in operating, maintaining, and h'd 2 providing technical support to the plant to ensure nuclear safety: @ @ Q- e /. The individuals who train the oMrating staff, carry out AlealtrohvsfB. or perform qual (ity assurance functions may rabb6 report to the appropriate onsite manager: however, these ! gbM. individuals shall have sufficient organizational freedom to ensure their independence from operating pressures. 6.2.2 Unit staff The unit staff organization shall include the following:

a. A non licensed operator shall be assigned to each reactor containing fuel and an additional non licensed operator l

i l (continued) l 1 au 5.0 2 Rev 1. 04/07/95

O !

INSERT 2

d. The Executive Vice President Nuclear Generation Department will be the Senior l Nuclear Executive and have corporate responsibility for overall nuclear safety; and O

O c ,, , , , , , ,,, , y ,,,,m ,,,,,

leeue Number l 9j Affected Section {3.3.1 RTS l Affected UnHs CNS: l Yesl MNS: l Yes) Affected Pages ITS: 3.3-7, 8 ITS: 3.3-7, 8 l ITS Bases: 8 3.3-44 ITS Bases: B 3.3-45 CTS: 3/434 CTS: 3/43-7 DOCS: DOCS: NRG: 33-8.9 NRG: 3.3-8. 9 NRG Bases: 8 3.3-48.49 NRG Bases: 8 3.3-48.49 JFD: JFD: NSHC: NSHC: ! Description CTS 3.3.1 RTS Interiocks oeftons ollows more than one chonnel to be inoperable. Revim ITS oc+1ons consistent wtth CTS. 1

1 l l

RTS Instrumentation 3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME Q. One RTB train ------------NOTES------------ inoperable. 1. One train may be bypassed for up to 2 hours for surveillance testing, provided the other train is OPERABLE.

2. One RTB may be bypassed for up to 2 hours for maintenance on undervoltage or shunt trip mechanisms, provided the other train is OPERABLE.

Q.1 Restore train to I hour OPERABLE status. B Q.2 Be in MODE 3. 7 hours R. One or more channel (s) R.1- Verify interlock is I hour l inoperable. in required state for existing unit conditions. E R.2 Be in MODE 3. 7 hours (continued) O Catawba Unit 1 3.3-7 Supplement 1 l

RTS Instrumentation 3.3.1 ACTIONS (continutd) O CONDITION REQUIRED ACTION COMPLETION TIME l S. Oneormorechannel(s) S.1 Verify interlock is I hour inoperable. in required state for existing unit conditions. E S.2 Be in MODE 2. 7 hours T. One trip mechanism T.1 Restore inoperable 48 hours inoperable for one trip mechanism to RTB. OPERABLE status. M T.2. Be in MODE 3. 54 hours g U. Two RTS trains U.1 Enter LC0 3.0.3. Inunediately 1 inoperable. O 3.3-8 Supplement 1 l Catawba Unit 1

..1

Ji RTS Instrumentation 3.3.1 ACTIONS (continued) O- CONDITION REQUIRED ACTI0fi COMPLETION TIME Q. One RTB train ---_____----NOTES------------ inoperable. 1. One train may be bypassed for up to 2 hours for surveillance testing, provided the other train is OPERABLE. l

2. One RTB may be bypassed for up to 2 hours for maintenance on undervoltage or shunt trip mechanisms, provided the other train is OPERABLE. ;

Q.1 Restore train to I hour OPERABLE status. E I Q.2 Be in MODE 3. 7 hours . 1 R. One or mere channel (s) R.1 Verify interlock is 1 hour l inoperable. in required state for existing unit j conditions. M R.2 Be in MODE 3. 7 hours (continued) O Catawba Unit 2 3.3.) Supplement 1 l

RTS Instrumentation 3.3.1 ACTIONS (continued) O. CONDITION REQUIRED ACTION COnetETION TIME l S. One or more channel (s) S.1' Verify interlock is I hour inoperable. in required state for existing unit conditions. DR S.2 Be in MODE 2. 7 hours F T. One trip mechanism T.1 Restore inoperable 48 hours inoperable for one trip mechanism to RTB. OPERABLE status. DE T.2. Be in MODE 3. 54 hours U. Two RTS trains U.1 Enter LC0 3.0.3. Immediately inoperable. l O 3.3-8 Supplement 1 l- Catawba Unit 2

RTS Instrumentation B 3.3.1 BASES O ACTIONS R.1 and R.2 (continued) Condition R applies to the P-6 and P-10 interlocks. With l one or more channel (s) inoperable for one-out-of-two or two-out-of-four coincidence logic, the associated interlock must be verified to be in its required state for the existing unit condition within 1 hour or the unit must be placed in MODE 3 within the next 6 hours. Verifying the interloc'k status, by visual observation of the control room status lights, manually accomplishes the interlock's Function. The Completion Time of 1 hour is based on operating experience and the minimum amount of time allowed for manual operator actions. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging unit systems. The 1 hour and 6 hour Completion Times are equal to the time allowed by LCO 3.0.3 for shutdown actions in the event of a complete loss of RTS Function. S.1 and S.2 Condition S applies to the P-7, P-8, P-9, and P-13 Ol ' interlocks. With one or more channel (s) inoperable for one-out-of-two or two-out-of-four coincidence logic, the associated interlock must be verified to be in its required state for the existing unit condition within 1 hour or the unit must be placed in MODE 2 within the next 6 hours. These actions are conservative for the case where power level is being raised. Verifying the interlock status, by visual observation of the control room status lights, manually accomplishes the interlock's Function. The. Completion Time of 1 hour is based on operating experience and the minimum amount of time allowed for manual operator actions. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 2 from full power in an orderly manner and without challenging unit systems. T.1 and T.2 Condition T applies to the RTB Undervoltage and Shunt Trip Mechanisms, or diverse trip features, in MODES 1 and 2. With one of the diverse trip features inoperable, it must be (continued) l Catawba Unit 1 B 3.3-44 Supplement 1

RTS Instrumentation B 3.3.1 BASES k ACTIONS T.1 and T.2 (continued) restored to an OPERABLE status within 48 hours or the unit must be placed in a MODE where the requirement does not apply. This is accomplished by placing the unit in MODE 3 within the next 6 hours (54 hours total time). With both diverse trip features inoperable, the reactor trip breaker is inoperable and condition Q is entered. The Completion ' Time of 6 hours is a reasonable time, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging unit systems. With the unit in MODE 3, the MODES 1 and 2 requirement for this function is no longer required and Condition C is entered. The affected RTB shall not be bypassed while one of the diverse features is inoperable except for the time required to perform maintenance to one of the diverse features. The allowable time for performing maintenance of the diverse features is 2 hours for the reasons stated under Condition Q. The Completion Time of 48 hours for Required Action T.1 is reasonable considering that in this Condition there is one

 ,s                  remaining diverse feature for the affected RTB, and one (d T                OPERABLE RTB capable of performing the safety function and given the low probability of an event occurring during this interval.

El With two RTS trains inoperable, no automatic capability is available to shut down the reactor, and intnediate plant shutdown in accordance with LC0 3.0.3 is required. SURVEILLANCE The SRs for each RTS Function are identified by the SRs REQUIREMENTS column of Table 3.3.1-1 for that Function. A Note has been added to the SR Table stating that Table 3.3.1-1 determines which SRs apply to which RTS Functions. Note that each channel of process protection supplies both trains of the RTS. When testing Channel I, Train A and Train B must be examined. Similarly, Train A and Train B q (continued) (/ Catawba Unit 1 B 3.3-45 Supplement 1 l

RTS Instrumentation B 3.3.1 n BASES ( SURVEILLANCE must be examined when testing Channel II, Channel III, and REQUIREMENTS Channel IV (if applicable). The CHANNEL CALIBRATION and (continued) COTS are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies. SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION. O Agreement criteria are determined by the unit staff based on Q

a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels. SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output every 24 hours. If the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is not declared inoperable, but must be adjusted. If the NIS channel output cannot be properly adjusted, the channel is declared inoperable. (continued) l Catawba Unit 1 B 3.3-46 Supplement 1

RTS Instrumentation B 3.3.1 BASES ACTIONS R.1 and R.2 (continued) Condition R applies to the P-6 and P-10 interlocks. With l one or more channel (s) inoperable for one-out-of-two or two-out-of-four coincidence logic, the associated interlock must be verified to be in its required state for the existing unit condition within 1 hour or the unit must be placed in MODE 3 within the next 6 hours. Verifying the interlock status, by visual observation of the control room status lights, manually accomplishes the interlock's Function. The Completion Time of 1 hour is based on operating experience and the minimum amount of time allowed for manual operator actions. The Completion Time of 6 heurs is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging unit systems. The 1 hour and 6 hour Completion Times are equal to the time allowed by LCO 3.0.3 for shutdown actions-in the event of a complete loss of RTS Function. S.1 and S.2 Condition S applies to the P-7, P-8, P-9, and P-13 - l interlocks. With one or more channel (s) inoperable for one-out-of-two or two-out-of-four coincidence logic, the associated interlock must be verified to be in its required state for the existing unit condition within 1 hour.or the unit must be placed in MODE 2 within the next 6 hours. These actions are conservative for the case where power i level is being raised. Verifying the interlock status, by visual observation of the control room status lights, manually accomplishes the interlock's Function. The Completion Time of 1 hour is based on operating experience and the minimum amount of time allowed for manual operator actions. -The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 2 from full power in an orderly manner and without challenging unit systems. T.1 and T.2 Condition T applies to the RTB Undervoltage and Shunt Trip Mechanisms, or diverse trip features, in MODES 1 and 2. With one of the diverse trip features inoperable, it must be (continued) O l Catawba Unit 2 B 3.3-44 Supplement 1

RTS Instrumentation B 3.3.1 BASES ACTIONS T.1 and T.2 (continued) restored to an OPERABLE status within 48 hours or the unit must be placed in a MODE where the requirement does not apply. This is accomplished by placing the unit in MODE 3 within the next 6 hours (54 hours total time). With both diverse trip features inoperable, the reactor trip breaker is inoperable and condition Q is entered. The Completion Time of 6 hours is a reasonable time, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging unit systems. With the unit in MODE 3, the MODES 1 and 2 requirement for this function is no longer required and Condition C is entered. The affected RTB shall not be bypassed while one of the diverse features is inoperable except for the time required to perform maintenance to one of the diverse features. The allowable time for performing maintenance of the diverse features is 2 hours for the reasons stated under Condition Q. The Completion Time of 48 hours for Required Action T.1 is reasonable considering that in this Condition there is one remaining diverse feature for the affected RTB, and one I OPERABLE RTB capable of performing the safety function and given the low probability of an event occurring during this interval. ll.d With two RTS trains inoperable, no automatic capability is available to shut down the reactor, and intnediate plant shutdown in accordance with LC0 3.0.3 is required. SURVEILLANCE The SRs for each RTS Function are identified by the SRs REQUIREMENTS column of Table 3.3.1-1 for that Function. A Note has been added to the SR Table stating that Table 3.3.1-1 determines which SRs apply to which RTS Functions. Note that each channel of process protection supplies both trains of the RTS. When testing Channel I, Train A and Train B must be examined. Similarly, Train A and Train B I (continued) Catawba Unit 2 B 3.3-45 Supplement 1 I l

l RTS Instrumentation B 3.3.1 BASES SURVEILLANCE must be examined when testing Channel II, Channel III, and REQUIREMENTS Channel IV (if applicable). The CHANNEL CALIBRATION and (continued) COTS are performed in a manner that is consistent with the assumptic .s used in analytically calculating the required channel accuracies. SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other ] channels. It is based on the assumption that instrument channels monitoring the same parameter should read , approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of j something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the l instrumentation continues to operate properly between each CHANNEL CALIBRATION. O Agreement criteria are determined by the unit staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. l The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LC0 required channels. SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output every 24 hours. If l the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is not declared inoperable, but must be adjusted. If the NIS channel output cannot be properly adjusted, the channel is declared inoperable. (continued) l Catawba Unit 2 B 3.3-46 Supplement 1

I!!l I{!{ 7 C O[ . p7 $ pP 7~ y a u c - [t U g h Sy O gA 2 R S g hh S. S

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                                                                                                                                           %h., s.3.1 TABLE 3.3-1 (Continued)

O CACTIC # h!TetA ACTION STATEMEllTSfcontinued) g e g ,,gt . p % ACTION f 'Wtth th umber of OPE LE cnannes e tess than tn Humber f Channels, nRTUP and/or otal Aro ded the fall ER OPERATION proceed na conditions re catitHad-Ll g

~~

The inoperable channel is placed in the tripped condition M.) within 6 hours @ N m,, \ $ L'L .84 'a maest3, go, \ (rne nure-- ch==<s Op san m.reaut r= t h ==t-tle inoperable channel say be bypassed for up to 4' hours ei . . 1 e4**'y

    '           en Gi es*5                              Y &efor survelliance aar m car _ mtesting   a x n . of other channel              t gp ACTION        -

witn tpr number ---W

                                                                                                                        @ tom en.penWM an* ia<<eama "ne Jn nr-- -

hehandi norn ffr vrt_-?"* m I: # c.t i.L res~ tore the inoperab'

                                                                                                                                                                        .I e -- - %. ,,

to OPERA 8LE status within 6. hours or he ' n at least HOT stale 8 mow

within the next 6 hours fElWEP4one - --_may be ownansaa 35 d'" A' pg-wa Q iorprovided un so e nours the other ror ance su;rventest Ms

                                                                                                                                               ~ - ^

g op st.2 oc M _J"LE. ( i. n rs ACT - M ;-- maa thF Mint-i's. 6 1 hour ___.o M nannels G 5 8 4h g thin ,,,, ' 7 sq.7 - rat =* mE t E. that the interlock 's-------_.e y ee " peet 't ,,e

                                 ^
                                           ~                 the existing plant condition n ,u,1 ;,, e.uons.0.3           in its required state fo sa=Indle.

( '" i

               ' 8'% @actromM Co.e nts 6,                ,.w
                                                        ~

[th nmHJa within 6 hours; F En" rbvrT -- - e' -- = aaa =----*G-..^

                                                                                                      . i ^__ be in at least HOT STannay Manws @

m h(appeeei2noursrorsurveiiiancetesti-..yxie enanjne aay se oypassed for up to ,&*,'"j"' ) r73 EAa g,. provided the other channel is 0

-~

n coron *.an . a w , oc ; ' j 9da.3a. a ch.- 4J11

                      =L            ACTION          -

itn th 2r of u j {g'- ,, k=-- g< oornamer p- m- in aneeess snan symnMit g y . f 5 restore the inoperanie channel 1 i to OPERABLE status within 48 hours or open the Reactor trio A8 g g.j"asy

                                         ,T"g) breakers within the next hourg. ;,.r* q7+1ucuae 96 4 Cha 3                              ACTION h                          in ; m r or vrmootE cnannel O. L Aak.a wdam.,                                                                                                      ess than tt# Total JediEe'h f eb yIs. oseratk6n say conti

=adadrthi inoperable f**** 4* 'M

            '"                                              channels are placed in the Irlpped condition within 6 hours.
                         'r ACTION                 -

With one of the diverse trip features (Uttdervoltage or shunt trip attachment) inoperable, restore it to OPERA 8LE status M go f98 within_48 hours ordieclap.nu orpeter inograble an64ppTjl> %g,,e 3 ima.mbe.pne o er sna I n se oypassed yntle one of w e4+=19 \ me.tTE iversytrip fe res is i rable except for the time gq 1 Iae*yNZ 'ged for wete t./ L a -[-_ _b_ _ _, _ _ 4 PF8h*'r etat .I otsina mal maanCe to restare the hre er_ t With th(bronar bynassdd. apply ACT/f0N 9.] g mu u - h any reactor tripft,. ass breaker inoperable / restore tne ypass breaker to OyLE status prior to pf1ing it in service- . CATAW8A - UNIT 3/4 3-6 Amendment No. 148 Q hcwon t ; o e d rpe.i.,am. res

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                                                                                                                                                                                                                        "-84 b%                3. 3.1          -

TABLE 3.3-1 (Continued 1 g ACTIONSTATENERIS'lContinued) (ACT10 g - AfeletD ^8. Mad la'#"' ACTION @- {NWith theChannels. Per ofSTA OPERA 8 channels P and/or POW one ess than the otal dumber rovi d the following onditions ar atisfied: OPERATION proceed E @

                                                                    ~~d                   The inoperable channel is placed in the tripped condition g                                                                         within 6 hours,@

f4ess Tb Q ghe nistnum Channars OPraan' F r&&Gtramaat

              .98.,                                   ACU*W                                                                                                                           i samen - _ .r .3 E .'      6a*^                                                            fthe InoperasTe channel may be bypassed for up to 4 hours T popg g i. G h*wv.                      E.L.N##877                         (for sumaillance test                     other channel g                                                      j Al           _f ucatfhe 4 3/1.I ACTIONS -                lW1th"ynumber t            o_ fpr.zABLE Le              tr'
                                                                                            . s OPE"^^' vr =- I                 heelsAne less tham.he mW tJewstori sne 1noperas                                            aw               .

j g 7, to OPERASLE within the next 6 hours; status - within JBne -- 6 Tiours or be in at least-NOT STALES

              ~l
               ~

jger,,,, p /for up to 4 .-. . ror m.. wi _ -- be r meu i y. , very esti [ts.Mr.n provided the C is . . 4.2. . &. ,a Aor.5 ACTION -

in 7 % r ne y a= a a cas

                                                                                         =1                                                                                                =----           cAa =ae4, f,,1, %                                                                                  efalt that the Taterlock is in its required state fo '7,

the existing plant condition,y r .

S14'" 10N O Dith t number of C r.t en 8"

                   "' "7(o,s. era w i enatha                                                    OPERABLE d a=4 -

1 g wtcnin a r.e n : -- -Y ~'s ha

                                                                                                                - 2rone enannel may e nypasses for_up tr 4 *watnets r$ded             e$ ?-                         ?-                                                                    k'

foa. srrn 871 ACTION9Gs.

a

                    '                                                             ten the
                    * *$ $                              gy ,                       .4                erofOPERAg8 OPERAarr      reaut         channels
                                                                                                                           +1 onedess than the eint g                                                                                                             restore the snoperante channel                                                    .

Y,,,y to urtNABLE status within 48 hours or onen the Daartar tria breakers within the next hourg g&p, e , y t % .* P,- 2 f**6 j l o.2 fah n * "*** ACTION - bithtph e6- -seunber oftt- OPE*"2 *4s channels casti-~~ pss than thuAotal Number) Ad Pl -- =vvided/the ' noperaDie hannels are p'l aced in the tripped conettson within 6 hours. ACTION GP,=, CV With one of the diverse trip attachment) in trip features (Undervoltage or shunt _

          @'            on.
                                                                             "ithin      48    hours    or       reble, restore it to OPERABLE status ig,, ,',, Megr 3 g,, 9             4r .,4To aa*Yk, og'M                                      -
                                                                                                                   ==4mm nr==*a -= u r erse +.. aisir wg,, ,, A.,,,

Acmu e, e = abto Aa, L,.,, dse s breater nais not se assee uni w one un

    .sett 2 9 .n u                              vers          ip featu        is inoperab except for the erW4mc+A p <ce4uses,     ir                                              for pe vining mainten                  to restore (o%<naid.
                                                  .An Al Is                              LE statu           with tha b svaker ar livoassadra v ACTION fETidRi3                    With any             actor trip b              breaker inopers                                            restore the reaker to 0F                status prior e                                              og it in A.4
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CATAWBA - UNIT 2 15 Ac% r, 3/4 3-6 Amendment No. 14R 9* ,c ,,o , u *. on c~. a :a., a. n. TasteN. P*wer. *r 4 6- 1 on. is A. ow. s.a .m :. ' ^ . " - \ Q.1r r,ct.o.o or : B.s huhe C o ra.*

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I RTS Instrumentation 3.3.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIE

94. One RTB train NOTES - - -

inoperable. 1. One train may be bypassed for up to 2 hours for surveillance testing, provided the other train is OPERABLE.

2. One RTB w N bypassed for u;> in f twe fv -

maintes w w re u h ;iL4geIer $4ri trip seche !'uss, p%vided the other train,is OPERABLE. G 4.1 Restore train to I hour OPERABLE status. , 4.2 Be in MBE 3. 7 hours ' ; Or- N 0't,) p 6 Rt. anne 1 4.1 Verify interlock is inoperable.(S) I hour in required state for existing unit conditions.

                                                               .2       Be in MODE 3.         7 hours (continued)

W)iHIT5i 3.3 8 Rev 1. 04/07/95 cawk ~ s e

RTS Instrumentation 3.3.1 ACTIONS (continued) COWITION REQUIRED ACTION C0dPLETION TIME or mwi) g 05 S/. One nnelOS inoperable. X.1 Verify interlock is 1 hour in required state for existing unit conditions. IE s T.2 Be in MODE 2. 7 hours T

      'T4. One trip mechanism     V.1 inoperable for one            Restore inoperable      48 hours R17.

trip mechanism to i OPERABLE status. h M.2 Be in MODE 3. 54 hours

U. 2 Open R . hours Y g. Two RTS trains

                                   .1     Enter LC0 3.0.3.

inoperable. Ismediately WDGPI$5 3.3 9 Ca %k Rev 1. 04/07/95 G

                                    ,                  No chan                 &is g                                                   .

Ince on w arS r,istru.entation V 3.3.1 Tabte 3.3.1 1 (page $ of e) teacter Trip system Instrimentation APPLICASLE MODES OR OTHER SPECIFIED ateutaED FLAICTION Comifl0HS survtILtAmCE ALLO m BLE ft! CHANNELS COWITIONS REeu!AEBENTS VALLE SETPOIN lQ,,W. i e '-i-W(4 Eh e. P jana m3 103 h /N st 3.3.1.10 t t st 3.3.1.15 ps pois

b. Turtpine Step 1NI 4 5 /'O Vatwo Ctesure st 3.3.1.10 ejthopen Jtd seen at 3J.1.15 7 l [ I. Safety 1,2 2 trains injection (SI) ff SR 3J.t.14 h4 mA levast free engineered s.fety Feature Actuation System (ESFAs) k h /. Beester felp Systes Intertecke

e. Intermedlete 2 Renee Neutrun 2

                                                                           /R         an 3.3.i.it     =\e-st7         etce.10,J- @

f hat, P-6 at 3 ? M.13 esp ange

b. Lau Peuer Reacter Trips 1 1 per trein f.T ISR labg.1'.1 Ng na mA Bleek, P-F

c. Power Range 1 4 Weutron Flm, 33 3.3.1.11 . g gyp P-8 a 3.3.1.13 p

d. Power Range 1 &

heutron Flm , [$ st 3J.t.tt g g g gyp P9 sa 3.3.1.13 g

e. Peuer targe 1,2 6 boutron Ftm , /( at 3.3.1.11 3 . I t 1 E RTP P-10 at 3.3.1.13 RTP and I 5 12. E

f. furtitne tapulse 1 2 [5 h RTP 3

Pressure, P 13 (tse }<fMD EK J.J.l. % skt2.dz turttine @V skt0fg turt>Ene l BR 3.3.1.t3

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_ c leptementattens ap(cantain enty Attenut>[value elapending en Se nt d f4 seteu the P-6 (Interensiste aanse neutron Flux; intertects. (jj Ahew the P-9 (Power tense boutron 7 tun) Intertect. STS 3.3 19 C4 M A Rev 1. 04/07/95 l t o (V

RTS Instrumentation B 3.3.1 BASES ACTI0lts and 2 (continued) OP accion AI.1) is rwanext 6 p Time letion . Theof 6 hours (Required le considering that in this Condition. the runeining OPEMBLE train is to perfore the safety function and given the low an event during this interval. The Completi Time of ility~of 6 hours (Reair red Action #.2) is reasonable, based on f .ong orderly men @ner and without challenging unit systems.iones h The alicusRequired b Actions have been modified by a Note that testing,ypassing one train up tq,t4Phours for surveillance provided the other tratft is OPERABLE. h.1an .2 ' Condition lies to the RTBs in MODES 1 andThese 2. RTBs. . With one train inoperable. I hour is allousd' toacti restortinthe placed M 3train withinco GPERNKE the next 6 hours.status The Comp or the unit'aust he' 1stion Time of 6 hours is reasonable, based on operating -* experience. to reach MODE 3 from full power in an orderly

      ,                 menner and without challenging unit systems. The 1 hour and r                       6 hour Completion Times are equel to the time alloued by II0   3.0.3 fbr shutdown actions in the event of a complete loss of RTS Function.

he riquirweent for 21s paPlacing the unit in M 3 removes lar Funct1on.,

The Required Actions Note 1 allows one been modified by two Notes. oto be bypassed for 1 for surveillance testing, provided the other 2 hours lh 2 hours for maintenance on undervolNoteup to Q one RTB OPERABLE. \O1 2 allows to be bypa or shunt trip sechanisms if thew other RTB train-is- < The 2 hour --

time limit is . justified in Reference ?.

h1and 2 (or Awe one tion 6 applies to the P 6 and P.10 interlocks. With coinci l/ inoperable for one out of two or two out-of four

              @ to bei n its required state for the existing unit cond (continued)

M B 3.3 48 cm L L- Rev 1. 04/07/95 i l (O

RTS Instrumentation B 3.3.1 BASES

 ,    ACTIONS    bg.1a         2 (continued) within 1 hour or the unit must be placed in m 3 within the next 6 hours. Verifying the interlock statustaanually        fj %g a     lishes the interlock s Function. The Completion Itse                            -.

o is based on operating experience and the minimus amount of time allowed for manual operatv actions. The Completion Time of 6 hours is reasonabi h sed on operating k DSM W'** experience, to reach M 3 from full pw in an orderly l'M> > manner and without challenging unit systems. The 1 hour and 6 hour Caspletion Times are equal to the time allowed by LCO 3.0.3 for shutdown actions in the event of a complete loss of RTS Function.

                 @f.1and 2                                                                                l
               . Condition     applies          P-   8 P 9, rA P 13 interlocks. With are              i   able for one out of two or two out of four coincidence 1      , the associated                               l interlock must be verified to be n its required state for the existing unit condition within 1 hour or the unit must be placed in MODE 2 within the natt 6 hours. These-actions are conservative for the case where power level is being raised. Verifying the interlock ste r--        -11 v g

a 11shes the interlock's Function. The Completion Time

       .           6f 1       is based on operating experience and the minisua amount of time allowed for manual operator actions. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 2 from full power in an orderly manner and without challenging unit systems.          -'
                     .1/~DE          4.24 Condition         lies to the RTB Undervoltage and Shunt Tri Hechanisms. or diverse trip featuresr in-MIBES 1 and 2. +p     -       -    ---

With one of the diverse trip features 1 able, it must be restored to an OPERA 8LE status within 48 s or the unit must be placed in a MIEE where the requirement dJes not apply. This is accomplished by placing the unit in MOE 3 within tu na* 5 hours (54 hours tota tianinarlowed a s ,ino -;be nius in Mdditionsi .nour as trudt otal t == . -

          -       The completion Time of 6 hours is a reasonable time.-based on operating experience.'to reach MODE 3 from full power in an orderly manner and without challenging unit systems.           g g fmot 1 4

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     %                                 8 3.3 49                    Rev 1. 04/07/95 irwge Ve. a. E.

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leeue Number l 12l Affected Section l3.3.3 Post Accident Monitoring ] Affected Unite CNS: l Yesl MNS: l Yesj ANected *ogos IT5: ITS: ITS Bases: B 3.3-110 ITS Bases: B 3.3-109 CTS: CTS: DOCS: DOCS: NRG: NitG: NRG Bases: B 3.3-124 NRG Bases: B 3.3-124 JFD: JFD: NSHC: NSHC: Descriphon Added clarificotton to LCO Bases to clarify the intent of told channels versus required channels since there are often more totd channels than required channels.

l PAM Instrumentation B 3.3.3 O BASES V APPLICABLE

Determine if a gross breach of a barrier has occurred; SAFETY ANALYSES and (continued)

Initiate action necessary to protect the public and to estimate the magnitude of any impending threat. . ( PAM instrumentation that meets the definition of Type A in Regulatory Guide 1.97 satisfies Criterion 3 of 10 CFR 50.36 (Pef. 4). Category I, non-Type A, instrumentation must be

                        .tained in TS because it is intended to assist operators in
                        'nimizing the consequences of accidents. Therefore,
                      .ategory I, non-Type A. variables are important for reducing public risk.

4 LC0 The PAM instrumentation LC0 provides OPERABILITY requirements for Regulatory Guide 1.97 Type A monitors, which provide information required by the control room { operators to perform certain manual actions specified in the unit Fmergency Operating Procedures. These manual actions { ensure that a system can accomplish its safety function, and are credited in the safety analyses. Additionally, this LC0 .

 ,O                   addresses Regulatory Guide 1.97 instruments that have been designated Category I, non-Type A.

The OPERABILITY of the PAM instrumentation ensures there fs sufficient information available on selected unit parameters to monitor and assess unit status following an accident. This capability is consistent with the recomendations of Reference 1. ! LC0 3.3.3 requires two OPERABLE channels for most Functions. Two OPERABLE channels ensure no single failure prevents operators from getting the information necessary for them to determine the safety status of the unit, and to bring the unit to and maintain it in a safe condition following an accident. Furthermore, OPERABILITY of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. . In some cases, the total number of channels exceeds the l number of required channels, e.g., pressurizer level has a e s (continued) Catawba Unit 1 B 3.3-109 Supplement 1 l +

PAM Instrumentation B 3.3.3 (3 BASES d LCO total of three channels, however only two channels are (continued) required OPERABLE. This provides additional redundancy beyond that required by this-LCO, i.e., when one channel of pressurizer level is inoperable, the required number of two channels can still be met. The ACTIONS of this LC0 are only entered when the required number of channels cannot be met. Type A and Category I variables are required to meet Regulatory Guide 1.97 Category I (Ref. 2) design and qualification requirements for seismic and environmental qualification, single failure criterion, utilization of emergency standby power, intnediately accessible display, continuou's readout, and recording of display. Listed below are discussions of the specified instrument Functions listed in Table 3.3.3-1. 1, 2. Reactor Coolant System (RCS1 Hot and Cold Lea Temoeratures RCS Hot and Cold Leg Temperatures are Category I variables provided for verification of core cooling and long term surveillance. RCS hot and cold leg temperatures are used to determine RCS subcooling margin. RCS subcooling margin will allow termination of safety injection (SI), if still in progress, or reinitiation of SI if it has been stopped. RCS subcooling margin is also used for unit stabilization and cooldown control. In addition, RCS cold leg temperature is used in conjunction with RCS hot leg temperature to verify the unit conditions necessary to establish natural circulation in the RCS. Reactor coolant hot and cold leg temperature inputs are provided by a fast response resistance element in each loop. RCS Hot and Cold Leg Temperature are diverse indications of RCS temperature. Core exit thermocouples also provide diverse indication of RCS temperature. (continued) l Catawba Unit 1 B 3.3-110 Supplement 1 l

PAM Instrumentation B 3.3.3 O BASES V APPLICABLE ' Determine if a gross breach of a barrier has occurred; SAFETY ANALYSES and (continued)

Initiate action necessary to protect the public and to 1 estimate the magnitude of any impending threat.

PAM instrumentation that meets the definition of Type A in Regulatory Guide 1.97 satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4). Category I, non-Type A, instrumentation must be retained in TS because it is intended to assist operators in minimizing the consequences of accidents. Therefore, Category I, non-Type A, variables are important for reducing public risk. LCO The PAM instrumentation LCO provides OPERABILITY requirements for Regulatory Guide 1.97. Type A monitors, wnich provide information required by the control room operators to perform certain manual actions specified in the unit Emergency Operating Procedures. These manual actions ensure that a system can accomplish its safety function, and . are credited in the safety analyses. Additionally, this LCO addresses Regulatory Guide 1.97 instruments that have been - O. designated Category I, non-Type A. i The OPERABILITY 'of the PAM instrumentation ensures there is sufficient information available on selected unit parameters to monitor and assess unit status following an accident. This capability is consistent with the recommendations of l Reference 1. LCO 3.3.3 requires two OPERABLE channels for most Functions. Two OPERABLE channels ensure no single failure prevents operators from getting the information necessary for them to determine the safety status of the unit, and to bring the unit to and maintain it in a safe condition following an accident. Furthermore, OPERABILITY of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. In some cases, the total number of channels exceeds the number of required channels, e.g., pressurizer level has a (continued) Catawba' Unit 2 B 3.3-109 Supplement 1 l l

F PAM Instrumentation B 3.3.3 I BASES i l c ) LCO total of three channels, however only two channels are (continued) required OPERABLE. This provides additional redundancy beyond that required by this LCO, i.e., when one channel of i , pressurizer level is inoperable, the required number of two channels can still be met. The ACTIONS of this LC0 are only entered when the required number of channels cannot be met. Type A and Category I variables are required to meet Regul M ory Guide 1.97 Category I (Ref. 2) design and qualitication requirements for seismic and environmental J qualification, single failure criterion, utilization of emergency standby power, immediately accessible display, continuous readout, and recording of display. Listed below are discussions of the specified instrument Functions listed in Table 3.3.3-1. 1, 2. Reactor Coolant System (RCS) Hot and Cold Lea l Temoerat ggi ' RCS Hot and Cold Leg Temperatures are Category I ,p variables provided for verification of core cooling Q and long term surveillance. RCS hot and cold leg temperatures are used to determine RCS subcooling margin. RCS subcooling margin will allow termination of safety injection (SI), if still in progress, or reinitiation of SI if it has been stopped. RCS subcooling margin is also used for unit stabilization and cooldown control. ' In addition, RCS cold leg temperature is used in conjunction with RCS hot leg temperature to verify the unit conditions necessary to establish natural circulation in the RCS. Reactor coolant hot and cold leg temperature inputs are provided by a fast respon?3 resistance element in each loop. RCS Hot and Cold Leg Temperature are diverse indications of RCS temperature. Core exit thermocouples also provide diverse indication of RCS temperature. i 1 (continued) ! l Catawba Unit 2 B 3.3-110 Supplement 1 1

PAM Instrumentation B 3.3.3 0 - LC0 Furthermore, OPERABILITY of two channels allows a CHANNEL (continued) CHECK during the post accident _ phase to confirm the validity of displayed information.) Inan two a ...is oe- ' ( rN;. W units Y t unit specific Regu ry k 1.97.i lyses (Ref, determined that fa' ure of one gpe accident itoring che results in informa on ambigotty (that i the redundant isplays disagree) t 3 - could lead 3pera s to defeat or -ail to accomplish a ired safet) [f_u n, t <wi ~ dhe -- _-uun i. ... is unn n n. i Isolation Val (CIV) Position.tnetwocr.o.i1.Uh1 I case, the important informati s the status of the nennt penetrations. The LCO ires one position indi for each active CIV This i ficient to redundantly ify the isolation S stat of each isolable penetrat either via indicated sta of the active valve and or knowledge of a pa ve v , or via system bounder tus. If a normally ive V is known to be closed deactivated, position indicatien is not needed deterwine status. fore, the position indicati or values in tMa <tata j equired to be OPERAB y n bo ; T wiw 5.a.4- provides a list or , lauive wn i or t ~ identified y the unit specific latory Guide 1.97 (Ref.1) alyses. Table 3.3.3 n unit specific TS ld list al Type A and Category ariables identified the unit ific Regulatory Gui 1.97 analyses, as by the 's SER. O2 a vne nnmCategory I variables are required to meet Regulatory Guide 1.97 Category I (Ref. 2) design and qualification requirements for~ seismic and environmental qualif' cation, single failure criterion, utilization of emergency standby power immediately accessible display,

                 ,  continuous readout, and recording of display.

Listed below are discussions of the specified instrument Functions listed in Table 3.3.31. JThese uncussi ari

                   /1nt r.G.G ayexamples of what sDluld be provided f (Functior)4en the unit speciKc list is preoarett                      ach h 1

(continued) WJr51T B 3.3 124 Rev 1, 04/07/95 (deGutst O

1 INSERT ,O In some cases, the total number of channels exceeds the number of required channels, e.g., pressurizer level has a total of three channels, however only two channels are required OPERABLE. This provides additional redundancy beyond that required by this LCO, i.e., when one channel of pressurizer level is inoperable, the required number of two channels can still be met. The ACTIONS of this LCO are only entered when the required number of channels cannot be met. O l l l l l l l I l INSERT Page B 3.3-124 ; Catawba

leeue Number l 15l Affected Section p.5.1 Accumulators I Affected Uraits CNS: l Yesl MNS: l_ Yesl Affected Pogos ITS: 3.5-3 ITS: 3.5-3 ITS Bases: B 3.5-2. 8 ITS Bases: B 3.5 2. 8 CTS: 3/45-3 CTS: 3/45-2 DOCS: M-1 DOCS: M-1 NRG: 3.5 3 NRG: 3.5-3 NRG Boeos: B 3.5-2, 8 NRG Bases: B 3.5-2. 8 JFD: TS 1. 81 JFD: TS 1. 81 NSHC: NSHC: Description Boses are revised to indicate that the occumulator isolation volves are Mt considered to be within the scope of IEEE-279 since the volves are opened manually at 1000 and power remove % therefore, the devices are possive. The Bases are revised to be consistent with the UFSAR 5.5.12 discussion. SR 3.5.1.5 is dso revised to require power removoi at 1000 psig consistent with Westinghouse Nuclear Sofety Advisory Letter for Mode 4 LOCA andysis.

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify power is removed from each 31 days accumulator isolation valve operator wnen pressurizer pressure is > 1000 psig. l 0 - O Catawba Unit 1 3.5-3 Supplement 1 l

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify power is removed from each 31 days accumulator isolation valve operator when pressurizer pressure is > 1000 psig. l i i 1 l O l 1 l O Catawba Unit e 3.5-3 Supplement 1 l I

Accumulators B 3.5.1 1 i I BASES BACKGROUND This interlock also prevents inadvertent closure of the l (continued) valves during normal operation prior to an accident. The i valves will automatically open, however, as a result of an SI signal. The isolation valves between the accumulators and the Reactor Coolant System are required to be locked open during unit operation. In that the subject valves are normally open and do not serve as an active device during a 1 LOCA, the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref. 1) is not applicable in this situation. Therefore, the subject I valve control circuit is not designed to this standard. The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA. 0 V APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY AL fSES small break LOCA analyses at full power (Ref. 2). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits. In performing the LOCA calculations, conservative assumptions are 7de concerning the availability of ECCS flow. In the eariy stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow . until f.he emergency diesel generators start, come to rated i speed, and go through their timed loading sequence. In cold ' leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break. The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump. During (continued) l Catawba Unit 1 B 3.5-2 Supplement 1

Accumulators B 3.5.1 BASES APPLICABLE this event, the accumulators discharge to the RCS as soon as SAFETY ANALYTES RCS pressure decreases to below accumulator pressure. (continued) As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting, the valves opening, and the pumps being loaded and delivering full flow. The delay time is conservatively set to account for SI signal generation, diesel start-up, pump start-up, and valve opening. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA. The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators, safety injection pumps, and centrifugal charging pumps all play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators O continues to decrease until they are not required and the centrifugal charging pumps become solely responsible for terminating the temperature increase. This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 3) will be met following a LOCA:

a. Maximum fuel element cladding temperature is s 2200*F;

b. Maximum cladding oxidation is s 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water i reaction is s 0.01 times the hypothetical amount that ,

would be generated if all of the metal in the cladding l cylinders surrounding the fuel, excluding the cladding j surrounding the plenum volume, were to react; and i

d. Core is maintained in a coolable geometry. 2 Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute directly to the long term (continued)

Catawba Unit 1 B 3.5-3 Supplement 1 l

Accumulators B 3.5.1 ( BASES APPLICABLE cooling requirements of 10 CFR 50.46. However, the boron SAFETY ANALYSES content of the accumulator water helps to maintain the (continued) reactor core subcritical after reflood, thereby eliminating fission heat as an energy source for which cooling must be provided. For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained I water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. The large and small brRk LOCA analyses are , performed with accumulator volumes trat are consistent with ' the LOCA evaluati values of

30 ftgn models.

To all w for operating margin, are specified. The minimum boron concentration setpoint is used in the post ; LOCA sump boron concentration calculation. The calculation i is performed to assure reactor subcriticality in a post LOCA ! envieonment. Of particular interest is the large break LOC 4 since no credit is taken for control rod assembly , insertion. A reduction in the accumulator minimum boron ' concentration would produce a subsequent reduction in the { p available containment sump concentration for post LOCA e i shutdown and an increase in the maximum sump pH. The V maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. The large and small break LOCA analyses are performed with accumulator pressures that are consistent with the LOCA evaluation models. To allow for operating margin and accumulator design limits, a range from 585 psig to 678 psig is specified. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity. The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Ref. 4). The accumulators satisfy Criterion 3 of 10 CFR 50.36 (Ref.5). (continued) u ; l Catawba Unit 1 B 3.5-4 Supplement 1 l

Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.4 REQUIREMENTS (continued) The boron concentration should be verified to be within required limits for each accumulator every 31 days since the static design of the accumulators limits the ways in which the concentration can be changed. The 31 day Frequency is adequate to identify changes that could occur from mechanisms such as stratification or inleakage. Sampling the affected accumulator within 6 hours after a 75 gallon increase will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST). because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the reconrnendation of NUREG-1366 (Ref. 6). , i SR 3.5.1.5 Verification every 31 days that power is removed from accumulator isolation valve operators for NI54A, NI658, NI76A, ( e [ and NI888 when the pressurizer pressure is > 1000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. Since pner is removed and circuit breakers padlocked under administrative control, the 31 day Frequency will provide adequate assurance that power is removed. This SR allows power to be supplied to the motor operated l isolation valves when pressurizer pressure is s 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves. Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA. OV (continued) l Catawba Unit 1 B 3.5-8 Supplement 1

Accumulators B 3.5.1 O BASES b BACKGROUND This interlock also prevents inadvertent closure of the (continued) valves during normal operation prior to an accident. The valves will automatically open, however, as a result of an SI signal. The isolation valves between the accumulators and the Reactor Coolant System are required to be locked open during unit operation. In that the subject valves are normally open and do not serve as an active device during a LOCA, the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref. 1) is not applicable in this situation. Therefore, the subject valve control circuit is not designed to this standard. The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators i are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following.a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA. (d APPLICABLE SAFETY ANALYSES The accumulators are assumed OPERABLE in both the large and small break LOCA analyses at full power (Ref. 2). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits. In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss : of offsite power, the accumulators provide the sole source ; of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator ' are assumed to be lost through the break. The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump. During n (continued) V l Catawba Unit 2 B 3.5-2 Supplement 1

Accumulators > B 3.5.1 ! BASES O i APPLICABLE this event, the accumulators discharge to the RCS as soon as SAFETY ANALYSES RCS pressure decreases to below accumulator pressure. (continued) As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting, the valves opening, and the pumps being loaded and delivering full flow. The delay time is conservatively set to account for SI signal generation, diesel start-up, pump start-up, and valve opening. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA. ; The worst case small break LOCA analyses also assume a time I delay before pumped flow reaches the core. For the larger { range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped . flow then providing continued cooling. As break size decreases, the accumulators, safety injection pumps, and centrifugal charging pumps all play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators ( continues to decrease until they are not required and the . 4 centrifugal charging pumps become solely responsible for terminating the temperature increase. This LC0 helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 3) will be met following a LOCA:

a. Maximum fuel element cladding temperature is s 2200'F;

b. Maximum cladding oxidation is s 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is s 0.01 times the hype".netical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and

d. Core is maintained in a ( )lable geometry. ,

I Since the accumulators discharge during the blowdown phase { of a LOCA, they do not contribute directly to the long term (continued) Catawba Unit 2 B 3.5-3 Supplement 1 l

Accumulators B 3.5.1 BASES APPLICABLE cooling requirements of 10 CFR 50,46. However, the boron

    . SAFETY ANALYSES content of the accumulator water helps to maintain the (continued)  reactor core subcritical after reflood, thereby eliminating -

fission heat as an energy source for which cooling must be provided. For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. The large and small break LOCA analyses are performed with accumulator volumes that are consistent with the LOCA evaluati valuesof*30ftgnmodels. To allow for operating margin, are specified. The minimum boron concentration setpoint is used in the post LOCA sump boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control . rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA , .O shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. The large and small break LOCA analyses are performed with accumulator pressures that are consistent with the LOCA evaluation models. To allow for operating margin and accumulator design limits, a range from 585 psig to 678 psig is specified. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity. The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Ref. 4). The accumulators satisfy Criterion 3 of 10 CFR 50.36 (Ref. 5). (continued) l Catawba Unit 2 B 3.5-4 Supplement 1

Accumulators B 3.5.1 ( BASES V] SURVEILLANCE SR 3.5.1.4 REQUIREMENTS j (continued) The boron concentration should be verified to be within a required limits for each accumulator every 31 days since the static design of the accumulators limits the ways in which the i concentration can be changed. The 31 day Frequency is adequate to identify changes that could occur from mechanisms { such as stratification or inleakage. Sampling the affected accumulator within G hours after a 75 gallon increase will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), I because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the reconmendation of NUREG-1366 (Ref. 6). SR 3.5.1.5 Verification every 31 days that power is removed from accumulator isolation valve operators for NI54A, NI658, NI76A, , l and NI888 when the pressurizer pressure is > 1000 psig ensures A that an active failure could not result in the undetected V ' closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. Since power is removed and circuit breakers padlocked under administrative control, the 31 day Frequency will provide adequate assurance that power is removed. This SR allows power to be supplied to the motor operated l isolation valves when pressurizer pressure is s 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or , shutdowns. Even with power supplied to the valves, i inadvertent closure is prevented by the RCS pressure interlock i associated with the valves. Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA. (continued) l Catawba Unit 2 8 3.5-8 Supplement 1

~ '

6&A.% 3,g,, EMERGENCY CORE COOLING SYSTEMS . O SURVEILLANCE REOUIREMENTS (Continued) A .i SR. 3.5.ls @ At least _ence per 31 days when the Reactor Coolant System ressure is p,I above @S)psig@ verifQ tha_t power is removed from 1 valve operators >n ves MI54 I65B NII6 sotatio _wl Qespective circuit akers a adlocked *pd -and N187B and tngt_ T d.~ At least o ce per 18 month by verifying that ach colo leg injec toni accumulat r isolation valv opens automatical under each of th followin conditions: L'#

1) Wha an actual or a inulated Reactor C clant System press re si taal exceeds the -11 (Pressurizer P ssure Block of Sa ty I ; ection) Setpoin , and

2) receipt of Safety injection t t signal.

f4.5.1.2 Eac cold les inject n accumulator water level and pressu channel sh all be d strated OPERA 8 : - I i a, least once per days by the perfo e of an - Ot,.I TIONAL TEST and CHANNEL

b. At least onca r 18 months by the rformance of a EL CALIBRATION.

 /t O

G

                                        .                                                                            \

CATAWBA - UNIT 1 3/4 5-3 Amendment No. 148 l m : U 6p o l

S

St aAAes 3,5, s EMERGENCY CORE COOLING SYSTENS SURVEILUUICE RE0UIREMENTS (Continued) AJ N 36'I'g@ Ataboveless $once per9 31 days when the Reactor Coolant frna System

                                                                                             . ssure is p,j                              psig     verifyGb   that power    is removed                   sola+ta=cp.W valve anaratart An V        es MIb4A,      one, nun , and N188Vana thalf thA CennectiveAlrcuit        eakers are ,ydiocked;ndf                                 g i d. At least o e per 18 months by ertfying that e h cold leg inject on accumulat      isolation valve o s automatically under each of the followl conditions                       ,

L. 1) an actual or a s lated Reactor C lant Systes press re s' al exceeds the P 1(PressurizerP ssu.~e Block of Sa ety ection)Setpoint and i % receipt of a fety Injection st stenal. 4.5.1.2 E cold leg injec on accumulator we r level and pressu channeM shall be d strated OPGtAB :

a. one days by the pe reance of an CHAfslEl.

18 months by th performance of a n; d A go P El. b ' I l 1 CA1AWBA - UNIT 2 3/4 5-3 Amendment No. 142 1 I 6'c 3.03 i O

I Discussien of ChangIs ' S:ctien 3.5 - Emerg;ncy C:ra C mling System (ECCS) I v TECHNICAL CHANGES - MORE RESTRICTIVE M.1 Not used. M.2 CTS 4.5.1.1.c requires that the power be removed from the accumulator isolation valve when RCS pressure is greater than 2000 psig. ITS SR 3.5.1.5 requires that the power be removed when RCS pressure is greater than 1000 psig. Westinghouse Nuclear Safety Advisory Letter (NSAL) 97-003 has recomended changes to the Bases based on shutdown LOCA considerations. These recomendations relate to the consideration of operating bypasses and compliance with IEEE 279-1971. The automatic unblock feature of these valves is not considered on operating bypass as described in the UFSAR i since the valves are manually opened and power is removed. The } Bases of ITS 3.5.1 has been revised consistent with the UFSAR. l The accumulator isolation valves are manually opened when RCS 1 pressure is greater than 1000 psig. The requirement to remove power to the valves at 1000 psig is more restrictive, but is consistent with existing practice and with the operability assumptions described in the subject NSAL for shutdown LOCA. o -

O Catawba Units 1 and 2 Page M - 14 Supplementis/2c/97l

Accumulaters 3.5.1 SURVEILLANCE REQUIRENENTS (continued) V SURVEILLANCE FREQUENCY SR 3.5.1.5 Verify power is removed from each 31 days accumulator isolation valve ator when pressurizer pressure i psig

                                &          5&E I

I l i

   %                                  3.5 3                 Rev 1. 04/07/95 CA liwh A.

O

J::stificatica fcr Deviatitns S;cticn 3.5 - Emerg:ncy Ctre Cooling Systems (ECCS) TECHNICAL SPECIFICATIONS

9. NUREG SR 3.5.1.5 requires that the power be removeo from the accumulator isolation valve when RCS pressure is greater than 2000 psig consistent with CTS 4.5.1.1.c. ITS SR 3.5.1.5 will reqaire that the power be removed when RCS pressure is greater than 1000 psig.

Westinghouse Nuclear Safety Advisory Letter (NSAL) 97-003 has recommended changes based on shutdown LOCA considerations. The more restrictive requirement is discussed in DOC M.2. O l O . Catawba Units 1 and 2 22 Supplement 15/20/07l

Accumulators B 3.5.1 /3 BASES U BACKGROUND This interlock also prevents inadvertent closure of the (continued) valves during normal operation The valves will aut==tically open. prior to an as however. accident. a result of SI sional. IThese atures ensure that he valves meet the requirements of Institute of El rical and Electroni hM5M ---' J Engineers (IE Standard 279 1971 f. 1) for "operat bypasses

a that the accumula will be available or injection hout reliance on ator action.

The acetsuulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accLaulator will be lost via the RCS pipe break during the blowdown phase of the LOCA. APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY AhALYSES small break LOCA analyses at full power (Ref. 2). These are the Design Basis Accidents (DBAs) that establish the-acceptance limits for the accumulators. Reference to the analyses fy these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits. O' In performing the LOCA calculations, conservative assumptions are made c:ncerning the availability of ECCS flow. In the early stages of a LOCA. with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by lations and conservatively imposes a delay wherein the E pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go thwgh their timed loading sequence. In cold leg break scenar ' are asstaued to the entire contents of one acetsuulator v t through the break. The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant . During this event, the acetamulators discham to the as soon cs PCS pressure decrenes to below accumulator pressure. (continued)

   }$XPSTS                                 B 3.5 2                                                   Rev 1. 04/07/95 c a w b a.

O

INSERT @ The isolation valves between the accumulators and the Reactor Coolant O System are required to be locked open during unit operation. In that the subject valves are normally open and do not serve as an active device during a LOCA, the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref.1) is not applicable in this situaticn. Therefore, the subject volve control circuit is not designed to this standard, c . GQQ INSERTPage B 3.5-2

Accumulators B 3.5.1 BASES SURVEll1ANCE SR 3.5.1.4 REQUIREENTS (continued) The boron concentration should be verified to be within required limits for each accumulator every 31 day since the static design of the accumulators limits the ways in which the concentration can be changed. The 31 day frequency is adequate to identify changes that muld Accur from mechanisms such as stratification or inleakager"3amo11ng tre arrecteo '1Cyl6 5 accumulator within 6 hours after d(ifarchne] increase will identify whether inleakage has causea a reouction in boron concentration to below the required limit. It is not necessary to verify boron concentration.1f the added water inventory is from the refueling water storage tank (RWST). because the water contained in the RWST is within the accumulator boron concentration requirements. This is & consi tent with the recommendation of NlREG 1366 (Ref. . f SR 3.5.1.5 (3.w Pe % -MMIMIU8 s*1, Nr K6 g Verificati very 31 days that s removed fron Qr y accumulator pressure 1 lation valve oper when the pressuri@zer.3) S psig ensures tha an active failure could not result in t1e undetected closure of an accumulator motor k operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single -@ failure coincident with a LOCA. Since power is removedfunder M ONU administrative control, the 31 day Frequency will provide adequate assurance that power is removed. LLoken paleM)J This SR allows power to be supplied to the no rated isolation valves when pressurizer pressure is G gs psig, 1000 g thus allowing operational flexibility by avoiding unnecessary I delays to manipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves. Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves muld open a closed valve in the event of a LOCA. (continued) M 8 3.5 8 Rev 1, 04/07/95 ca w ba iO

l Justificatica f r Deviations S:ctien 3.5 - Emerg:ncy Cara Cooling Systems (ECCS) l (3

 %.)  g NOTE:         The first five justifications for these changes from NUREG-1431 were generically used throughout the individual Bases section markups. Not all generic justifications are used in each section.

i

1. The brackets have been removed and the proper plant specific information or value has been provided.

2. Editorial change for clarity or for consistency with the Improved Technical Specifications (ITS) Writer's Guide.

3. The requirement / statement has been deleted since it is not applicable to this facility. The following requirements have been renumbered, where applicable, to reflect this deletion.

4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

O Q 5. Bases have been modified to reflect changes made to the technical specifications. I

6. Westinghouse Nuclear Safety Advisory Letter (NSAL) 97-003 has recomended a review of the ITS Bases for accumulator isolation valves (NUREG 3.1.5) based on shutdown LOCA considerations. These recomendations of the NSAL relate to the consideration of operating bypasses and compliance with IEEE 279-1971. The automatic open feature of these valves is not considered on operating bypass as described in the UFSAR since the valves are manually opened and power is removed.

The Bases of ITS 3.5.1 has been revised consistent with the UFSAR and current licensing basis. 4 O lCatawbaUnits1and2 14 Supplement 15/20/97

leeue Number l 16} Affected Socilon @.9.5. 3.9.6 (3.9A 3.9.5 for CNS) RHR Loops l Affected Units CNS: l Yesl MNS: l Yesi Affected Pogos ITS: 1 ITS: ITS Bases: 8 3 9-t 7. 21 ITS Boeos: B 3 9-18. 22 CTS: CTS: DOCe: DOCS: NRG: NRG: NRG Bases: B 3 9-18. 22 NRG Bones: B 3.9-18. 22 JFD: B 1, 2 JFD: B1 NSHC: NSHC' Descripilon Boses 3.7.6 and 3.7.7 under Applicabdtty contains the centence: 'In MODES 5 & 6, the OPERABluTY requirernents of the CCW/NSWS are deterrnined by the system it supports." Contrary to this statement, Bones 3.9.5 and 3.9.6 do not contoln any further information recording operobility requirements for CCW and NSW in Mooes 5 & 6. Added the following sentence to Bases 3.9.5 and 3.9.6 (3.9.4 and 3.9.5 for CNS):ihe operobility of the operating RHR train and the supporting heat sink is dependent upon the ability to maintoin the desired NC system temperature? O

RHR and Coolant Circulation-High Water Level B 3.9.4 ( BASES LCO Only one RHR loop is required for decay heat removal in MODE 6, with the water level 2 23 ft above the top of the reactor vessel flange. Only one RHR loop is required to be OPERABLE, because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one RHR loop must be OPERABLE and in operation to provide:

a. Removal of decay heat;

b. Mixing of borated coolant to minimize the possibility of criticality; and

c. Indication of reactor coolant temperature.

An OPERABLE RHR loop includes an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an , OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. The operability of the operating RHR train and the supporting heat sink is dependent on the ability to maintain the desired RCS l temperature. The LC0 is modified by a Note that allows the required operating RHR loop to be removed from service for up to I hour per 8 hour period, provided no operations are oermitted that would cause a reduction of the RCS boron concentration. Boron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve ! testing. During this I hour period, decay heat is removed i by natural convection to the large mass of water in the refueling cavity. APPLICASILITY One RHR loop must be OPERABLE and in operation in M00t 6, with the water level 2 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft water level was selected because it corresponds to the 23 ft requirement established for fuel movement in LC0 3.9.6,

                " Refueling Cavity Water Level ." Requirements for the RHR (continued)

Catawba Unit 1 B 3.9-17 Supplement 1 l

RHR and C:olant Circulation-High Water Level B 3.9.4 BASES U APPLICABILITY System in other MODES are covered by LCOs in Section 3.4, (continued) Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (FCCS). RHR loop requirements in MODE 6 with the water level < 23 ft are located in LC0 3.9.5, " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level ." ACTIONS RHR loop requirements are met by having one RHR loop OPERABLE and in operation, except as permitted in the Note to the LCO. M If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations cannot occur by the addition of water with a lower boron concentration than that contained in the RCS because all of unborated water sources are isolated. / \ V u If RHR loop requirements are f ot met, actions shall be taken innediately to suspend loading of irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core. A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is a prudent action under this condition. M If RHR loop requirements are not met, actions shall be initiated and continued in order to satisfy RHR loop requirements. With the unit in MODE 6 and the refueling water level 2 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated immediately. (continued)

%-)

l Catawba Unit 1 B 3.9-18 Supplement 1

1 RHR and Coolant Circulation-Low Water Level B 3.9.5 l BASES l LC0 Additionally, one loop of RHR must be in operation in order

(continued) to provide: i; a. Removal of decay heat;; b. Mixing of borated coolant to minimize the possibility of criticality; and; c. Indication of reactor coolant temperature.

An OPERABLE RHR loop consists of an RHR pump, a heat exchanger, valves, piping, instruments and controls to < ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. The operability of the operating RHR train and the supporting heat sink is dependent on the ability to maintain the desired RCS temperatum. Both RHR pumps may be aligned to the Refueling Water Storage l Tank to support filling the refueling cavity or for { performance of required testing. ! O APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loop must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flar.ge, to provide decay heat removal. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level 2 23 ft are located in LCO 3.9.4, " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level." i ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, ; action shall be immediately initiated and continued until ' the RHR loop is restored to OPERABLE status and to operation or until 2 23 ft of water level is established above the reactor vessel flange. When the water level is 2 23 ft above the reactor vessel flange, the Applicability changes i to that of LCO 3.9.4, and only one RHR loop is required to (continued) Catawba Unit 1 B 3.9-21 Supplement 1 l

RHR and Coolant Circulation-Low Water Level a B 3.9.5 ( BASES ACTA 0NS A.1 and A.2 (continued) be OPERABLE and in operation. An inmediate Completion Time is necessary for an operator to initiate corrective actions. IL1 If no RHR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations cannot occur by the addition of water with a lower boron concentration than that contained in the RCS, because all of the unborated water sources are isolated. fLZ If no RHR loop is in operation, actions shall be initiated inmediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously, u If no RHR lcop is in operation, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours. With the RHR loop requirements not met, the potential exists for the coolant to boil and release radicactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded. SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient ! decay heat removal capability, prevent vortexing in the suction of the RHR pumps, and to prevent thermal and boron stratification in the core. The RCS temperature is (continued) ( l Catawba Unit 1 B 3.9-22 Supplement 1

RHR and Coolant r.irculation-Lcw Water Level B 3.9.5 BASES SURVEILLANCE SR 3.9.5.1 (continued) REQUIREMENTS determined to ensure the appropriate decay heat removal is maintained. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The { Frequency of 12 hours is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator for monitoring the RHR System in the control room. ; l SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat remova; and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. REFERENCES 1. UFSAR, Section 5.5.7.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

t O Catawba Unit 1 B 3.9-23 Supplement 1 l

RHR and Coolant Circulation-High Water Level B 3.9.4 BASES LC0 Only one RHR loop is required for decay heat removal in MODE 6, with the water level 123 ft above the top of the reactor vessel flange. Only one RHR loop is required to be OPERABLE, because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one RHR loop must be OPERABLE and in operation to provide:

a. Removal of decay heat;

b. Mixing of borated coolant to minimize the possibility of criticality; and

c. Indication of reactor coolant temperature.

An OPERABLE RHR loop includes an RHR pump, a heat exchanger, valves, piping, instruments, a.nd controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. The operability of the operating RHR train and the supporting heat sink is dependent t the ability to maintain the desired RCS temperature. l The LCO is modified by a Note that allows the required operating RHR loop to be removed from service for up to I hour per 8 hour period, provided no operations are permitted that would cause a reduction of the RCS boron concentration. Boron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve testing. During this 1 hour period, decay heat is removed by natural convection to tne large mass of water in the refueling cavity. APPLICABILITY One RHR loop must be OPERABLE and in operation in MODE 6, with the water level 2 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft l water level was selected because it corresponds to the 23 ft requirement established for fuel movement in LC0 3.9.6, I " Refueling Cavity Water Level." Requirements for the RHR (continued) O Catawba Unit 2 B 3.b17 Supplement 1 l

RHR and Co31 ant Circulation-High Water Level B 3.9.4 BASES J APPLICABILITY System in other MODES are covered by LCOs in Section 3.4, (continued) Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level < 23 ft are located in LCO 3.9.5, " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level." ACTIONS RHR loop requirements are met by having one RHR loop OPERABLE and in operation, except as permitted in the Note to the LCO. M If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations cannot occur by the addition of water with a lower boron concentration than that contained in the RCS because all of unborated water sources are isolated, u If RHR loop requirements are not met, actions shall be taken imediately to suspend loading of irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core. A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is a prudent action under this condition. U If RHR loop requirements are not met, actions shall be initiated and continued in order to satisfy RHR loop requirements. With the unit in MODE 6 anJ the refueling water level :t 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated imediately. ( (continued) l Catawba Unit 2 B 3.9-18 Supplement 1

I RHR and Coolant Circulation-Low Water Level B 3.9.5 BASES LCO Additior, ally, one loop of RHR must be in operation in order (continued) to provide:

a. Removal of decay heat;

b. Mixing of borated coolant to minimize the possibility of criticality; and i

c. Indication of reactor coolant temperature.

An OPERABLE RHR loop consists of an RHR pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. The operability of the operating RHR train and the supporting heat sink is dependtnt on the ability to maintain the desired RCS temperature. Both RHR pumps may be aligned to the Refueling Water Storage Tank to support filling the refueling cavity or for performance of required testing. O APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loop must be in operation in' MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with Je water level 2 23 ft are located in LCO 3.9.4, " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level." ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action shall be inrnediately initiated and continued until the RHR loop is restored to OPERABLE status and to operation or until 2 23 ft of water level is-established above the reactor vessel flange. When the water level is 2 23 ft above the reactor vessel flange, the Applicability changes to that of LCO 3.9.4, and only one RHR loop is required to (continued) Catawba Unit 2 B 3.9-21 Supplement 1 l I

RHR and Co31 ant Circulation-Low Water Level B 3.9.5 BASES ACTIONS A.1 and A.2 (continued) be OPERABLE and in operation. An intnediate Completion Time is necessary for an operator to initiate corrective actions. B.d If no RHR loop is in operation, there will be no forced j circulation to provide mixing to establish uniform boror. ; concentrations. Reduced boron concentrations cannot occur by the addition of water with a lower boron concentration than that contained in the RCS, because all of the unborated water sources are isolated. ILZ If no RHR loop is in operation, actions shall be initiated intnediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and B ! concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously. (q/ aa If no RHR loop is in operation, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours. With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded. i l SURVEILLANCE SR 3.9.5.1 REQUIREMENTS , This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability, prevent vortexing in the 1 suction of the RHR pumps, and to prevent thermal and boron stratification in the core. The RCS temperature is (continued) l Catawba Unit 2 B 3.9-22 Supplement 1

l RHR and Coolant Circulation-Low Water Levol B 3.9.5 BASES o SURVEILLANCE SR 3.9.5.1 (continued) REQUIREMENTS determined to ensure the appropriate decay heat removal is maintained. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Frequency of 12 hours is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator for monitoring the RHR System in the control room. SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper , breaker alignment and power available to the required pump. { The Frequency of 7 days is considered reasonable in view of ' other administrative controls available and has been shown to be acceptable by operating experience. t REFERENCES 1. UFSAR, Section 5.5.7.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

1 (continued) Catawba Unit 2 B 3.9-23 Supplement 1 l 1

RHR and Coolant Circulation-High Water Level B3.94 BASES b APPLICABLE reducti . Thereforf. the RHR en is ret a

SAFETY ANALYSES Spect cation. / (continued) LCO Only one RHR loop is required for decay heat removal in H0DE 6, with the water level a 23 ft above the top of the reactor vessel flange. Only one M loop is required to be OPERABLE. because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one RHR loop must be OPERABLE and in' operation to provide:

a. Removal of decay heat;

b. Mixing of borated coolant to minimize the possibility '

of criticality: and

c. Indication of reactor coolant temperature.

An OPERABLE RHR loop includes an RR pump, a heat exchanger. valves, piping. instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.Qgg r The LCO is modified by a Note that allows the required- ) operating RHR loop to be removed from service for up to

,\                 1 hour per 8 hour period, provided no. operations are

[Q permitted that would cause-a reduction of the RCS boron concentration. Boron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor i vessel hot leg nozzles and RCS to RHR isolation valve i testing. During this I hour period, decay heat is removed I by natural convection to the large mass of water in the refueling. cavity. APPLICABILITY One RHR loop must be OPERABLE and in operation in MODE 6 I with the water level a 23 ft above the top of the reacter i vessel flange . to provide decay heat removal. The 23 ft water level us selected beceuse it corresponds to the 23 ft (continued) I W3G STS B 3.9 10 Rev 1. 04/07/95 MAujBA ) O

INSERT The operability of the operating RHR train and the supporting heat sink is dependent on the ability to maintain the desired RCS temperature. O INSERT Page B 3.9-18 l

RHR and Coolant Circulation-Low Watir Level B 3.9. p BASES O LCO Additionally, one loop of RHR must be in operation in order (continued) to provide:

a. Removal of decay heat;

b. Hixing of borated coolant to minimize the possibility of criticality; and

c. Indication of reactor coolant temperature. !

An OPERABLE R)R loop consists of an RHR pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the low end h temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.V p gg7 { ' APPLICABILITY Two RfR loops are required to be OPERABLE. and one RHR loop must be in operation in H00E 6. with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the RHR S MODES are covered by LCOs in Section 3.4.ystem in other Reactor Coolant

,                    System (RCS), and Section 3.5. Emergency Core Cooling Systems (ECCS). RHR loop requirements in W 6 with the             S water level 2 23 ft are located in LCO 3.9.$ " Residual Heat Removal (RHR) and Coolant Circulation-Highwater Level."

ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE. ' action shall be immediately initiated and continued until the RHR loop is restored to OPERABLE status and to operation - or until k 23 ft of water level is established above the reactor vessel flange. When the water level is a 23 ft above the reactor v ssel flange the Applicability changes O", to that of LCO 3.9 . and only one RHR loop is required to be OPERK)LE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions. 4 i (continued) WOG STS B 3.9 22 Rev 1. 04/07/95 hTAW64 ) 1

INSERT l-The operability of the operating RHR train and the supporting heat sink is dependent on the ability to maintain the desired RCS temperature.

 !                                          INSERT 2 Both RHR pumps may be aligned to the Refueling Water Storage Tank to support filling the refueling cavity or for performance of required testing.

i O l 4 l o 1RSERr Rage e 3.g.22

Justificatien f&r Deviations S:cticn 3.9 - Refueling Oper tions BMES. NOTE: The first five justifications for these changes from NUREG-1431 were generically used throughout the individual Bases section markups. Not all generic justifications are used in each section.

1. The brackets have been removed and the proper plant specific information or value has been provided.

2. Editorial change for clarity or for consistency with the Improved Technical Specifications (ITS) Writer's Guide.

3. The requirement / statement has been deleted since it is not applicable to this facility. The following requirements have been renumbered, where applicable, to reflect this deletion.

4. Changeshavebeenmade(additions, deletions,and/orchangestothe NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

5. The Bases have been modified to reflect changes made to the technical specifications.

6. A clarifying statement has been added to the Bases of ITS 3.9.1 to provide plant interpretation of the term " positive reactivity addition" for refueling operations. This statement in the bases clarifies that cooldown effects with a positive MTC is not considered a " positive reactivity addition" because it does not significantly affect the core reactivity condition over the refueling temperature range.

7. The boron dilution accidents are analyzed in all modes, however, the statement that MODE 5 is most limiting may be cycle dependent. This statement was deleted from the Bases of ITS 3.9.1.

8. This change reflects a generic change to NUREG-1431 proposed by the industry owners groups. The justification for this change is contained in Technical Specification Task Force (TSTF) change number TSTF-21.

9. The Applicability discussion in the Bases for 3. 7.6 and

3. 7. 7 indicates that the operability requirements for CCW and NSWS in MODES 5 and 6 are determined by the systems supported. A statement is added to the Bases for 3.9.4 and 3.9.5 for RHR loops to indicate that the RHR loops I

lCatawbaUnits1and2 14 Supplement 15/20/97

4 Justific tien f:r Deviations Section 3.9 - Refueling Operations T Ed.SEE operability requirements are also based on the ability to maintain RCS temperature._ This statement is an editorial enhancement desired by the plant staff to make a logical tie ta the support system relationship provided by NSWS and CCW to RHR in MODE 6. l i l l i 1 O l Catawba Units 1 aid 2 24 Supplement 15/20/07l

leeue Number [ 17j Affected Section l3.7.7 (3.7.8 CNS) Nuclear Service Water System l Affected Units CNS: l Yes] MNS: l Yes! /.;sected Pages ITS: 1 IT5: B 3.7-38 t ITS Bones: B 3 7-42 i ITS Boeos: CTS: I CTS: DOCS: DOCS: NRG: NSG: NSG Bones: B 3.7-42 NRG Boeos. B 3.7-42 JFD: B1 JFD: B1 NSHC: NSHC: Description 3.7.7 (3.7.8 CNS) Boses Applicctste Safety Anotysis, lost parograph, line 3 hos incomplete sentence. See 3.7.6 BASES for comparison. Change '(Ref. 3) entry condmons' to '(Ref. 3) from RHR entry conditions.'

f l NSWS B 3.7.8 q BASES lO i BACKGROUND Additional information about the design and operation of the l (continued) NSWS, along with a list of the components served, is presented in the UFSAR, Section 9.2.1 (Ref.1). The l principal safety related function of the NSWS is the removal of decay heat from the reactor via the CCW Systeni. APPLICABLE The design basis of the NSWS is for one NSWS train, SAFETY ANALYSES in conjunction with the CCW System and a containment spray . system, to remove core decay heat following a design basis I LOCA as discussed in the UFSAR, Section 6.2 (Ref. 2). This ) prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System by the ECCS pumps. The NSWS is designed to perform its function with a single failure of any active component, assuming the loss of offsite power. The NSWS, in conjunction with the CCW System, also cools the unit from residual heat removal (RHR), as discussed in the Ol UFSAR, Section 5.4 (Ref. 3), from RHR entry conditions to MODE 5 during normal and post accident operations. The time required for this evolution is a function of the number of CCW and RHR System trains that are operating. Thirty six hours after a trip from RTP, one NSWS train is sufficient to remove decay heat during subsequent operations in MODES 5 and 6. This assumes a maximum NSWS temperature, a simultaneous design basis event on the other unit, and the loss of offsite power. The NSWS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4). ! LC0 Two NSWS trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power. An NSWS train is considered OPERABLE during MODES 1, 2, 3, and 4 when: l (continued) l l Catawba Unit 1 B 3.7-42 Supplement 1

NSWS B 3.7.8 BASES po BACKGROUND Additional information about the design and operation of the (continued) NSWS, along with a list of the components served, is presented in the UFSAR, Section 9.2.1 (Ref. 1). The principal safety related function of the NSWS is the removal of decay heat from the reactor via the CCW System. APPLICABLE The design basis of the NSWS is for one NSWS train, 1 SAFETY ANALYSES in conjunction with the CCW System and a containment spray system, to remove core decay heat following a design basis LOCA as discussed in the UFSAR, Section 6.2 (Ref. 2). This 1 prevents the containment sump fluid from increasing in i temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System ; by the ECCS pumps. The NSWS is designed to perform its j function with a single failure of any active component, i assuming the loss of offsite p.ower. j The NSWS, in conjunction with the CCW System, also cools the unit from residual heat removal (RHR), as discussed in the l UFSAR, Section 5.4 (Ref. 3), from RHR entry conditions to fs V MODE 5 during normal and post accident operations. The time required for this evolution is a function of the number of CCW and RHR System trains that are operating. Thirty six hours after a trip from RTP, one NSWS train is sufficient to remove decay heat during subsequent operations in MODES 5 and 6.. This assumes a maximum NSWS temperature, a simultaneous design basis event on the other unit, and the loss of offsite power. The NSWS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4). LCO Two NSWS trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions te remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power. An NSWS train is considered OPERABLE during MODES 1, 2, 3, and 4 when: (continued) .O l Catawba Unit 2 8 3.7-42 Supplement 1 l

B BASES APPLICABLE The SAFETY ANALYSES in conjunction with thejCCW System. also cools the unit from residual he (continued) $5AR. Section [5.4@at removal ARHR), as discussed in the I i during normal and post accident operations.(Ref. The time 3)Aentry conditions to MO j required for this evolution is a function of the ni=har af

                    ~A CCW and RHR System trains that are operating rJine rain                       % g .j,n 3 is sufficient to remove decay heat during subsequen 7

ooerations in MODES 5 and 6. mn < n - This assu::rds. aimuwon _ * +we Fa~ (temperature of [951*F. Vrat Inade a d he system.1 occurring simultaneously with maximum 7; L- arf m y The ~7 m mag, .em dl%U satisfies Criterion 3 of tne wmlicy sestd

                                             ~                           lorpe.Tc. 34,4 cesf.44M fj         fu, g s As ske-        )

V ~ a LCO Two SWS tratns are required to be OPERABLE to provide the i 'gnk J krs. less + required redundancy to ensure that the system functions to remove post accident heat loads assuming that the wurst case single of ffsite active failure occur,s coincident with the loss power. An SW5 t).210. is considered OPERABLE during MODES 1. 2, 3 and 4 when:

      @#O                   a. Ilsemar=p is #tMBLE: Anol

b. _

 / ^.                               The associated piping, valves, ht essfUiiEiEB and sgtr i                    instrumentation and controls requirea to perform the safety related function are OPERABLE.

APPLICABILITY In N0 DES 1, 2. 3. and 4. the system that is is a normally rating equipment servi tred to support the OPERABIL of the these MODES. by the_1SWS and required to be OPERABLE in In NODES 5 and 6. the OPERABILITY requirements of t are detenmined by the systems it supports. l ACTIONS

                         &J f

If one SWS train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this Condition. (continued) B 3.7 42 Rev 1. 04/07/95 O

Justificatica ftr Deviatic:s Section 3.7 - Plcnt System O V aeEi NOTE: The first five justifications for these changes from NUREG-1431 were generically used throughout the individual Bases section markups. Not all generic justifications are used in each section.

1. The brackets have been removed and the proper plant specific information or value has been provided.

2. Editorial change for clarity or for consistency with the Improved Technical Specifications (ITS) Writer's Guide.

3. The requirement / statement has been deleted since it is not applicable to this facility. The following requirements have been renumbered, where applicable, to reflect this deletion.

4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

/] 5. The Bases are modified to reflect changes made to the technical (/ specifications.

6. Not used.

I

1. The Bases for NUREG 3.7.8 have been revised to add the missing text for l the Applicable Safety Analysis section consistent with the language ;

contained in NUREG 3.7.7 on RHR entry conditions. This change is retained in ITS 3.7.8. This change is an administrative correction to an omission within the NUREG and does not represent a technical change. O lCatawbaUnits1and2 1+ Supplement 1 5/20/97

/ i leeue Number l 18l Affected Section l3.3.2 ESFAS: 346 Containment Spray j Affected Unite CNS: l_Yes) MNS: [ Yesj Affected Pages ITS: 3.3-31.32 i ITS: 3.3-30. 31 ITSBoese: B 3.3-66: B 3.6 39 1 ITSBases: B 3.3-69: B 3.6-39 CTS: 3/4316 i CTS: 3/4 3-18.19 DOCe: DOCS: NRG: 3.3-33. 34 NRG: 3.3-33. 34 NRG Bases: B 3.a-/o. 77: B 3.647 NRG Boese: B 3.3-76. 77: 0 3.6-87 JFD: TS 3 JFD: TS 3 NSHC: NSNC: Description ITS Table .* 1 4 ' item 2.0 Contohynent Spoy-Monual and item 3.b Phase B-Manual initiation. Required chantrais re'ars to ' . & oin, 2 trains'stilch is incorrect. The design is *1 per train,2 trains'. The ITS Bases dso refers to the 6ncorrect ;; e train,2 trains,'l.e *A monud octuation...two swttches.' Also,3.6.6 Bones refers to hcorrect design. O

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (pa3e 1 of 5) O Engineered Safety Feature Actuation System Instrumentation APPLICAPLE MODES 01 OTetER SPECIFIED RE0diRED SURVEILLANCE ALLOWA8LE TRIP FUNCTION CONDITIONS CHANNEL 5 CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Safety Injection

a. Manual Initiation 1,2,3,4 2 B SR 3.3.2.8 NA NA

b. Autwatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA

              ' Actuation Logic                                                                                                                                    SR    3.3.2.4 and Actuation                                                                                                                                      SR 3.3.2.6 Relays

c. Containment 1,2,3 3 D SR 3.3.2.1 s 1.4 psig s 1.? psig Pressure - High SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

d. Pressurizer 1,2,3(a) 4 D SR 3.3.2.1 a 1839 psig a 1845 psig Pressure - Low SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

2. Containment Spray

a. Manual Initiation 1,2.3.4 1 per B SR 3.3.2.8 NA NA train, 2 l

trains [~h b. Automatic Actuation Logic 1,2,3,4 2 trains C SR 3.3.2.2 SR 3.3.2.4 NA NA and Actuation SR 3.3.2.6 Relays

c. Containment 1,2,3 4 E SR 3.3.2.1 s 3.2 psig s 3.0 psig l Pressure . SR 3.3.2.5 i' High High SR 3.3.2.9 SR 3.3.2.10 !

3. Contairment Isolation

a. Phase A Isolation (1) Manual 1,2,3,4 2 B SR 3.3.2.8 NA NA Initiation (2) Automatic 1.2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (3) Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements.

(continued) (a) Above the P-11 ' Pressurizer Pressure) interlock. Catawba Unit 1 3.3-31 Supplement 1 l

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 bage 2 of 5) Engineered Safety Feature Actuation System Instrumentation

APPLICA8LE MODES OR OTHER SPECIFIED REQUIREO SURVEILLANCE ALLOWA8LE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

3. Containment Isolation (continued)

b. Phase 8 !soletion l (1) Manual Initiation 1,2,3,4 1 per 8 SR 3.3.2.8 NA NA train, 2 trains (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays (3) Containment 1,2,3 4 E SR 3.3.2.1 s 3.2 psig s 3.0 psig Pressure - High SR 3.3.2.5 High SR 3.3.2.9 SR 3.3.2.10

4. Steam Line Isolation

a. Manual Initiation (1) System 1,2(b) 3(b)

                                                                                        ,       2 trains        F                      SR 3.3.2.8         NA                   NA (2) Individual                                             1,2(b)3(b)
                                                                                       ,          1 per        G                       SR 3.3.2.8        NA                    NA line

b. Automatic Actuation 1,2(b)3(b)

                                                                                       ,        2 trains       H                       SR 3.3.2.2        NA                    NA Logic and Actuation                                                                                               SR 3.3.2.4 Relays                                                                                                            SR 3.3.2.6

c. Containment 1,2(b) , 4 E SR 3.3.2.1 s 3.2 s 3.0 psig Pressure - High SR 3.3.2.5 psig High 3(b) SR 3.3.2.9 SR 3.3.2.10

d. Steam Line i Pressure b) ,

(1) Low 3 per 0 SR 3.3.2.1 a 744 psig a 775 psig 3(a)(b) steam SR 3.3.2.5 line SR 3.3.2.9 SR 3.3.2.10 (continued) (a) Above the P-11 (Pressuriter Pressure) interlock. (b) Except when all MSIVs are closed and de-activated. l Catawba Unit 1 3.3-32 Supplement 1

ESFAS Instrumentatien 3.3.2

                                                                            .able 3.3.21 (pace 1 of 5)

Engi w red Safety Feature Actuation System Instrumentation Oi APPLICA8LE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONOITIONS REQUIREMENTS VALUE SETPOINT

1. Safety injection

a. Manual Initiation 1.2.3.4 2 8 SR 3.3.2.8 NA NA

b. Automatic 1.2.3.4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation 5R 3.3.2.6 Relays

c. Containment 1.2.3 3 0 SR 3.3.2.1 s 1.4 psig s 1.2 psig Pressure - High SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

d. Pressurizer 1.2.3(a) 4 0 SR 3.3.2.1 a 1839 psig a 1845 psig Pressure - Low SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

2. Containment Spray

a. Manual Initiation 1.2.3.4 1 per B SR 3.3.2.8 NA NA l train. 2 trains

b. Automatic 1.2.3.4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 j and Actuation Relays SR 3.3.2.6 i

c. Containment 1,2,3 4 E SR 3.3.2.1 s 3.2 psig s 3.0 psig i Pressure - 51 3.3.2.5 High High SR 3.3.2.9 ;

SR 3.3.2.10

3. Containment isolation i

a. Phase A Isolation l (1) Manua1 1.2.3.4 2 B SR 3.3.2.8 NA NA Initiation l l

(2) Automatic 1.2.3.4 2 trains C SR 3.3.2.2 NA NA j Actuation SR 3.3.2.4 l Logic and SR 3.3.2.6 i Attuation Relays (3) Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements. (continued) ! (a) Above the P-11 (Pressurizer Pressure) interlock. I 1 0 Catawba Unit 2 3.3-31 Supplement 1 l

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 5) Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWA8LE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

3. Containment Isolation (continued)

b. Phase 8 Isolation l (1) Manual Initiation 1.2.3.4 1 per B SR 3.3.2.8 NA NA train. 2 trains (2) Automatic 1.2.3.4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays (3) Containment 1.2.3 4 E $R 3.3.2.1 s 3.2 psig s 3.0 psig Pressure High SR 3.3.2.5 High SR 3.3.2.9 SR 3.3.2.10

4. Steam Line Isolation

a. Manual Initiation (1) System 1.2(b) ,3(b) 2 trains F SR 3.3.2.8 NA NA (2) Individual 1.2(b) 3(b)

                                                                   ,                                       1 per       G        SR 3.3.2.8         NA           NA line

b. Automatic Actuation 1.2(b) ,3(b) 2 trains H SR 3.3.2.2 NA NA Logic and Actuation SR 3.3.2.4 Relays SR 3.3.2.6

c. Containment 1.2(b) . 4 E SR 3.3.2.1 s 3.2 s 3.0 psig Pressure - High SR 3.3.2.5 psig High 3(b) SR 3.3.2.9 SR 3.3.2.10

d. Steam Line Pressure (1) Low 3 pe- D SR 3.3.2.1 a 744 psig a 775 psig 3(a)(b) steam SR 3.3.2.5 line SR 3.3.2.9 SR 3.3.2.10 (continued)

(a) Above the P-11 (Pressurizer Pressure) interlock. (b) Except when all MSIVs are closed and de-activated. i d (]& l Catawba Unit 2 3.3-32 Supplement I t .. . . . .

ESFAS Instrumentation B 3.3.2 BASES O APPLICABLE 2. Containment Sorav SAFETY ANALYSES, LCO, and Containment Spray provides two primary functions: APPLICABILITY (continued) 1. Lowers containment pressure and temperature after an HELB in containment; and

2. Reduces the amount of radioactive iodine in the containment atmosphere.

These functions are necessary to:

Ensure the pressure boundary integrity of the containment structure; and a Limit the release of radioactive iodine to the environment in the event of a failure of the containment structure.

The containment spray actuation signal starts the containment spray pumps and aligns the discharge of the pumps to the containment spray nozzle headers in the upper levels of containment. Water is initially drawn from the RWST by the containment spray pumps. Ol When the RWST reaches the low low level setpoint, the spray pump suctions are manually shifted to the containment sump if continued containment spray is required. Containment spray is actuated manually or by Containment Pressure-High High.

a. Containment Sorav - Manual Initiation There are two manual Containment Spray switches, one per train, in the control room. Turning the switch will actuate the associated containment spray train in the same manner as the automatic actuation signal. Two Manual Initiation l switches, one per train, are required to be OPERABLE to ensure no single failure disables the Manual Initiation Function. Note that Manual Initiation of containment spray also actuates Phase B containment isolation. Two train actuation requires operation of both Train A and Train B manual containment spray switches.

(continued) l Catawba Unit 1 B 3.3-66 Supplement 1

I ESFAS Instrumentatien f B 3.3.2 BASES O APPLICABLE 2. Containment Sorav SAFETY ANALYSES, LCO, and Containment Spray provides two primary functions: APPLICABILITY (continued) 1. Lowers containment pressure and temperature after an HELB in containment; and

2. Reduces the amount of radioactive iodine in the containm2nt atmosphere.

These functions are necessary to:

Ensure the pressure boundary integrity of the containment structure; and

                           . Limit the release of radioactive iodine to the environment in the event of a failure of the containment structure.

The containment spray actuation signal starts the containment spray pumps and aligns the discharge of the pumps to the containment spray nozzle headers in the upper levels of containment. Water is initially drawn from the RWST by the containment spray pumps. (' l When the RWST reaches the low low level setpoint, the spray pump suctions are manually shifted to the containment sump if continued containment spray is required. Containment spray is actuated manually or by Containment Pressure-High High.

a. Containment Sorav - Manual Initiation There are two manual Containment Spray switches, l one per train, in the control room. Turning the switch will actuate the associated containment spray train in the same manner as the automatic actuation signal. Two Manual Initiation l switches, one per train, are required to be OPERABLE to ensure no single failure disables the Manual Initiation Function. Note that Manual

! Initiation of containment spray also actuates Phase B containment isolation. Two train actuation requires operation of both Train A and Train B manual Containment Spray switches. (continued) O 8 3.3-66 Supplement 1 l Catawba Unit 2

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ESFAS Instrumentation 3.3.2 tehte 3.3.2 1 case. 2 of si Enettuared Safety Feature Actation System Ittetrtmanntation Apeticants m aEs et OfW BsGCIFim agallem W im SWRVE!LLAst* ALLindhSLE ft!P telfIng reanum a CibSIfles ageWItemets Waus 8tts01 f 1, Safety injeetten ~ teentfresad)

s. sich Stems F in 1,2,3883 2 por e SR 3.3.2.1 fue stems L to) (f) stone sa 3.3.2.S Itno se 3.3.2.9 sa 3.3.2.10 Colne utth 1,2,3(d3 1 per 9t 4tne se 3J.2.1 a tab 5)(c t (6751 pets stems sa 3.3.2.5 pole

                             = Leu                             tine                     = 3J.2.9

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2. Centalment sprer

a. menet inteletten 1,2,3,4 e se 3J.2.5 un na trotne

b. Autamatie 1,2,3,4 2 trotne Actustlen Legie C at 3.3.2.2 un an and Actuetten et 3.3.2.6 Rotere SR 3.3.2.6

s. Centalw

                   *reemde 9 1,2,3             4             e        at 3.3.2.1 9                                                        = 3.3.2.

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                                                                                       = 3aJ.,s             ...               .

at 3.3.2.10 'Q uten =3 ti PtenteI esp 1,2,3 este a at 3.3.2.1 s (12.333 s 12.053 of (23 3.3.2.5 peie pmie 3.3.2.9 at 3.3.2.10 teentivesed) Os fa3 ese guer' C .. settiedeleer used s spectfts tapt the witt. Tene may centefh Alleuchte Welue depend en Setpoint 5t Ic) Time eerstente in the leadflag retter are t, t [d) Abewe the P.1 f., = Lees Law) inter accende arid tg 3 (5) . te) Lees then L to e fwietten ined as W 9 Ire Linmorty free (441 futt stems fleu et ;ns to 16431 fuit atoes tem heteu [201E teed. and to - . . 2:ng te (11411 . . E tend to (11411 fuit one fleu et (10011 L (f) Lees t i etene fleu shove SOE teed. er espast to a fisicti defined me p . . 2;ng to (4011 futt see flou'aetusen (03E Lead (203E (_ 11 M E then toed.a- e inereest Linearly free (403%4tems fleu et (201E tend e (1103E futt etese ft et 'j l 3.3 33 Rev 1. 04/07/95 CamM i i i

ESFAS Instrumentation 3.3.2 febte 3.3.2 1 (mees 3 et 33 tragineered Safety Feeture Actuotten Systes Instrueentation A99LitA8Lt last8 CR 0f188 sPECIFisD WSJIRm stayE!LLANCE AL' rhael 8 ft! PtalCTION Caelflees CunemLB CeBSITime meest w f5 mLe WTFOI

3. Centeemmet leetetten

e. Phase A leetetten (1) nerosol 1,2,3,4 2 e at 3J.2.8 m m innttetten (2) Auteemets 1,2,3,4 2 treiro C st 3.3.2.2 m a4 Actetten sa 3.3.2.4 Lesle and at 3J.2.6 Astuotten )

teleye (3) sefety eefer to twetion 1 (sofety injestlan) for ett inteletten injectlen twettens and reestrensits.

h. Phase e testetten (1) marsal 1,2,3,4 e m 3.3.2.8 m Inittetten m n, 2 .,

trotru (2) Auteestle 1,2,3,4 2 treine C et 3.3.2.2 m me Astetten Laple are a 3.3.2.4 Actuotten sa 3.3.2.4 Betsys (3) Centainment preneure hO& 1,2,3 I4[ E a 3.3.2.1 1 s lii:: st 3J.2.10 4. N_b5km j Stees Line lastatten 1,2( h noroastinitiet6eni .' I 2 F ist 3.3.2.8 m m

b. Auteestle 1,2 3,3( ) 2 treine su 3J.2.2 m M Actuation togts at 3J.2.4 and Actuotten M 3J.R.6 setsys

(*d) b l',y.) ,l s a N I N* remd) h y g f ee t_ _..cnone soy sprite 6n enty AttaueDWyetum no err setpelkt

    ) Encept ihn ett mive are etened and              activated

WGMTS 3.3 34 Rev 1. 04/07/95 cal *h s

l Justific tia fcr Deviatiras Section 3,3 - I strumentatta G h TECHNICAL SPECIFICATIONS

15. A surveillance has been added to ITS 3.3.3 to perform channel calibrations on the hydrogen monitors every 92 days. The CTS currently requires performance of an ACOT every 31 days and channel calibration I 92 days on a staggered test basis. Operating history for this '

equipment does not support an 18 month surveillance interval. Therefore, the ACOT is deleted and replaced by a channel calibration once per 92 days. j I

16. Not used.

l

17. NUREG Surveillance Requirement 3.3.4.2 is deleted. The components comprising the remote shutdown system at Catawba cannot be individually {

transferred to the remote shutdown panel. The transer is accomplished using a " bulk" component transfer switch. This results in a complete transfer of control from the control room. Routine testing of transfer and control switches, therefore, rannot be accomplished without placing i a significant burden on the statien during outage conditions. This requirement is not contained withia the current TS and is not proposed for inclusion within the ITS. i d 18. NUREG LCO 3.3.7, Action B.2 which permits placing both control rooin filter trains in operation when the actuation instrumentation is inoperable is deleted. The design at Catawba precludes the starting of individual components such as the filter trains and pressurizing fans. The entire train, including air handlers and chillers are started using a key switch which permits only one train in operation at any time. l Upon receipt of a safety injection signal, the pressurizing fans and filter train which is not inservice are started, however, this action cannot be accomplished manually without using the key switch. Noise levels in the control room when two full air handling trains are in operation make this option undesirable. The alternatives provided in NUREG 3.3.7 actions B.1.1 and B.1.2 are sufficient.

19. NUREG LCO 3.3.2, Table 3.3.2-1, Functions 2.0 and 3.b reflect containment spray and phase B isolation manual initiation designs which utilize two switches per train. ITS Table 3.3.2-1 is revised to reflect that Manual initiation of containment spray and phase B ,

isolation is accomplished via individual train specific switches. ! O 1 lCatawbaUnits1and2 33 Supplement 15/20/07

ESFAS Instrumentation B 3.3.2 O - APPLICABli 2. Containment Sorav SAFETY ANALYSES. LCO. and ' Containment Spray provides primary functions: APPLICABILI1Y (continued) 1. Lowers contairment pressure and temperature after an IELB in containment:@

2. Reduces the amount of radioactive iodine in the contairment a

1. . we pn or water in the ir d h a[uSlations r fter a large - LOCA. ,

These functions are necessary to: Ensure the pressure boundary integrity of the containment structurw: g Limit the release of radioactive iodine to the envirorust in the event of a failure of the contairment strtsturegg!gg) P Minia correston or 4 ..s anuAystem)s

       .                                                                             ( ins              containment , lowing a LOCA. /                   >

The containment spray actuation signal starts the containment spray pups and aligns the discharge of O the pumps to the containment spray nozzle headers in the upper levels of corttainment. Water is initia11 drawn from the RMST by the contairunent spray m rei En V soG1 m-- m ==>mian a rns t ive td*1 Mun tie RMST reaches the low low-level setpoint, the spray pup suctions are shifted to the containment samp if continued contairment spray is required h r yressure. Contairunent spray is actuated manuallypy yssent Friessure-pran 3 AlContairment

                                                                                                   -mgn myl.                                             Q@

a. fatainment Sorav-Manual Initiation FThe up= can initiate sai.- 6 spr any time the cont.ao room by simulta 1 k iturning contai snra_v_ actuation th

                                                                                            \in t       = + " % > Because I                                      ineuvertent actuati       f contairment      ay cmuld have      ch seri      consequences,        switches must       turnedj (continued)

W)lPS1T B 3.3 76 Rev 1. 04/07/95 C=cfm4 5 O l 1

1 1 ESFAS Instrumentation 1 8 3.3.2 BASES APPLIClati a. Containment Soray-Manual Initiation (continued) SAFETY ANALYSES. LCO. and APPLICABILITY {c pg6i= controi d * - - - d v *a dai+4 roomT W . are w qswit t* * = t=4- --/ w s h qm p dn the we y

                                                             .-_ A              ing the me 3           -

spra n manner automatic actuation si 1. Two Manual M ] par ,- Initiation swi train rare required to

u. urt-xx to ensure no single failure disables 4

NM8 */&*4% M O.* b. the Manual Initiation Function. Note that Manual Initiation of containment spray also actuates Y Tees *e $ noj 'ftas$ Phase B containment isolation d i

       #tedwal htA A            . Containment Soray-Anda=atic Actuation tmic and                    s dfreySM$g&ag,~                Actuation Relavs
           ~'                                                                                            l Automatic actuation logic and actuation relays
 /

consist of the same features and operate in the se e manner as described for ESFAS Function 1.b. Manual and autobic initiation of containment spray must be OPERABLE in MI)ES 1. 2. andJ what

     -                             there is a potential for an accident to occur.

and sufficient in the primary or secondary systems to pose a St to contairulant integrity to overpr====w conditions.1 , i nitiatioris also twoIItred in MG)E . ever I M M enmatie deeda'tinn 4e nne rirwt ' In '

                             @ s.agy    nuutar  adequate     time   is available act.uate required components in the event of a  to manually DBA. However, because of the large nuutier of                             4 I

components actuated on a containment spray, actuation is simplified by the use of the manual actuation push buttons. Automatic actuation 4 l logic and actuation relays must be OPERABLE in MODE 4 to support system level manual initiation. In MODES 5 and 6. there is insufficient energy in the primary and secondary systems to result in containment overpressure. In MODES 5 and 6 there is also te time for the o evaluate unit itions and respond,perators to to sitigate the consequences of abnormal conditions by manually starting individual wwds. (continued) WQG4TS B3.37f Rev 1. 04/07/95 Calawl. O

Containment Spray System B 3.6.6 O BASES (continued) BACKGROUND capability of the Containment Spray System during the (continued) injection phase. In the recirculation mode of o>eration, heat is removed from the containment sump water z Containment Spray System and RHR heat exchangers.y Each the train of the Containment Spray System, supplemented by a train of RHR spray, provides adequate spray coverage to meet the system design requirements for containment heat removal. For the hyp3thetical dounle-ended rupture of a Reactor coolant System pipe, the pH of the sump solution (and, consequently, the spray solutto.i) is raised to at least 8.0 within one hour of the onset of the LOCA. It is possible to adjust the pH of the sump solution using the chemical mixing tank and the charging pumps, should it become necessary. The alkaline pH of the containment sump water minimizes the evolution of iodine and the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid. The Cor.tainment Spray System is actuated either automatically by a containment pressure high-high signal or manually. An automatic actuation opens the containment spray pump discharge valves, starts the two containment spray pumps, and begins the injection phase. A manual (4ru,a rela d actuation of the Containment Spray System requires the

               ~    operator to actuate two separat*Wwitches on the main control board to beo<n tne samt senuencet' The injection
 ' oC N b *e }      phase cont'nues until     an RWST ' eve' Low-Low alarm is e at b h b ]       received. The Low-Low alam for the RWST signals the operator to manually align the system to the recirculation mode. The Containment Spray System in the recirculation mode maintains an equilibriin temperature between the containment atmosphere and the recirculated sump water.

Operation of the Containment Spray System in the recirculation mode is controlled by the operator in accordance with the emergency operation procedures. The RHR spray operation is initiated manually, when required by the emergency operating Core Cooling System (ECCS) procedures, is operating afterrecirculation in the the Emergency mode. The RHR sprays are available to supplement the Containment Spray System, if required, in limiting containment pressure. This additional spray capacity would typically be used after the ice bed has been depleted and in the event that containment pressure rises above a (continued) O Catawba Unit 1 B 3.6-39 Svpplem 4 1 5/2^/^7

Containment Spray System-B 3.6.6 BASES (continu BACKGROUND (continued \ s of the Containment Spray System during the

                                <n phase.

In the recirculation mode of o>eration, amoved from the containment sump water a

.t*

mt Spray System and RHR heat exchangers.y the trat. the Containment Spray System, supplemented by a Each l train .i RHR spray, provides adequate spray coverage to meet the system design requirements for containment heat removal. For the hypothetical double-ended rupture of a Reactor Coolant System consequently, t sipe, the pH of the sump solution (and, w spray solution) is raised to at least 8.0 within one hour of the onset of the LOCA. It is possible to { adjust the pH of the sump solution using the chemical mixing tank and the charging pumps, should it become necessary. The alkaline pH of the containment sump water minimizes the evolution of iodine and the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid. The Containment Spray System is actuated either automatically by a containment pressure high-high signal or manually. An automatic actuation opens the containment spray pump discharge valves, starts the two containment m spray pumps, and begins the injection phase. A manual inm, rt W ) actuation of the Containment Spray System ~ requires the

                 '  anarator to actuate two separaterswitches on the main "o f M b.^         control board to haain the gaRTsequenc# The injection phase continues until an RWST '

t Ath% evel Low-Low alam is received. The Low-Low alarm for the RWST signals the

         -          operator to manually align the system to the recirculation mode.

The Containment Spray System in the recirculation mode maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. Operation of the Containment Spray System in the recirculation mode is controlled by the operator in accordance with the emergency operation procedures. The RHR spray operation is initiated manually, when required by the emergency operating Core Cooling System (ECCS) isprocedures, after operating in thethe Emergency recirculation mode. The RHR sprays are available to supplement the Containment Spray System, if required, in Itaiting containment pressure. This additional spray capacity would typically be used after the ice bed has been depleted and in the event that containment pressure rises above a (continued) Catawba Unit 2 B 3.6-39 l d Jurfem 4/20/ E

/ Contairnent Spray System ML-W@ B 3.6.6@ @ BASES BACKGROUM) determining the heat removal capability of the Containment (continued) Spray System during the injection phase. In the recirculation mode of operation, heet is removed from the containment stmip water by the Containment Spray System and RHR heat exchangers. Each train of the Containment Spray System, supplemented by a train of RHR spray, provides adequate spray coverage to meet the system design requirements for containment heat removal. f The Spray tive System injects a sodiisa hydro de(NaOH f solution i o the spray. The resulting alkali pH of the

     %/44#                   spray         es the ability of the spray to s enge iodine B4cx44,uuo           fission odtr;ts from the containment atmos           e. TheNaOH;l added       the spray also ensures an alkali pH for the s lut'      recirculm+ d da *ha ennt=i m .       =a l The alkaline pH of the containment sum water minimizes the evolution of iodine and the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid.                gg The Containment Spra automaticallybya/ySystemisactuatedeithe)r ontainment 40 rid fressurelsignal or             @

manually. An automatic actuation opens the containment

]                           Spray pump discharge valves, starts the two containment

'j . spray ptmps, and begins the injection phase. A manual fwd L _MaIf h actuation of the Contairment Soray System requires the r 4 (d Iw o M 8a, w o wr w actuate two separasa gswitches on the main control board to begin the mW The injection pnose cononues unn i an mi seves Low-Low alarm is h ' Lac W received. iyes TheLow-LowalarmfortheRWSifactuatge+=4 waa n.-.sjpray pump suetion to the al 3yve _] signals the operator to manually align the sy em to the recirculation mode. The Containment Spray System in the recirculation mode maintains an equilibrita temperature between the containment atmosphere and the recirculated sum water. Operation of the Containment Spray Systen in the recirculation mode is controlled by the operator in accordance with the emergency operation procedures. The RHR spray operation is initiated manually, when required by the emer cy operating Core Cooling, stem (ECCS) procedures, is operating after in the the Emergency recirculation mode. The MM sprays are available to supplement the Containment Spray System, if required, in limiting (continued) E CS- B 3.6-87 Rev 1. 04/07/95 a ' i i

I feeue Nurnber [ 22j ANected Section ICTS SR 4 8.1.1.2.c.4 i ANocted UnNs CNS: l Yesj MNS: l Yes} ANected Pogos ITS: ITS: ITS Boese: B 3.8-16 ITS Bones. B 3.8-16 1 CTS: I CTS: DOCS: LA-1 i DOCS: LA-1,2 NRG: NRG: NRG Bones: B 3.816 NRG Boeos: B 3.8-16 JFD: JFD: NSHC: NSHC: Dordotion Revised DOC LA-4 and Bmes for Section 3.8 to rnove DG storting signols to Bones for SR 3.8.1.2 and 3.8.1.7. This is the oppropriate locotton for this material rather than the SLC.

1 1 i

AC S:urces-Operating B 3.8.1 BASES U) SURVEILLANCE Where the SRs discussed herein specify voltage and frequency REQUIREMENTS tolerances, the following is applicable. The minimum steady (continued) state output voltage of 3740 V is 90% of the nominal 4160 V output voltage. This value allows for voltage drop to the terminals of 4000 V motors whose minimum operating voltage is specified as 90% or 3600 V. It also allows for voltage drops to motors and other equipment down through the 120 V level where minimum operating voltage is also usually specified as 90% of name plate rating. The specified minimum and maximum frequencies of the DG are 58.8 Hz and 61.2 Hz, respectively. These values are equal to

2% of the 60 Hz nominal frequency and are derived from the recomendations given in Regulatory Guide 1.9 (Ref. 3).

SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads

 ,m                                                        are connected to their preferred power source, and that appropriate independence of offsite circuits is maintained.

(U) The 7 day Frequency is adequate since breaker position is not likely to change without the operator being aware of it and because its status is displayed in the control room. SR 3.8.1.2 and SR 3.8.1.7 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition. l To minimize the wear on moving parts that do not get l lubricated when the engine is not running, these SRs are modified by a Note (Note 2 for SR 3.8.1.2) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior ; to loading. For the purposes of SR 3.8.1.2 and SR 3.8.1.7 testing, the DGs are started from standby conditions using a manual start, loss of offsite power signal, safety injection signal, or loss of offsite power coincident with a safety G (continued) U l Catawba Unit 1 B 3.8-16 Supplement 1

AC Sources-Operating B 3.8.1 O BASES V SURVEILLANCE SR 3.8.1.2 and SR 3.8.1.7 (continued) REQUIREMENTS injection signal. Standby conditions for a DG mean that the l diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recomendations. In order to reduce stress and wear on diesel engines, the manufacturer recomends a modified start in which the starting speed of DGs is limited, warmup is limited to this lower speed, and the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 3, which is only applicable when such modified start procedures are recomended by the manufacturer. SR 3.8.1.7 requires that, at a 184 day Frequency, the DG starts from standby conditions and achieves required voltage and frequency within 11 seconds. The 11 second start requirement supports the assumptions of the design basis LOCA analysis in the UFSAR, Chapter 15 (Ref. 5). The 11 second start requirement is not applicable to ) (' q SR 3.8.1.2 (see Note 3) when a modified start procedure as s described above is used. If a modified start is not used, the 11 second start requirement of SR 3.8.1.7 applies.- Since SR 3.8.1.7 requires a 11 second start, it is more restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2. This is the intent of Note 1 of SR 3.8.1.2. l The normal 31 day Frequency for SR 3.8.1.2 is consistent with Regulatory Guide 1.9 (Ref 3). The 184 day Frequency for SR 3.8.1.7 is a reduction in cold testing consistent with Generic Letter 84-15 (Ref. 8). These Frequencies provide adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing. SR 3.8.1.3 This Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads greater than or equal to the equivalent of the maximum expected accident loads. A minimum run' time f. f 60 minutes is required to stabilize engine temperatures, j (continued) Catawba Unit 1 B 3.8-17 Supplement 1 l 1

l AC Sources-Operating B 3.8.1 BASES SURVEILLANCE Where the SRs discussed herein specify voltage and frequency REQUIREMENTS tolerances, the following is applicable. The minimum steady (continued) state output voltage of 3740 V is 90% of the nominal 4160 V output voltage. This value allows for voltage drop to the terminals of 4000 V motors whose minimum operating voltage is specified as 90% or 3600 V. It also allt.as for voltage drops to motors and other equipment down through the 120 V level where minimum operating voltage is also usually specified as 90% of name plate rating. The specified minimum i and maximum frequencies of the DG are 58.8 Hz and 61.2 Hz, respectively. These values are equal to i 2% of the 60 Hz nominal frequency and are derived from the recomendations given in Regulatory Guide 1.9 (Ref. 3). SR 3.8.1.1 j This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their preferred power source, and that O appropriate independence of offsite circuits is maint3tned. The 7 day Frequency is adequate since breaker positior, is not likely to change without the operator being aware of it l and because its status is displayed in the control room. '

SR 3.8.1.2 and SR 3.8.1.7 i

These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition. l To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 2 for SR 3.8.1.2) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior to loading. For the purposes of SR 3.8.1.2 and SR 3.8.1.7 testing, the DGs are started from standby conditions using a m ual start, loss of offsite power signal, safety inju :n signal, or loss of offsite power coincident with a safety i p (continued) V l Catawba Unit 2 B 3.8-16 Supplement 1 ; 1

AC S:urces-Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 and SR 3.8.1.7 (continued) REQUIREMENTS injection signal. Standby conditions for a DG mean that the l diesel engine coolant and oil are being conting'usly circulated and temperature is being maintained consistent with manufacturer recommendations. In order to reduce stress and wear on diesel engines, the manufacturer reconsnends a modified start in which the starting speed of DGs is limited, warmup is limited to this lower speed, and the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 3, which is only applicable when such modified start procedures are reconinended by the manufacturer. SR 3.8.1.7 requires that, at a 184 day Frequency, the DG starts from standby conditions and achieves required voltage and frequency within 11 seconds. The 11 second start requirement supports the assumptions of the design basis LOCA analysis in the UFSAR, Chapter 15 (Ref. 5). The 11 second start requirement is not applicable to SR 3.8.1.2 (see Note 3) when a modified start procedure as CO described above is used. If a modified start is not used, the 11 second start requirement of SR 3.8.1.7 applies. Since SR 3.8.1.7 requires a 11 second start, it is more restrictive than SR 3.8.1.2, and it may be perfonned in lieu of SR 3.8.1.2. This is the intent of Note 1 of SR 3.8.1.2. The normal 31 day Frequency for SR 3.8.1.2 is consistent with Regulatory Guide 1.9 (Ref. 3). The 184 day Frequency for SR 3.8.1.7 is a reduction in cold testing consistent with Generic Letter 84-15 (Ref. 8). These Frequencies provide adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing. 1 SR 3.8.1.3 This Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and l accepting loads greater than or equal to the equivalent of { the maximum expected accident loads. A minimum run time of I 60 minutes is required to stabilize engine temperatures, i c (continued) i Catawba Unit 2 B 3.'t-17 Supplement 1 l

Discussion cf Ch:nges Section 3.8 - Electrical Power Systems n TECHNICAL CHANGES - REMOVAL OF DETAILS LA.1 Not used. LA.2 CTS 3.8.1.1 Action g requires a DG operating at greater than 5750 kW be reduced to 5 5750 kW within one hour. This requirement has l been moved from the TS to the Selected Licensee Commitments (SLC) (UFSAR Chapter 16) consistent with NUREG-1431. This level of detail is not necessary for inclusion within the TS. Changes to the SLC are reviewed and controlled in accordance with the requirements of 10 CFR 50.59. These controls are adequate to assure any change is properly reviewed. LA.3 CTS 3.8.1.1 Action h and SR 4.8.1.1.2d requires the Cathodic Protection System to be OPERABLE for a DG. These requirements have been moved from the TS to the Selected Licensee Commitments (SLC)(UFSAR Chapter 16) consistent with NUREG-1431. This level of detail is not necessary for inclusion within the TS. Changes to the SLC are reviewed and controlled in accordance with the i requirements of 10 CFR 50.59. Thsse controls are adequate to l assure any change is properly reviewed. O O LA.4 CTS SR 4.8.1.1.2a 4) requires the DG to be started for a test by one of four methods. The requirement has been moved from the TS to the Bases for ITS 3.B.J.Sciccted Licen:ce C ritment: (SLC) (UFSA" Ch;ptcr 15) consi; tent with "UREC 1931. This level of detail is not necessary for inclusion within the TS. Requirements for operability of the DG are still retained in ITS 3.8.1. Changes to the RGBases are reviewed and controlled in accordance with the requirements of the ITS Section 5.0 Bases Control Program and require an evaluation in accordance with 10 CFR 50.59. These controls are adequate to assure any change is properly reviewed. LA.5 Not used. CTS SR 4.3.1.1.2: 5) require; th: DC tc be verified aligned to the :: ciated cmcrgency bus. The requirement h:: been meved frc the TS tc the Selected Licen:cc C =it=cnt: (SLC) (UFSARChapter15) consistent with NURSC 1431. This level Of detail is not nece;;;ry for inclusien within the TS. Ch;nge; to the SLC are reviewed and controlled in accordance with the requirc cnt; cf 10 CFR 50.50. Tht:c controh arc adequate to

urc any change is properly revicwed.

m U l lCatawbaUnits1and2 Page LA - 18 5/20/07 Supplement 1 i i l

AC Sourcss-Operating 8 3.8.1 O -s SUR4EILLANCE REQUIREMENTS (continued) (hNSM84.114ef.11)) allows for voltage drop to the tea 6nais or 4000 y motors whose minimum operating voltage is specified as 90% or 3600 V. It also allows for voltage h drops to motors and other equipment down through the 120 V level where minimum operating voltage is also usually

                 ,   specified as ear of nlate ratina._ nne.specmn

[he i ana a no f ma phe t s that for ghtly loacea di ibution sys t tage at the inals of 4000 is no .} t anxima ra .ooeratino vol .I The smetried minima and saximum rrequencies or tne DE art 58. I Hz and 61.2 Hz. respectively. These values are 1 to i 2% of

                   'the 60 Hz nominal frequency and are deri          from the recomeendations given in Regulatory Guide 1.9 (Ref. 3).

SR 3.8.1.1 This SR ensures proper circuit continuity for the.offsite E electrical power supply to the onsite distribution network and availability of offsite E electrical power. The breaker alignment verifies that each breaker is in 13s correct position to ensure that distribution buses and loads are connected to their preferred power source, and that appropriate independence of offsite circuits is maintained. The 7 day Frequency is adequate since breaker position is not likely to change without the operator being aware of it and because its status is displayed in the control room. SR 3.8.1.2 and SR 3.8.1.7 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition. To minimize the wear on moving parts that do not get lubricated when the engine is not running, these $Rs are modified by a Note (Note 2 for SR 3.8.1.2) to indicate that all OG starts for these Surve111ances any be preceded by an engine prelube period and followed by a warsup period prior to loading. For the purposes of SR 3.8.1.2 and SR 8.1.7 testing. the DGs are started from staney condition Stan@y conditions (continued) WpirTTS ' 8 3.8 16 Rev 1. 04/07/95 CAhAN IMthg Athomm.\I M ,Ioss3-C eCtrik pte SiiA sab Af d,4, r,f4,,. liu e .W4 P w Cadd64 w M ,s ,gecJ,,:,da.l. O w J

l l leeue Number l 30J Affected Section l3;3.1, 3.3.2, 3.3.3 Instrumentaten i ANected Unlle CNS: l Yes) MNS: l Yesl Affected Popes ITS: 1 ITS: 1 ITS Saees: B 3.3-33. 88.120 l ITS bases: B 3.3-34. 90.119 CTS: I CTS: DOCS: 1 DOCS: N#G: NRG: NRG Bases: 8 3 3-37. 104,133 NRG Bases: 8 3.3-37,104.133

                     #D:           B2. B3                                   # D:           B2,B3                           '

NSHC: NSHC: Descripilon Thrd porograph of the ES'AS ACTIONS Bones indcotes that functons wNch are on a per item basit e.g., per $d, per steam line, etc., are allowed separate condtton entry. This clonfies that the octual function is singular. That It the function

Steam Line Pressure Low"is o separate function for each steam line. This is consistent wtth the CTS, orthough in a different format. The CTS uses toths and totd channels and mrwnum channels to gNe the some meaning. Revise the RTS Bases and PAM Bases to odo this some clarfficotton wNch is applicab6e to these functions consistent with the CTS.

1 I i l l l

RTS Instrumentation B 3.3.1 BASES ( APPLICABLE 19. Automatic Trio Loaic SAFETY ANALYSES, LCO, and The LC0 requirement for the RTBs (Functions 17 and 18) APPLICABILITY and Automatic Trip Logic (Function 19) ensures that (continued) means are provided to interrupt the power to allt,w the rods to fall into the reactor core. Each RTB is equipped with an undervoltage coil and a shunt trip coil to trip the breaker open when needed. Each RTB is equipped with a bypass breaker to allow testing of the trip breaker while the unit is at power. The reactor trip signals generated by the RTS Automatic Trip Logic cause the RTBs and associated bypass breakers to open and shut down the reactor. The LCO requires two trains of RTS Automatic Trip Logic to be OPERABLE. Having two OPERABLE channels ensures that random failure of a single logic channel will not prevent reactor trip. These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip Functions must be OPERABLE when the RTBs and associated bypass breakers are closed, and the CRD System is capable of rod withdrawal. The RTS instrumentation satisfies Criterion 3 of 10 CFR 50.36 (Ref. 6). ACTIONS A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.1-1. When the Required Channels in Table 3.3.1-1 are specified (e.g., on a per steam line, per loop, per SG, etc., basis), then the Condition may be entered separately for each steam line, loop, SG, etc., as appropriate. In the event a channel's Trip Setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument loop, signal processing electronics, or bistable is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the LCO Condition (s) entered for the protection Function (s) affected. (continued) V Catawba Unit 1 B 3.3-33 Suppleme n t1 l

ESFAS Instrumentation B 3.3.2 BASES (3 APPLICABLE 10. Nuclear Service Water System Suction Transfer - Low SAFETY ANALYSES, Pit Level LCO, and , APPLICABILITY Upon an emergency low pit level signal from either NSWS pit, interlocks isolate the NSWS from Lake Wylie, align NSWS to the standby nuclear service water pond, ; close particular crossover valves, and start the NSWS pumps. This function is initiated on a two-out-of-three logic from either NSWS pump pit. This function must be OPERABLE in MODES 1, 2, 3, and 4 to ensure cooling water remains available to essential components during a DBA. In MODES 5 and 6, the sufficient time exists for manual operator action to l realign the NSWS pump suction, if required. The ESFAS instrumentation satisfies Criterion 3 of 10 CFR 50.36 (Ref. 6). ACTIONS A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each O Function listed on Table 3.3.2-1. When the Required V Channels in Table 3.3.2-1 are specified (e.g., on a per steam line, per loop, per SG, etc., basis), then the Condition may be entered separately for each steam line, , loop, SG, etc., as appropriate. ' In the event a channel's Trip Setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument Loop, signal processing electronics, or bistable is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the l LCO Condition (s) entered for the protection Function (s) l affected. , When the number of inoperable channels in a trip function exceed those specified in one or other related Conditions associated with a trip function, then the unit is outside the safety analysis. Therefore, LCO 3.0.3 should be intnediately entered if applicable in the current MODE of operation. (continued) \_/ l Catawba Unit 1 B 3.3-88 Supplement 1

PAM Instrumentaticn B 3.3.3 BASES ACTIONS Note 1 has been added in the ACTIONS to exclude the MODE change restriction of LCO 3.0.4. This exception allows entry into the applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require unit shutdown. This exception is acceptable due to the passive function of the instruments, the operator's ability to respond to an accident using alternate instruments and methods, and the low probability of an event requiring . these instruments. Note 2 has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed on Table 3.3.3-1. When the Required Channels in Table 3.3.3-1 are specified (e.g., on a per steam line, per loop, per SG, etc., basis), then the Condition may be entered separately for each leam line, loop, SG, etc., as appropriate. The Completion Time (s) of the inoperable channel (s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function. A.d

    )                              Condition A applies to all PAM instrument Functions.
 's                                Condition A addresses the situation when one or more required channels for one or more Functions are inoperable.

The Required Action is to refer to Table 3.3.3-1 and take the appropriate Required Actions for the PAM instrumentation affected. The Completion Times are those from the referenced Conditions and Required Actions. L1 Condition B applies when one or more Functions have one required channel that is inoperable. Required Action A.1 requires restoring the inoperable channel to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. p) (continued) l Catawba Unit 1 B 3.3-120 Supplement 1

RTS Instrumentation B 3.3.1 BASES C APPLICABLE 19. Automatic Trio Loaic SAFETY ANALYSES, LCO, and The LCO requirement for the RTBs (Functions 17 and 18) APPLICABILITY and Automatic Tri? Logic (Function 19) ensures that (continued) means are provideo to interrupt the power to allow the rods to fall into the reactor core. Each RTB is equipped with an undervoltage coil and a shunt trip coil to trip the breaker open when needed. Each RTB is equipped with a bypass breaker to allow testing of the trip breaker while the unit is at power. The reactor trip signals generated by the RTS Automatic Trip Logic cause the RTBs and associated bypass breakers to open and shut down the reactor. The LC0 requires two trains of RTS Automatic Trip Logic to be OPERABLE. Having two OPERABLE channels ensures that random failure of a single logic channel will not prevent reactor trip. These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip Functions must be OPERABLE when the RTBs and associated bypass breakers are closed, and /~'N the CRD System is' capable of rod withdrawal. O The RTS instrume'tation satisfies Criterion 3 of 10 CFR 50.36 (Ref. 6). I ACTIONS A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.1-1. When the Required Channels in Table 3.3.1-1 are specified (e.g., on a per ! steam line, per loop, per SG, etc., basis), then the l Condition may be entered separately for each steam line, loop, SG, etc., as appropriate. In the event a channel's Trip Setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument loop, signal processing electronics, or bistable is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the LCO Condition (s) entered for the protection Function (s) affected. (continued) Catawba Unit 2 B 3.3-33 Supplement 1 l

ESFAS Instrumentaticn B 3.3.2 BASES O O APPLICABLE 10. Nuclear Service Water System Suction Transfer - Low SAFETY ANALYSES, Pit Level LCO, and APPLICABILITY Upon an emergency low pit level signal from either NSWS pit, interlocks isolate the NSWS from Lake Wylie, o align NSWS to the standby nuclear service water pond, close particular crossover valves, and start the NSWS pumps. This function is initiated on a two-out-of-three logic from either NSWS pump pit. This function must be OPERABLE in MODES 1, 2, 3, and 4 to ensure cooling water remains available to essential components during a DBA. In MODES 5 and 6, the sufficient time exists for manual operator action to realign the NSWS pump suctior., if required. The ESFAS instrumentation satisfies Criterion 3 of 10 CFR 50.36 (Ref. 6). ACTIONS A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed on Table 3.3.2-1. When the Required

Channels in Table 3.3.2-1 are specified (e.g., on a per steam line, per loop, per SG, etc., basis), then the Condition may be entered separately for each steam line, loop, SG, etc., as appropriate.

In the event a channel's Trip Setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument Loop, signal processing electronics, or bistable is found inoperable, then all affected Functions provided by that channel must ce declared inoperable and the LC0 Condition (s) entered for the protection Function (s) l affected. When the number of inoperable channels in a trip function i exceed those specified in one or other related Conditions i associated with a trip function, then the unit is outside i the safety analysis. Therefore, LC0 3.0.3 should be i imediately entered if applicable in the current MODE of operation. (continued) l Catawba Unit 2 B 3.3-88 Supplement 1

PAM Instrumentation ( B 3.3.3 BASES ACTIONS Note 1 has been added in the ACTIONS to exclude the MODE change restriction oi LCO 3.0.4. This exception allows entry into the applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require unit shutdown. This exception is acceptable due to the. passive function of the instruments, the operator's ability to respond to an accident using alternate instruments and methods, and the low probability of an event requiring these instruments. Note 2 has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed on Table 3.3.3-1. When the Required Channels in Table 3.3.3-1 are specified (e.g., on a per steam line, per loop, per SG, etc., basis), then the Condition may be entered separately for each steam line, loop, SG, etc., as appropriate. The Completion Time (s) of the inoperable channel (s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function. O Condition A applies to all PAM instrument Functions. Condition A addresses the situation when one or more required channels for one or more Functions are inoperable. ;

                                       .The Required Action is to refer to Table 3.3.3-1 and take the appropriate Required Actions for the PAM instrumentation affected. The Completion Times are those from the referenced Conditions and Required Actions.

S.d Condition B applies when one or more Functions have one required channel that is inoperable. Required Action A.1 requires restoring the inoperable channel to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel, the passive nature of the i instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval . (continued) l Catawba Unit 2 B 3.3-120 Supplement 1

RTS Instrtamentation B 3.3.1 t' /3 BASES APPLICABLE g SAFETY ANALYSES 21. Automatic Trio Loaic icontinued) LCO. and APPLICABILITY These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. .In MODE 3. 4. or 5. these RTS trip Functions must be OPERABLE when the RTBs and associated bypass breakers are closed. and the CRD System is capable of rod withdrawal. The RTS instmeentation satisfies Criterion 3 of @ U N ** # E Q dCFC @ b (R(0,(p g ACTIONS A Note has been added to the ACTIONS.to clacify the application of Coopletion Time rules. The Conditions of this ification as be entered independently for each Functi listed in T le3.3.11. Q ggg7 , In the event a channel's Trip setpoint is found t nonconservative with respect to the Allowable Value, or the transmitter, instrument loop, si or bistable is found inoperable gnal processing electronics, then all affected Functions provided by that channel must ta declared inoperable and the LCD Condition (s) entered for the protection Function (s) affected.

 .              -        Een the raaber of inoperable channels in a trip Function exceed those specified in one or other related Conditions

/ associated witn a tri the safety analysis. p Function, then the unit is outside t iemediatel Therefore. LCO 3.0.3 must be operation.y entered if applicable in the currenc MODE of Reviewer' approv the te: Certain LCO ical reports. order letion Times tre for a 11 use on 7 ines, the licen t justify the ired by the st ion Times Safety Evaluation (SER) for he topical report. i . Ad Condition A applies to all RT3 protection Functions Condition A addresses the situation where one or oor. e required channels for one or more Functions are inoperable ' (continued) B 3.3 37 a a .- Rev 1. 04/07/95 O .

INSERT When the Required Channels in Table 3.3.1-1 are specified (e.g., on a per steam line, per loop, per SG, etc., basis), then the Condition may be entered separately for each steam line, loop, SG, etc., as appropriate. O

l l INSERT Page 83.3-37 O Catavba

l ESFAS Instrumentation I B 3.3.2 ' N d BASES (continued) ACTIONS A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Ccaditions of this Specification ray be entered independently for each Function listed on Table 3.3.21.W In the event a channel's Trip Setpoint is found nonconservative with to the Allowable Value, or the transmitter, instrument signal processing electronics, or bistable is found inoperable, then all affected Functions provided by that channel must be declared inoperable and the LCO Condition (s) entered for the orotection Function (s) _ affectad. m. the Required Channels in ISDle 3.3.Z 1 are ' specineo

                                 ;e.g.. on a per steam line, per loop, per SG.

etc., basis), then the Condition may be entered separately for each steam line, loop. SG etc...as appropriate. idhen the number of inoperable channels in a trip function exceed those specified in one or other related Conditions associated with a tri the safety analysis. p function, then the unit is outside Therefore. LCO 3.0.3 should be immediately operation. entered if appl,1 cable in the current H00E of

Reviewer

te: Certain LCO e

letion Times are sed on appm ical reports. In r for a 11 to use these mes, the licensee t justify the Camp tion Times / as ired by the staff ety Evaluation / topical report. (SER) for (- O Ad Condition A applies to all ESFAS protection functions. Condition A addresses the situation where one or more channels or trains for one or more Functions are inoperable at the same time. The Required Action is to refer to Table 3.3.21 and to take the Required Actions for the protection functions affected. The Completion Times are

                  .those from the referenced Conditions and Required Actions.

1 (continued) B3.310k ca Rev 1. 04/07/95 l 0 - I s

PAM Instrumentation B 3.3.3 BASES s LCO Auxiliary Feedwater Flow (continued) At some units. AFW flow is a Type variable because rator act is required to ttle flow during a acci to prevent the in runout pumps from operati itions. AFW fl is also used by , a to verify that the System is delt ing l flow to each

  @                          indi tion used           the However, the pri ry

( is SG 1 1. / stor to ensure an te

                                                                                                  )
                                                                                                                          ]

APPLICABILITY The PAM instrumentation LCO is applicable in NODES 1, 2 and 3. These variables are related to the diagnosis and prv planned actions required to sitigate DBAs. The applicable DBAs art assmed to occur in NODES 1. 2. and 3. In H0 DES likelihood 4.event of an 5 and 6.would that unit conditions twquire PAMart such that the instrumentation is low: is not required to be OPERABLE in these MODES.therefore the PAN instrum

                                                                                         = -
                                                                                                 %        34y
                                                                                       "         ,M 9M h (4.        .

ETIOP!S , Hote I has been added in the ACTIONS to exclude 'orenA6Le the MODE kmsoe S} change restriction of LCO 3.0.4. This exception allows s entry into the applicable MODE while relying on the ACTIONS even though the ACTIONS may eventually require Unit shutdown.

This exception is acceptable due to the passive function of the instruments, the operator's ability to res to att accident usi alternate instr w ents and met . and the low instrtsnents. lity of an event requiring these Note 2 has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered i ntly for each Function listed on Table 3.3.3 1.1f r
etion Time (s) of ,

the inoperable channel (s) of a function wil be tracKeo 7

                  . separately for each Function r.tarti free the time the              (l#3 M I)

Condition was entered for that Furic' t on. r (continued) B 3.3 133 Rev I. 04/07/95 Co M de~-

1 INSERT 2 When the Required Channels in Table 3.3.3-1 are specified (e.g., on a per steam line, per loop, per SG, etc., basis), then the Condition may be entered separately for each steam line, loop, SG, etc., as appropriate. O INSERT Page B3.3-133 I O Catawba

Justificaticn far DeviaticIs 5 ction 3.3 - Instrument:tien A U fLLEji

10. The Bases for SR 3.3.1.2 have been clarified to indicate that the 2%

agreement between the calorimetric and the excore detectors is only applicable during steady state conditions. The surveillance requirement is only required to be performed on a daily basis, however, operators have indications in the control room that identify mismatches. Older plants, for which this SR was originally written, may not have this capability and do not detect the indicated mismatches during power maneuvering. The instrumentation may indicate a mismatch due to rod shadowing effects, neutron flux shifts, and changes in RCS temperature. These effects on the indication are not expected after the unit returns to steady state conditions.

11. Statements in the NUREG-14313.3.1 and 3.3.2 Bases regarding as found and as left calibration data have been deleted. During the implementation for WCAP-10271, the station provided the required certifications for calibration data. These statements v;ere accepted and approved by the NRC in the plant specific SER on this WCAP. These statements are not consistent with current plant practice and represent the imposition of a more restrictive change to the plant license via the Bases.

12. The Bases for 3.3.1 and 3.3.2 refer to specific methodologies for preparing setpoints and allowable values. Setpoints and allowable values are determined by a number of calculations rather than existing in one specific document. These calculations are prepared by the station rather than by the vendor. The station does not desire to maintain lists of calculation files within the Bases for the Technical Specifications. Additional changes are made to the Bases dicussions for these topics to clarify how these parameters 6t* treated at this i station.

13. Changes to the Bases have been made to incorporate detail which has been relocated from the CTS. These changes are consistent with the CTS and are justified in the discussion of changes.

14. The NUREG 3.3.2 ACTIONS Bases, second paragraph, states that functions which are specified on a once per steam Iine, loop, steam generator, etc., are allowed separate condition entry for each steam line, loop, or steam generator, as applicable. This is consistent with CTS Table 3.3-3 which presents functions in a multiple column format which O

Catawba Units 1 and 2 2B Supplement 15/20/07l 4

Justificcticn fcr Deviaticns l 5 cticn 3.3 - Instrumentttien BASES identtfies the total channels and minimum channels required operable on a per steam line, loop, or steam generator basis. This is also applicable to the reactor trip system and post recident monitoring ' functions as shown in CTS Tables 3.3-1 and 3.3-JiL The NUREG clarification is added to the ACTIONS Bases for ITS 3.3.1 and 3.3.3, consistent with the current licensing basis. The statement is also moved in the Bases for ITS 3.3.2 to indicate that this is a clarification of the separate condition entry note in the LCO Actions. l 1 4 i O lCatawbaUnits1and2 32 Supplement 15/20/97

l leeue Number l 31l Arected Section l CTS SR 4.8.1.1.2.o.6 l Affected Units CNS: l Yesl MNS: l Yes! Affected Pages ITS: l ITS: ITS Somes: I ITS Boeos: CTS: 3/48-5 CTS: 3/4 8-4 DOCS: L-8. 9: LA-1 DOCS: L-8; LA-1. 2 NIG: NRG: NR G Bases: NRCSomes. JFD: JFD: NSHC: 48.49 NSHC: 46.47 DescrisWlon CTS SR 4.8.1.1.2.o.6 requires vermying DG is aligned to provide standby power to ossociated emergency bus. This was relocated to SLC by LA-5 in the ITS submittal. The DGs con only connect to the ossociated bus and this is vermed through other survedances which stort the DG and tie to bus. Revise CTS to show this change as L25.

I.I ELECTRICAL POWER SYSTEMS O SURVEILLANCE RE0UIREMENTS (Continued)

                                                %s, cir) h gg         54 32 IJ @

Verify @ ine ==== man is synchronized, loaded and operates at 5600 - 5750 kms for at men minutes, and , M3.?M @

                             % cn:aum;x'r ~s@@

At least once per 31 days (fRiter each ----1 #~' G. sus.,vrioa or onersfion was creater$an! or tion of theMdpal eaual hw hv> ch==-r for ana removg aw.-ulates water from g!p tank; g [* g,3' g.

                           }ater from.the fuel oli storage tanks;At                     least
                                                                                    -.mny       =_- per 41 aays o accumulate      *

d. By verifying that t verifying: athodic Prvtection System is 0 E by

1) At le a once per 60 days that cathodic pro ction rectifiers.,

E and have been inspected in e ordance'with the ufacturer's inspection procedures, At sionleast once perin12 is provided months tha dequat~e protection from ce acconian t procedures, with manufacturer's inspect e7 Bytionsagling to storagenew fuel tanks oil in accordance with ASTM-04057 prior to and: 1) By verifying in acconlance with the tests specified in ASTM-0975-81 sample has: prior to addition to the storage tanks that the a) An API Gravity of within 0.3 degrees at 60*F or a specific Adhs5 /by gravity of within 0.0016 at 60/60*F, when com, pared to the rrs 3.t3 supplier's certificate, or an absolute specific gravity at 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees; b) A kinematic viscosity at 40*C of greater than or equal to 1.9 contistokes, but less than or equal to 4.1 centistokes (alternatively, Saybolt viscosity, SUS at 100*F of greater than or equal to 32.6, but less than or equal to 40.1), if gravit supplier'yswas not detenmined by comparison with the certification;

               @0iesel generator loadings may be done in accordancee with ggf S R 3.2. 8.3       manufacturer's recommendations. The purpose of the load range is to prevent overloading the engine, and momentary excursions outside of th eeres #+ 2.        load range shall not invalidate the test.

CATAWBA - UNIT 1 3/4 8-5 Amendment No.155 O n* c'

SpuMS, 3.2.1 ELECTRICAL P0uER SYSTEMS SURVElltANCE RE00!REMENTS (Continued) SR 3 5.15 @ Verify h pt'L ttseRT 10 5600 - 750 k for Ef' GiR'n 60 minutes, andissynchroniZed,loade l 6) QD l Vert fy g the diesel gefera6er is al to the associa}s6 emergency bu s. to providepadby 5 [.

                $A 3S.I. @

At least once per 31 dals5nd>fter each oper4 tion of tils'diese

                                 @ere checkqWk persever      ooeration(was annate? tKan or en"Wto 1 hogy' for am remov accumulated water from                cann;
           /h A To           c.                                                                                     ;

(sa 24 38 water from the fuel oil storage tanks; iAt ~iim once per .n cays by

d. By Ye~rityt verifyi that the Cathodic Protection tem is OPERABLE by

1) t least once per 60 days that g AS odic protection rectifiers ' '

are OPERABLE and have been in manufacturer's inspection p ted in accordance with the edures, and

2) At least once per 12 mon s that adequate protection f ston is provided in ac niance with manufacturer's ins corro-procedures, tion e.

tion to storage tanks ano:By sampilng new tuel ot < ut acconlance w ' O 1) By verifying in accordance with the tests specified in ASTM-0975-81 s le has- prior to -addition

                                                           ----   -~~ -   to the
                                                                              ' ' - storage
                                                                                     - ~ ~ ~ ~- tanks that ,the a) gggp                             An API Gravity of within 0.3 degrees at 60'F, or a specific g 3,3,3                     gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60*F of greater than or eeni to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to 27 degrees but less than or equal to 39 degrees; b)

A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes (alternatively, Saybolt viscosity SUS at 100*F of greater than or equal to 32.6, but less than or equal to 40.1), if gravity was not determined by comparison with the supplier's certification:

                   @IDiesel generator loadings may be done in accordance with the 093 Ngg M       ' ,,3 manufacturer's recosamendations.        The purpose of the load range is to prevent eserloading the engine, and momentary excursions outside of th load range shall not invalidate the test 7

CATAWBA - UNIT 2 3/4 8-5 Amendment No.147 p to of 2I 1

Discussitn of Ch nges Section 3.8 - Electrical Power Systems TECHNICAL CHANGES - LESS RESTRICTIVE use of an actual or simulated actuation signal for testing purposes. This change permits credit to be taken for unplanned events (actual signals) which provide the necessary data to satisfy the SR. The actual signal is what is credited within the safety analysis and is sufficient for demonstrating compliance with the SR. This change is consistent with NUREG-1431. L.24 Added to the requirements of CTS LCOs 3.8.3.1 and 3.8.2.1 is an allowance that one "or more" of the required buses or channels may be inoperable. ITS LCO 3.8.9, " Distribution Systems - Operating" provides in Action A for one or mare AC buses, in Action B for one or more AC vital buses, in Action C for one or more DC channels, and in Action D for one or more DC trains to be inoperable. ITS LCO 3.8.9 also adds Action F which requires if two or more of the required buses or channels are inoperable that results in a loss of safety function, LC0 3.0.3 must be entered immediately. With the additional allowance of "or more", ITS LC0 3.8.9 Action F, ITS LCO 3.0.6, and the Safety Function Determination Program, the i proposed change will ensure that with a loss of ary electrical power distribution system, no loss of function will occur without the appropriate action being taken. Therefore, this less restrictive change will have negligible impact on safety and is consistent with NUREG-1431. L.25 CTS 4.8.1.1.2.a.6 requires verification that each DG is aligned to provide standby power to the associated emergency buses once per 31 days. ITS 3.8.1 requires the DGs to be OPERABLE and specifies associatec' Surveillance Requirements to ensure the DGs are maintained OPERABLE, but does not include this requirement. The definition of OPERABILITY and procedural controls on DG standby alignment are sufficient to ensure the DG remains aligned to provide standby power. This type of requirement is addressed by plant specific processes which monitor plant conditions to ensure that changes in status of plant equipment requiring entry into Actions are identified in a timely manner. This verification is on implicit part of using Technical Specifications and determining appropriate actions in the event of equipnt inoperability. Equipment status is routinely monitored by operating personnel and documented in records and logs. The monitoring process includes reevaluation of compliance with Technical Specification requirements when equipment becomes inoperable using the records O and logs as an aid. Therefore, the explicit requirement to Catawba Units 1 and 2 Page L - 88 5/20/97 Supplement 1l j

Discussien of Ch:nges Secticn 3.8 - Electrical Power Systems TECHNICAL CHANGES - LESS RESTRICTIVE periodically verify that each DG is aligned to provide standby power to the associated emergency buses is considered to be unnecessary for ensuring compliance with *he existing Technical Specification OPERABILITY requirements n,1d it deleted. O l 1 l l l 9 Catawba Units 1 and 2 Page L - 98 JB0/47 Supplement 1l l

Discussicn of Charg:s S:ctica 3.8 - Elcctrical Power Systems TECHNICAL CHANGES - REMOVAL 0F DETAILS LA.1 Not used. LA.2 CTS 3.8.1.1 Action g requires a DG operating at greater than 5750 kW be reduced to 5 5750 kW within one hour. This requirement has been moved from the TS to the Selected Licensee Commitments (SLC) (UFSAR Chapter 16) consistent with NUREG-1431. This level of detail is not necessary for inclusion within the TS. Changes to the SLC are reviewed and controlled in accordance with the requirements of 10 CFR 50.59. These controls are adequate to assure any change is proparly reviewed. LA.3 CTS 3.8.1.1 Action h and SR 4.8.1.1.2d requires the Cathodic Protection System to be OPERABLE for a DG. These requirements j have been moved from the TS to the Selected Licensee Connitments (SLC)(UFSAR Chapter 16) consistent with NUREG-1431. This level of detail is not necessary for inclusion within the TS. Changes to the SLC are reviewed and controlled in accordance with the requirements of 10 CFR 50.59. These controls are adequate to assure any change is properly reviewed. ( LA.4 CTS SR 4.8.1.1.2a 4) requires the DG to be started for a test by one of four methods. The requirement has been moved from the TS to the Bases for ITS 3.6.1.Sc!ccted Licen:ce Corit= nt: (SLC) (UFSfM Chapter 15) censistent with NUREC l'31. This level of

detail is not necessary for inclusion within the TS. Requirements ) for operability of the DG are still retained in ITS 3.8.1. l Changes to the RC-Bases are reviewed and controlled in accordance with the requirements of the ITS Section 5.0 Bases Control Program and require an evaluation in accordance with 10 CFR 50.59. These ! controls are adequate to assure any change is properly reviewed. LA.5 Not used. CTS SR d.S.1.1.2 5) rcquirc; th: OC t be verified Oligned t0 th :::Oci;ted c= rgency bus. The requirc= nt h:: been Oved frc: the TS t0 the Sekcted Licen c0 00r.it=cnt; (SLC) (UFSf.R Chaptcr 15) consistent-with NUREC l'31. This icvel of detail 1 not nece;;;ry fcr inc sien ithin the TS. Change; to the SLC are reviewed and contr0lkd : Ordance with the ) requirement: Of 10 CFR 50.50. The:c controh are cd quate to

',ure any chang: is properly revicwed.

r

 \
                                                                                     \

l lCatawbaUnits1and2 Page LA - 18 5/20/07 Supplement 1 ! j

No Significtnt H:z:rds C nsideraticn Secticn 3.8 - Elcctrical Power Systems LESS RESTRICTIVE CHANGE L.25 The Catawba Nuclear Station is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1431, " Standard Technical ' Specificottons, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive. Below is the description of this less restrictive change and the No

Significant Hazards Consideration for conversion to NUREG-1431. 1 CTS 4.8.1.1.2.a.6 requires verification that each DG is aligned to provide standby power to the associated emergency buses once per 31 days. ITS 3.8.1 requires the DGs to be OPERABLE and specifies associated Surveillance Requirements to ensure the DGs are maintained OPERABLE, but does not include this requiremer'. The definition of OPERABILITY and procedural controls on DG st.;ndby alignment are sufficient to ensure the DG remains aligned to provide standby power. This type of requirement is addressed by plant specific processes which monitor plant conditions to ensure that changes in status of plant equipment requirity entry into Actions are identified in a timely manner. This verification is ' on implicit part of using Technical Specifications and determining appropriate actions in the event of equipment inoperability. . J Equipment status is routinely monitored by operating personnel and documented in records and logs. The monitoring process includes reevaluation of compliance with Technical Specification , requirements when equipment becomes inoperable using the records l and logs as an aid. Therefore, the explicit requirement to periodically verify that each DG is aligned to provide standby power to the associated emergency buses is considered to be unnecessary for ensuring compliance with the existing Technical Specification OPERABILITY requirements and is deleted. In accordance with the criteria set forth in 10 CFR 50.92, the Catawba i Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change removes the requirement to explicitly verify that each DG is aligned to provide standby power to the associated emergency buses every 31 days. This verification is not assumed Catawba Units 1 and 2 Page 48% of 53% 5l20lC7 Supplement 1l J

No Signific:nt H2zards Crnsideratien Section 3.8 - Electrical Power Systems O to be the initiator of any analyzed event, therefore, the probability of an event previously evaluated is not affected. Verification that the DG is capable of providing standby power is adequately demonstrated by the existing surveillance requirements and by plant processes which monitor the status of plant equipment. Operability of the DG is not affected by this change, therefore, the consequences of an event are not affected.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not permit operation in a new or different mode, or permit the installation of a new or different type of equipment. The proposed change only removes on explicit verification which is considered redundant to existing operability requirements. The DG remains capable of performing the ,, design safety function. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those previously evaluated.

3. Does this change involve a significant reduction in a norgin of safety?

The proposed change continues to require standby DG power for the required equipment necessary to mitigate analyzed events. Existing requirements for DG operability and monitoring of plant and equipment status ensure that appropriate Technical Specification actions are entered when required equipment is inoperable. Therefore, this change does not involve a significant reduction in a margin of safety. 1 O V lCatawbaUnits1and2 Page 49 % of 53 M 5/20/97 Supplement 1

issue Number l 33j Affected Section [3.3.2 ESFAS, CPCS l Affected Units CNS: l Yes) MNS: l Nol Affected Pogos ITS: 3.3-27 i ITS: ITS Aosw: 8 3.3-99 ITSBasee: CTS: 3/4 3-19.27 CTS: DOCS: M-4. 6 DOCS: NitG: 3 3-28 NltG: NitG Bones: B 3.3-114 NltG Boeos: JFD: IS 4 JFD: NSHC: NSHC: Desertplion The CTS 3.3.2 Action 16b requirements for on inoperable Conto 6nment Pressure Control channel are amb6guous. The ) octions require o channel to be placed in trip in I hour. However, the CPCS hos both a stort permissive and l terminate function. The a:: tion is unclear os to which state the channel should be placed. If the channel were placed in the permissive stote, no protection from inodvertent octuotton is provided. If placed in terminate, the function would not octuate during the DBA. The oppropriate action is to cascode to the supported system function immedlotely, similar to the actions in the McGuire CTS. Revise ITS 3.3.2 Action P to require immediate cascode, j I i i O

l ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME

0. One channel 0.1 Verify interlock is I hour 4 inoperable. in required state for J existing unit

[ condition. 2 0.2.1 Be in MODE 3. 7 hours M 0.2.2 Be in MODE 4. 13 hours P. One or more P.1 Declare affected Insnediately Containment Pressure Support System Control System inoperable. channel (s) inoperable. Q. One Nuclear Service Q.1 --------NOTE--------- Water Suction The inoperable Transfer-Low Pit Level channel may be channel in one or more bypassed for up to 2 pits inoperable. hours for surveillance testing of other channels. Place channel in 4 hours trip. M Q.2.1 Be in MODE 3. 10 hours M Q.2.2 Be in MODE 5. 40 hours (continued) O Catawba Unit 1 3.3-27 Supplement 1 l

ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME R. Two Nuclear Service R.1 Align the Nuclear 4 hours Water Suction Service Water System Transfer-Low Pit level for Standby Nuclear channels in one or Service Water Pond more pits inoperable. recirculation. Ofl R.2.1 Be in MODE 3. 10 hours AE R.2.2 Be in MODE 5. 40 hours

O - l I O 3.3-28 Supplement 1 l Catawba Unit 1

ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME

0. One channel 0.1 Verify interlock is I hour inoperable. in required state for existing unit condition.

DE 0.2.1 Be in MODE 3. 7 hours M 0.2.2 Be in MODE 4. 13 hours P. One or more P.1 Declare affected Inunediately Containment Pressure supported system Control System inoperable, channel (s) inoperable. Q. One Nuclear Service Q.1 --------NOTE--------- Water Suction The inoperable Transfer-Low Pit Level channel may be channel in one or more bypassed for up to 2 pits inoperable. hours for surveillance testing of other channels. Place channel in 4 hours trip. QR Q.2.1 Be in MODE 3. 10 hours M Q.2.2 Be in MODE 5. 40 hours 1 (continued) I i l O Catawba Unit 2 3.3-27 Supplement 1 l

ESFAS Instrumentation 3.3.2 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME R. Two Nuclear Service R.1 Align the Nuclear 4 hours Water Suction Service Water System Transfer-Low Pit Level for Standby Nuclear channels in one or Service Water Pond more pits inoperable. recirculation.. Ofl R.2.1 Be in MODE 3. 10 hours AND R.2.2 Be in MODE 5. 40 hours l O I O l Catawba Unit 2 3.3-28 Supplement 1

i ESFAS Instrumentation i B 3.3.2 C BASES (>} ACTIONS 0.1. 0.2.1 and 0.2.2 (continued) Condition 0 applies to the P-11 and P-12 interlocks. With one channel inoperable, the operator must verify that the interlock is in the. required state for the existing unit condition. This action manually accomplishes the function of the interlock. Determination must be made within 1 hour. The 1 hour Completion Time is equal to the time allowed by ) LCO 3.0.3 to initiate shutdown actions in the event of a complete loss of ESFAS function. If the interlock is not in the required state (or placed in the required state) for the existing unit condition, the unit must be placed in MODE 3 I within the next 6 hours and MODE 4 within the following j 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. Placing the unit in MODE 4 removes all requirements for OPERABILITY of these interlocks. P.1. P.2.1. and P.2.2 O \ Condition P applies to the Containment Pressure Control System Start and Terminate Permissives. With one or more channels inoperable, the affected containment spray, containment air return fans, and hydrogen skinsner fans must be declared inoperable innediately. The supported system LCOs provide the appropriate Required Actions and Completion Times for the equipment made inoperable by the inoperable channel. The immediate Completion Time is appropriate since the inoperable channel could prevent the supported equipment from starting when required. Additionally, protection from an inadvertent actuation may not be provided if the terminate function is not OPERABLE. (continued) Catawba Unit 1 B 3.3-99 Supplement 1 l l

ESFAS Instrumentation B 3.3.2 BASES ACTIONS 0.1. 0.2.1 and 0.2.2 (continued) Condition 0 applies to the P-11 and P-12 interlocks. With one channel inoperable, the operator must verify that the interlock is in the required state for the existing unit i condition. This action manually accomplishes the function of the interlock. Determination must be made within 1 hour. The 1 hour Completion Time is equal to the time allowed by LCO 3.0.3 to initiate shutdown actions in the event of a complete loss of ESFAS function. If the interlock is not in the required state (or placed in the required state) for the existing unit condition, the unit must be placed in MODE 3 within the next 6 hours and MODE 4 within the following 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner ; and without challenging unit systems. Placing the unit in MODE 4 removes all requirements for OPERABILITY of these interlocks. j P.I. P.2.1. and P.2.2 Ox Condition P applies to the Containment Pressure Control J System Start and Terminate Permissives. With one or more channels inoperable, the affected containment spray, containment air return fans, and hydrogen skimer fans must be declared inoperable immediately. The supported system LCOs provide the appropriate Required Actions and Completion Times for the equipment made inoperable by the inoperable channel. The imediate Completion Time is appropriate since the inoperable channel could prevent the supported equipment from starting when required. Additionally, protection from an inadvertent actuation may not be provided if the terminate function is not OPERABLE. I (continued) Catawba Unit 2 B 3.3-99 Supplement 1 l

7

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   .-                       Deelsse se$4ehul 8*Y#8'                                 j _ABLE (5ys4ste,, ane me            s     u Mt ^- ~' I                                (Continued) g~                                  m.110N STATEMENT $'(Continued 1                              k(,ka.
                                                                                                                                                                ^    '

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                               / ACTION @                    - fWith N

e nunoer of LRA5Lt, Gnan s one less t er of Channe STARTUP and the T D I ovided the fo owina conditi e a~ POWER OPE d'FL: may oceed gg' c. p. T g , k..rs). ~ FThe inoperable chan y M ithin I ho_ur is placed in t ripped c iti

                                          }               $ / w ..                           in ---i.

g ,,,opf,@f: . amr tionE channe nornaiu v, 7 mg y be bypassed for

                                                                                                                                                  .r           _' _[

for surveillance te ng of other channe to 2 hours pose w - 1TJCEtion 4 4# D. m g.3.k ACTION 17 e With less than the Nintmum Channels opera- OPERA Nxhaust valve < are maintained closed. ACTION - Itn sn t ggg hanne r of v

                                                                                               ,  -n t chanGris one ler.r than % -un e 4,.;,

Qa.pe w,e. opt-a*LE status within 48noperab' hours or be in at e channe1~to within following the 30 hours. next 6 hours and in COLD SHUTD0let within 1 TION $ -

                                                       'With tpi number of ~ =ABLE chann N"-            of Channel            TARTUP and one less th                      to e,A g

ec'9

                                        .Q               pro ded the fell inq conditi                                    POWER OPERATI are satisfi               .

may p w . se e;.na=64' ,3 u > The hin inoperable 6 hourg channel is placed in the tripped condition h ,.im ,o

e. in. ^

[ Ar.**d b ' Qka..*r A _.._ N

                            *                                                                                                                          -c
                                                                                                                                                                == . m for surve onM           channel        may        be     bypassed          for   up to 4 hours            .

11ance tutin ('" _ J o Qg"~~~'"~ ~~ ~~ g of other channelsqD _ ACTION f

                                                -     mi -          -

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                                                                                                               ==.7 ure----        Qnecweaeaop.steQ 5 i.d.
              .fe(,.g,,
                     ,% mooeoek i, li3keun).x 164%    determina>Dy o5ipar.uun er une av                                              -

dwithin 1 hour d aht is1 ,wrai.wvs stat wemsung@swa= t the condition, interlock isorranarvin < ts requi A m eme -

                                                                                                                                         "ec   state     for   the                   t ACT!

wenis@ i n.3 (

                                                    -_p        tn tne L _i 31 or y 2 - ~ ~
                                                                                                                                          - r===

DM M unannelvore-ame r r:c-.r = --=-

            .                       'M                                                               n+f      restore         the   inoperable GPtRABLE status within 6 hours or be in at least NOT ST
                                    /M               within following the      6 hours to 4 hours for sur; vet ' 5) next 6 hours and in at least HOT SHilTnna"' w ance, one -- - may be bypasses for up k mit               provided the other Mis OPERA 8testitig                                      isEhoecmeauaru,r.m
                          ~                                                                                             L.E.

twee 1=

m. 3.18 ACTION Zla - With the number of OPERABLE channels n mum one le i

6 hours and in at least HOT SHUT 00WN with hours; however, one channel may be bypassed other channel is OPERABLE. N I

CATAW8A - UNIT L 3/4 3-27 Amendment No. 148 ANI Ne. % .b reedl<. ,6teslere, tvw:, 46 OPEp.ASLE SW ea 6 bd o ., , , u.m. - w. ,,_, % # ws H 4%*ws Ts. saes,N sa 4ts Q pesvAI Mt sher +nio: G Ut#0E8' k A .'N At.tw T gneclaa.t ia ero.Ut. Plate et. y

g i n 12 h *"'5 ^*oM : The. 4w,w,i ,, trip b hours or- be..a M0cE3 seveille, a 4esk,3 fonu cw.cta ,

S e.(a t $p saJ C- 9 4 9 ke (=- w

[p ; y'M 'l e o .

g r i 3, n k d n 9 ) ( p a 1 l s n o hQ [H f j t. [/ E

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E g , , N I. uD 3 n 3 3 3 O I - i 3 O I uM , t , Lr f, T 2 2 a A r b i 2 2 2 T , . (1 ( , t N A l - 1 i E A n I 1 1 M L R t a (1 2 4 T . s e 1 S SE N mLLEB g g n i t . I l MNA c o M INR .c i p e n n n N NAE sa m f i i i E T IHP t e . u 2 n a a a t

d S MCO s e p I r r r 2 n e Y /np / t t t e u S 3ic 1 ft / / m n p - 3 3 d N . f n i t O a e n I t m s Ta A

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            -   (

T A U E n e g g s i T n a 2 C I E i n n A lT ,yt p r _ s. 3 A ana m o i a i e [(d E L t T r m i HO CT t ar . s

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1 t a r 1 B A I b A T E F a )5 e Y L n. T E R a r E )n Q F A S O@S E I A H C a t

                                                     .                 )p im 2                       n n i i a

r r ta i a 2 9 1 3 D s t p r 4 E R E T / 4 P 4

t S 4 t

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4 t

1

3 E N c I G N E

4 l o r f i g o t L

                                     )      t                                                 n                                  s ny d

e aP W( ) 4 C o oa _ u n - n e n il n rh is P o e v o te i og ap ( r r i aR t ti Mmu i t u i s e t u n aH p c s s a tn o r- lP i e s i n t a i co . C eh e t Ai ( ng l Ar j n r m oi r d w i ca t ei e I P e t e r iu p GH i-r ft oa o y t P a e n f Ma tt ac 2 M e ml ae ev pd ie re w t c a f n e m t ra i m y r r l u n mA o td T I N U n te L TF e a n t e a a un R S im S EI i b r ae tt ns T i l M Aa A i ( f. T u . oy x B C T c e

                                                                                 .         .           .      . u                      W N                               f.                    f     CS      a    b        A                       A U       .                                                                                                  T F

r6 f /. K. a A C 6 O 9 a o- e c

                                                                                                                              ,e i

4; J[

er neere *^ -? _

1 f9t80

                 *lf c 4 I s,eh ewt ..pwdie.
                           ' peelect e&Apetu b                        ,W 3 % e sgr:n- 3.3.2.
                                 "Y                  #

Continued) \ e w-wm TABL g ACTION STATEMENTS (Continued 1 f 3

                                 ' ACTION @ -                  sii th      e number of 0        BLE channels           ess than the Tata                I 0_!,                                                 Nu er of Channels STARTUP and/or                       OPERATIONmayjroceed gg a W.          3    -   %rs                              ovided the fo wino conditinne                   eaticffad?      /

m 5 *n 37 he=er. filie InoperaDie c nel is placed i he tripped cond o L within I hou

                              /A0 0        _

g MimimarrhanneIs uPPKARI F#e" 4 r- mt i e ame - neweven, h uchannel may i h 'a w ,y ypassed for up t hours ! for J ~=i11ance testing ther channels

                                                                ~ 3  4*fon a x 1L                                                       -

M0T U CTION 173 With less than the Minicia Channels OPERA 8LE requirement, opera-14 33 tion may continue provided the containment purge supply an

                              .                                - hmust valves are maintained closed.

ACTI 16 - ith the n r of OP^ cnannels one 13sf than the2Hslimui> del o*' fhm'inel ERABLE ree tI restore the Inopersoie c tannel to ~

                                 ,,,,d'^d'j                    OPERA 8LE status within 48 hours or be in at least HOT STAlW8Y within the next 6 hours and in ' COLD SHUTD0l81 within the                          Al c,s.Q @                      following 30 hours.                                              m ,u.y            3 t* .g\            A.3'l       ACTION IS -                With the n          r of OPERABLE               is one less        the TotaT"*Psetisle J Number           annels, STARTUP          /or POWER OPERA         any proceed h%

o.or a i. is eM ~ the following c tions are satisfi : g ,

                                                           , [provi     The inoperable channel is placed in the tripped condition M4 ..if k es A within 6 pg 9                               GK~        The Mif- - cha.in.7 s opreame r ; =, .            't is one e

% AJ g,,a og p tr Q q-r

                                                                                         - channel may be bypassed for up to 4 hours -
                                                                        .o Z :--survatilance testing of other channels gp                               s g                                         *

(On +1 inopui c. ACTION - Wen t est _ +h>= 14 iiins ra-=1s OPED'"9,_within 1 hour u _ ,

7 " *"8 .e. _ . _ . m e . _ 1ss,.e _ ._ _m.2.>s f-

                                                                         $/fliat the interlock is in its requ1            e state for the LiaA *. ,, ,, * '" G 6 st' ng Mcondition, orsyvir arm in.edm.o.v3                                          j
                                               @                                - w se           _

ACTION @ - the n f OPERABLE - ---i< aa- @ "- "" *a i -9 annels AaLE raauiM*f, restore the inoperable ! {^. h a'*Fg OPERA 8LE status within 6 hours or be in at least HOT S@ ' y within the next 6 h goand in at least HOT SHUTDOWN within the following 6 hours; ow., a one CEEETeay be bypassed for up 3 AcT4*al d*I to 4 hours for surveillance testing ser sneemcattarw.J.uo y provided the other CIMMH is OPERA 8LE.

                                    ~ACTION Zla -               With the number of OPERAS [E channels one less than the Ninth Channels OPERA 8LE requirement, be in at least H0T STAND 8Y withirf M                                                   6 hours and in at least H0T SHUTD0l81 withiji the following 6 ff5 MI                                             hours; however, one channel any be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided thej other channel is OPERA 8tE.                                           "

CATAWBA - UNIT 2 3/4 3 27 Amendment No. 142 l ACnon L Ech

h NOW $'N !*
h"

                                                         % ocY*t.,I*'W'.iace J.2.'s-                                                           69,de,.

or 4 to w es. verr : ci,,c % .

                                                                   + a e s 4- ws.va.aa. w1;y p ales si,a. n ,a,.,                  y;, mQ
                                 $.C1)nJ 3               0s cea-ael ioper> *dL. We+ CLr^~a l **Lk%s or                          b a:e .e eue 3

o2. was. unrs: % e wcl ~1 e- 4t=<a/ &ph Dr W 4 ,w, Z,e s.,en;!s s.ea.

                                                                                                 ~.f.<=w-h d.y oS & .*hsands.

P.1y. A et- tz.

Discussitn cf Ch:ng:s 5:ction 3.3 - Instrumentation TECHNICAL CHANGES - MORE RESTRICTIVE M.13 CTS Table 3.3-3 Action 16a for RWST level-low require:; the inoperable channel to be placed in bypass and operation may continue. ITS 3.3.2 requires the channel to be placed in bypass. within 6 hours. This change is more restrictive since no time limit is specified in the CTS. When the CTS actions cannot be completed, an LCO 3.0.3 entry is required. ITS 3.3.2 adds a shutdown requirement if the actions cannot be met within the required Completion Time. The shutdown requirements are for the plant to be in mode 3 within the following 6 hours and mode 5 within the following 30 hours. This change is more restrictive, because with the CTS LCO 3.0.3 requirement, one additional hour is ! allowed prior to beginning the shutdown. These changes are consistent with NUREG-1431. M.14 Not used. M.lb US Table 3.3-3 actions 155 =d 29 is r: :ntered when one channel l of the associated ESFAS Function is inoperable. When actions 1 cannot be completed, an LC0 3.0.3 entry is required. ITS 3.3.2

 ,                                       adds a shutdown requirement if the actions cannot be met within the required Completion Time. The shutdown requirements are for the plant to be in mode 3 within the following 6 hours and mode 5 within the following 30 hours. This change is more restrictive, because with the CTS LCO 3.0.3 requirement, one additional hour is allowed prior to beginning the shutdown. This change is consistent with NUREG-1431.

M.16 Not used. M.17 Not used. M.18 CTS 3.3.2 Control Room Area Ventilation Operation is applicable in all modes. ITS 3.3.7 adds the applicability of "during movement of irradiated fuel assemblies" and "During CORE ALTERATIONS" to cover those times when a potential accident could occur. Fuel handling accidents are events that could cause radiation levels to rise in the control room. Other related changes involve actions to alleviate these potential hazards with various levels of Control Room ventilation degradation. These changes in ITS 3.3.7 are additional restrictions placed on the plant and are considered more restrictive. This change is consistent with NUREG-1431. Catawba Units 1 and 2 Page M - 48 Supplement 15/20/07l l t

Discussion &f Ch:nges Secticn 3.3 - Instrumentaticn TECHNICAL CHANGES - MORE RESTRICTIVE M.24 Not used. M.25 CTS Table 3.3-3 action 16b requires that an inoperable Containment Pressure Control System channel be placed in trip in one hour. ITS 3.3.2 Action P requires that the affected supported system be declared inoperable inunediately. With a channel inoperable, the supported system may not start when required and protection from an inadvertent start may not be provided. The CTS would allow continued operation with the channel placed in a tripped condition, but does not specify whether the channel is tripped to the start permissive state or the terminate state. The ITS actions are more restrictive and ensure that unit operation is limited to the completion time associated with the supported system when the CPCS is inoperable.

l l i O Catawba Units 1 and 2 Page M - 658 Supplement 15/20/07l i ..

l i I DiFASInstrune%gt 4 ACTIONS COICITION REQUIRED ACTION COMPLETION TIME (coatinued) .2.1 Be in NODE 3. 12 hours

                            .2.2     Be in MODE 5.          42 hours
                                                                               )

W . One channel .1 verify interivek is 1 hour inoperable. in required state for existing unit condition. 2.1 Be in H00E 3. 7 hours

                           .2.2     Be in MODE 4.           13 hours O  *QWT       -

I i I 3.3 28 Rev 1. 04/07/95 Ca w A s O

INSERT I i CONDITION REQUIREU ACTION COMPLETION TIME l P. One or more P.1 Declare affected Imediately Containment Pressure supported system j g Control channel (s) inoperable. j inoperable. " Q. One Nuclear Service Q .1 -- ---- - - -- - NOTE - - - --- - - - - i Water Suction The inoperable channel Transfer-Low Pit may be bypassed for up to )j Level channel in one 2 hours for surveillance { or more pits testing of other inoperable. channels. Place channel in trip. 4 hours i Q.2.1 Be in MODE 3. 10 hours l M Q.2.2 Be in MODE 5. 40 hours Two Nuclear Service R.1 Align the Nuclear Service \ R. 4 hours Water Suction Water System for Standby Transfer-Low Pit Nuclear Service Water l Level channels in one Pond recirculation. or more pits inoperable. E R.2.1 Be in MODE 3. 10 hours M R.2.2 Be in MODE 5. 40 hours INSERT Page 3.3-28 i O Catawba

                                                                    *n
                                                 \nQr %k yO.dkhi$ &

3 Esas Instnenentaten 3.3.2 Table 3.3.21 (pose 8 ef g) Engineered Safety Feature Actuation Systes Instrumentation Apet t ' OfMft FWCfl0N SPECIFIED eteultED amyggLLANCg .ar e nuass e Couptfl0NS CNANNELS Costflous teeuttesagTS yggp g VALME SETPolsfW f - j T. Autasette 1 erit to Centefrument J (continued)

c. ihist t - Lew t,2,3,4 6 L K st 3.3.2.1 t (1515 t (1831 st 3.3.2.5 at 3.3.2.9 st 3.3.2.10 coincident alth tefer to timet Safety Injection fwictions 1 (Sefety Injectlan) for att initiation ragsfrasents.

and Coinciennt with 1 , ,4 4 Cwiteinment simp E st 3.3.2 m (303 In. t t 3 in. Lovet - Nish st 3.3 .5 stume SR 3 .2.9 above ! se et. [755) it et. t 2ft

                                                                                                   .3.2.10

8. ESFAS Intertecke

s. Reacter Trip, P-4 1,2,3 1 per train, 2 i SR 3.3.2. 54 lut trains g [qq

b. Pressuriser 1,2,3 Pressure, P.11 3

                                                                                   /D             #9

( 3R 3.3.2.5 at 3.3.2.9 s psi Is

c. f,-Leu Lev, P-12 1,2,5 1 per

                                                                                   /0                        a    5g*r          a 53 7 st 3    .2.9 th        ic      sonntettens may con n only Attomable Yet ing en SetedIint Study s

q, bbM b85** h %%\ 5 3skm 4 $bd %ss< i,2.. 5,4 4 er M P g,y,l . g o.qr ps.J f ON t ied 6,, "Ttemi b 8, t. 3, 9 V" P 3a , . IO- WI *' 58^'" C sud,e,., T, d-- g,1,3,9 3 pe, pif G,R 3 5,k. l

                                                                                                              -> 5s'2.9 ft 1 SM N l*W FOut.\                                                               3( 3,3,t,4 S t. 1 4.1.11 3.3 39 Rev 1. 04/07/95
  ..O                                                                                                                                            l
                                                                                                                                                 \

l

Justific:tien fcr Deviaticas S:cticn 3.3 - Instrumentatien TECHNICAL SPECIFICATIONS 20.'NUREG 3.3.? is revised to include a requirement for the Containment Pressure control System. The requirements for this functton are consistent with the CTS, however, the CTS actions are modified to require that the supported system be declared inoperable inmediately when a channel is inoperable. These more restrictive actions are described in the Discussion of Changes (M.25).

21. Not used.

O i

o Catawba Units 1 and 2 43 Supplement 15/20/07l l

Q C. 8L

                                     \h                                      ESFASInstrumegt BASES g        @.1.         .2.1 and      .2.2 (continued)

LCO 3.0.3 complete loss ofto .~nitiate ESFAS shutdown function. actions in the event of a the required existing state (or placed in the required state) for theIf the interlo unit condition within the next 6 hours,and H00E 4 within the followinthe unit must b 6 hours. The allowed Com on operating experience, pletion Times.a.e reasonable,g based to reach the required unit O5 and without challenging unit systems. conditions from full power c (LMWtr H00E 4 removes all requirements for OPERABILITY of thesePlacing the u interlocks. s SURVEILLANCE REQUIREENTS The SRs for each ESFAS Function are identified by the SRs-column of Table 3.3.21. A Note has been added to the SR Table to clarify that Table 3.3.21 deterutnes wl)ich Functions. ' SRs apply to which ESFAS { Note that each channel of process protection supplies' both trains of the ESFAS. train B must be examined.When testing channel I, train A and must be examined when testinSimilarly, train A and train 8 channel IV COTS are per(if appitcable). The CHANEL g CALIBRATION channel II, and channel III, and formed in a manner that is consistent with the channel accuracies. assumptions used in analytically calculating the required Reviewer topi

Note: Certain F ::1es are based on b reports. v ti , the licensee susIn order ustify or_a the licensee to use these Fre b he staff SER for tred topical report.gaencies as FR 3.3.2.1 i

                      ' Performance of the CHANM1 CHECK once every 12 hours ensures CHANNEL CHECK is normally a comparison                                  A       of the param indicated on one channel to a similar parameter on other channels.

channels monitoring the same parameter should readIt is based; l (continued) M B3.311k CalnA Rev 1. 04/07/95 i

m (V

i INSERT O V P.1. P.2.1 and P.2.2 Condition P applies to the Containment Pressure Control System Start and Terminate Permissives. With one or more channels inoperable, the affected containment spray, containment air return fans, and hydrogen skimmer fans must be declared inoperable immediately. The supported system LCOs provide thc appropriate Required Actions and Completion Times for the equipment made inoperable by the inoperable channel. The immediate Completion Time is appropriate since the inoperable channel could prevent the supported equipment from starting when' required. Additionally, protection from an inadvertent actuation may not be provided if the terminate function is not OPERABLE. 0.1. 0.2.1. and 0.2.2 With one channel of NSWS Suction Transfer - Low Pit Level inoperable in one or more NSWS pits, 4 hours are allowed to restore the channel to OPERABLE status or to place it in the tripped condition. The failure of one channel places the Function in a two-out-of-two configuration. The failed channel must be tripped to place the Function in a one-out-of-two configuration that satisfies redundancy requirements. O Failure to restore the inoperable channel to 0PERABLE status or place it in ~ the tripped condition within 4 hours requires the unit be placed in MODE 3 l within the following 6 hours and MODE 5 within the next 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, this Function is no longer required OPERABLE. R.I. R.2.1. and R.2.2 With two channels of NSWS Suction Transfer - Low Pit Level inoperable in one or more pits, one channel must be restored to OPERABLE status or the NSWS must be aligned to the Standby NSWS Pond within 4 hours. Failure to restore one channel or to accomplish the realignment within 4 hours requires the unit be placed in MODE 3 within the following 6 hcurs and MODE 5 within the next 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, this Function is no , longer required OPERABLE. INSERT Page B 3.3-114 O Catawba

issue Number l 34l Affected Section [322 ESFAS. RWST Swopover } Affected UnNs CNS: l Yesj MNS: l Noj Affected Pogos ITS: I ITS: 1 ITS Bases: B 3.3-98 i ITS Lases: CTS: CTS: DOC:: DOCS: NRG: NRG: NRG Bases: B 3.3-113 NRG Bones: JFD: JFD: NSHC: NSHC: Descripilon Lost parograph n1 STS Action K (ITS Action N) allows channel to be bypassed for 4 hours. CTS 3.3.2 ordy ollows 2 hours for the RWST swapover function. Correct STS rnorkup consistent with CTS and typed ITS. I l l

ESFAS Instrumentation B 3.3.2 BASES ACTIONS- N.I. N.2.1 and N.2.2

         .(continued)

Condition N applies to:

RWST Level-Low Low Coincident with Safety Injection.

RWST Level-Low Low Coincident With SI provides actuation of switchover to the containment sump. Note that this Function requires the bistables to energize to perform their required action. 'The failure of up to two channels will not prevent the operation of this Function. However, placing a failed channel in the tripped condition could result in a premature switchover to the sump, prior to the injection of the minimum voluma from the RWST. Placing the inoperable channel in bypass results in a two-out-of-three logic configuration, which satisfies the requirement to allow another failure without disabling actuation of the , switchover when required. Restoring the channel to OPERABLE status or placing the inoperable channel in the bypass condition within 6 hours is sufficient to ensure that the Function remains OPERABLE, and minimizes the time that the Function may be in a partial trip condition (assuming the inoperable channel has failed high). The 6 hour Completion . Time is justified in Reference 7. If the channel cannot be L O. returned to OPERABLE status or- placed in the' bypass condition within 6 hours, the unit must be brought to MODE 3 within the following 6 hours and MODE 5 within the rext 30 hours. The allowed Completion Times-are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, the unit does not have any analyzed transients or conditions that require the explicit use of the protection functions noted above. The Required Actions are modified by a Note that allows I placing a second channel in the bypass condition for up to J 2 hours for surveillance testing. The total of 12 hours to reach MODE 3 and 2 hours for a second channel to be bypassed is acceptable based on the results of Reference 7. l l l (continued) O B 3.3-98 Supplement 1 l Catawba Unit 1

ESFAS.. Instrumentation B 3.3.2

   . BASES ACTIONS         N.1. N.2.1 and N.2.2
       .(continued)

Condition N applies to:

                     *. RWST Level-Low Low Coincident with Safety Injection.

RWST Level-Low Low Coincident With SI provides actuation' of switchover to the containment sump. Note that this Function requires the bistables to energize to perform their required action. The failure of up to two channels will not prevent the operation of this Function. However, placing a failed channel in the tripped condition could result in a premature switchover to the sump, prior to the injection of the minimum volume from the RWST. Placing the inoperable channel in bypass results in a two-out-of-three logic configuration, which satisfies the requirement to allow another failure without disabling actuation of the switchover when required. Restoring the channel to OPERABLE status or placing the inoperable channel in the bypass condition within 6 hours is sufficient to ensure that the Function remains OPERABLE, and minimizes the time that the Function may be in a partial trip condition (assuming the inoperable channel has failed high). The 6 hour Completion

                    . Time is justified in Reference '7. If the channel cannot be O                     returned to OPERABLE status or placed in the bypass condition within 6 hours, the unit must be brought to MODE 3 within the following 6 hours and MODE 5 within the next 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, the unit does not have any' analyzed transients or conditions that require the explicit use of the protection functions noted above.

The Required Actions are modified by a Note that allows placing a second channel in the bypass condition for up to 2 hours for surveillance testing. The total of 12 hours to reach MODE 3 and 2 hours for a second channel to be bypassed is acceptable based on the results of Reference 7. (continued) l Catawba Unit,2 B 3.3-98 Supplement 1

ESFAS Instrumentation B 3.3.2 g BASES { ACTIONS .1. .2.1 and 2.2 (continued) requires the bistables to energize to perform their required action. The failure of up to two channels will not prevent the operation of this Function. However, placing a failed channel in the tripped condition could result in a premature switchover to the sump. ior to the injection of the ministas voluna from tne . Placing the inoperable channel in bypass results in a two out of three logic configuration, which satisfies the requirement to allow another failure without disabling actuation of the switchover when required. Restoring the channel to OPERABLE status or placing the inoperable channel in the bypass condition within 6 hours is sufficient to ensure that the Function remains OPERABLE. and minimizes the time that the Function may be in a partial tri condition (assiming the inoperable channel has failed hi ). The 6 hour Completion Time is justified in Ref . If the channel cannot be returned to urEHABt1 status or placed in the bypass condition within 5 hours, the unit must be broucht to MODE 3 within the following 6 houts and MODE 5 within the next 30 hours. The allowed Completion Times are reasonable. based on operating experience, to reach the reautred.. unit

     +               conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5. the unit does not have any analyzed transients or conditions that require the explicit use of the pmtection functions noted above.

D j

 /                   The Required Actions are modified by a Note that allows acing a second channel in the bypass condition for up to hours for survei lance testing. The total of 12 hours        j to reach MODE 3 and      hours for a second channel to be ~

bypassed is acceptable ased on the results of Referencef bl.1. .2.1 an M Condition applies to the P 11 and P 12((arpf JrD

                   . interlocks.

With one channel inoperable, the operator must verify that the interlock is in the required state for the existing unit condition. This action manually accomplishes the function ! of the interlock. Determinattart must be made within 1 hour. The 1 hour Completion Time is equal to the time allowed by I l (continued) 1 W0y5T5 B 3.3 11'3 Rev 1. 04/07/95 Cad

g issue Number l 3 51 Affected Section (3.3.1 Stsom Generator Water Level (CNS unit 1) { Affected Unite CNS: l Yesl MNS: l Nol Affected Pogos ITS: 3 3-16 (Unit 1 only) ITS: ITS Bones: ITS Bases: CTS: 3/4 3-12 (Unit 1 only) CTS: DOCS: A-18 DOCS: NRG: NitG: NRG Bones: NRG Bones: JFD: JFD: NSHC: NSHC: i v**a (S Table 4.3-1, Note 13 provides a filter time constant for the SG water levet function for unit 1. A firne constant is mt in the Unit 2 CTS. This note was added as port of on omendment request in 9/30/86 to delay function octuotton and ossist in reducing the number of spurious trips. The tirne constant has never actuoty been utilized in the plant and is proposed for deletion. l O v

RTS Instrumentation 3.3.1 Table 3.3.1 1 (page 3 of 7) O peactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCT]DN CON 0!TIONS CHANNELS CONDITION 5 REQUIREMENTS VALUE SETPOINT

11. Undervoltage 1(') I per bus L SR 3.3.1.9 a 5016 V a 5082 V RCPs SR 3.3.1.10 with a 0.7 with a 0.7 SR 3.3.1.16 sec response sec response.

time time

12. Underfrequency I')

I 1 per bus L SR 3.3.1.9 a 55.9 Hz a 56.4 Hz RCPs SR 3.3.1.10 with a 0.2 with a 0.2 SR 3.3.1.16 sec response sec response time time

13. SteamGenerator(SG) 1.2 4 per SG E $R 3.3.1.1 a 9% of a 10.7% of Water Level - Low SR 3.3.1.7 narrow rme narrow rme Low SR 3.3.1.10 span ur spanUI SR 3.3.1.16

14. Turbine Trip

a. Stop Valve EH IId) 4 N SR 3.3.1.10 a 500 psig a 550 psig Pressure Low SR 3.3.1.15

b. Turbine Stop I II) 4 0 SR 3.3.1.10 a 1% open a 1% open valve Closure SR 3.3.1.15

/ 15. Safety injection (51) 1.2 2 trains P SR 3.3.1.14 NA NA Input from Engineered Safety Feature Actuation System (ESFA5) * (continued) (e) Above the P.7 (Low Power Reactor Trips Block) interlock. l (t) Not used. (j) Above the P-9 (Power Range Neutron Flux) interlock. l I O l Catawba Unit 1 3.3-16 Supplement 1 L

ZC . f% 4

              .                                                  3                                                                aed 8     a .

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O IR i 2 f

HU WS1 1 1 f I I f I f 2 1 1 T _ NS _ OE _

  '                         II                                                                                                                  _

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9 ) 0 n I L l. R G A 3 3 1 I ie N 1 1 t D R l u N I TET CA O I A. s f t T H L s u

Wh A. A. A. 8 d s IRT VES EPE N A 4 n @ N l f N 4 1 a R W

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l a u n a a m DOT h L s a A l 5

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l f o

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s d n e n sri M . . . . m i t A webP A. A. A. A. A. ( A n s OOi S( N N N c l a u t I I N N Q h"A

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e J d A I. b9W L-1 5 3 9 l L L l t li o 3 i r C L [ s. a 1 R L E t E g-Q9@@ I . s s I 3 a A T n I n W W I C

M f A A. N O E' 3 r s P a n L E 3

0 1 4 I I N A. A. A. A. A. A. A. A. 3 R N N N N N T l A N N N t H i C u t e s l. n r k A t a u c E l e l o e s o ee o n R v o r l r l r o r ( e C r u C f e t L r o s t n u r o t s e t n o e e t c a e r v p u I r N l t t a c a e P a n m P u e R V i e e g . W e N t . i l n R - E p n sex a r o o ytu e R 1 o - y e t i Sal g t a e cs np v S t iF n8 r9 T ec a-pl pd e- I r g em ia e i en RP wP l T e a uu rV n j rmo o l U I n e t qP T n Trr r , P , l a Gw l e p b I et ex x - l o o v rt fn eo nt r u y rtu one wu wu L mL rs ra ol ol A iS T t TIN PF LF 8 A a- ep el b e c W N ew dm do r fF a A O to nu no u . e . I T SL UP UC Ta b SE aS Ra y. r. T A C C d g C. I . . N I F 'M W. '5 M r. J J 2 3 1 1 1 1 4 l

                                                                                        's l

bl o 4 _ O b s-

S Pet'bcAbs' '5.3.I

                                                                  .            11-1 ABLE          _ (Continued)                                    ~

TABLE $TATI0lds f Only if the R ctor Trip System breakers ap n to be closed and the Cont Rod Drive 5 em is capable of rod wit rawa

                            #       Above P     (Reactor Trip on Turbine                ip Interlock) Setpoint.

H Be P-6 (Intermediate Range tronFluxInterlock)Setpoint. { ass a.1 o in u nw setnaint P r Ranoe fleutron r1um fatarlachi t. int. If not performed in previous 31 f~ sa. 3.3.1.2. @ e-rina i . calorimetric to encore power indicationMvp45% of ftAT OAl ( 5

               ,,, g ,                        ,g.         .j ust -- - . . -

_...._. venis w ....n .-m-

                                                                                                                            - /
                                                                                                                              -rm amp d"'

set.3.).t.t /g _--t uts differmera in aramese *"- 9dri Ine Jfovision examei f fsd'!ian a a Vsem an+ andlleakle far 4*rw inth=W 7 a 1 ) f SR L1I- @ $1nale noip* of incore to encore anfa' flux d fference%hovM50 1 /g iha= _-- 1 ==r= n .. If. the abso' ute dif"erence n grug iNdg Ethma ar --9 ta 3%.I neu onlicahWforAntrW intMvnsivisa er2 or Ify__sricas - 4.U.9 7 ..,g g

n. eve t g , (llestron detectors may be excluses tres OlNOIEL CALIBRATIch 4A3 5) DetF :fr pla c---- saan be obta,%. aval ="=e and c Jers.; t fa*#-me's pm3 For the Internestate Etie -

l

              / gm 2,                                                                                                                 ex                 i g Nwma.1JI channels             into IIODEthe 2 orprevisions

1. j of Specification %ge and Power Itanee SR 3llla Incore - Excore Calibration gp@41.a, y .m m jaDove ar.

j of RATED

                                                                                          -mm.                  POWER. ine
                                                                                                         , e (. . .

vi-or i u Each train shall be tested at Isast every days on a STAGGERED TEST BAS!$.

        / urs ro e a='m

( g 39,g 6cation that 111ance an ---- c. include verifi-ruissives P F-lo are . ard

                                                                                           ' n theirMshall r9tps' re@d       state for existing           conditions                                           wissp s stasu v ivow              -

g#

                  ,g Setpoint verification is not applicable.

i

11) T/ TRIP ACTilAT *;1VICE OPERATM TEST include 'ad t verifD A2I @ Antion of the r r 4H ity of the undervolta and Sheet tr:ys.

i j

hwUt& Sat 1 1,s.tu. t filt # tisBr constant b W is adjustH t l N W15' ' - " value s v.tWaJr [ pmr.,,k J 3 I CATAWBA - UNIT 1 3/4 3-12 Amendment No. 148 l O g j2. o[ '1 \

Discussitn of Ch:nges S:cticn 3.3 - Instrumentatien ADMINISTRATIVE CHANGES A.67 CTS Table 4.3-1 requires a TADOT on the manual initiation function for reactor trip, reactor trip bypass breakers, and SI input to reactor trip. ITS SR 3.3.1.14 indicates that the verification of setpoint is not required for this surveillance. The TADOT definition includes setpoint verification, however, these are manual actuations with no associated setpoints. This change is administrative and consistent with NUREG-1431. A.68 (Unit 1 only) Note 13 in CTS Table 4.3-1 concerning the filter ti?? con ~ tant in the Unit 1 steam generator low-low level reactor t,"; circuitry is being deleted. Unit 1 presently has no filter time constant associated with this circuitry. This note was added to the Unit 1 Technical Specifications on Sep; ember 30, 1986, via license amendments 13 and 5 for Units 1 and 2, respectively. The purpose of the time constant was to assist in reducing the number of spurious low-low steam generator level reactor trips that occurred early in the plant operating history. The subject filter time constant was never actually implemented. Deleting the subject note is considered an administrative change. 3 (O Catawba Units 1 and 2 Page A - 18M Supplement 15/20/07l l

1 RTS Instrumentation ' 3.3.1 O' fehle 3.3.1 1 (psee 6 of 8) teacter Trip System instrumentatten ApplicAatt esses cm ofum WECIFlm ageWim M i maas s Malcflou CINetitems reassama n astEILLAst2 gaty (g ettggus meWgmarts yhtig agtpont II. Geester Cootent pwp (str) e Peef tien

e. single esp 1(h) por 0 N 3J .M M

b. 1 Lemps 1III 1 per N 88 .3.1. 4 M M Il Unitervet tese

                 "'                        1 per
                                                                           )ll.      M 3.3.1.9        8          W      k             V "lii:l*            ar .,. x           :::WW )

gL)4 saunerfregency

                =                          tdi h per                pl,.      Se 3J.1.9                  Its t

sa 13 A C stone lii:: st, tat t " M 4a=at - a. t 1.2 Attmer a a 3.3.1.1 eenereter (SS)

Water Lovet -Lee SGC M 3J.1.T k s$ t w gI myttraab et w EE J. . y rfma 1 85 Uhter 1,2 2 per SS M t 135.435 t (32.315 at 3J.1.te M 3J.1.M telnetemt wt 1,2 3 per stems fled _g a 3J.1.1 s I s 06e s Feeshester a 3.3.1.7 futt futt steen

              .llesetch                                                             a 3.3.1,13      fL et sty flees et sty p

a 3J.1.4 . _ (eentinued) Revleuer un le ; tens apr senteln Atteumhte Yalue slapend en Setpoint

  . a) ~ t.e - , . ee ,s            r  et.r ,,,s .Lo , ,nte,4         .
  $ f/) Aheve the P.8 (*eme a: ige muutron Flm) Intertect.

h G4 Above the P.T (Lau pesar teacter trips Steekt interleek and hetes the P.S (Pemer Renee seutrue f tm) I d a M M k/ Jb $4. U- W Is a I.T fet. h '

        %.ws3                                                                                                                                ,

4XhT5 3.3 18 Rev 1. 04/07/95 O

C 8 g leeue Number l 36j Affected Section l3.3.2 CTS Function 17 l Affected UnNe . CNS: I Yesl MNS: l No! ANocied Pogos ITS: ITS: ITS Saees: ITS Boase: CTS: 3/4 3-25, 37, 54 CTS: DOCS: L-7; LA-4. 5 DOCS: NRG: NRG: NRG Bones: NRG Booms: JFD: JFD: NSHC: 48,49 NSHC: DescripNon The IIS sutrnittoi proposed to relocate the DG ventilation operation function to the SLC (LA17). The existing actions , which are to be relocated are more restrk:tive than if the octual DG were inoperable. DG Ventilation is a support system and should not specify more restrictive actions than the function supported. L23 odded to justify deletion of this requirement. 1 1 I

v" y a"- g i n t O ' hn Y a i i t a o O F 6I(f M 8 1 1 2 i i t n o d. j 'd 3 #h 2 E 4 4 O a r e p [ / L B . . r ' N O % AE L0 i0

                             -       3         3 t

o 3 3 3 3 3 I , a , . . .

                %. r  tN                                 r 2
                                                                                                 /.

T ' 2 2 2 2 2 A r - e T , n . a [ N - 1 1 e 1 1 1 l E - G t s - l . 8 t R l ( ie nn 4 1 T SE s A t g S MLL e a a

N I M E T I UEB MNA NAE IHP NR 2 Mi D y c n 2 2 3 2 t

3 s h c a e s N t o n d S NCO e e e Y g m u S r e . d n - n i N m g. e t O E nn m n I

                        /                                      ,

ie A o T S l s . g

       )  C   A     LP                 .                lt a

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        - i i

T C EI NR r - t s e p 2 A NT 1 1 o ; 2 2 2 2 os A fi - t,d Q t r a HO CT eu ve i E u oe br

                                                                               -      s             .

L T n B A " a e A E .d g T F S n 5a . Y L s T dE 1 nn s f t O E E s 5 F I eo / 2 A S A H C 2 ti It 3 3 7 3 n r a r r m I D' hi i c r o t S( h 4

3 n k([(7 G g a s - N o o r sk - E i L s e ec . r _ t a ny e n ro e e u oa G ul tr r u [, t _ t o n il te ae s s 1 2 4 n l et s s - -

                                     !      aR       e            Fn      e      e         P       r e       t      u       s                 I   r      r   P o                           _

V a tn e y P P . , t - i co i t m p a _ Ai D ee r r - i d t i n ca t yn co ft as e z e z T mirT r . TT l I iu tt ni Sy i S r w r 1 - Ti i c et d u o o T s un a A ga en s L t I - Bo u rr ro sl - c N i n a u ee mp ei eI e w a U [t r lt et r- o e ea sr M Aa EO na PP P L R - iu O ee gt A - I T ip DO nc B

                                        .    .        .           EA                                                 W
                                                                                          /.

C N . a b c . /. c A T U 7 d b A A F 1 J C

f h - f O F ,P g

Na ch %',,, g y o,, bd'- 3pec k ed3a M .1. 4 2.-t TABLE . (Continued) O m.is- 0 ACTION STATEMENTS fContinued) kLka-a4.I Iaw ACTION @ fWith e nunner or uuL5Lt. cnan s one less th U . A '" '# 3 '" 3 L.,s N er of Channe STARTUP and POWER OPERA the may 8 oceed ov_ided the fo owine conditi e a~ - M ' d :

    '. p oe i b The inoperable chan N                         within I hour,j@ nel is placed in the tripped condition                                k h            u ._

i ----i , norna ni v q. - - - - 1_ ,%m w.tg c g ,g for survelliance testing of other channelsg)- ==runan hannel m w v. _ m g.3,s ACTION 17 - With less than the Minimum Channels OPERA 8LE requirement, opera-tion may continue provided the containment purge supply and w xhaust valv=< are maintained closed. ACTION - 51tn Ene r of 9 a s one le r than 'M

                    !qg@[s                     rn                me r 8reme. nt     cnan it the noperab'e channel to u +,.;,,

Qa.rs=tae., taA8LE status within 48 hours or be in at least NOT STAlW8Y within following the next 6 hours and in COLD SHUTD0681 within the - 30 houu. ODACTION $ 'With number of AIow i . A, 9, e - m. one less th To N of Channels /s=_--?nE STARTUP and chann POWER OPERA may p T.tepae ia Sa 81 p Att n ad the falldirine conditi e ar* <mtisff w I af,,,,/ The inoperable channel is placed in the tripped condition A oeas* % n u t within 6 hour g E% N* channel may be bypassed for up to 4 hours O i for surve'11ance testing of other channelsqD (q--

       % io ,wooe 3 .,, w.c7:

ACTION g%--- m- - deterni== ey

                                                             -------~~r==---

o go .ao,..qg us+ ---- n wrv.u n er une ar .-- - -..w.ithin 1 hour Qar tne interlock is in ts requirse state ss. T l . . 11

       .,,4
       '    g, m.oey ;. in <t} Lr              amt for tne    LA.i s emzu ny gasus condition. arranniw -                             o r=-
                                                                                                                 ' =J.
                                                                       ~

ACT! -

                                                                                    ^'
                                              -- - I           aner 72 _.- -
                          % hin                                                           restore the inoperable                    to g,tnABLE status within 6 hours or be in at least HOT ST LO   "*                                                                                                              '"

OA.31j within the next 6 hours and in at least NOT carTnnam' within the following 6 hours; 65) * (M 1 to 4 hours for survetT ance,testing one cW'uty be bypasses for up asnn.cancannin.e.a ( me provided the other Mis OPERABLE. m.m= 1= ACTIUR Zla - With the number of OPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, be in at least HOT STAND 8Y withi in 3;gf 6 hours and in at least HGT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERA 8LE. _ J o CATAWBA - UNIT 1 3/4 3-27 Amendment No. 148 ACUW I One. buiws ape,J4. , s2e J,,,. h w:. 46 OPE p cLt. S W s 6 4s=d of' W. in MooG 1 n e t. l swL N6Tk?* OnaAre: % Gs PnWE O vie b 4 haw"A 4 , rary. h se M3 6 Pr*wid / 4 k * % ,- +Qr.

                        @COi          h l

Q (A a s NP l l A. l a , 3 E L T r B A v u n I I N a r o N r o R f e 6v _ 2 s AEs f f s TF u t i l T e e (3 _a Y LA v T ER o E NR I b $ 9 O F Il a . . 4 A Al ) . t. 5 S NA A. . - CC d e

                                                          .                          A.               5                  3,  J                        u    3 1

n N N r a r r L E NE i t n n u I t s e 1 n mm ee tr e Q 3 R X / An c u

3 4 m NG o . Ii I AE C e . eq u . . . G N ( e A. A. a N E CQ n S M N ee SR s N A. N

                                                                                                                                      -      A.

N i o sk ec _ t n ro - a o ul _ r n no t r , 2 e n o i ae , lt 1 p o et - gO t i c ea Fn - n i t i sr I P it t c r a gn ee y l , ds e ge l oo ip t n I l u j np Li DO ee r- , . g, t ft l a uh i n iO n cnaou yr as eP z T d I co Sy ie 1 Sx 1 n yE y f o l a t s ea nt S rr w T t ui a m y gr d en uu ss ; o I rd e a n u ou Bt n re ro ss L N ae f ; - U i r l e i t xl S a ll ei st en a ucn M AAa ttd E e m ei et na iu ee rr PP g L w o A ui . gt B ie . . . nc _ AF c DV a . W (n b c EA 4 c A T Mb F

f. [

7 1 g A C 8 (s y y

                                              ,g g$                             L gg L

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        .                                          R                                                                  h v

ti eo EY s l a at TAT . ) v Wro SLS AEE NRT A. N ( 1 M**o d a r o ep c N O I NS OE T / l f o s p m u ic vi rf ei T IT n p sc A T T AC o rp e N UI . ) i e as E TG t h e-N CO A. 1 ( a t t AL N

                                                                                                                  . u                    i. n i              8 U                                                           M                   S t                f                  n            4 R                                                                               I       c           o            a r u               1 T

L S a S N G A A B g n o efh n I N n . N O i o M I I T i S d t a Ere ye t N

ES TT PUIRT TET ACA )? - E u Tl t u om t n dS N ITVES 2- . c c se . e e u YE RCEPE dy D x a nb4 m i t n S NR OI E f TADOT 3 3

                                                                               #                    S N

E e R E g G n l a n e utr ruso d n e

    ,                    nI U                        L                   2                          O         Gi                             em,                 m oT Q                        A 3

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s J-a r t A N RT ES E n a ol u

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orE u e L L I OT N Q A T B s r y e c x e i s t T E N a r n 0 8 A V O d o g o AER s n m n TF U I n mei r y T E Y S LA ER NB T

                                                                                         /            Q3w     y r

e s a i t s e oc cns ai O%F NI t nl 5 F ' AL e e olt 5 . A HA A. v a y ii - S e t a ten 3 CC N A D t o l l e evu mr 4 E s v r ue / N R E E L E NK l a ef o f o Sh t 3 N NC . I G AE H A. tt a is t n4 N E CQ sk s N S i s =

                                                                                                                                             =ro ec                                       d                   s ro ul t

e n s sJ o ct S n

a. 3 n

                                                                                                                                                    ,l y

t r e r Fn ae et P 4 r t i* s = l

                                                                                                                                               -2m,o I

e ai a e y b nt h ,t tm p l s w s i.1 . s y ee i gz'g . n LDES l . ft r a nl as T h ei Or Sy S r e G s ud t) s =Met 1 d o n e- e a T en ro t i r tJ_f W I N c a e r ym y e. 1 L ei a s U E et 2 t ei l e;i N na R nt h t . i -

                                                 -  N           iu                                          h       t          t      e .t v A             gt                                         c nd n                              ir            A H             nc                                         a on o i

ne B W C EA I. Ena M 2uS E. R

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t a l E s L BS AE 4 3 3 4

                                                      ,      r o              3

[ 3 3 3 f 3 a a CD t l IO , , a , , , , , T LM 2 2 r 2 2 2 2 2 P e A T P A 1 1

                                                      ,      n e              1 1

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                 ]t sLLEB i

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                                                                                                                  / a pt eos 0

8 M E NAE c t IMP n n i T MCO e e d S g m e Y r . d m m S , n e n i l i e g m t G A s a I T l . i t A S lt e a U L ane t r . 0-

        )(C 1

T C EI IR rm s es pt A eT l A 1 o fm 2 2 2 2 o 1 t HO ee l I 2M 1 C r e CT vg oe _ u f. 5 r br n T a a A A T E F

                                                               .d n                                                  ,

5a Y 1 s a t mn 5 O u s r eo 3, / 2 g 2 2 ti 3 3 . 4 A MM it . 3 S I C c e 4 D en ^ / r eu w ., 3 (W a r r G u l n i c g t r o a Sf s

                                                                                     / G$

I o o r sk l E i L e ec , . r t a ny e s n ro ul e r e r t e l oa G tr u u 2 a n ae s s 4 W il

i 1 t o te l et s s - r n aR e Fn e e V e i t a tn i co i u s e y tm I P r

                                                                                     'P r            P
                                                                                                       ,Pg    . t o

g ee r r i g n t i n Ai ca t D yn ft as e z e t T mir T d i iu ce Sy i 2 1 tt ac nt et d S r u w o to r T mA o ga rr en ro s s1 L

                                                                                                      -     c e

s I N l td ee mp ei e1 e w a e U lt un et na r- r o e ea Aa EO PP P L R - sr iu _@ ee A gt ip nc 8 DO . . . EA . . W a p A ch b

                                  .                                    .                                                                        T f                                      s         b                                                                 A
                                                                    )                                                                           C

[1 f I t' g g_ O . i lj llll

tk w 3. p.p.

se - e o- .3. 3.:. F y I o.,)3 TABL continued)

                                                                                                ,                          (,1-s:~ d.kapset &

g,lf g ACTION STATEMENTS (Continued) ACTION @ - ith e number of 0 SLE channels on less than the Tata 0.3 N r of Channel TARTUP and/or R OPERATION ma roceed g a stal 3 !a ~14. rs ovided the fo wine cond"4 aae <2'4*ff*d? q geos .i ,*n 37 hec. r The inoperable channel is placed in the tripped condition L - within 1 hourj @ t w s t') 4=- ,e te - -

                                  #               @ (tw mi-i-                  , - - i s     un-== F4-                                 no ven,           j
                       /Q"              @c t.p,6 channel may be bypassed for up to 2 hours                                                                I f       --411aare testing of other channels e                                                 i' ecif.i dtion a_3.er m, h04 D                     T!ON 17            With less than the Minimum Channels OPERABLE requirement, opera ld 3 34                                   tion may continue provided the containment purge supply and                                           '
                        .                           *-k= <t valves are maintained closed.                  _

JS - (With the n ^"r of Grunn

                                                                                                                                      ; Mnieus)

ACT! A.I F reami.pr. cnannels U restoreone tnehasti than the ie c unnel to

                          -o ,,.Mf"p ff'----Il      OPERA 5M status within 48 hours or be in at least NOT STAM8Y v..

(,..peq within the next 6 hours and in COLD SHUT 00let within the Al follot';g 30 hours. m ,%.y3 3

                                                                                                                                                *mlk 1 t+ .0          A.34     ACTION IS -          @iththen                r of OPERABLE                    is one uss              he TotaT Number                    1s, STARTUP             /er POWER OPERA             may proceed h.,.,or3 f,n.,.%.

prow the following c tions are satisfi : g The inoperable channel is placed in the tripped condition g ,, w 4 '.. it k.esk within 6 hourg The Rur c'-- 7 s GPfkf u =21 - 3 is meti hsGeQ P3 A3 g.rg jos 7-o.jy Q(p; _

                                                                       - -- channel may be bypassed for up to 4 hours ~
                                                            .or surveillance testing of other channels gp g                                              *
                                                                                                             @,w, ama.aa*l eaoperme c ACTION        -      ha Iman +ka- M utas                          r--     -
                                                                                                           < dr6 -- . within 1 hour _ u.2.

1 ,,, fwoDE 3 ," y y""8 ** M dl W oW ue=='T= M ENE UT J;ne 5lisoClaus pef 91ssives.6ua i LS^ ^ b m ,, + '' * * ='8 DMIiat the < nterlect < s in its requires state for tne l ex st'no N condition, or-- i r a.=u m u --2.va. 1

                                     @                                      w .,.          _

ACTION O - the a f OPERABLE _ - I< - M * ' -- "- "f a ' S annels am r -4.- , ', restore the inoperable {a."**g

                                 *"                OPERABM status within 6 hours or be in at least H0T S within the next 6                    and in at least Hof SHUTnnWM within the Y           .

1 M) one CllMEiney be bypassed for up l g3 u' fol. to lowing hours6 hours; for survemance testin, -r -c-i ~.- provided the other @ s OPERABLE. i gy.,.

ACTION Zla - With the number of OPERAS [E channels one less than the Mini % I

         /0#g        p                               Channels OPERA 8LE requirement, be in at least HOT STAN08Y withirf 84                                        6 hours and in at least H0T SHUTD0l#1 within the following 6 ff5 U#                                   hours; however, one channel may be bypas;sd for up to 2 hours for surveillance testing per Specification 4.3.2.1 provi]ded t other channel is OPERABLE.

CATAWBA - UNIT 2 3/4 3 27 Amendment No. 142

                                               ><., +rn.'. ."a*p.ra.11c 4 %,v + M a h .VEM&f .s+ds [a /* As *

acq 7,,-/t.x.. .rsors:c>,, Ac,y,og I or or r . - p 6pa.K. 4 to

                                                        + so    es   v     w   .s.va.,.a.         u  ,y   p  ies  r  ,,  w   m   ;,, ;, ,, m ,
                                                                                                                     '    *N3 3r *:t 06 rsaant$ A s                         08 <- GNLA*6$'                                            # M*CE 3 (CTn) f
                                                                                   .s.       e.r t

O ,_ _. . . , f t >e *4 + ~ m .r Maaac/s. D 4 . fg f.c s m'/'aees. E"'] e4 M At3< t4 of- 'i '2.

Jfl

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d n e o t i s t u a sj r id e p i r a h O T gs S r ii T l H N o e t I t a v en O r e t a P L at T e Rs E n r n S e e eo G t vc P a i I l W t e R e ae T N O I l O s e t r o a r

                                                                                                    '     N f

et

                                                                                                             - s ei               2 T

y rh 4 A c ut 1 T n s N e g st E E r . ea o s U es m as rh N Pt a L A me ee

d T V Eu l g g tu Sl ee nr t n e S e s N l a i i a iu m s E lV s s lV Ls I a L a p p l S n d i M B e P ael N me n e t E A rl 4

O a -

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Discussion cf Changes Sectit:n 3.3 - Instrumentation TECHNICAL CHANGES - LESS RESTRICTIVE LCO 3.0.3 would be required. ITS 3.3.8 requires one ABFVES train to immediately be placed in emergency filtration operation and to insnediately enter the applicable Conditions and Required Actions for the other ABFVES train made inoperable by the inoperable ABFVES actuation instrumentation, or to immediately place both trains in operation. The, first set of actions are acceptable because they accomplish the actuation instrumentation function and places the unit in a conservative mode of operation consistent with a single inoperable train. The second action is acceptable, because, by placing both trains in operation, it ensures the ABFVES function can be pcrfonned even in the oresence of a single failure. This change is consistent with the intent of NUREG-1431. L.22 Not used. L.23 CTS Table 3.3-3, 3.3-4, and 4.3-2 specify requirements for Diesel Building Ventilation operability. These requirements are not included in ITS 3.3.2. CTS Table 3.3-3 actions 18 requires an inoperable manual start switch be restored to operable status within 48 hours and action kla requires the unit be shutdown if O the automatic actuation logic for the ventilation system is inoperable. CTS 3.8.1.1 and ITS 3.8.1 requires that on inoperable diesel generator be restored to operable status in 72 hours. The ventilation system is a support system for the diesel generator and does not directly mitigate any analyzed event, therefore, a separate more restrictive TS requirement for the ventilation system operability independent of the diesel generator is not necessary and is not consistent with other mechanical system specifications (e.g., ECCS pumps) which also require ventilation. Each diesel generator is provided with separate ventilation systems and the loss of one ventilation system does not offect the other diesel generator. The definition of operability requires that all necessary attendant support equipment required for the diesel generator to perform its safety function also be capable of performing the support function. The ITS 3.8.1 requirement for on operable diesel generator is sufficient to ensure that the diesel generator building ventilation system is also operable to support diesel generator operation. O Catawba Units 1 and 2 Page L - 77 Supplement 15/20/07l s _ _ - _ E

Discussitn of Ch nges S:ctica 3.3 - Instrumentaticn TECHNICAL CHANGES REMOVAL OF DETAILS LA.14 CTS Tables 3.3-3, 3.3-4, and 4.3-2 Function 1, Safety injection,

      ' includes all the functions initiated by a safety injection signal     i (e.g., Reactor Trip, Feedwater Isolation, etc.). This descriptive information is moved to the Bases of ITS 3.3.2. This descriptive information is more appropriate in the Bases of the TS. The ESFAS Instrumentation in ITS Table 3.3.2-1 lists the signals which produce an ESFAS signal. Changes to the Bases are controlled in accordance with the Administrative Controls in ITS Chapter 5.0 and     )

1 require a 10 CFR 50.59 evaluation to change. The 10 CFR 50.59 evaluation ensures that any changes to these require.nents are appropriately reviewed. This change is' consistent with NUREG-It31. LA.15 CTS Tabic 3.3-3 AFW functions contain descriptive information about which initiating signal starts the motor or steam driven pumps. This detail is moved to the' Bases for ITS 3.3.2. This level of detail is not necessary for the TS and is more appropriate to the Beses. Changes to the Bases are cortrolled in accordance with the Administrative Controls section, ITS Chapter 5.0, and require a 10 CFR 50.59 evaluation. The 10 CFR 50.59 - evaluation ensures that any changas to these requirements are appropriately reviewed. This change is consistent with NUREG-1431. , LA.16 CTS Table 3.3-4 Feedwater Isolation and Auxiliary Feedwater for the SG Level-High High (P-14) and Auxiliary Feedwater for the SG Level-Low Low listed in the trip setpoints and allowable values of

       " narrow range instrument span," are moved to the Bases of ITS LCO 3.3.2. This level of detail is not necessary for the TS and-is more appropriate to the Bases. Changes to the Bases are controlled     i by the Administrative Controls, ITS Chapter 5.0, and require a 10      .

CFR 50.59 evaluation to change. This evaluation ensures that l changes to these requirements are appropriately reviewed. This i change is consistent with NUREG-1431. LA.17 Not used. CTS T:ble 3.3 3, 3.3 4, :nd 4.3 2 for 010;:1 Sailding ! Voetil:ti0n h:: 5 :n r:100:ted t0 the Selected Licen;;; C;--it=nt; $n=l(SLC)(UFS?,R Ch:ptcr 15) . The =ntil: tion y te: , i: : : pp;rt ;y t f0r the dit;:1 ; ncr t r, h = = r, th: :y:te- l i: n:t ::=:d t mitig:t : DSf, or tr:n icnt. Relocati:n of dese O Catawba Units 1 and 2 Page LA - 49 Supplement 15/20/07l

Discussicn cf Ch ng:s S:cticn 3.3 - Instrument:tien / TECHNICAL CHANGES - REMOVAL OF DETAILS V type: cf n : tem; cut:id: Of the Tedrie:' Specif t:: tion: i

n i;teni -ith the phil : phe f JREC l31 and with th:

10 ti:n criteri: f;r inclu icn within the Technical Spt:1fi :tieer LA.18 CTS 3.3.3.6, Tables 3.3-10 and 4.3-7 for Accident Nonitoring Instrumentation, contains information which describes the plant unique identifier for some functions. This level of detail is not necessary within the Technical Specification and is moved to the appropriate plant procedures. The plant procedures are controlled by administrative controls which ensure that changes to there requirements are appropriately reviewed. This change is consistent with NUREG-1431.

LA.19 The specific information located in CTS SR 4.6.4.1 which requires the Channel Calibration to be performed using a sample gas containing hydrogen and the specific calibration points is moved to the Bases. Changes to the Bases are controlled by the Administrative Controls, ITS Chapter 5.0, an.* reyuire a 10 CFR , n 50.59 evaluation to change. This evaluation ensures that changes to these requirements are appropriately reviewed. lhis change is consistent with NUREG-1431. LA.20 CTS 3.3.3.5 Table 3.3-9 lists the Readout Location for remote shutdown instrumentation. The change moves this level of detail information to plant procedures. This type of information is aot necessary in the Technical Specifications. The requirements l retained in IT3 3.3.4 provides sufn cient controls to ensure I operability of remate shutdown systems. Changes to plant procedures are controlled by Administrative Controls which ensure that any changes to this information are appropriately reviewed. This change is consistent with NUREG-1431. LA.21 The specific detail in ClS 3.3.3.11, that the Boron Dilution Mitigation System operates with Alarm Setpoints at 4 times the steady-state count rate is moved to the Bases for ITS 3.3.9. This level of detail is not necessary for inclusion within the actions. Changes to the Bases are controlled by ITS Chapter 5.0, l "Admini trative Controls," and require a 10 CFR 50.59 evaluation. The 10 CFR 50.59 evaluation provides adequate assurance that any G V lCAtawbaUnits1and2 Page LA - 59 Supplement 1 5/20/97

l l No Significant H22:rds C:nsid:raticn ! S:cticn 3.3 - Instrument: tion LESS RESTRICTIVE CHANGE L.23 The Catawba Nuclear Station is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1431, " Standard Technical Specifications, Westinghouse Plants." The proposed change involves making the current Technical Specifications (CTS) less restrictive. Below is the description of this less restrictive change and the No l Significant Hazards Consideration for conversion to NUREG 1431. CTS Table 3.3-3, 3.3-4, and 4.3-2 specify requirements for Olesel ButIding Ventilatton operability. These requirements are not included in ITS 3.3.2. CTS Table 3.3-3 actions 18 requires an inoperable manual start switch be restored to operable status l within 48 hours and action 21a requires the un't be shutdown if 1 the automatic actuation logic for the ventilation system is inoperable. CTS 3.8.1.1 and ITS 3.8.1 requires that an inoperable diesel generator be restored to operable status in 72 hours. The ventilation system is a support system for the diesel generator and does not directly mitigate any analyzed event, therefore, a separate more restrictive TS requirement for the ventilation system operability independent of the diesel generator is not necessary and is not consistent with other mechanical system specifications (e.g., ECCS pumps) which also require ventilation. Each diesel generator is provided with separate ventilation systems and the loss of one ventilation system does not offect the other diesel generator. The definttton of operability requires that all necessary attendant support equipment required for the diesel generator to perform its safety function also be capable of performing the support function. The ITS 3.8.1 requirement for an operable diesel generator is sufficient to ensure that the diesel generator building ventilation system is also operable to support l diesel generator operation. l l In accordance with the criteria set forth in 10 CFR 50.92, the Catawba Nuclear Station has evaluated this proposed Technical Specifications change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?

The proposed change increases the completion time when the diesel generator building ventilation system is inoperable to 72 hours and eliminates the separate TS requirement for the system since it lCatawbaUnits1and2 Page 47W of 52% 5/20/97

N2 Significant H z rds Crnsider:ticn S:cticn 3.3 - Instrumentaticn is required by the definttton of operability. This change will not affect the probability of an accident. The diesel generator building ventilation system instrumentation are not initiators of any analyzed events. The consequences of an accident are not affected by this change. The function of the ventilation system is to provide cooling to an operating diesel generator. The definttton of operability for the diesel generator already requires that the ventilation system also be capable of performing its support function. The change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, this change will not involve a significant increase in the probability or consequence of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

This change will not physically alter the plant (no new or different type of equipment will be installed). The changes in methods governing normal plant operation are consistent with current safety analysis assumptions. Therefore, the change does not create the possibility of a new or differe:t kind of accident O from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety is not affected by this change because the diesel generator, which provides the primary safety function, already permits a 72 hour completion time. The functton of the diesel generator building ventilation instrumentation is to actuate the ventilation trains to provide diesel generator cooling. The proposed change accomplishes this by recognizing that the requirements for ventilation system operability are already part of the requirement for diesel generator operability. The safety analysis assumpttons will stiil be maintained, therefore, the change does not involve a significant reduction in a margin of safety.

 .Q Catawba Units 1 and 2       Page 4844 of 5260                                 5/20/97l

, leeue Nunher l 37l Affected Section l3.8.3 DG Fuel Oil Lube 00, and Storting Air ] Affected Units CNS: fM MNS: [ Nol Affected Pogos ITS: ITS: ITSBases: B 3.8-44 ITS Bases: CTS: CTS: DOCS: DOCS: NRG: NRG: NRG Bases: B 3.8-47 NitG Boeos: JFD: JFD: NSHC: , NSHC: Deecription .. The Boses of SR 3.8.3.3 requiresDG fuel ou particulate determinations be made in accordance with Adb D2276. Th!s standard is for jet fuel and requires a 0.8 micron filter. Clarify the Bases that a 3 micron filter is osowed for hr OG fuel oil porticulate determinations. This was in the McGuire Boses at the time of the ITS submittal. O O , I

Diesel Fuel Oil, Luba Oil, and Starting Air B 3.8.3 g BASES y SURVEILLAWCE SR 3.8.3.3 (continued) REQUIREMENTS Failure to meet any of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO concern since the fuel oil is not added to the storage tariks. Within 31 days following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM 0975-b1 (Ref. 8) are met for new fuel oil when tested in accordance with ASTM D975-81 (Ref. 7), except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 (Ref. 7) , or ASTM D2622-82 (Ref. 7). The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an inmediate ' effect on DG operation. This Surveillance ensures the availability of high quality fuel oil for the DGs. Fuel oil degradation during long term storage shows up as an increase in particulate, due mostly to oxidation. The i presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment, . however, which can cause engine failure. Particulate concentrations should be determined based on ASTM D2276-78, Method A (Ref. 7). The test described in this Standard is for jet fuel. It is therefore permissible to determine particulate concentration using a 3 micron filter instead of the 0.8 micron required oy the Standard. This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/1. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing. For those designs in which the total stored fuel oil volume is . contained in two or more interconnected tanks, each tank must be considered and tested separately. The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals. (continued) d l Catawba Unit 1 B 3.8-44 Supple;nent 1

Diesel Fuel 011. Lube Oil, and Starting Air B 3.8.3 BASES O. 5 SURVEILLANCE SR 3.8.3.3 (continued) REQUIREMENTS Failure to meet ry of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LC0 cor.cern since the fuel oil is not added to the storage tanks. Within 31 days following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM D975-81 (Ref. 8) are met for new fuel oil when tested in accordance with ASTM D975-81 (Ref. 7), except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 (Ref. 7) or ASTM D2622-82 (Ref. 7). The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limitr, would not have an innediate effect on DG operation. This Surveillance ensures the availability of high quality fuel oil for the DGs. Fuel oil degradation during long term storage shows up as an increase in particulate, due mostly to oxidation. The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause (~N fouling of filters and fuel oil injection equipment, V however, which can cause engine failure. Particulate concentrations should be determined based on ASTM D2276-78, Method A (Ref. 7). The test described in this Standard is for jet fuel. It is therefore permissible to determine particulate concentration using a 3 micron filter instead of the 0.8 micron required by the Standard. This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/1. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing. For those designs in which the total stored fuel oil volume is contained in two or more interconnected tanks, each tank must be considered and tested separately. The Frequency of this test takes into cor. sideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between j Frequency intervals. ) l 1 Montinued) l 0 l Catawba Unit 2 B 3.8-44 Supplement 1

Dies 21 Fuel 011. Lube 011. and Starting e B 3.8.3 g BASES t SURVEILLANCE SR 3.8.3.3 (continued) REQUIREMENTS Failure to meet any of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO concern since the fuel oil is not added to the storage tanks. 7 Within 31 days following the initial new fuel oil sample, g( the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM 0975j Ref. ! are met for new fuel oi' when tested in accordance w th l ' _( ASTM irur may DE 0975 g sorsedD P(Ref fwith in accordance ), ASTM except that the analysis for D1552,4 or ASTM D2622i C PtRefJ5).1Th_e 31 day period e l accep Decause re Tuel oil properties or interest. @ even if they were not within stated limits, would not have an immediate effect on DG operation. This Surveillance ensures the availability of high quality fuel oil for the DGs. Fuel oil degradation during long ters storage shows up as an increase in particulate, due mostly to oxidation. The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment. however, which can cause engine failure. y ,g Ol g P

                      ,q.articula':e
                           .;    .m     enneantrations should be determinedIS
                                     .eus ASTM D2276O }f Method A (Ref. S. This (O

W N. method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/1. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing

                     >qFor those designs in which the total stored fuel oil volume is contained in two or more interconnected tanks, each tank must be considered and tested separately.}e The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.
                                                                    *Tk. h a fussh u.-p u. e Wa           n S0aN n

Pt.f 4 MVa, h k N at c.e4 vuk; <. s mparbinde- v.JA. Mah d el b.! nlw is9 J& b }kg, W u l cont 3 l B 3.8 47 Rev 1. 04/07/95 cu O

I leeue Number l 38l Affech>d Section l3.3.1 RTS, OP Detto T and OT Detto T l Affected Units CNS: [ Yes) MNS: l Noj Affected Pages ITS: 3.3-14. 15 ITS: ITS Bones: ITS Soses: CTS: CTS: DOCS: DOCS: NRG: NRG: NRG Bones: NRG Bases: q JFD: JFD: NSHC: NSHC: , Descripolon Allowable Value and Trip Setpotnt columns refer to notes on potes 21 and 22 of Tobie 3.3.1-1 of typed ITS. TNs shoukt be pages 18 ond 19, i I i

RTS Instrumentation 3.3.1 Cs Table 3.3.1 1 (page 1 of 7) 1 Reactor Trip System Instrumentation ( APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNEL 5 CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Manual Reactor Trip 1,2 2 B SR 3.3.1.14 NA NA 3(a),4(a),$(a) 2 C SR 3.3.1.14 NA NA

2. Power Range Neutron Flux

a. High 1,2 4 D SR 3.3.1.1 s 110.9% RTP s 109% RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16

b. Low 1(b) 2

                                                  .             4            E        5R 3.J.1.1      s 27.1% RTP       s 25% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16

3. Power Range Neutron Flux High Positive Rate 1.2 4 0 SR 3.3.1.7 s 6.3% RTP s 5% RTP SR 3.3.1.11 with time with time constant constant a 2 sec a 2 sec

4. Intermediate Range 1(b)2ICI 2 F.G 5R 3.3.1.1 s 31% RTP s 25% RTP

[q Neutron Flux SR 3.3.1.8 4 - SR 3.3.1.11 2 IdI 2 H SR 3.3.1.1 s 31% RTP s 25% RTP SR 3.3.1.8 SR 3.3.1.11

5. Source Range Neutron 2(d) 2 1,J 5R 3.3.1.1 s 1.4 E5 cps s 1.0 E5 cps Flus SR 3.?.1.8 SR 3.3.1.11 3(a),4(a), $(ak 2 JK SR 3.3.1.1 s 1.4 E5 c'ps SR 3.3.1.7 cps SR 3.3.1.11

6. Overtemperature AT 1,2 4 E SR 3.3.1.1 Refer to Refer to SR 3.3.1.3 Note 1 (Page Note 1 (Page l SR 3.3.1.6 3.3-18) 3.3-18)

SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.17 (continued) (a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal. (b) Below the P-10 (Power Range Neutron Flux) interlocks. (c) Above the P-6 (Intermediate Range Neutron Flux) interlocks. (d) Below te P 6 (Intermediate Range Neutron Flux) interlocks.

l Catawba Unit 1 3.3-14 Supplement 1

i RTS Instrumentation 3.3.1 Table 3.3.1 1 (page 2 of 7) [ Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

7. Overpower AT 1.2 4 E SP 3.3.1.1 Refer to Refer to SR 3.3.1.3 Note 2 (Page Note 2 (Page ,

SR 3.3.1.6 3.3 19) 3 3 19) l j SR 3.3.1.7 I SR 3.3.1.10 l SR 3.3.1.16 SR 3.3.1.17

8. Pressurizer Pressure I')

I 4 L SR 3.3.1.1 a 1938II) a1945(f) SR 3.3.1.7 psig psig SR 3.3.1.1C SR 3.3.1.16

b. High 1.2 4 E $R 3.3.1.1 s 2399 psig s 2385 psig l SR 3.3.1.7 SR 3.3.1.10 5' 3.3.1.16

9. Pressurizer Water 1(') 3 L SR 3.3.1.1 s 93.8% s 92%

Level - High SR 3.3.1.7 SR 3.3.1.10

10. Reactor Coolant O Flow. Low -

a. Single Loop II9) 3 per M $R 3.3.1.1 a 89.7% a 91% ,

loop SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

b. Two Loops 1(h) 3 per L $R 3.3.1.1 a 89.7% a 91%

loop SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 j (continued) (e) Above the P.? (Low Power Reactor Trips Block) interlock. I (f) Time constants utilized in the lead. lag controller for Pressurizer Pressure . Low are 2 seconds for lead and 1 second for lag. (g) Above the P.8 (Power Range Neutron Flum) interlock. (h) Above the P.7 (Low Power Reactor Trips Block) tr... lock and below the P-8 (Power Range Neutron Flum) interlock. O Catawba Unit 1 3.3-15 Supplement 1 l r

RTS Instrumentation 3.3.1 Table 3.3.1 1 (page 1 of 7) i Reactor Trip System Instrumentation C APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Manual Reactor 1.2 2 B SR 3.3.1.14 NA NA 3(a),4(e),$(a) 2 C SR 3.3.1.14 NA NA ;

2. Power Range Neutron Flu

a. High 1.2 4 D SR 3.3.1.1 s 110.3% RTP s 109% RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16

b. Low- 1(b) .2 4 E SR 3.3.1.1 s 27.1% RTP s 25% RTP SR 3.3.1.8 SR 3.3.1.11 !

SR 3.3.1.16

3. Power Range Neutron Flux High Positive Rate 1.2 4 0 SR 3.3.1.7 s 6.3% RTP s 5% RTP SR 3.3.1.11 with time with time constant constant a 2 sec a 2 see -

   )   4. Intermediate Range        IIU)2IC)               2                F.G      SR 3.3.1.1 SR 3.3.1.8 s 31% RTP      s 25% RTP Neutron Flux SR 3.3.1.11 2(d)                2                 H       SR 3.3.1.1             s 31% RTP      s 25% RTP SR 3.3.1.8 SR 3.3.1.11

! 5. Saurce Range 2 Id) 2 1.J SR 3.3.1.1 s 1.4 ES cps s 1.0 ES Neutron Flux SR 3.3.1.8 cps SR 3.3.1.11 l 3(a),4(a). S I*) 2 J.K SR 3.3.1.1 s 1.4 E5 s 1.0 ES SR 3.3.1.7 cps cps SR 3.3.1.11 , 6. Overtemperature AT 1.2 4 E SR 3.3.1.1 Refer to Refer to l 9 3.3.1.3 Note 1 (Page Note 1 ) l SR 3.3.1.6 3.318) (Page ' l SR 3.3.1.7 3.318) l SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.17 (continued) (a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal. (b) Below the P-10 (Power Range Neutron Flux) interlocks. (c) Above the P-6 (Intermediate Range Neutron Flux) interlocks. I (d) Below the P-6 (Internedtate Range Neutron Flux) irterlocks. ) O ! l Catawba Unit 2 3.3-14 Supplement 1

l RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 7) Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIRENENTS VALUE SETPOINT

7. Overpower AT 1.2 4 E SR 3.3.1.1 Refer te Refer to SR 3.3.1.3 Note 2 Note 2 SR 3.3.1.6 (Page (Page SR 3.3.1.7 3.319) 3.3-19) l SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.17

8. Pressurizer Pressure

a. Low 1(') 4 L SR 3.3.1.1 a 1938(I) a1945(f)

SR 3.3.1.7 psig psig SR 3.3.1.10 SR 3.3.1.16

b. High 1.2 4 E SR 3.3.1.1 s 2399 psig s 2385 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

9. Pressurizer Water 1(') 3 L SR 3.3.1.1 s 93.8% s 92%

Level . High SR 3.3.1.7 SR 3.3.1.10 k 10. Reactor Coolant Flow Low

a. Single Loop II9) 3 per M SR 3.3.1.1 2 89.7% a 91%

loop SR 3.3.1.7 { SR 3.3.1.10 SR 3.3.1.16

b. Two Loops IIh) 3 per L SR 3.3.1.1 2 89.7% a 91%

loop SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 (continued) (e) Above the P-7 (Low Power Reactor Trips Block) interlock. (f) Time constants utilized in the lead-lag controller for Pressurizer Pressure - Low are 2 seconds for lead and I sece'id for lag. (g) Above the P-8 (Power Range Neutron Flux) interlock. (h) Above the P-7 (Low Power Reactor Trips Block) interlock and below the P-8 (Power Range Neutron Flux) interlock. p Catawba Unit 2 3.3-15 Supplement 1 l

issue Numbot l 39j ANected Section l3.9.2 l Affected Unite CNS: l Yes! MNS: l Noj ANected Pages ITS: 3.9-2, 3 ITS: ITS Boeos: 83.9-7.8.9 ITS Bases: CTS: 3/49-2.3 CTS: DOCS: DOCS: NRG: g-4. 5 NRG: NRG Bases: 83.99,10 NRG Boess: JFD: JFD: NSHC: NSHC: Description The ITS 3.9.2 octens for BDMS were incorrectly wrttten as rnore restrtetive than the actions of CTS 3.9.2.1. Revise the actions to be consistent with the CTS. The ITS octions would not allow fuel movement while relying on source ronge mordfors which is allowed by the CTS. l l

Nuclear Instrumentation 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation LCO '3.9.2 Two Boron Dilution Mitigation System (BDMS) trains shall be OPERABLE.

                           ----------------------------NOTE----------------------------

Automatic actuation of the BDMS may be blocked during core reloading until two assemblies are loaded into the core. APPLICABILITY: MODE 6. l T.CTIONS j CONDITION REQUIRED ACTION COMPLETION TIME l { l A. One or both BDMS A.1.1 Suspend CORE Inunediately trains inoperable. ALTERATIONS. I m Ol - A.1.2 Suspend positive reactivity additions. Immediately I m l A.1.3 Verify unborated I hour . water source isolation valve (s) i are closed and l secured. ! I DE l A.2.1 Verify two Source Ininediately Range Neutron Flux Monitors are OPERABLE. I m l A.2.2 Verify Reactor. Makeup 1 hour Water Pumps combined flow rates are within the limits specified in the COLR. l Catawba Unit 1 3.9-2 Supplement 1

Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l l l SR 3.9.2.1 Perform CHANNEL CHECK. 12 hours i SR 3.9.2.2 Perform COT. 31 days SR 3.9.2.3 Verify each automatic valve moves to the % mths correct position and Reactor Makeup Water t pumps stop upon receipt of an actual or simulated actuation signal. SR 3.9.2.4 -------------------NOTE------ ------------ Only required to be performed when used to satisfy Required Action A.2.1. l , Perform CHANNEL CHECK on the Source Range 12 hours - Neutron Flux Monitors. SR 3.9.2.5 -------------------NOTE-------------------- Only required to be performed when used to satisfy Required Action A.2.1. l Perform COT on the Source Range Neutron 7 days Flux Monitors. SR 3.9.2.6 -------------------NOTE-------------------- l Only required to be performed when used to satisfy Required Action A.2.2. l Verify combined flowrates from both Reactor 7 days , Makeup Water Pumps are s the value in the l COLR. O ; Catawba Unit 1 3.9-3 Supplement 1 l

Nuclear Instrumentation 1 3.9.2 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation LCO 3.9.2- Two Boron Dilution Mitigation System (BDMS) trains shall be OPERABLE.

                    ----------------------------NOTE----------------------------

Automatic actuation of the BDMS may be blocked during core reloading until two assemblies are loaded into the core. -i

I APPLICABILITY: MODE 6. J ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME l A. One or both BDMS A.1.1 Suspend CORE Inunediately trains inoperable. ALTERATIONS. l M . l A.1.2 Suspend positive Inmediately reactivity additions. I E l A.1.3 Verify unborated 1 hour water source isolation valve (s) are closed and secured. I a l A.2.1 Verify two Source Inunediate,1y Range Neutron Flux Monitors are OPERABLE. I m l A.2.2 Verify Reactor Makeup 1 hour l Water Pumps combined flow rates are within the limits specified in the COLR. l Catawba Unit 2 3.9-2 Supplement 1 i i

Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. 12 hours SR 3.9.2.2 Perform COT. 31 days SR 3.9.2.3 Verify each automatic valve moves to the 18 months correct position and Reactor Makeup Water pumps stop upon receipt of an actual or simulated actuation signal. SR 3.9.2.4 -------------------NOTE-------------------- Only required to be performed when used to satisfy Required Action A.2.1. l Perform CHANNEL CHECK on the Source Range 12 hours Neutron Flux Monitors. SR 3.9.2.5 --------------.----NOTE-------------------- Only required to be performed when used to satisfy Required Action A.2.1. l Perform COT on the Source Range Neutron 7 days Flux Monitors. SR 3.9.2.6 .------------------NOTE-------------------- Only required to be performed when used to satisfy Required Action A.2.2. l Verify combined flowrates from both Reactor 7 days Makeup Water Pumps are s the value in the COLR.

O l Catawba Unit 2 3.9-3 Supplement 1 l

i Nuclear Instrumentation B 3.9.2 BASES APPLICABILITY MODES 2, 3, 4, and 5, this same installed BDMS and (continued) associated circuitry is also required to be OPERABLE by LCO 3.3.9, " Boron Dilution Mitigation System (BDMS)". ACTIONS A.1.1. A.1.2. A.1.3. A.2.1. and A.2.2 l With only one cr no Boron Dilution Mitigation System trains available, the system is considered inoperable and CORE ALTERATIONS and positive reactivity additions must be suspended immediately. In addition, valva NV-230 must be closed and secured within 1 hour to isolate the unborated water source. Performance of Required Actions A.1.1 and A.1.2 shall not preclude completion of movemert of a component to a safe position. An option to isolating the_unborated water source is provided to allow alternate methods of monitoring core reactivity conditions and controlling boron dilution incidents. This includes the utilization of the two Source Range Neutron Flux Monitors. These monitors must be verified to operate with alarm setpoints less than or equal l O to one-half decade (square root of 10) above the steady-state count rate, each with continuous visual indication in In addition, the combined flowrate from the control room. both Reactor Makeup Water Pumps must be verified to be within the limits specified in the COLR in 1 hour. Once these options are verified, CORE ALTERATIONS and positive reactivity changes can continue.

 -SURVEILLANCE   SR   3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is a        !

comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading l and core geometry can result in significant differences, but each train should be consistent with its local conditions. The Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LC0 3.3.9. (continued) Catawba Unit 1 B 3.9-7 Supplement 1 l

Huclear Instrumentation B 3.9.2 BASES SURVEILLANCE SR 3.9.2.2 REQUIREMENTS (continued) SR 3.9.2.2 is the performance of the CHANNEL OPERATIONAL TEST for the Boron Dilution Mitigation System, which is the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of requi. red alarm, interlock, display, and trip functions. The COT also includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. This surveillance must be performed once per 31 days. The frequency is based on operating experience, which has shown to be adequate. SR 3.9.2.3 SR 3.9.2.3 is performed on the Boron Dilution Mitigation System to verify the actuation signal actually causes the appropriate valves to move to their correct position and the Reactor Makeup Water Pumps to stop to mitigate a boron p dilution accident. The 10 month frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown these components ! usually pass the Surveillance when performed at the 18 month ! Frequency. ! SR 3.9.2.4 SR 3.9.2.4 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences, but each channel should be consistent with its local conditions. A note is provided to clarify that the CHANNEL CHECK only needs to be performed on the Source Range Neutron Flux l Monitors when used to satisfy Required Action A.2.1. (continued) l Catawba Unit 1 B 3.9-8 Supplement 1

1 1 Nuclear Instrumentation B 3.9.2 O G BASES SURVEILLANCE SR 3.9.2.4 (continued) REQUIREMENTS The Frequency of 12 hours is consistent with the CHANNEL 1 CHECK Frequency specified similarly for the same instruments 1 in LC0 3.3.1. SR 3.9.2.5 SR 3.9.2.5 is the performance of the CHANNEL OPERATIONAL TEST for the Source Range Neutron Flux Monitors, which is i the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT also includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. This SR is only required when the Source Range Neutron Flip Monitors are used to satisfy Required Action A.2.1. Thh l surveillance must be performed prior to placing the monitors l in service and once per 7 days thereafter. The 7 day i Frequency is based on operating experience, which has been j shown to be adequate. O , SR 3.9.2.6 SR 3.9.2.6 verifies the combined flow rates from the both Reactor Makeup Water Pumps are s the value in the COLR. This surveillance is only required when implementing Required Action A.2.2. It ensures the assumptions in the l analysis for the baron dilution event under these conditions are satisfied. This surveillance must be performed once per 7 days and is based on engineering judgement and the unlikely event that a boron dilution will occur during this time. i REFERENCES 1. UFSAR, Section 15.4.6

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

O Catawba Unit 1 B 3.9-9 Supplement 1 l

Nuclear Instrumentation B 3.9.2 BASES APPLICABILITY- MODES 2, 3, 4, and 5, this same installed BDMS and (continued) associated circuitry is also required to be OPERABLE by LCO 3.3.9, " Boron Dilution Mitigation System (BDMS)". I ACTIONS A.1.1. A.1.2. A.1.3. A.2.1. and A.2.2 l With only one or no Boron Dilution Mitigation System trains available, the system is considered inoperable and CORE ALTERATIONS and positive reactivity additions must be suspended intnediately. In addition, valve NV-230 must be closed and secured within 1 hour to isolate the unborated water source. Performance of Required Actions A.1.1 and A.1.2 shall not preclude completion of movement of a component to a safe position. An option to isolating the unborated water source is provided to allow alternate methods of monitoring core reactivity conditions and controlling boron dilution incidents. This includes the utilization of'the two Source Range Neutron Flux Monitors. These monitors must be ; verified to operate with alarm setpoints less than or equal l to one-half decade (square root of 10) above the steady-state count rate, each with continuous visual indication in the control room. In addition, the combined flowrate from both Reactor Makeup Water Pumps must be verified to be within the limits specified in the COLR in I hour. Once these options are verified, CORE ALTERATIONS and positive reactivity changes.can continue. SURVEILLANCE SR 3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences, but each train should be consistent with its local conditions. The Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.9. (continued) Catawba Unit 2 B 3.9-7 Supplement 1 l

Nuclear Instrumentation ] B 3.9.2 l [ BASES SURVEILLANCE SR 3.9.2.2 REQUIREMENTS (continued) SR 3.9.2.2 is the performance of the CHANNEL OPERATIONAL TEST for the Boron Dilution Mitigation System, which is the injection of a simulated or actual signal into the channel i as close to the sensor as practicable to verify the 1 OPERABILITY of required alarm, interlock, display, and trip functions. The COT also includes adjustments, as necessary, of the required alann, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. i This surveillance must be performed once per 31 days. The frequency is based on operating experience, which has shown to be adequate. 4 SR 3.9.2.3 , 4 SR 3.9.2.3 is performed on the Boron Dilution Mitigation System to verify the actuation signal actually causes the appropriate valves to move to their correct position and the Reactor Makeup Water Pumps to stop to mitigate a boron i dilution accident. The 18 month frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.9.2.4 SR 3.9.2.4 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading cnd core geometry can result in significant differences, but each channel should be consistent with its local conditions. A note is provided to clarify that the CHANNEL CHECK only needs to be performed on the Source Range Neutron Flux l Monitors when used to satisfy Required Action A.2.1. (continued) l Catawba Unit 2 B 3.9-8 Supplement 1

Nuclear Instrumentation B 3.9.2 BASES SURVEILLANCE SR 3.9.2.4 (continued) REQUIREMENTS The Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.1. ( SR 3.9.2.5 SR 3.9.2.5 is the performance of the CHANNEL OPERATIONAL TEST for the Source Range Neutron Flux Monitors, which is the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT also includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. This SR is only required when the Source Range Neutron Flux Monitors 'are used to satisfy Required Action A.2.1. This l surveillance must be performed prior to placing the monitors in service and once per 7 days thereafter. The 7 day Frequency is based on operating experience, which has been shown to be adequate. t SR 3.9.2.6 SR 3.9.2.6 verifies the combined flow rates from tne both Reactor Makeup Water Pumps are s the value in the COLR. This surveillance is only required when implementing Required Action A.2.2. It ensures the assumptions in the l analysis for the boron dilution event under these conditions are satisfied. This surveillance must be performed once per 7 days and is based on engineering judgement and the unlikely event that a boron dilution will occur during this time. REFERENCES 1. UFSAR, $ection 15.4.6

2. 10 CFR 50.36. Technical Specifications, (c)(2)(11).

O Catawba Unit 2 B 3.9-9 Supplement 1 l

Speet[cadeon .3,Q 3.9 REFUELING OPERATI M

                                                                ~ RNW4dMh.

g G2T1P2 [5NSTRUMENTATION O I

(15M!TTilG CalefTION T91MfPERAD Ad (IdC 3.1.M GELED us vainirgsg fwo trains of the Boron Dilution Mitigation System shall be OPERABLEjenu upwf ating wi 6na adnu en or uv.u nergin m arm yuva aus .. i s s~p dica onquasIo'4timeg1the in +hm raa+ st* raa=.1 ** enunt rate. each 4fth raa+4a ga g APPLICA8tLITY: M0bE6 Weckd , lur'sCa<t. rtI*'#'d xaN Iw* j Af)),Q3: Q aswwM.'at se<. loaded th SM cowomou A (a) With one or both trains of the oron Ollution Mitigation System inoperable ereat anerm in 4,g g% g, g (1) lausediately suspend all operations involving ORE ALTEDATIONS or 1 Arfw positive reactivity changes, and verify that valve is closed and secured within the next hour or 44, l

               ),l.2                           (2) verify that two Source Range Neutron Flux Monitors are OPERA 8LE 1

And o- sung witu nu suspam63 s a snan or h* 4* 39 i one f decade ( re root of.1 vei su bove the s ady-state count (ra each with tinuous visual ndication i the control ro bib d1 YoI5nESnIa'fE_h vYr' fhat the combined flowrate from both Reactor Makeup Water Pumps is less than or equal to the Reactor Makeup Water Punr1 flowrate u cresen

                                        ..       . hm.ted in the Core Operating Limits Repor+ vithin the next b With both trains of the Boron Dilution Mitigation System inoperable gm

[ '

-- .. mand one of the Source Range Neutron Flux Monitors inoperable or not operating famediately suspend all operations !

J involving CORE ALTERATIONS or positive reactivity changes and verify a that val eQs d and secured within the next hour. c) With both srawn of sne soron Diluti Mitigation System inop ab g'/ or not oper ung and both of the So e Range Neutron Flux nitors inoperabi or not operating, dete ine the boron concentr on of th W1/fnf M eat /er colant 5 stem at least nee per 12 hours and we fy that ('valve Reactor -230 is ekosed red andwithin se the next hou . SotttYt 160/44tV1 I CATAWBA - UNIT 1 3/4 9-2 Amendment No. 148 O Pne. idi

Y W'AC= 6t" 3.9.2. REFUELING OPERATIONS SURVEILLANCE RE0VIREMENTS l 4.9.2.1.1 Each train of the Baron Otlutt'o'n Hitigation System shall be demon-strated OPERABLE by performance of:

        @ a v. 2 g     .fs7 A CHANNEL CHECK at least once per 12 hours, fni                   C            !    a j,er' An              C          OPERATIONAL TEST at tesst-tmce per 31 Tiays.
                      ,,,4d7 At least on          er 18 months the BOMS shall be demonstrated OPERABLE by:

(1) Vertfying that each automatic valve actuat by the BONS moves to its correct position upon receipt of signal, and

            @ 3.4.2 3)

(2) Verifying each reactor makeup water pump stops, as dest ned, upon receipt of g ignal. . .

                      .El & If using tne soursu n . . -,.w vn r ium mui m . .o-meet the
      <MM         requirements of Technical Specific 7ttion 3.9 2, each Source Range Neutron Flux Nonttor shall be demonstrated OPERA 8LE by performance of:                                  j A CHANNEL CHECK at least once per 12 hours,
          @ 3.9.2.f) b) An ANALOG              L OPERAll       lui witnin        urs prior to tHe-s                       initia     tart of CORE        RATIONS or with     I hour     aftar plarin            f 64 A9.4 5) -          the         0!LUT10N MI       TION' SYSTEM in    rabl e . =d            ,
                        @ An         f             EL OPERATIONAL TEST at least once per 7 days.                       {
    '                                              owrate from both Reactor Makeup Water Pumps shall be Q3,9.2.4}      (d) . The combine verified as less than or equal to the Reactor Makeup Water Pump flowrate presented in the Core Operating Limits Report at least once per 7 days.

3/4 9-3 Amendment No. 148 CATAWBA - UNIT 1 i O p, , .n

5/W:d M M 39.2 REFUELING OPERATIONS Q M7TDb kSTRUMENTATICN N# dm mier enainettom ropH9paea r renn co .5Sh a 1. s = = ni fwo trains of the Baron Dilution Mitigation System shall be OPERA 8LEj na opera 6is -i di anu ouw=u no ry n m a m ns 6 va a w 6p 6 ira nan or ". iv , times the ndicati[,n the control ro eady. state ennat rate. sach with centhucu_ APPLICABItITY: MODE 6 F .I[ko1Ct

                                                                                                                                                                   -           AuMi akh." 4 K4- 604.s UO Nectsal- W eert. n l E s % l1 6 .
                                                                                                                                                                                                                     ]k
 ,                                                                                                                    AC. TION.

855emkis ort leJ E 4 Me m.

                                                                                                                           @ With one or both trains of t                   oren Dilution Mitigation System inoperableenr #enaaer1AF.

g7iauco Qg 1 (1) immediately suspend all operations involving CORE ALTERATIONS or g

                                                                                                                             ,             positive reactivity changes, and verify that a valve                   is f

closed and secured within the next hour or [(2) verify that two Source Range Neutron Flux Monitors are OPERABLE ene operau n un mam aw6puin6m f,[,1i 33 6nen ur ugues 6 i hne-halfde the en (f guare root o'f ndication 10)[above the steady-s ec i gj i rate. each tinunut visual in the e ten 1 -* - l- A'b*. g n.n.711pinvics5i i'n 6ne r" 4- vnd,ndicaticiintheraadv

                                                                                                                                                                                                  - - - - - - - , =

i ia==at/and verify that the v. h cea. aed flowrote from both Reactor Makeup Water Pumps is less than or equal to ile Reactor Makeup Water Pump flowrate presented..in the Cars. Operating Limits Report within the next hour. With hath trains of the Boron Dilution Mitigation System inoperable [\ Ad'* naf an* and one of the Source Range Neutron flux Monitors noperable@or not operating immediately suspend all operations involving CO LTERATIONS or positive reactivity changes and verify thatfale d and secured within the next hour. c) With both t ins of the Baron Di'ution itigation System inope o or not op ating and both of the Sour Range Neutron Flux M itors inopera e or not peratin9, determ e the baron concentrat on of th Reset Coolant System at . east o e per 12 hours and ver y tha g g g / t/ tm /fr al NY-230 is closed and secur within the next hour. sarce isoMan i.4 r A.) CATAW8A - UNIT 2 3/4 g.2 Amendment No. 142 O P.qe / of 2.

M3 be iM. pp VY5 043 4eadai<m 3.9.z REFUELING OPEQaff0NS SURVEILLANCE REOUTREMENTS 4.9.2.1.1 Each train of the Boron 011ution Mitigation System shall be demon. strated OPERABLE by performance of:

                                        , p y,9 ,2 ./] @ A CHANNEL CHECK at least once per 12 hours, C                  a M1                    g An                                                                   C   L OPERATIONAL TEST at least once per 31 da s.

g At least a per 18 months the BDMS shall be demonstrated OPERABLE by: . (1) Verifying that each autom..ic valve actuated by the BOMS moves g,g to its correct. position upon receipt of r1 signal, and ((2) Verifying each reactor makeu water puso stops, as designed, upon receipt of @ signa . gg 3? + t r usi ng% ", . 6= nenge neutron.r eux nun.wrs n meet the requirements of Technical Specification 3.9.2. each Source Range Neutron F1~ nitne ch=11 he demonstrated OPERABLE by performance of: MED A CHANNEL CHECK at leatt once per 12 hours, (b) An ANAL ~ CHANNEL OPERATI TE5T with 8 hours prior to e

                                                                           - - initia start of CORE A ERATIONS or                                                               thin i hour after d laring     -

(54 3.9.2 5) the B ON DILUTION M1 ATION SYST inoperable, and

              .                                                           @ An                                                                      EL OPERATIONAL TEST at least once per 7 days.

(V) g 3,9 g,g h The combined flowrate from both Reactor Makeup Water Pumps shall br. verified as less than or equal to the Reactor Makeup Water Pump flowrate presented in the Core Operating Limits Report at least once per 7 days. l l CATAWBA - UNIT 2 3/4 g-3 Amendment No. 142 a,e 2,n

Nuclear Instrumentation 3.9.0

  /*g     3.9 REFUELING OPERATIONS 3.9. Nuclear Instrumentation aan D;/a6n Mil'neboke Sy sleM (BMO dw&

LCO 3.9. Two emarca ranoe neutron 41ux annif;r3 shall be OPERABLE.

                                ~hG~         3 ;! 'g 'g I g#p2.- ~ h'.4/Eks.L K6k -

APPLICABILITYi T- - N -- - ' ' - ~ ACTIONS REQUIRED ACTION C0ffLETION TIME A. One equ1reaj souprs A.1 Suspend CORE !.nediately '

                           ,r 6n nu/I ALTERATIONS.                                     O a   inoperable.

or b jBPM6 fya,ggf AJ Suspend positive Imediately g reactivity additions. 4 {/NfMr I) B. Two[re ired) source range eutron flux B.1 Initiateatt'nto lamediately restore o w ource moni rs inoperable, range neut n flux

      '                                             monitor t OPERABLE f'                                                 status.

b g B.2 Per orm SR 3.9.1.1. 4 hours M -

Once r 12 ours t reafter

                                         /                                 /

WOG STS 3.9 4 Rev 1. 04/07/95 ha.vbA / b G 1 I I 1

1 INSERT 1 @ CONDITION REQUIRED ACTION COMPLETION TIME A Verify unborated I hour water source isolationvalve(s) are closed and secured. k

Q.2.1 Verify two Source a muy luued[.hl Range Neutron Flux Monitors are J

OPERABLE.

                   @2.2    Verify Reactor Makeup  1 hour Water Pungs combined                                -

r ( flow rates are within l the limits specified in the COLR. 1 l l Insert Page 3.9-4 l 1 0hibA- !

O

Na"Y >thaY* *h tk 99 Nuclear Instrumentation l 3.9.G 8 /N SURVEILLANCE REQUIREMENTS V mvuuANI t FREQUENCY I i SR 3.9. Perform CHANNEL CHECK. . 12 hours

        &     3.9.3.2.. ...................N0rE.

Neutron detectors are e luded from CHANNEL f CALIBRATION. l Perform CHANNEL IBRATION.

18) months I

       %sar >>          D l
                                                                                                   )

{ t f' i ( i i I l WOG STS 3.9 5 Rev 1, 04/07/95 (Chwh~ O

INSERT 3 SR 3.9.2.2 Perform COT. 31 days SR 3.9.2.3 Verify each automatic valve moves to 181nonths their correct position and Reactor Makeup Water pumps stop upon receipt of an actual or simulated actuation signal. SR 3.9.2.4 ----------------NOTE-------------------- Only required to be performef,/herLysed to satisfy Required Action A.f: Perform CHANNEL CHECK on the Source 12 hours Range Neutron Flux Monitors. SR 3.9.2.5 ----------------NOTE-------------------- Only required to be perfonned ed to satisfy Required Action A Perform COT on the Source Range Neutron 7 days Flux Monitors. SR 3.9.2.6 ----------------NOTE-------------------- Only required to be performed when used to satisfy Required Action A. Verify combined flowrstes from both 7 days Reactor Makeup Wr.ter Pumps are s the ! value in the COLR. { Insert Page 3.9-5 calawba. O

I i b bY Nuclear Instrumentation FYT % B3.9.8g BASES (continued) (p.,/1/4,o A1./,g6 g/M

i APPLICABILITY In MODE 6. the # e=?A="*raam ur -=*a9 must be Og OPERABLE to determine changes,in core reactivity 4

                                    ,,,,,o.._      . - - - - _       -- . . . _ _         . .
                                    .1
                                 .--.--   r In NODES 2. nd m-    3. circuitry

4. and 5. tamagrIEFTritt anggahu r Mirec to meo De M*f6 ' d-OPERABLE by LCO 3.3.

w w .mt w ..'

                                                                     -- N n w - un                                 - - -

- ,g &,,,e D,1&R 4#iabn $ #MGDvC] ACTIONS With only o source range neatron flux monit redundancy LE. oni rji s been lost. Since thsse inst s are the 3 means of monitoring c. ore react ty conditions,

          < fear de:r/suj,      C     A'.

itionsired mustAction be sus TIONS and immediately. positive roactivity. Performance of A.1 hl tha 1 t preclude completion of mov a s e position. t of a component to r With no source range neutro ux monitor OPERABLE.. action to restore a monitor to OP "LE status shall be initiated j innediately. Once initi ed. action shall be continued i until a source range ne on flux monitor is restored to ) OPERA 3LE status. IL2 With no source ange neutron flux monitor OPERABLE there are no direc means of detecting changes in core reacti ty. However, si CORE ALTERATIONS and positive reactivi additions re not to be made, the core reactivity ition is stab ized untti the source range neutron fl .an: tors are OP LE. This stabilized condition is det aire by perf ing SR 3.9.1.1 to ensure that the requ co boron tration exists. Completion Time of 4 hours is sutfi ent to obtain and analyze a reactor ecolant s le for ron concentration. The Frequency of or.ce per 12 rs sures that unplanned ! changes in boron concentration d be identified. The 12 hour Frequency is reasonable onsidering the low (continued) WOG STS B 3.9-9 7 Rev 1. 04/07/95

O

I I r INSERT ACTIONS _(] _./ A,1,1. A.1.2. A.1.3. A.2.1. and A.2.2 l l With only one or no Boron Dilution Mitigation System trains available the ' systen: is considered inoperable and CORE ALTERATIONS and positive reactivity additiocs must be suspended mediately. In addition, valve NV230 must he closed and secured within 1 hour to isolate the unborated water source. Performance of Required Actions A.1.1 and A.1.2 shall not preclude completion l of movement of a component to a safe position. An option to isolating the unborated water source is provided to allow alternate methods of monitoring core reactivity conditions and controlling boron dilution incidents. This includes the utilization of the two Source Range Neutron Flux Monitors. These monitors must be verified to operate with l alarm setpoints less than or equal to one-half decade (square root of 10) above the steady-state count rate, each with continuous visual indication in the control room. In addition, the combined flowrate from both Reactor Makeup Water Pumps must be verified to be within the limits specified in the COLR in 1 hour. Once these options are verified CORE ALTERATIONS and positive , reactivity changes can continue.

0 G l l Insert Page B 3.9-9 1 l b & (L o O

' f*I Nucisar Instrumentation F V J @k'* B 3.9. BASES ACTIONS. continued)Q

                   @          obability of a change in[re reactivity during period.

f SURVEILLANCE 3.9. 1 REQUIRENENTS @ y SR SR 3.9.5 1 15 the performance of a CHANNEL CECK. which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry c_an recult in significant differences w _ rr- ain; but each.criplPIO should be consistent wit its local conditions, g YheFrequencyof12hoursisconsistentwiththeCHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.9. A M NRAT N SR 3.9.3. s the performance of a 18 mont This SR is modified by a PKimmuun eve % stating that neutro detectors are excluded from CHANNEL CALIBRATION,

The O EL CALIBRATION for the s ce range neutron flux mon ors consists of obtaining t detector plateau or pr amp discriminator curves, ev uating those curves, and ring the curves to the

{ facturer*5 data. The ! 8 month Frequency is based the need to perform this Surveillance under the tions that apply during a ant outage. Operating experi has shown these compo ts ) j

                         %usually emnev , pass  the Survei   ane*  when   neefnmart  at  t 18  mon                i
                           ,              Q u[ m .t., Je d e n 6.4. D A                                                                                    l REFERENCES      9.               50. Appendit A1GDC 13. GDC 2]K GDC 28. and

29. 1 -

2 3 ($mf saetiorff15.2A1L tyg so.%,Teckn,'c \ Spec 8'cabW,(gX2b i WOG STS B 3.9 10 Rev 1. 04/07/95 b O

INSERT SR SR 3 9.2.2 SR 3.9.2.2 is the performance of the CHANNEL OPERATIONAL TEST for the Boron Dilution Mitigation System, which is the injection of a simulated or actual signal into the channel.as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT also includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and

     . accuracy.

This surveillance must be performed once per 31 days. The frequency is based on operating experience, which has shown to be adequate. l SR 3.9.2.3 { i SR 3.9.2.3 is performed on the Boron Dilution Mitigation System to verify the actuation signal actually causes the appropriate valves to move to their correct position and the Reactor Makeup Water Pumps to stop to mitigate a boron dilution accident. The 18 month frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. 0perating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. O i SR 3.9.2.4 SR 3.9.2.4 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It .is based on the assumption that the two indication channels should be consistent ~with core conditions. Changes in fuel loading and core geometry can result in significant differences, but each channel should be consistent with its local conditions. A note is provided to clarify that the CHANNEL CHECK only nee :s to be performed on the Source Range Neutron Flux Monitors when used to satisfy l.RequiredActionA.2.1. The Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LC0 3.3.1. Insert Page B 3.9-10a C&uu I I

INSERT SR (continued) SR 3.9.2.5-SR 3.9.2.5 is the performance of the CHANNEL OPERATIONAL TEST for the Source Range Neutron Flux Monitors, which is the injection of a simulated or actual I signal into the channel as close to the sensor as practicable to verify the 1 OPERABIt.ITY of required alarm,' interlock, display, and trip functions. The COT a?so includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. This SR is only required when the Source Range Neutron Flux i' Monitors are used to satisfy Required Action A.2.1. This surveillance must be l performed prior to placing the monitors in service and once per 7 days thereafter. The 7 day Frequency is based on operating experience, which has been shown to be adequate. SR 3.9.2.6 SR 3.9.2.6 verifies the combined flow rates.from the both Reactor Makeup Water Pumps are s the value in the COLR. This surveillance is only required when implementing Required Action A.2.2. It ensures the assumptions in the l analysis for the boron dilution event under these conditions are satisfied. This surveillance must be performed once per 7 days and is based on O v engineering judgement and the unlikely event that a boron dilution will occur during this time. I i l Insert Page B 3.9-10b bb d

k issue Number l 42l ANected Section lSR 3.0.1 Boses l ANected Unlis CNS: l Yesl MNS: l Yesj ANected Pages ITS: ITS: ITS Bones: B 3.011 ITS Bases: B 3.0-11 CTS: CTS: DOCS: DOCS: NRG: NilG: NRG Bones: NRG Bases. JFD: JFD: NSHC: NSIC DescripHon SR 3.0.1 Bases page B 3.0-11 hos o typo in the first line on the page. .' ACTIONS define the remediol measures that oply.' Should be 'opply.' l l l

SR Applicability B 3.0 D [V BASES SR 3.0.1 because the ACTIONS define the remedial measures that apply. l (continued) Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status. Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed. SR 3.0,? SR 3.0.2 establishes the requirements for meeting the (V specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per . . ." interval. SR 3.0.2 oermits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities). The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS, O V (continued) Catawba Unit 1 B 3.0-11 Supplement 1 l

1 SR Applicability B 3.0 4-BASES SR 3.0.1 . because the ACTIONS define the remedial measures that apply. l (continued) Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status, i Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This-includes ensuring applicable Surve111ances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been establisned. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed. SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the ' specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required. Action on a "once per . . ." interval. SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenanceactivities). The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in j the individual Specifications. The requirements of regulations take precedence over the TS. (continued) Catawba Unit 2 B 3.0-11 Supplement 1 l

-www issue Number l 43l ANected Sociton l3.2.4 Required Action A.4 l ANected Unlis CNS: l Yes) MNS: l Noj ANected Pogos ITS: 3.2-12 ITS: j ITS Boeos: ITS Booss: _ CTS: CTS: DOCS: DOCS: NRG: NRG: NRG Bases: NRG Bones: , JFD: JFD: NSHC: NSHC: Descripilon The Completion Time for Required Action A.4 in the typed ITS is 6 hours. This is incorrect. It should be 72 hours os shown in the CTS markup (L10) and the NUREG Markup. l

I l l I l l I I l l l l

QPTR 3.2.4 L ACTIONS CONDITION REQUIRED ACTION COMPLETION' TIME l A. (continued) A.4 Reduce Power Range 72 hours Neutron Flux - High 1 rip Setpoint a 3% for each 1% of QPTR > 1.02. Alf A.5 Reevaluate safety Prior to analyses and confirm increasing i results remain valid THERMAL POWER for duration of above the more operation under this restrictive ! condition, limit of Required Action A.1 or A.2 ale I L A.6 --------NOTE--------- l Perform Required Action A.6 only after Required Action A.5 is completed. Calibrate excore Prior to detectors to show increasing zero QPT. THERMAL POWER above the more restrictive limit of Required Action A.1 or A.2 Alm (continued) I I l l l i O l Catawba Unit 1 3.2-12 Supplement 1

QPTR 3.2.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l A' . (continued) A.4 Reduce Power Range 72 hours Neutron Flux - High Trip Setpoint 2 3% for each 1% of QPTR > 1.02. M A.5 Reevaluate safety Prior to analyses and confirm increasing results remain valid THERMAL POWER for duration of above the more operation under this restrictive condition. limit of Required Action A.1 or A.2 M A.6 --------NOTE--------- Perform Required O - Action A.6 only after Required Action A.5 is completed. Calibrate excore Prior to detectors to show increasing zero QPT. THERMAL POWER above the more restrictive limit of Required Action A.1'or A.2 M (continued) 1 l I O ' l Catawba Unit 2 3.2-12 Supplement 1

issue Number l M[ ANected Section lSR 3.6.6.7 Boses ] l Affected UnMs CNS: l Yesl MNS: l Yesl Affected Poges ITS: ITS: ITS Bones: B 3.6-44 ITS Bases: B 3.6 45 CTS: CTS: DOCS: DOCS: HRG: NilG: NilG Bones: B 3.6-92 NRG tones: 836-92 JFD: __ JFD: NSHC: NSHC: DescripHon The Boses for the sroy nozzle surveiBonce. SR 3.6.6.7, does not describe current testing methodologies and CTS allowance. The CTS ollows on air flow test, but does not specify direction. The test is currently done by drowing a vacuum on the nozzle and drowing air through the nozzle in lieu of a blower. This test method prevents the srnali omounts of contaminated water in the lines from being bicwn into the containrrunt atmosphere. Reference UFSAR 6.5.4. Add discussion to ITS Bases consistent with UFSAR.

 .s'                                                                                                                      .

i I o

Containment Spray System B 3.6.E BASES (continued). SURVEILLANCE SR 3.6.6.3 and SR 3.6.6.4 (continued) REQUIREMENTS Frequency was concluded to be acceptable from a reliability standpcint. The surveillance of containment sump isolation valves is also required by SR 3.6.6.3. A single surveillance may be used to satisfy both requirements. { j i SR 3.6.6.5 and SR 3.6.6.6 l 1 These SRs require verification that each containment spray pump discharge valve opens or is prevented from opening and each containment spray pump starts or is de-energized and prevented from startina upon receipt of Containment Pressure Contol System start ar. terminate signals. The CPCS is described in the Bases for LCO 3.3.2, "ESFAS." The 18 month I Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage. SR 3.6.6.7 With the containment spray inlet valves closed and the spray header draiiled of any solution low pressure air or smoke can be blown through test connections. The spray nozzles can also be periodically tested using a vacen blower to induce air flow through each nozzle to verify unobstructed flow. This SR ensures that each spray nozzle is , unobstructed and that spray coverage of the containment during an accident is not degraded. Because of the passive design of the nozzle, a test at 10 year htervals is considered adequate to detect obstructio* Of the spray i nozzles. l 1 I REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.

2. UFSAR, Section 6.2. ,

3. 10 CFR 50.49.

4. 10 CFR 50, Appendix K.

(continued) l Catawba Unit 1 B 3.6-44 Supplement 1

Containment Spray System B 3.6.6 BASES (continued) REF R S 5. 10 CFR 50.36. Technical Specifications, (c)(2)(fi).

6. ASME, Boiler and Pressure Vessel Code, Section XI.

O , i i O Catawba Unit 1 B 3.6-45 Supplement 1 l

Hydrogen Recombiners B 3.6.7 8 3.6' CONTAINMEllT SYSTEMS B 3.6.7 Hydrogen Recombiners BASES BACKGROUND _ The function of the hydrogen recombiners is to eliminate the potential breach of containment due to a hydrogen oxygen reaction. 3 Per 10 CFR 50.44, " Standards for Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref.1), and l 1 ' GDC 41, " Containment Atmosphere Cleanup" (Ref. 2), hydrogen recombiners are required to reduce the hydrogen concentration in the containment following a loss of coolant accident (LOCA). The recombiners accomplish this by recombining hydrogen and oxygen to form water vapor. The vapor remains in containment, thus eliminating any discharge to the environment. The hydrogen recombiners are manually initiated since flammable limits would not be reached until several days after a Design Basis Accident (DBA). Two 100% capacity independent hydrogen recombiner systems I m are provided. Each consists of controls located outside containment in an area not exposed to the Post-Loss of-Coolant Accident environment, a power supply and a recombiner. Recombination is accomplished by heating a hydrogen air mixture above 1150"F. The resulting water vapor and discharge gases are cooled prior to dischuge from the recombiner. A single recombiner is capable of maintaining the hydrogen concentration in containment below the 4.0 volume percent (v/o) flaninability limit. Two recombiners are provided to meet the requirement for redundancy and independence. Each recombiner is powered from a separate Engineered Safety Features bus, and is provided with a separate power panel and control panel. APPLICABLE The hydrogen recombiners provide for the capability of SAFETY ANALYSES controlling the bulk hydrogen concentration in containment to less than the lower flaninable concentration of 4.0 v/o following a DBA. This control would prevent a containment wide hydrogen burn, thus ensuring the pressure and temperature assumed in the analyses are not exceeded. The limiting DBA relative to hydrogen generation is a LOCA. (continued) l Catawba Unit 1 B 3.6-46 Supplement 1 i

Hydrogen Recombiners B 3.6.7 BASES APPLICABLE Hydrogen may accumulate in containment following a LOCA as a SAFETY ANALYSES result of: (continued)

a. A metal steam reaction between the zirconium fuel rod cladding and the reactor coolant;

b. Radiolytic decomposition of water in the Reactor Coolant System (RCS) and the containment sump; j

c. Hydrogen in the RCS at the time of the LOCA (i.e.,

hydrogen dissolved in the reactor coolant and hydrogen gas in the pressurizer vapor space); or i

d. Corrosion of metals exposed to containment spray and Emergency Core Cooling System solutions.

l To evaluate the potential for hydrogen accumulation in l containment following a LOCA, the hydrogen generation as a j function of time following the initiation of the accident is calculated. Conservative assumptions reconinended by > Reference 3 are used to maximize the amount of hydrogen l' calculated. A '

!j                 Based on the conservative assumptions used to calculate the hydrogen concentration versus time after a LOCA, the              i hydrogen concentration increases at different rates               I depending on the region of the containment being measured.

The initiation of the Air Return System and Hydrogen Skintner System along with the hydrogen recombiners will maintain the hydrogen concentration in the primary containment below flammability limits. The hydrogen recombiners are designed such that, with the conservatively calculated hydrogen generation rates, a single recombiner is capable of limiting the peak hydrogen concentration in containment to less than 4.0 v/o (Ref. 3). The hyc'rogen recombiners satisfy Criterion 3 of 10 CFR 50.36 (Ref. 4). I pkJ . (continued) j Catawba Unit 1 B 3.6-47 Supplement 1 l l

4 Hydrogen Recombiners B 3.6.7 BASES (continued) LCO Two hydrogen recombiners must be OPERABLE. This ensures , operation of at least one hydrogen recombiner in the event j of a worst case single active failure. ~ Operation with at least one hydrogen recombiner ensures that , the post LOCA hydrogen concentration can be prevented from l exceeding the flammability limit. l 1 I i APPLICABILITY In MODES 1 and 2, two hydrogen recombiners are required to l control the hydrogen concentration within containment below j its flamability limit of 4.0 v/o following a LOCA, assuming a worst case single failure. In MODES 3 and 4, both the hydrogen production rate and the total hydrogen produced after a LOCA would be less than -that calculated for the DBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the hydrogen recombiners is low. Therefore, the hydrogen recombiners are not required in MODE 3 or 4.

A In MODES 5 and 6, the probability and consequences of a LOCA Q are low, due to the pressure and temperature limitations in l these MODES. Therefore, hydrogen recombiners are not i required in these MODES. ! ACTIONS Ad With one containment hydrogen recombiner inoperable, the inoperable recombiner must be restored to OPERABLE status within 30 days. In this condition, the remaining OPERABLE hydrogen recombiner is adequate to perform the hydrogen control function. However, the overall reliability is reduced because a single failure in the OPFa' tE recombiner could result in reduced hydrogen control capability. The , 30 day Completion Time is based on the availability of the other hydro , occurring (gen recombiner, that would generatethe an small amountorobability of hydrogenof that a LOCA exceeds the flannability limit), and the amount of time , available after a LOCA (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flamability limit. O (continued) U l Catawba Unit 1 B 3.6-48 Supplement 1

Hydrogen Recombiners B 3.6.7 BASES ACTIONS Ad (continued) Required Action A.1 has oeen modified by a Note that states l the provisions of LC0 3.0.4 are not applicable. As a ' result, a MODE change is allowed when one recombiner is inoperable. This allowance is based on the availability of the other hydrogen recombiner, the small probability of a LOCA occurring (that would generate an amount of hydrogen that exceeds the flaninability limit), and the amount of time available after a LOCA (should one occer) for operator action to prevent hydrogen accumulation from exceeding the flaninability limit. B.d If the inoperable hydrogen recombiner(s) cannot be restored to OPERABLE status within the required Completion Time, the A nt must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to 1 at least MODE 3 within 6 hours. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.7.1 REQUIREMENTS Performance of a system functional test for each hydrogen recombiner ensures the recombiners are operational and can attain and sustain the temperature necesiary for hydrogen recombination. In particular, this SR verifies that the minimum heater sheath temperature increases to a 700*F jn l s 90 minutes. After reaching 700'F, the oower is increased 1 to maximum povice for approximately 2 minutes and power is verified to be 2 60 kW. Industi, operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued) Catawba Unit ? 8 3.6-49 Supplement 1 l

Hydrogen Recombiners B 3.6.7 BASES SURVEILLANCE SR 3.6./.2 REQUIREMENTS 1 (continued) This SR ensures there are no physical problems that could affect recombiner operation. Since the recombiners are mechanically passive, they are not subject to mechanical failure. The only credible failure involves loose wiring or structural connections, deposits of foreign materials, etc. A visual inspection is sufficient to determine abnormal conditions that could cause such failures. The 18 month Frequency for thi: SR was developed considering the incidence of hydrogen recombiners failing the SR in the past is low. SR 3.6.7.3 This SR requires performance of a resistance to ground test for each heater phase to ensure that there are no detectable grounds in any heater phase. This SR should be performed following SR 3.6.7.1. This is accomplished by verifying that the resistance to ground for any heater phase is

t 10,000 ohms.

The 18 month Frequency for this Surveillance was developed considering the incidence of hydrogen recombiners failing the SR in the past is low. REFERENCES 1. 10 CFR 50.44.

2. 10 CFR 50, Appendix A, GDC 41.

3. UFSAR Section 6.2. )

4. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

l l Catawba Unit 1 B 3.6-50 Supplement I l

HSS 1 B 3.6.8 I q B 3.6 CONTAINMENT SYSTEMS V B 3.6.8 Hydrogen Skinrner System (HSS) BASES BACKGROUND The HSS reduces the potential for breach of containment due tc a hydrogen oxygen reaction by providing a uniformly mixed post accident containment atmosphere, thereby minimizing the potential for local hydrogen burns due to a pocket of hydrogen above the flamable concentration. Maintaining a uniformly mixed containment atmosphere also ensures that the hydrogen monitors will give an accurate measure of the bulk hydrogen concentration and give the operator the capability of preventing the occurrence of a bulk hydrogen burn inside cor.tainment per 10 CFR 50.44, " Standards for Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref.1), and 10 CFR 50 GDC 41, " Containment Atmosphere Cleanup" (Ref.2). The post accident HSS is an Engineered Safety Feature (ESF) and is designed to withstand a loss of coolant accident (LOCA) without loss of function. The System has two

independent trains, each consisting of two fans with their t

Q Q own motors and controls. Each train is sized for 4260 cfm. There is a normally closed, motor-operated valve on the hydrogen skintner suction line to prevent ice condenser bypass during initial b!owdown. The two trains are initiated automatical'y on a containment pressure high-high signal. The automatic action is to open the motor-operated valve on the hydrogen skinrner suction line after a 9

1 minute delay. Once the valve has fully opened, the hydrogen skimmer fan will start. Each train is powered from a separate emergency power supply. Since each train fan can provide 100% of the mixing requirements, the System will provide its design function with a limiting single active )

i failure. Air is drawn from the dead ended compartments by the mixing fans and is discharged toward the upper regions of the containment. This complements the air patterns established by the containment air return fans, which take suction from the operating floor level and discharoe to the lower regions of the containment, and the containreit spray, which cools the air and causes it to drop to lower elevations. The systems work together such that potentially stagnant areas , where hydrogen pockets could develop are eliminated. (continued) ) Catawba Unit 1 B 3.6-51 Supplement 1 l , l

HSS B 3.6.8 p BASES (continued) ! V 1 l APPLICABLE The HSS provides the capability for reducing the local 1 SAFETY ANALYSES hydrogen concentration to approximately t'ne bulk average concentration. The limiting DBA relative to hydrogen concentration is a LOCA. Hydrogen may accumulate in containment following a LOCA as a result of: l a- A r_tal steam reaction between the zirconium fuel rod cladding and the reactor coolant;

b. Radiolytic decomposition of water in the Reactor l Coolant System (RCS) and the containment sump;

c. Hydrogen in the RCS at the time of the LOCA (i.e.,

l hydrogen dissolved in the reactor coolant and hydrogen gas in the pressurizer vapor space); or

d. Corrosion of metals exposed to containment spray and Emergency Core Cooling System solutions.

To evaluate the potential for hydrogen accumulation in containment following a LOCA, the hydrogen generation as a - i

   '                     function of time following the initiation of the accident is     i calculated. Conservative assumptions reconinended by             )

Reference 3 are used to maximize the amount of hydrogen calculated. The HSS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4). LC0 Two HSS trains must be OPERABLE, with power to each from an independent, safety related power supply. Each train consists of one fan with its own motor and controls and is automatically initiated by a containment pressure high-high 1 signal. Operation with at least one HSS train provides the mixing necessary to ensure uniform hydrogen concentration throughout containment. l l (continued) l Catawba Unit 1 8 3.6-52 Supplement 1

HSS l B 3.6.8 r BASES (continued) j b] APPLIC/BILITY In MODES 1 and 2, the two HSS trains ensure the capability l to prevent localized hydrogen concentrations above the ! flamability limit of 4.0 volume percent in containment assuming a worst case single active failure. In MODE 3 or 4, both the hydrogen production rate and the total hydrogen produced after a LOCA would be less than that calculated for the DBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the HSS is low. Therefore, the HSS is not required in MODE 3 or 4. Ir,' MODES 5 and 6, the probability and consequences of a LOCA or steam line break (SLB) are reduced due to the pressure and temperature limitations in these MODES. Therefore, the HSS is not required in these MODES. ACTIONS L.1 With one HSS train inoperable, the inoperable train must be restored to OPERABLE status within 30 days. In this / G Condition, the remaining 0PERABLE HSS train is adequate to d perform the hydrogen mixing function. However, the overall reliability is reduced because a single failure in the OPERABLE train could result in reduced hydrogen mixing capability. The 30 day Completion Time is based on the availability of the other HSS train, the small probability of a LOCA or SLB occurring (that would generate an amount of hydrogen that exceeds the flamability limit), the amount of time available after a LOCA or SLB (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flamability limit, and the availability of the hydrogen recombiners and hydrogen ignitors. Required Action A.1 has been modified by a Note that states the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when one HSS train is inoperable. This allowance is based on the availability of , the other HSS train, the small probability of a LOCA or SLB ) occurring (that would generate an amount of hydrogen that i exceeds the flamability limit), and the amount of time available after a LOCA or SLB (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flamability limit. (continued) V Catawba Unit 1 B 3.6-53 Supplement 1 l

l HSS B 3.6.8 BASES !f] b ACTIONS B.J (continued) If an inoperable HSS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.8.1 REQUIREMENTS Operating each HSS train for 215 minutes ensures that each train is OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan and/or motor failure, or excessive vibration can be detected for corrective action. The 92 day Frequency is consistent with Inservice Testing Program Surveillance Frequencies, operating experience, the known reliability of the fan motors and controls, and the two train redundancy available. SR 3.6.8.2 Verifying HSS fan motor current at rated speed with the motor operated suction valves closed is indicative of overall fan motor performance and system flow. Such ! inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of 92 days was based on operating experience which has shown this Frequency to be acceptable. SR 3.6.8.3 l This SR verifies the operation of the motor operated suction valves and HSS fans in response to a start permissive from the Containment Pressure Control System (CPCS). The CPCS is described in the Bases for LC0 3.3.2, "ESFAS." The l p (continued) V l Catawba Unit 1 B 3.6-54 Supplement 1

HSS B 3.6.8 BASES SURVEILLANCE SR 3.6.8.3 (continued) REQUIREMENTS Frequency of 92 days was based on operating experience which has shown this Frequency to be acceptable. SR 3.6.8.4 This SR ensures that each HSS train responds properly to a - containment pressure high-high actuation signal. The ; Surveillance verifies that each fan starts after a delay of 1

t 8 minutes and s 10 minutes. The Frequency of 92 days ]

conforms with the testing requirements for similar ESF j equipment and considers the known reliability of fan motors

 ,                   and controls and the two train redundancy available.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50.44.

2. 10 CFR 50, Appendix A, GDC 41.

3. Regulatory Guide 1.7, Revision 1. l 4' . 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

i i O I Catawba 'Init 1 B 3.6-55 Supplement 1 l l

HIS B 3.6.9 [ B 3.64 CONTAINMENT SYSTEMS i B 3.6.9 Hydrogen Ignition System (HIS). BASES l BACKGROUND The HIS reduces the potential for' breach of primary containment due to a hydrogen oxygen reaction in post accident environments. The HIS is required by 10 CFR 50.44,

                            " Standards for Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref.1), and Appendix A.

GDC 41, " Containment Atmosphere Cleanup" (Ref. 2), to reduce the hydrogen concentration in the primary containment following a degraded core accident. The HIS must be capable of handling an amount of hydrogen equivaler.c to that generated from a metal water reaction involving 75% of the fuel cladding surrounding the active fuel rydon (excluding theplenumvolume). l 10 CFR 50.44 (Ref.1) requires units with ice condenser ! containments to install suitable hydrogen control systems that would acconinodate an amount of hydrogen equivalent to that generated from the reaction of 75% of the fuel cladding n with water. The HIS provides this required capability. ( This requirement was placed on ice condenser units because - of their small containment volume and low design pressure (compared with pressurized water reactor dry containments). l Calculations indicate that if hydrogen equivalent to that generated from the reaction of 75% of the fuel _ cladding with water were to collect in the primary containment, the resulting hydrogen concentration would be far above the lower flammability limit such that, if ignited from a random ignition source, the resulting hydrogen burn would seriously challenge the containment and safety systems in the

containment.

l The HIS is based on the concept of controlled ignition using thermal ignitors, designed to be capable of functioning in a post accident environment, seismically supported, and l capable of actuation from the control room. A total of 70 ignitors are distributed throughout the various regions ! of containment in which hydrogen could be released or to which it could flow in significant quantitiet,. The ignitors are arranged in two independent trains such that each containment region has at least two ignitors, one from each train, controlled and powered redundantly so that ignition p (continued) V l Catawba Unit .1 B 3.6-56 Supplement 1 l

HIS B 3.6.9 'O BASES U BACKGROUND would occur in each region even if one train failed to (continued) energize. When the HIS is initiated, the ignitor elements are energized and heat up to a surface tenperature 21700'F. At this temperature, they ignite the hydrogen gas that is present in the airspace in the vicinity of the ignitor. The HIS depends on the dispersed location of the ignitors so that local pockets of hydrogen at increased concentrations would burn before reaching a hydrogen concentration significantly higher than the lower flaninability limit. Hydrogen ignition in the vicinity of the ignitors is assumed to occur when the local hydrogen concentration reaches , 8.5 volume percent (v/o) and results in 100% of the hydrogen present being consumed. APPLICABLE The HIS causes hydrogen in containment to burn in a l SAFETY ANALYSES controlled manner as it accumulates following a degraded i coreaccident(Ref.3). Burning occurs at the lower i flaninability concentration, where the resulting temperatures and pressures are relatively benign. Without the system, . , N hydrogen could build up to higher concentrations that could l / l 's result in a violent reaction if ignited by a random ignition source after such a buildup. l The hydrogen ignitors are not included for mitigation of a Design Basis Accident (DBA) because an amount of hydrogen equivalent to that generated from the reaction of 75% of the fuel cladding with water is far in excess of the hydrogen calculated for the limiting DBA loss of coolant accident (LOCA). The hydrogen concentration resulting from a DBA can l be maintained less than the flaninability limit using the j hydrogen recombiners. The hydrogen ignitors, however, have

been shown by probabilistic risk analysis to be a l significant contributor to limiting the severity of accident j sequences that are commonly found to dominate risk for units l

l with ice condenser containments. As such, the hydrogen ; ignitors satisfy Criterion 4 of 10 CFR 50.36 (Ref. 4). l l l LC0 Two HIS trains must be OPERABLE with power from two l independent, safety related power supplies. ! (continued) Catawba Unit 1 B 3.6-57 Supplement 1 i

HIS B 3.6.9 A BASES bl LCO For this unit, an OPERABLE HIS train consists of 34 of 35 (continued) ignitors energized n the train. Operation with at least one HIS train ensures that the hydrogen in containment can be burned in a controlled manner. Unavailability of both HIS trains could lead to hydrogen buildup to higher concentrations, which could result in a violent reaction if ignited. The reaction could take place fast enough to lead to high temperatures and overpressurization of containment and, as a result, breach containment or cause containment leakage rates above those assumed in the safety analyses. Damage to safety related equipment located in containment could also occur. APPLICABILITY Requiring OPERABILITY in MODES 1 and 2 for the HIS ensures its immediate availability after safety injection and scram actuated on a LOCA initiation. In the post accident environment, the two HIS subsystems are required to control the hydrogen concentration within containment to near its flammability limit of 4.0 v/o assuming a worst case single failure. This prevents overpressurization of containment

,                                                 and damage to safety related equipment and instruments i

located within containment. In MODES 3 and 4, both the hydrogen production rate and the total hydrogen production after a LOCA would be significantly less than that calculated for the DBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the HIS is low. Therefore, the HIS is not required in MODES 3 and 4. In MODES 5 and 6, the probability and consequences of a LOCA are reduced due to the pressure and temperature limitations of these MODES. Therefore, the HIS is not required to be OPERABLE in MODES 5 and 6. ACTIONS A.1 and A.2 With or,e HIS train it. operable, the inoperable train must be restored to OPERABLE status within 7 days or the OPERABLE train must be verified OPERABLE frequently by performance of SR 3.6.9.1. The 7 day Completion Time is based on the low .p (continued) t l Catawba Unit 1 B 3.6-58 Supplement 1 I

HIS B 3.6.9 1 !A BASES b ACTIONS A.1 and A.2 (continued) probability of the occurrence of a degraded core event that ! would generate hydrogen in amounts equivalent to a metal i water reaction of 75% of the core cladding, the length of time after the event that operator action would be required to prevent hydrogen accumulation from exceeding this limit, j and the low probability of failure of the OPERABLE HIS train. Alternative Required Action A.2, by frequent surveillances, provides assurance that the OPERABLE train continues to be OPERABLE. IL1 Condition B is one containment region with no OPERABLE hydrogen ignitor. Thus, while in Condition B, or in Conditions A and B simultaneously, there would always be ignition capability in the adjacent containment regions that would provide redundant capability by flame propagation to the region with no OPERABLE ignitors. Required Action B.1 calls for the restoration of one r I hydrogen ignitor in each region to OPERABLE status within i 7 days. The 7 day Completion Time is based on the same i l reasons given under Required Action A.1.

L1 ,

The unit must be placed in a MODE in which the LC0 does not apply if the HIS subsystem (s) cannot be restored to OPERABLE status within the associated Completion Time. This is done by placing the unit in at least MODE 3 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. i SURVEILLANCE SR 3.6.9.1 i REQUIREMENTS This SR confirms that a 34 of 35 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements. Therefore, energizing provides (continued) Catawba Unit 1 B 3.6-59 Supplement 1 l l I l l

l HIS B 3.6.9 BASES l b SURVEILLANCE SR 3.6.9.1 (continued) l REQUIREMENTS assurance of OPERABILITY. The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise redundancy in that region, the containment regions are interconnected so that ignition in one region would cause burning to progress to the others (i.e., there is overlap in each hydrogen ignitor's effectiveness between regions). The Frequency of 92 days has been shown to be acceptable through operating experience. SR 3.6.9.2 This SR confirms that the two inoperable hydrogen ignitors allowed by SR 3.6.9.1 (i.e., one in each train) are not in the same containment region. The Frequency of 92 days is acceptable based on the Frequency of SR 3.6.9.1, which provides the information for performing this SR. SR 3.6.9.3 A more detailed functional test is performed every 18 months to verify system OPERABILITY. Each glow plug is visually examined to ensure that it is clean and that the electrical circuitry is energized. All ignitors (glow plugs), including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized. Additionally, the surface temperature of each glow plug is measured to be 2 1700"F to demonstrate that a temperature j sufficient for ignition is achieved. The 18 month Frequency l is based on the need to perform this Surveillance under the i conditions that apply during a plant outage and the l potential for an unplanned transient if the Surveillance { were performed with the reactor at power. Operating ; experience has shown that these components usually pass the ! SR when performed at the 18 month Frequency, which is ' based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (continued) l Catawba Unit 1 B 3.6-60 Supplement 1

HIS B 3.6.9 BASES (continued) REFERENCES 1. 10 CFR 50.44.

2. - 10 CFR 50, Appendix A, GDC 41..

! 3-. UFSAR, Section 6.2. l 10 CFR 50.36 Technical Specifications, (c)(2)(ii). 4. i' l f I O O Catawba Unit 1 B 3.6-61 Supplement 1 l l ,

AVS B 3.6.10 ()

Q 8 3.6 CONTAINMENT SYSTEMS B 3.6.10 Annulus Ventilation System (AVS)

BASES BACKGROUND The AVS is required by 10 CFR 50, Appendix A, GDC 41,

                        " Containment Atmosphere Cleanup" (Ref. 1), to ensure that radioactive  materials into the reactor  buildingthat leak fromcontainment (secondary   the primary)following containment a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The containment has a secondary containment called the reactor building, which is a concrete structure that surrounds the steel primary containment vessel. Between the containment vessel and the reactor building inner wall is an annulus that collects any containment leakage that may occur following a loss of coolant accident (LOCA) or rod ejection-accident. This space also allows for periodic inspection of the outer surface of the steel containment vessel.

                      ' The AVS establishes a negative pressure in the annulus between the reactor building and the steel containment p

vessel. Filters in the :;ystem then control the release of (] radioactive contaminants to the environment. Reactor building OPERABILITY is required to ensure retention of primary containment leakage and proper operation of the AVS. The AVS consists of two separate and redundant trains. Each train includes a heater, a prefilter/ moisture separator, upstream and downstream high efficiency particulate air (HEPA) filters, an activated charcoal adsorber section for removal of radiciodines, and a fan. Ductwork,valvesand/or dampers, and instrumentation also form part of the system. The moisture separators function to reduce the moisture content of the airstream. A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case of failure of the main HEPA filter bank. Only the upstream HEPA filter and the charcoal adsorber section are credited in the analysis. The system initiates and maintains a negative air pressure in the reactor building annulus by means of filtered exhau.c ventilation of the reactor building annulus following receipt of a safety injection (SI) signal. The sy. stem is described in Reference 2. p (continued) b l Catawba Unit 1 B 3.6-62 Supplement 1

AVS B 3.6.10 l BASES BACKGROUND The prefilters remove large particles in the air, and the (continued) moisture separators remove entrained water droplets present, l l to prevent excessive loading of the HEPA filters and charcoal absorbers. Heaters are included to reduce the relative humidity of the airstream. Continuous operation of each train, for at least 10 hours per month, with heaters on, reduces moisture buildup on their HEPA filters and

adsorbers.

The AVS reduces the radioactive content in the annulus atmosphere following a DBA. Loss of the AVS could cause . site boundary doses, in the event of a DBA, to exceed the i values given in the licensing basis. l APPLICABLE The AVS design basis is established by the consequences of i SAFETY ANALYSES the limiting DBA, which is a LOCA. The accident analysis i (Ref. 3) assumes that only one train of the AVS is l functional due to a single failure that disables the other : train. The accident analysis accounts for the reduction in ! airborne radioactive material provided by the remaining one train of this filtration system. The amount of fission , l 7N products available for release from containment is ! i determined for a LOCA. The modeled AVS actuation in the safety analyses is based {' upon a worst case response time following an SI initiated at the limiting setpoint. The total response time, from exceeding the signal setpoint to attaining the negative l pressure of 0.5 inch water gauge in the reactor building annulus, is 1 minute. This response time is composed of signal delay, diesel generator startup and sequencing time, system startup time, and time for the system to attain the required pressure after starting. The AVS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4). LCO In the event of a DBA, one AVS train is required to provide i the minimum particulate iodine removal assumed in the safety l l analysis. Two trains of the AVS must be OPERABLE to ensure that at least one train will operate, assuming that the other train is disabled by a single active failure. (continued) Catawba Unit 1 B 3.6-63 Supplement 1 l l l i

AVS B 3.6.10 BASES (continued) APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could lead to fission product release to containment that leaks to the reactor building. The large break LOCA, on which this system's design is based, is a full power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and Reactor Coolant System pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low. In MODES 5 and 6, the probability and consequences of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the AVS is not required to be OPERABLE. ACTIONS A.J With one AVS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days. The 7 day Completion Time is based on consideration of such factors as c the availability of the OPERABLE redundant AVS train and the low probability of a DBA occurring during this period. The Completion Time is adequate to make most repairs. B.1 and B.2 With one or more AVS heaters inoperable, the heater must be restored to OPERABLE status within 7 days. Alternatively, a report must be initiated within 7 days per Specification 5.6.6, which details the reason for the heater's inoperability and the corrective action required to return the heater to OPERABLE status. The heaters do not affect OPERABILITY of the AVS filter trains because charcoal adsorber efficiency testing is performed at 30*C and 95% relative humidity. The accident analysis shows that site boundary radiation doses are within 10 CFR 100 limits during a DBA LOCA under these conditions. p (continued) \ O l Catawba Unit 1 B 3.6-64 Supplement 1

L AVS. B 3.6.10 BASES ACTIONS C.1 and C.2 (continued) If the AVS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least l MODE 3 within 6 hours and to MODE 5 within 36 hours. The l allowed Completion Times are reasonable, based on operating l experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. i SURVEILLANCE SR 3.6.10.1 REQUIREMENTS Operating each AVS train from the control room with flow through the HEPA filters and charcoal adsorbers ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for 210 continuous hours eliminates moisture on the adsorbers O - and HEPA filters. Experience from filter testing at operating units indicates that the 10 hour period is adequate for moisture elimination on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls, the two train redundancy available, and the iodine removal capability of the Containment Spray System and Ice Condenser. SR 3.6.10.2 This SR verifies that the required AVS filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The AVS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 5). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. l l (continued) l Catawba Unit 1 B 3.6-65 Supplement 1 l

 ,                                                                                                         AVS B 3.6.10 r    BASES V]

SURVEILLANCE SR 3.6.10.3 REQUIREMENTS f (continued) The automatic startup on a safety injection signal ensures that each AVS train responds properly. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore the Frequency was concluded to be acceptable from a reliability standpoint. Furthermore, the SR interval was developed considering that the AVS equipment OPERABILITY is demonstrated at a 31 day Frequency by SR 3.6.10.1. SR 3.6.10.4 The AVS filter cooling electric motor-operated bypass valves are tested to verify OPERABILITY. The valves are normally closed and may need to be opened to initiate miniflow cooling through a filter unit that has been shutdown following a DBA LOCA. Miniflow cooling may be necessary to O limit temperature increases in the idle filter train due to . decay heat from captured fission products. The 18 month Frequency is considered to be acceptable based on valve reliability and design, and the fact that operating experience has shown that the valves usually pass the Surveillance when performed at the 18 month Frequency. REFERENCES 1. 10 CFR 50, Appendix A, GDC 41.

2. UFSAR, Section 9.4.

3. UFSAR, Chapter 15.

4. 10 CFR 50.36 Technical Specifications, (c)(2)(ii).

5. Regulatory Guide 1.52, Revision 2.

O l Catawba Unit 1 B 3.6-66 Supplement 1

I ARS l B 3.6.11 1 i B 3.6 CONTAINMENT SYSTEMS v B 3.6.11 Air Return System (ARS) BASES l BACKGROUND The ARS is designed to assure the rapid return of air from I the upper to the lower containment compartment after the . initial blowdown following a Design Basis Accident (DBA). The return of this air to the lower compartment and ' l subsequent recirculation back up through the ice condenser assists in cooling the containment atmosphere and limiting )' post accident pressure and temperature in containment to less than design values. Limiting pressure and temperature reduces the release of fission product radioactivity from 1 containment to the environment in the event of a DBA. The ARS also promotes hydrogen dilution by mixing the hydrogen with containment atmosphere and distributing throughout the contain.nent. The ARS consists of two separate trains of equal capacity, each capable of meeting the design bases. Each train J n includes a 100% capacity air return fan, and associated l ty motor operated damper in the fan discharge line to the i containment lower compartment. The damper acts as a barrier j between the upper and lower compartments to prevent reverse flow which would bypass the ice condenser. The damper is normally closed and remains closed throughout the initial blowdown following a postulated high energy line break. The damper motor is actuated several seconds after the Containment High-High pressure setpoint is reached and a start permissive from the Containment Pressure Control System is present. A backdraft damper is also provided at the discharge of each fan to serve as a check valve. Each train is powered from a separate Engineered Safety Features (ESF) bus. The ARS fans are automatically started by the containment pressure High-High signal 9

1 minutes after the containment pressure reaches the pressure setpoint and a !

start permissive from the Containment Pressure Control System is present. The time delay ensures that no energy released during the initial phase of a DBA will bypass the i ice bed through the ARS fans. ; l

 'O                                                                         (continued)      I l

Catawba Unit 1 B 3.6-67 Supplement 1 l l l I

ARS B 3.6.11 BASES BACKGROUND After starting, the fans displace air from the upper (continued) compartment to the lower compartment, thereby returning the , air that was displaced by the high energy line break blowdown from the lower compartment and equalizing pressures throughout containment. After discharge into the lower compartment, air flows with steam produced by residual heat through the ice condenser doors into the ice condenser compartment where the steam portion of the flow is condensed. The air flow returns to the upper compartment through the top deck doors in the upper portion of the ice condenser compartment. The ARS fans operate continuously after actuation, circulating air through the containment volume. When the containment pessure falls below a predetermined value, the ARS fans are automatically de-energized. Thereafter, the fcns are automatically cycled on and off if necessary to control any additional containment pressure trans 6r.cs. The ARS also functions, after all the ice has melted, to circulate any steam still entering the lower compartment to the upper compartment where the Containment Spray System can cool it. The ARS is an ESF system. It is designed to ensure that the

 %                    heat removal capability required during the post accident period can be attained. The operation of the ARS, in conjunction with the ice bed, the Containment Spray System, and the Residual Heat Removal (RHR) System spray, provides the required heat removal capability to limit post accident conditions to less than the containment design values.

APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES temperature and pressure are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are , analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. DBAs are assumed not to occur simultaneously or i consecutively. The postulated DBAs are analyzed, in regard I to ESF systems, assuming the loss of one ESF bus, which is ! the worst case single active failure and results in one ; train each of the Containment Spray System, RHR System, and (continued) x l Catawba Unit 1 B 3.6-68 Supplement 1

L FRS B 3.6. 1 BASES APPLICABLE ARS being inoperable (Ref.1)'. The DBA analyses show that SAFETY ANALYSES the maximum peak containment pressure results from the LOCA (continued) analysis and is calculated to be less than the containment design pressure. For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather th:n maximize, t?] calculated transient containment pressures, in accordance with 10 CFR 50, Appendix K (Ref. 2). The analysis for minimum internal containment pressure (i.e., maximum external differential containment pressure) assumes inadvertent simultaneous actuation of both the ARS and the Containment Spray System. The modeled ARS actuation from the containment analysis is based upon a response time associated with exceeding the O containment pressure High-High signal setpoint to achieving O full ARS air flow. A delayed response time initiation provides conservative analyses of peak calculated containment temperature and pressure responses. The ARS-total response time of 600 seconds includes signal delays. The ARS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 3). LCO In the event of a DBA, one train of the ARS is required to provide the minimum air recirculation for heat removal assumed in the safety analyses. To ensure this requirement is met, two trains of the ARS must be OPERABLE. This will ensure that at least one train will operate, assuming the worst case single failure occurs, which is in the ESF power supply. (continued) Catawba Unit 1 B 3.6-69 Supplement 1 l

ARS B 3.6.11 BASES (continued) k APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ARS. Therefore, the LC0 is applicable in MODES 1, 2, 3, and 4. In' MODES 5 and 6 the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ARS is not required to be OPERABLE in these MODES. ACi?ONS L1 If one of the required trains of the ARS is inoperable, it must be restored to OPERABLE status within 72 hours. The 72 hour Completion Time was developed taking into account the redundant flow of the OPERABLE ARS train and the low probability of a DSA occurring in this period. B.1 and B.2 O

If the ARS train cannot be restored to OPERABLE status C/ within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE 3R 3.LjL1 REQUIREMENTS Verifying tnat each ARS fan starts on an actual or simulated actuation signal, after a delay 2 8 minutes and s 10 minutes, and operates for 215 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly. It also ensures that blockage, fan and/or motor failure, or (continued) l l Catawba Unit 1 B 3.6-70 Supplement 1

ARS B 3.6.11 BASES SURVEILLANCE SR 3.6.11.1 (continued) REQUIREMENTS excessive vibration can be detected for corrective action. The 92 day Frequency was developed considering the known reliability of fan motors and controls and the two train redundancy available. SR 3.6.11.2 Verifying ARS fan motor current to be at rated speed with the return air dampers closed confirms one operating condition of the fan. This test is indicative of overall fan motor performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnonnal perfonnance. The Frequency of 92 days conforms with the testing requirements for similar ESF equipment and considers the known reliability of fan motors and controls and the two train redundancy available. . SR 3.6.11.3 O Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exi!;t when the fan is started. This Surveillance also' tests the circuitry, including time delays to ensure the system operates properly. The Frequency of 92 days was developed cons!dering the importance of the dampers, their location, physical environment, and probability of failure. Operating experience has also shown this Frequency to be acceptable. SR 3.6.11.4 and SR 3.6.11.5 Verifying the OPERABILITY of the check damper in the air return fan discharge line to the containment lower compartment provides assurance that the proper flow path will exist when the fan is started and that reverse flow can not occur when the fan is not operating. The Frequency of 92 days _was developed considering the importance of the dampers, their location, physical environment, and probability of failure. Operating experience has also shown this Frequency to be acceptable. (continued) L Catawba Unit 1 B 3.6-71 Supplement 1 l

t ARS B 3.6.11

 ,q    BASES (continued) i t

O j SURVEILLANCE SR 3.6.11.6 and SR 3.6.11.7 - l REQUIREMENTS (continued) These SRs require verification that each ARS motor operated

damper opens or is prevented from opening and each ARS fan l 1s prevented from starting upon receipt of Containment l Pressure Contol System start and terminate signals. The l

CPCS is described in the Bases for LCO 3.3.2, "ESFAS." The l 18 month Frequency is based on operating experience which I has shown it to be acceptable. REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50, Appendix K.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

O O B 3.6-72 Supplement 1 l Catawba Unit 1

o Ice Bed B 3.6.12

    ~B 3.6 CONTAINMENT SYSTEMS B 3.6.12 Ice Bed                                                                    3 BASES l    . BACKGROUND       The ice bed consists of over 2,475,252 lb of ice stored in 1944 baskets within the ice condenser. Its primary purpose is to provide a large heat sink in the event of a release of I                       energy from a Design Basis Accident (DBA) in containment.

L The ice would absorb energy and limit containment peak ! pressure and temperature during the accident transient. l Limiting the pressure and temperature reduces the release of-fission product radioactivity from containment to the i 1 environment in the event of a DBA. l { ! 1 l The ice condenser is an annular compartment enclosing ' approximately 300* of the perimeter of the upper containment compartment, but penetrating the operating deck so that a portion extends into the lower containment compartment. The i lower portion has a series of hinged doors exposed to the atmosphere of the lower containment compartment, which, for l normal unit operation, are designed to remain closed. At j the top of the ice condenser is another set of doors exposed l to the atmosphere of the upper compartment, which also j remain closed during nonnal unit operation. Intermediate j deck doors, located below the top deck' doors, form the floor l of a plenum at the upper part of the ice condenser. These I doors also remain closed during normal unit operation. The upper plenum area is used to facilitate surveillance and maintenance of the ice bed. ]

The ice baskets neld in the ice bed within the ice condenser are arranged to promote heat transfer from steam to ice.

l This arrangement enhances the ice condenser's primary function of condensing steam and absorbing heat energy released to the containment during a DBA. l l In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment. This allows air and steam to flow from the lower compartment into the ice condenser. The resulting pressure increase within the ice condenser causes the intermediate deck doors and the top deck doors to open, which allows the air to flow out of the ice condenser into the upper compartment. Steam condensation within the ice condenser limits the pressure and temperature buildup in (continued) ) Catawba Unit 1 B 3.6-73 Supplement 1 l l

Ice Bed B 3.6.12 (] LJ BASES BACKGROUND containment. A divider barrier separates the upper and ! (continued) lower compartments and ensures that the steam is directed into the ice condenser. l The ice, together with the containment spray, is adequate to I absorb the initial blowdown of steam and water from a DBA and the additional heat loads that would enter containment l during several hours following the initial blowdown. The additional heat loads would come from the residual heat in l the reactor core, the hot piping and components, and the secondary system, including the steam generators. During i the post blowdown period, the Air Return System (ARS) l returns upper compartment air through the divider barrier to the lower compartment. This serves to equalize pressures in containment and to continue circulating heated air and steam from the lower compartment through the ice condenser where the heat is removed by the remaining ice. As ice melts, the water passes through the ice condenser floor drains into the lower compartment. Thus, a second function of the ice bed is to be a large source of borated water (via the containment sump) for long term Emergency Core Cooling System (ECCS) and Containment Spray System heat removal functions in the recirculation mode. l A third function of the ice bed and melted ice is to remove fission product iodine that may be released from the core during a DBA. Iodine removal occurs during the ice melt J phase of the accident and continues as the melted ice is sprayed into the containment atmosphere by the Containment Spray System. The ice is adjusted to an alkaline pH that facilitates removal of radioactive iodine from the containment atmosphere. The alkaline pH also minimizes the ) l occurrence of the chloride and caustic stress corrosion on i mechanical systems and components exposed to ECCS and Containment Spray System fluids in the recirculation mode of operation. It is important for the ice to be uniformly distributed around the 24 ice condenser bays ar.d for open flow paths to exist around ice baskett. This is especially important during the initial blowdown so that the steam and water mixture entering the lower compartment do not pass through only part of the ice condenser, depleting the ice there while bypassing the ice in other bays. q (continued) NJ \

l Catawba Unit 1 B 3.6-74 Supplement I l

l l

I i Ice Bed B 3.6.12 ( ' BASES BACKGROUND Two phenomena that can degrade the ice bed during the long (continued) service period are:

a. Loss of ice by melting or sublimation; and; b. Obstruction of flow passages through the ice bed due to buildup of f.rost or ice. Both of these degrading phenomena are reduced by minimizing air leakage into and out of the ice condenser.
The ice bed limits the temperature and pressure that could i

be expected following a DBA, thus limiting leakage of fission product radioactivity from containment to the environment. j APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES temperature and pressure are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are

analyzed using computer codes designed to predict the l resultant containment pressure and temperature trans: ents.
fm DBAs are not assumed to occur simultaneously or !

l (") consecutively. l l Although the ice condenser is a passive system that requires i no electrical power to perform its function, the Containment Spray System, RHR Spray System, and the ARS also function to assist the ice bed in limiting pressures and temperatures. Therefore, the postulated DBAs are analyzed in regards to containment Engineered Safety Feature (ESF) systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the i Containment Spray System, RHR Spray System, and ARS being inoperable. The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. For certain aspects of the transient accident analyses, maximizing the calculated containment pressure is l not conservative. In particular, the cooling effectiveness i of the ECCS during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For (continued) Cattwba Unit 1 B 3.6-75 Supplement 1 l

l Ice Bed B 3.6.12 f BASES ! APPLICABLE' these calculations', the containment b'ackpres'sure is l SAFETY ANALYSES calculated in a manner designed to conservatively minimize,

          -(continued)  rather than maximize,' the calculated transient containment          {

pressures, in accordance with 10 CFR 50, Appendix K (Ref._2). The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed in the Bases for LCO 3.6.5, " Containment Air Temperature." In addition to calculating the overall peak containment pressures, the DBA analyses include calculation of the transient differential pressures that occur across . subcompartment walls during the initial blowdown phase of the accident transient. The internal containr/ent walls and structures are designed to withstand these local-transient pressure differentials for the limiting DBAs. The ice bed satisfies Criterion 3 of 10 CFR 50.36 (Ref. 3). LCO The ice bed LCO requires the existence of the required O quantity of stored ice, appropriate distribution of the ice and the ice bed, open flow paths through the ice bed, and appropriate chemical content and pH of the stored ice. The stored ice functions to absorb heat during a DBA, thereby limiting containment air te:nperature and pressure. The chemical content and pH of the ice provide core SDM (boron content) and remove radioactive iodine from the containment atmosphere when the melted ice is recirculated through the ECCS_and the Containment Spray System, respectively. APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed. Therefore, the LC0 is applicable in MODES 1, 2, 3, and 4. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ice bed is not required to be OPERABLE in these MODES.

   *                                                                        (continued) l Catawba Unit 1                      B 3.6-76                        Supplement 1 l

t Ice Bed B 3.6.12 BASES (continued) ACTIONS 8.d If the ice bed is inoperable, it must be restored to OPERABLE status within 48 hours. The Completion Time was developed based on operating experience, which confirms that due to the very large mass of stored ice, the parameters comprising OPERABILITY do not change appreciably in this time period. Because of this fact, the Surveillance Frequencies are long (months), except for the ice bed temperature, which is checked every 12 hours. If a degraded condition is identified, even for temperature, with such a large mass of ice it is not possible for the degraded condition to significantly degrade further in a 48 hour period. Therefore, 48 hours is a reasonable amount of time to correct a degraded condition before initiating a shutdown. 3.1 and 9 2 If the ice bed cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this - O, status, the plant must be brought to at least MODE 3 within 6 hours ano to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.12.1 REQUIREMENTS Verifying that the maximum temperature of the ice bed is s; 27'F ensures that the ice is kept well below the melting l point. The 12 hour Frequen:y was based on operating experience, which confirmed that, due to the large mass of , stored ice, it is not possible for the ice bed temperature ' to degrade significantly within a 12 hour period and was also based on assessing the proximity of the LC0 limit to the melting temperature. Furthermore, the 12 hour Frequency is considered adequate in viaw of indications in the control room, including the alarm, to alert the operator to an abnormal ice bed (continued) i 1 Catawba Unit 1 B 3.6-77 Supplement 1 l

,. Ice B:d B 3.6.12 BASES SURVEILLANCE SR 3.6.12.1 (continued) REQUIREMENTS temperature condition. This SR may be satisfied by use of the Ice Bed Temperature Monitoring System. SR 3.6.12.2 This SR ensures that the flow channels through the ice condenser have not accumulated an excessive amount of ice or

,                                                      frost blockage. The visual inspection must be made for two or more flow channels per ice condenser bay and must include the following specific locations along the flow channel:

a. Past the lower inlet plenum support structures and turning vanes;

b. Between ice baskets;

c. Past lattice frames;

d. Through the intermediate floor grating; and

e. Through the top deck floor grating.

The allowable 0.38 inch thick buildup of frost or ice is based on the analysis of containment response to a DBA with partial blockage of the ice condenser flow passages. If a flow channel in a given bay is found to have an accumulation of frost or ice > 0.38 inch thick, a representative sample of 20 additional flow channels from the same bay must be visually inspected. If these additional flow channels are all found to be acceptable, the discrepant flow channel may be considered single, unique, and acceptable deficiency. More than one discrepant flow channel in a bay is not acceptable, however. These requirements are based on the sensitivity of the partial blockage analysis to additional blockage. The Frequency of 9 months was based on ice storage tests and the I allowance built into the required ice mass over and above ; the mass assumed in the safety analyses. (continued) l Catawba Unit 1 B 3.6-78 Supplement 1

Ice Bed B 3.6.12 f) U BASES SURVEILLANCE SR 3.6.12.3 REQUIREMENTS i (continued) Verifying the chemical composition of the stored ice ensures that the stored ice has a boron concentration of at least 1800 ppm as sodium tetraborate and a high pH, 2 9.0 and ; s 9.5, in order to meet the requirement for borated water 1 when the melted ice is used in the ECCS recirculation mode of operation. Sodium tetraborate has been proven effective in mainta .;ing the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine. The high pH is required to enhance the effectiveness of the ice and the melted ice in removing iodine from the containment atmosphere. This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment Spray System fluids in the recirculation mode of operation. The Frequency of 9 months is based on operating experience. SR 3.6.12.4 The weighing program is designed to obtain a representative l O samole of the ice basFfts. The representative sample shall include 6 baskets from each of the 24 ice condenser bays and shall consist of one basket from radial rows 1, 2, 4, 6, 8, i and 9. If no basket from a designated row can be obtained for weighing, a basket from the same row of an adjacent bay ! shall be weighed. The rows chosen include the rows nearest the inside and outside walls of the ice condenser (rows 1 and 2, and 8 and 9, respectively), where heat transfer into the ice condenser is most likely to influence melting or sublimation. Verifying the total weight of ice ensures that there is adequate ice to absorb the required amount of energy to mitigate the DBAs. If a basket is found to contain < 1273 lb of ice, a representative sample of 20 additional baskets from the same bay shall be weighed. The average weight of ice in these 21 baskets (the discrepant basket and the 20 additional baskets) shall be 21273 lb at a 95% confidence level. (continued) Catawba Unit 1 B 3.6-79 Supplement 1 l

Ice B:d B 3.6.12

 /7        BASES V

SURVEILLANCE SR 3.6.12.4 (continued) REQUIREMENTS Weighing 20 additional baskets from the same bay in the event a Surveillance reveals that a single basket contains

                                    < 1273 lb ensures that no local zone exists that is grossly deficient in ice. Such a zone could experience early melt out during a DBA transient, creating a path for steam to pass through the ice bed without being condensed. The Frequency of 18 months was based on ice storage tests and the allowance built into the required ice mass over and above the mass assumed in the safety analyses. Operating experience has verified that, with the 18 month Frequency, the weight requirements are maintained with no significant degradation between surveillances.

SR 3.6.12,5 This SR ensures that the azimuthal distribution of ice is reasonably uniform, by verifying that the average ice weight in each of three azimuthal groups of ice condenser bays is within the limit. The Frequency of 18 months was based on ice storage tests and the allowance built into the required ice mass over and above the mass assumed in the safety analyses. Operating experience has verified that, with the 18 month Frequency, the weight requirements are maintained with no significant degradation between surveillances. SR 3.6.12.6 This SR ensures that a representative sampling of ice baskets, which are relatively thin walled, perforated cylinders, have not been degraded by wear, cracks, corrosion. ?r other damage. Each ice basket must be raised at least 12 feet for this inspection. The Frequency of 40 months, for a visual inspection of the structural soundness of the ice baskets is based on engineering judgment and considers such factors as the thickness of the basket walls relative to corrosion rates expected in their service environment and the results of the long term ice storage testing. (continued) l l Catawba Unit 1 B 3.6-80 Supplement 1

r T Ice Bed B 3.6.12 BASES (continued) REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50, Appendix K.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

i l I l o -

I l l l l O Catawba Uniit 1 B 3.6-81 Supplenent 1 l

Ice Condens2r Dosrs B 3.6.13 B 3.6 CONTAINMENT SYSTEMS B 3.6.13 Ice vsndenser Doors BASES BACKGROUND The ice condenser doors consist of the inlet doors, the intermediate deck doors, and the top deck doors. The functions of the doors are to:

a. Seal the ice condenser from air leakage during the lifetime of the unit; and

b. Open in the event of a Design Basis Accident (DBA) to direct the hot steam air mixture from the DBA into the {

ice bed, where the ice would absorb energy and limit containment peak pressure and temperature during the accident transient. Limiting the pressure and temperature following a DBA l reduces the release of fission product radioactivity from containment to the environment. The ice condenser is an annular compartment enclosing O, approximately 300* of the perimeter of the upper containment compartment, but penetrating the operating deck so that a J portion extends into the lower containment compartment. The inlet doors separate the atmosphere of the lower compartment from the ice bed inside the ice condenser. The top deck doors are above the ice beJ and exposed to the atmosphere of the upper compartment. The intermediate deck doors, located below the top deck doors, form the floor of a plenum at the upper part of the ice condenser. This plenum area is used to facilitate surveillance and maintenance of the ice bed. The ice basketc held in the ice bed within '.he ice condenser are arranged to promote heat transfer from steam to ice. This arrangement enhances tto ice condenser's primary function of condensing steam and absorbing heat energy released to the containment during a DBA. ) In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment. This allows air and steam to flow from the lower compartment into the ice condenser. The resulting ptessure increase within the ice condenser causes the intermediate deck doors and the top deck doors to open, (continued) O l Catawba Unit 1 9 3.6-82 Supplement 1

Ice Condenser Do::rs B 3.6.13 O BASES b BACKGROUND which allows the air to flow out of the ice condenser into (continued) the upper compartment. Steam condensation within the ice condensers limits the pressure and temperature buildup in containment. A divider barrier separates the upper and lower compartments and ensures that the steam is directed into the ice condenser. The ice, together with the containment spray, serves as a containment heat removal system and is adequate to absorb the initial blowdown of steam and water from a DBA as well as the additional heat loads that would enter containment during the several hours following the initial blowdown. The additional heat loads would come from the residual heat in the reactor core, the ht,t piping and components, and the secondary system, including the steam generators. During the post blowdown period, the Air Return System (ARS) returns upper compartment air through the divider barrier to the lower compartment. This serves to equalize pressures in containment and to continue circulating heated air and steam from the lower compartment through the ice condenser, where the heat is removed by the remaining ice. The water from the melted ice drains into the lower compartment where it serves as a source of borated water (via the containment sump) for the Emergency Core Cooling System (ECCS) and the Containment Spray System heat removal functions in the recirculation mode. The ice (via the Containment Spray System) and the recirculated ice melt also serve to clean up the containment atmosphere. The ice condenser doors ensure that the ice stored in the ice bed is preserved during normal operation (doors closed) and that the ice condenser functions as designed if called upon to act as a passive heat sink following a DBA. APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES pressure and temperature are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. j DBAs are assumed not to occur simultaneously or I consecutively. l 1 (continued) j O l Catawba Unit 1 B 3.6-83 Supplement 1 l

Ice Condenser Doors B 3.6.13 / BASES APPLICABLE Although the ice condenser is a passive system that requires SAFETY ANALYSES no electrical power to perform its function, the Containment (continued) Spray System and ARS also function to assist the ice bed in limiting pressures and temperatures. Therefore, the postulated DBAs are analyzed with respect to Engineered Safety Feature (ESF) systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the Containment Spray System and the ARS being rendered inoperable. The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the ECCS during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is , calculated in a manner designed to conservatively minimize, i rather than maximize, the calculated transient containment pressures, in accordance with 10 CFR 50, Appendix K (Ref. 2). O U The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed in the Bases for LC0 3.6.5, " Containment Air Temperature." An additional design requirement was imposed on the ice condenser door design for a small break accident in which ; the flow of heated air and steam is not sufficient to fully l open the doors. I For this situation, the doors are designed so that all of the doors would partially open by approximately the same amount. Thus, the partially opened doors would modulate the flow so that each ice bay would receive an approximately equal fraction of the total flow. This design feature ensures that the heated air and steam will not flow preferentially to some ice bays and deplete the ice there without utilizing the ice in the other bays. In addition to calculating the overall peak containment pressures, the DBA analyses include the calculation of the transient differential pressures that would occur across (continued) N l Catawba Unit 1 B 3.6-84 Supplement 1

Ice Condenstr Doors B 3.6.13 BASES [V3 APPLICABLE subcompartment walls during the initial blowdown phase of SAFETY ANALYSES the accident transient. The internal containment walls and (continued) structures are oesigned to withstand the local transient pressure differentials for the limiting DBAs. The ice condenser doors satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3). l LC0 This LC0 establishes the minimum equipment requirements to assure that the ice condenser doors perform their safety i function. The ice condenser inlet doors, intermediate deck I doors, and top deck doors must be closed to minimize air leakage into and out of the ice condenser, with its attendant leakage of heat into the ice condenser and loss of ice through melting and sublimation. The doors must be OPERABLE to ensure the proper opening of the ice condenser in the event of a DBA. OPERABILITY includes being free of any obstructions that would limit their opening, and for the inlet doors, being adjusted such that the opening and closing torques are within limits. The ice condenser doors function with the ice condenser to limit the pressure.and temperature that could be expected following a DBA. APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice condenser doors. Therefore, the LCO is ; applicable in MODES 1, 2, 3, and 4. The probability and consequences of these events in MODES 5 and 6 are reduced due to the pressure and temperature limitations of these MODESs Therefore, the ice condenser doors are not required to be OPERABLE in these MODES. ACTIONS A Note provides clarification that, for this LCO, separate Condition entry is allowed for each ice condenser door. I (continued) Catawba Unit 1 B 3.6-85 Supplement 1 l

Ice Condenser Dosrs B 3.6.13 BASES ACTIONS U (continued) If one or more ice condenser inlet doors are inoperable due to being physically restrained from opening, the door (s) must be restored to OPERABLE status within 1 hour. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LC0 3.6.1,

                  " Containment," which requires containment to be restored to OPERABLE status within 1 hour.

B.1 and B.2 If one or more ice condenser doors are detennined to be partially open or otherwise inoperable for reasons other than Condition A or if a door is found that is not closed, it is acceptable to continue unit operation for up to 14 days, provided the ice bed temperature instrumentation is monitored once per 4 hours to ensure that the open or inoperable door is not allowing enough air leakage to cause the maximum ice bed temperature to approach the melting point. The Frequency of 4 hours is based on the fact that O temperature changes cannot occur rapidly in the ice ud because of the large mass of ice involved. The 14 day Completion Time is based on long term ice storage tests that indicate that if the temperature is maintained below 27*F, there would not be a significant loss of ice from sublimation. If the maximum ice bed temperature is > 27'F at any time or if the doors are not closed and restored to OPERABLE status within 14 days, the situation reverts to ' Condition C and a Completion Time of 48 hours is allowed to , restore the inoperable door to OPERABLE status or enter into 1 Required Actions D.1 and D.2. Ice bed temperature must be verified to be within the specified Frequency as augmented by the provisions of SR 3.0.2. L.1 If Required Actions B.1 or B.2 are not met, the doors must be restored to OPERABLE status ano c.losed positions within 48 hours. The 48 hour Completion Tine is based on the fact that, with the very large mass of ice involved, it would not be possible for the temperature to increase to the melting D (continued) l Catawba Unit 1 B 3.6-86 Supplement 1

1

Ice Condenser Doors B 3.6.13 O O BASES l l l ACTIONS f.d (continued) point and a significant amount of ice to melt in a 48 hour period. l D.1 and D.2 If the ice condenser doors cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Comp 1t: tion Times are reasonable, based on operating experience, to reach the required plant conditions from full ; power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.13.1 REQUIREMENTS p- Verifying, by means of the Inlet Door Position Monitoring System, that the inlet doors are in their closed positions makes the operator aware of an inadvertent opening of one or more doors. The Frequency of 12 hours ensures that operators on each shift are aware of the status of the doors. SR 3.6.13.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris providet assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed. In determining if a door is impaired ty ice, the frost accumulation on the doors, joints, and hingas are to be considered in conjunction with the lifting force. limits of SR 3.6.13.7. The Frequency of 7 days is based on engineering judgment and takes into consideration such factors as the frequency of entry into the intermediate ice condenser deck, the time required for significant frost buildup, and the probability that a DBA will occur. (continued) Catawba Unit 1 B 3.6-87 Supplement 1 l

Ice' Condenser Doors B 3.6.13 BASES SURVEILLANCE SR 3.6.13.3 REQUIREMENTS 1 (continued) Verifying, by visual inspection, tha the ice condenser I inlet doors are not impaired by ice, frost, or debris l provides assurance that the doors are free to open in the l event of a DBA. For this unit, the Frequency of 18 months- 1 is based on door design, which does not allow water I condensation to freeze, and operating experience, which I indicates that the inlet doors very rarely fail to meet their SR acceptance criteria. Because of high radiation in the vicinity of the inlet doors during power operation, this Surveillance is normally performed during a shutdown. SR 3.6.13.4 Verifying the opening torque of the inlet doors provides assurance that no doors have become stuck in the closed - position. The value of 675'in-1b is based on the design opening pressure on the doors of 1.0 lb/ft 2

                                                                   . For this unit, the Frequency of 18 months is based on the passive nature of      J the closing mechanism (i.e., once adjusted, there are no          -

known factors that would change tN setting, except possibly O a buildup of ice; ice buildup ir C likely, however, because of the door design, whfch does not allow water condensation to freeze). Operating experience indicates that the inlet doors usually meet their SR acceptance ) criteria. Secause of high radiation in the vicinity of the inlet doors during power operation, this Surveillance is normally performed during a shutdown. SR 3.6.13.5 The torque test Surveillance ensures that the inlet doors have not developed excessive friction and that the return springs are producing a door return torque within limits. The torque test consists of the following:

1. Verify that the torque, T(OPEN), required to cause opening motion at the 40' open position is s 195 in-1b; (continued) l Catawba Unit 1 B 3.6-88 Supplement 1 ;

I i

Ice Condenser Doors B 3.6.13 O BASES

V SURVEILLANCE SR 3.6.13.5 (continued)

REQUIREMENTS

2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 40' open position is a 78 in-lb; and

3. Calculate the frictional torque, T(FRICT) = 0.5

{T(OPEN) - T(CLOSE)}, and verify that the T(FRICT) is s 40 in-lb.

l The purpose of the friction and return torque Specifications is to ensure that, in the event of a small break LOCA or SLB, all of the 24 door pairs open uniformly. This assures that, during the initial blowdown phase, the steam and water mixture entering the lower compartment does not pass through part of the ice condenser, depleting the ice there, while bypassing the ice in other bays. The Frequency of 18 months is based on the passive nature of the closing mechanism (i.e., once adjusted, there are no known factors that would change the setting, except possibly a buildup of ice; ice buildup is not likely, however, because of the door design, which does not allow water condensation to freeze). l Operating experience indicates that the inlet doors very l rarely fail to meet their SR acceptance criteria. Because l of high radiation in the vicinity of the inlet doors during l power operation, this Surveillance is normally perfonned , l during a shutdown. ' SR 3.6.13.6 Verifying the OPERA.BILITY of the intermediate deck doors provides assurance that the intermediate deck doors are free to open in the event of a DBA. The verification consists of visually inspecting the intermediate doors for structural deterioration, verifying free movement of the vent assemblies, and ascertaining free movement of each door when l lifted with the applicable force shown below: (continued) Catawba Unit 1 B 3.6-89 Supplement 1 l i

Ice Condenser Doors B 3.6.13 D [U BASES SURVEILLANCE SR 3.6.13.6 (continued) REQUIREMENTS QQE Liftina Force

a. Adjacent to crane wall < 37.4 lb

b. Paired with door adjacent to crane wall s 33.8 lb

c. Adjacent to containment wall s 31.8 lb i

d. Paired with door adjacent to containment s 31.0 lb wall The 18 month Frequency is based on the passive design of the l intermediate deck doors, the frequency of personnel entry I into the intermediate deck, and the fact that SR 3.6.13.2 confirms on a 7 day Frequency that the doors are not impaired by ice, frost, or debris, which are ways a door )

would fail the opening force test (i.e., by sticking or from increased door weight). SR 3.6.13.7 Verifying, by visual inspection, that the top deck doors are s in place and not obstructed provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA. The Frequency of 92 days is based on engineering judgment, which considered such factors as the following:

a. The relative inaccessibility and lack of traffic in the vicinity of the doors make it unlikely that a door l would be inadvertently left open, ;

i

b. Excessive air leakage would be detected by temperature monitoring in the ice condenser; and

c. The light construction of the doors would ensure that, in the event of a DBA, air and gases passing through the ice condenser would find a flow path, even if a door were obstructed.

l l i (continued) l Catawba Unit 1 B 3.6-90 Supplement i l i

Ice Condenser Doors B 3.6.13 BASES (continued) REFERENCES 1. UFSAR, Chapter 6.

2. 10 CFR 50, Appendix K.

3. 10 CFR 50.36, . Technical Specifications, (c)(2)(ii).

s l 1

l l f t i l l i l l O Catawba Unit 1 B 3.6-91 Supplement 1 l c

Divider Barrier Integrity B 3.6.14 8 3.6 CONTAINMENT SYSTEMS B 3.6.14 Divider Barrier Integrity BASES BACKGROUND The divider barrier consists of the operating deck and associated seals, personnel access doors, and equipment hatches that separate the upper and lower containment compartments. Divider barrier integrity is necessary to minimize bypassing of the ice condenser by the hot steam and air mixture released into the lower compartment during a Design Basis Accident (DBA). This ensures that most of the gases pass through the ice bed, which condenses the steam and limits pressure and temperature during the accident

                  .      transient. Limiting the pressure and tenperature reduces the release of fission product radioactivity from containment to the environment in the event of a DBA.

In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment. This allows air and steam to flow from the lower compartment into the ice condenser. The n resulting pressure increase within the ice condenser.causes t. the intermediate deck doors and the door panels at the top of the condenser to open, which allows the air to flow out of the ice condenser in'o the upper compartment. The ice condenses the steam as it enters, thus limiting the pressure and temperature buildup in containment. The divider barrier separates the upper and lower compartments and ensures that the steam is directed into the ice condenser. The ice, together with the containment spray, is adequate to absorb the initial blowdown of steam and water from a DBA as well as the additional heat loads that would enter containment over several hours following the initial blowdown. The additional heat loads would come from the residual heat in the reactor core, the hot piping and components, and the secondary system, including the steam generators. During the post blowdown period, the Air Return System (ARS) returns upper compartment air through the divider barrier to the lower compartment. This serves to equalize pressures in containment and' to continue circulating heated air and steam from the lower compartment through the ice condenser, where the heat is removed by the remaining ice. Divider barrier integrity ensures that the high energy fluids released during a DBA would be directed through the 7 (b (continued) l Catawba Unit 1 B 3.6-92 Supplement 1

Divider Barrier Integrity B 3.6.14 BASES BACKGROUND ice condenser and that the ice condenser would function as (continued) designed if called upon to act as a passive heat sink following a DBA. APPLICABLE Divider barrier integrity ensures the functioning of SAFETY ANALYSES the ice condenser to the limiting containment pressure and temperature that could be experienced following a DBA. The limiting DBAs considered relative to containment temperature and pressure are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. DBAs are assumed not to occur simultaneously or consecutively. Although the ice condenser is a passive system that requires no electrical power to perform its function the Containment Spray System, RHR Spray System, and the ARS also function to assist the ice bed in limiting pressures and temperatures. Therefore, the postulated DBAs are analyzed, with respect to containment Engineered Safety Feature (ESF) systems, assuming the loss of one ESF bus, which is the worst case O single active failure and results in the inoperability of onetraininboththeContainmentSpraySystgm,RHRSpray l System, and the ARS. Additionally, a 5.0 ft opening is conservatively assumed to exist in the divider barrier in the LOCA and SLB DBA analyses. The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. The maximum peak containment temperature results from the SLB analysis and is discussed in the Bases for i LCO 3.6.5, " Containment Air Temperature." In addition to calculating the overall peak containment i pressures, the DBA analyses include calculation of the ' transient differential pressures that occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand these local transient pressure differentials for the limiting DBAs. The divider barrier satisfies Criterion 3 of 10 CFR 50.36 (Ref.2). (continued) l Catawba Unit 1 B 3.6-93 Supplement 1 l

Divider Barrier Integrity B 3.6.14 p J BASES (continued) LC0 This LC0 establishes the minimum equipment requirements to ensure that the divider barrier performs its safety function of ensuring that bypass leakage, in the event of a DBA, does not exceed the bypass leakage assumed in the accident analysis. Included are the requirements that the personnel access doors and equipment hatches in the divider barrier are OPERABLE and closed and that the divider barrier seal is properly installed and has not degraded with time. An exception to the requirement that the doors be closed is made to allow personnel transit entry through the divider barrier. The basis of this exception is the assumption that, for personnel transit, the time during which a door is open will be short (i.e., shorter than the Completion Time of1hourforConditionA). The divider barrier functions , with the ice condenser to limit the pressure and temperature that could be expected following a DBA. l I APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the integrity of the divider barrier. Therefore, the LC0 is applicable in MODES 1, 2, 3, and 4. The probability and consequences of these events in MODES 5 and 6 are low due to the pressure and temperature limitations of these MODES. As such, divider barrier integrity is not required in these MODES. ACTIONS L1 If one or more personnel access doors or equipment hatches (other than the pressurizer enclosure hatch) are inoperable ; or open, except for personnel transit entry,1 hour is ; allowed to restore the door (s) and equipment hatches to OPERABLE status and the closed position. The 1 hour Completion Time is consistent with LC0 3.6.1, " Containment," which requires that containment be restored to OPERABLE status within 1 hour. Condition A has been modified by a Note to provide clarification that, for this LCO, separate Condition entry is allowed for each personnel access door or equipment hatch. (continued) l Catawba Unit 1 8 3.6-94 Supplement 1

Divid2r Barrier Integrity B 3.6.14 ! BASES ACTIONS B.d (continued) If the pressurizer enclosure hatch is inoperable or open, 6 hours are allowed to restore the hatch to OPERABLE status and the closed position. The 6 hour completion time is based on the need to perform inspections in the pressurizer compartment during power operatipn and analysis performed that shows an open hatch (7.5 ft bypass area) during a DBA does not impact the design pressure or temperature of the containment. C.d If the divider barrier seal is inoperable, I hour is allowed to restore the seal to OPERABLE status. The 1 hour Completion Time is consistent with LCO 3.6.1, which requires that containment be restored to OPERABLE status within 1 hour. D.1 and D.2 O . If divider barrier integrity cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.14.1 REQUIREMENTS Verification, by visual inspection, that all personnel access doors and equipment hatches between the upper and lower containment compartments are closed provides assurance that divider barrier integrity is maintained prior to the reactor being taken from MODE 5 to MODE 4. This SR is necessary because many of the doors and hatches may have ' been opened for maintenance during the shutdown. (continued) O Catawba Unit 1 B 3.6-95 Supplement 1 l l

Divider Barrier Integrity B 3.6.14 BASES SURVEILLANCE SR 3.6.14.2 REQUIREMENTS (continued) Verification, by visual inspection, that the personnel access door and equipment hatch seals, sealing surfaces, and alignments are acceptable provides assurance that divider barrier integrity is maintained. This inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.14.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for long periods of time. Those that use resilient materials in the seals must be opened and inspected at least once every 10 years to provide assurance that the seal material has not aged to the point of degraded performance. The Frequency of 10 years is based on the known resiliency of the materials used for seals, the fact that the openings have not been opened (to cause wear), and operating experience that confirms that the seals inspected at this Frequency have been found to be acceptable. I SR 3.6.14.3 Verification, by visual inspection, after each opening of a O U personnel access door or equipment hatch that it has been closed makes the operator aware of the importance of closing it and thereby provides additional assurance that divider barrier integrity is maintained while in applicable MODES. SR 3.6.14.4 Conducting periodic physical property tests on divider barrier seal test coupons provides assurance that the seal material has not degraded in the containment environment, including the effects of irradiation with the reactor at power. The required tests include a tensile strength test. The Frequency of 18 months was developed considering such factors as the known resiliency of the seal material used, i' the inaccessibility of the seals and absence of traffic in their vicinity, and the unit conditions needed to perform the SR. Operating experience hs shown that these , components usually pass the Steveillance when performed at i the 18 month Frequency. Therofore, the Frequency was ; concluded to be acceptable from a reliability standpoint. ! n (continued) ! U l Catawba Unit 1 B 3.6-96 Supplement 1

Divider Barrier Integrity B 3.6.14 BASES SURVEILLANCE SR 3.6.14.5 i REQUIREMENTS (continued) Visual inspection of the seal around the perimeter provides assurance that the seal is properly secured in place. The 1 Frequency of 18 months was developed considering such factors as the inaccessibility of the seals and absence of traffic in their vicinity, the strength of the bolts and ! mechanisms used to secure the seal, and the unit conditions needed to perform the SR. Operating experience has shown that these com)onents usually pass the Surveillance when performed at tie 18 month Frequency. .Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(11).

l l l 1 i O Catawba Unit 1 B 3.6-97 Supplement 1 l

Containment Recirculation Drains B 3.6.15 O B 3.6 CONTAINMENT SYSTEMS U-B 3.6.15 Containment Recirculation Drains BASES BACKGROUND The containment recirculation drains consist of the ice condenser drains and the refueling canal drains. The ice condenser is partitioned into 24 bays, each having a pair of inlet doors that open from the bottom plenum to allow the hot steam-air mixture from a Design Basis Accident (DBA) to enter the ice condenser. Twenty of the 24 bays have an ice condenser floor drain at the bottom to drain the melted ice into the lower compartment (in the 4 bays that do not have 1 drains, the water drains through the floor drains in the ) adjacent bays). Each drain leads to a drain pipe that. drops down several feet, then makes one or more 90* bends and exits into the lower compartment. A check (flapper) valve at the end of each pipe keeps warm air from entering during normal operation, but when the water exerts pressure, it i opens to allow the water to spill into the lower i compartment. This prevents water from backing up and interfering with the ice condenser inlet doors. The water l, delivered to the lower containment serves to cool the atmosphere as it falls through to the floor and provides a * ]<

 \                       source of borated water at the containment sump for long           i term use by the Emergency Core Cooling System (ECCS) and the        !

Containment Spray System durbg the recirculation mode of operation. The refueling canal drains are at low points in the refueling canal. During a refueling, valves are closed in , the drains and the canal is flooded to facilitate the ! refueling process. The water acts to shield and cool the ! spent fuel as it is transferred from the reactor vessel to i storage. After refueling, the canal is drained and the valves are locked open. In the event of a DBA, the refueling canal drains are the main return path to the lower compartment for Containment Spray System water sprayed into , the upper compartment. l The ice condenser drains and the refueling canal drains function with the ice bed, the Containment Spray System, and the ECCS to limit the pressure and temperature that could be expected following a DBA. (continued) l Catawba Unit 1 B 3.6-98 Supplement 1

Containment Recirculation Drains B 3.6.15 O BASES (continued) V APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES temperature and pressure are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. DBAs are assumed not to occur simultaneously or consecutively. Although the ice condenser is a passive system that requires no electrical power to perform its function, the Containment Spray System and the Air Return System (ARS) also function to assist the ice bed in limiting pressures and temperatures. Therefore, the analysis of the postulated DBAs, with respect to Engineered Safety Feature (ESF) systems, assumes the loss of one ESF bus, which is the worst case single active failure and results in one train of the Containment Spray System and one train of the ARS being j rendered inoperable. ! The limiting DBA analyses (Ref.1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. The maximum peak containment atmorphere temperature results from the SLB analysis and is discussed in the Bases for LCO 3.6.5, " Containment Air Temperature." t In addition to calculating the overall peak containment pressures, the DBA analyses include calculation of the transient differential pressures that occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand these local transient pressure differentials for the limiting DBAs. The containment recirculation drains satisfy Criterion 3 of 10 CFR 50.36 (Ref. 2). LC0 This LC0 establishes the minimum requirements co ensure that the containment recirculation drains perform their safety j functions. The ice condenser floor drain valve disks must be closed to minimize air leakage into and out of the ice condenser during normal operation and must open in the event , of a DBA when water begins to drain out. The refueling ! canal drains must have their plugs removed and remain clear l to ensure the return of Containment Spray System water to ' the lower containment in the event of a DBA. The containment recirculation drains function with the ice (continued) O Catawba Unit 1 8 3.6-99 Supplement 1 l

f Containment R; circulation Drains B 3.6.15 O (V BASES LC0 condenser, ECCS, and Containment Spray System to limit the (continued) pressure and temperature that could be expected following a DBA. APP:ICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature, which would require the operation of the containment recirculation drains. Therefore, the LC0 is applicable in MODES 1, 2, 3, and 4. The probability and consequences of these events in MODES 5 and 6 are low due to the pressure and temperature limitations of these MODES. As such, the containment recirculation drains are not required to be OPERABLE in i these MODES. l I ACTIONS M If one ice condenser floor drain is inoperable, I hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within j ' O. the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1,

                     " Containment," which requires that containment be restored to OPERABLE status within 1 hour.

M If one refueling canal drain is inoperable, I hour is ' allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LC0 3.6.1, which requires that containment be restored to OPERABLE status in 1 hour. C.1 and C.2 If the affected drain (s) cannot be restored to OPERABLE status within the required Completion Time, the plant must l (continued) l Catawba Unit 1 B 3.6-100 Supplement 1

Containment Recirculation Drains B 3.6.15 l BASES ACTIONS C.1 and c.2 (continued) be brought to a MODE in which the LCO does not acply. To j achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE 3R 3.6.15.1 and SR 3.6.15.2 REQUIREMENTS Verifying the OPERABILITY of the refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.15.1 confirms that the refueling canal drain valves have been locked open and that the drains are clear of any obstructions that could impair their functioning. In addition to debris near the drains, SR 3.6.15.2 requires attention be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation ar.d water O d is flowing to the drains. SR 3.6.15.1 must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the plugs have been removed and that no debris that could impair the drains was deposited during the time the canal was filled. SR 3.6.15.2 is performed every 92 days for the upper compartment and refuel canal areas. The 92 day Frequency was developed considering such factors as the inaccessibility of the drains, the i absence of traffic in the vicinity of the dr. ns, and the i i 9 'dancy of the drains. SR 3.6.15.3 Verifying the OPERABILITY of the ice condenser floor drains ; ensures that they will be able to perform their functions in i the event of a DBA. Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line following a DBA. Verifying that the drain lines ; are not obstructed ensures their readiness to drain water from the ice condenser. The 18 month Frequency was (q/ (continued) Catawba Unit 1 B 3.6-101 Supplement 1 l

Containment R: circulation Drains B 3.6.15 BASES SURVEILLANCE SR 3.6.15.3 (continued) REQUIREMENTS developed considering such factors as the inaccessibility of the drains during power operation; the design of the ice condenser, which precludes melting and refreezing of the ice; and operating experience that has confirmed that the drains are found to.be acceptable when the Surveillance is performed at an 18 month Frequency. Because of high radiation in the vicinity of the drains during power operation, this Surveillance is normally done during a shutdown. REFERENCES 1. UFSAR, Section 6.2. 2.. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). O I O l Catawba Unit 1 B 3.6-102 Supplement 1

Reactor Building B 3.6.16 B 3.6 CONTAINMENT SYSTEMS B 3.6.16 Reactor Building BASES BACKGROUND The reactor building is a concrete structure that surrounds the steel containment vessel. Between the containment vessel and the reactor building inner wall is an annular space that collects containment leakage that may occur following a loss of coolant accident (LOCA). This space also allows for periodic inspection of the outer surface of the steel containment vessel. The Annulus Ventilation System (AVS) Ettablishes a negative pressure in the annulus between the reador building and the steel containment vessel under post-accident conditions. Filters in the system then control the release of radioactive contaminants to the environment. The reactor building is required to be OPERABLE to ensure retention of containment leakage and proper operation of the AVS. APPLICABLE The design basis for reactor building OPERABILITY is a SAFETY ANALYSES LOCA. Maintaining reactor building OPERABILITY ensures that the release of radioactive material from the containment atmosphere is restricted to those leakage paths and associated leakage rates assumed in the accident analyses. The reactor building satisfies Criterion 3 of 10 CFR 50.36 , (Ref. 1). i LC0 Reactor building OPERABILITY must be maintained to ensure proper operation of the AVS and to limit radioactive leakage from the containment to those paths and leakage rates assumed in the accident analyres, i APPLICABILITY Maintaining reactor building OPERABILITY prevents leakage of l radioactive material from the reactor building. Radioactive material may enter the reactor building from the containment following a LOCA. Therefore, reactor building OPERABILITY (continued) Catawba Unit 1 B 3.6-103 Supplement 1 l

Reactor Building B 3.6.16 BASES APPLICABILITY is required in MODES 1, 2, 3, and 4 when a steam line break, (continued) LOCA, or rod ejection accident could release radioactive material to the containment atmosphere. In MODES 5 and 6 the probability and consequences of these events are low due to the Reactor Coolant System temperature and pressure limitations in these MODES. Therefore, reactor building OPERABILITY is not required in MODE 5 or 6. ACTIONS Ad In the event reactor building OPERABILITY ic not maintained, reactor. building OPERABILITY must be restored within 24 hours. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a Design Basis Accident occurring during this time period. B.1 and B.2 O If the reactor building cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.16.1 REQUIREMENTS Maintaining reactor building OPERABILITY requires maintaining each door in the access opening closed, except when the access opening is being used for normal transit entry and exit. The 31 day Frequency of this SR is based on engineering judgment and.is considered adequate in view of the other indications of door status that are available to the operator. I (continued) l Catawba Unit 1 B 3.6-104 Supplement 1

Reactor Building B 3.6.16 BASES SURVEILLANCE SR 3.6.16.2 REQUIRCMENTS (continued) The ability of a AVS train to produce the required negative pressure 2 0.5 inch water gauge during the test operation within 1 minute provides assurance that the building is adequately sealed. The negative pressure prevents leakage from the building, since outside air will be drawn in by the low pressure. The negative pressure must be established within the time limit to ensure that no significant quantity of radioactive material leaks from the reactor building prior to developing the negative pressure. The AVS trains are tested every 18 months on a STAGGERED TEST BASIS to er.sure that in addition to the requirements of LCO 3.6.10, " Annulus Ventilation System," either AVS train will perform this test. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. SR 3.6.16.3 This SR would give advance indication of gross deterioration

                       .of the concrete structural integrity of the reactor building. The 40 month Frequency is based on the requirement to perform two additional inspections at-approximately equal intervals between the Type A tests required by SR 3.6.1.1 performed on a 10-year interval. The verification is done during shutdown.

REFERENCES 1. 10 CFR 50.36 Technical Specifications, (c)(2)(ii).

                                                                                           -l O                                                                                          t Catawba Unit 1                       B 3.6-105                        Supplement 1  l l   .

CVIWS B 3.6.17 8 3.6 CONTAINMENT SYSTEMS A V B 3.6.17 Containment Valve Injection Water System (CVIWS) BASES BACKGROUND The CVIWS is required by 10 CFR 50, Appendix A, GDC 54,

                                 " Piping Systems Penetrating Containment" (Ref. 1), to ensure a water seal to a specific class of containmt isolation valves (double disc gate valves) during a LO/h, to prevent leakage of containment atmosphere through the gate valves.

Tht CVIWS is designed to inject water between the two seating surfaces of double disc gate valves used for Containment isolation. The injection pressure is higher than Containment design peak pressure during a LOCA. This will prevent leakage of the Containment atmosphere through the gate valves, thereby reducing potential offsite dose below the values specified by 10 CFR 100 limits following the postulated accident. During normal power operation, the system is in a standb, mode and does not perform any function. During accident situations the CVIWS is activated to perform its safety related function, thus limiting the release of containment atmosphere past specific containment isolation valves, in Os order to mitigate the consequences of a LOCA. Containment isolation valves, for systems which are not used to mitigate the consequences of an accident, will be supplied with CVIWS seal water upon receipt of a Phase A isolation signal. Containment isolation valves, for accident mitigating systems which are supplied with seal water from the CVIWS, have their seal water supplies actuated by a Containment Pressure - High High signal.

The system consists of two independent, redundant trains; one supplying gate valves that are powered by the A train diesel and the other supplying gate valves powered by the B train diesel. This separation of trains prevents the possibility of both containment isolation valves not sealing due to a single failure. l Each train consists of a surge chamber which is filled with water and pressurized with nitrogen. One main header exits the chamber and splits into several headers. A solenoid ' valve is located in the main header before any of the branch headers which will open after a 60 second delay on a I (continued) I l Catawba Unit 1 B 3.6-106 Supplement 1

CVIWS B 3.6.17 t BASES BACKGROUND Phase A isolation signal. Each of the headers supply (continued) injection water to containment isolation valves located in l I the same general location, and close on the same engineered I safety signal. A solenoid valve is located in each header which supplies seal water to valves closing on a Containment Pressure - High-High signal. These solenoid valves open after a 60 second delay on a Containment Pressure - High-High signal. Since'a Phase A isolation signal occurs before a Containment Pressure - High-High signal, the solenoid valve located in the main header will already be injecting water to Containment isolation valves closing on a Phase A isolation signal. This leaves an open path to the headers supplying injection water on a Containment Pressure - High- I High signal. The delay for the solenoid valves opening is to allow adequate time for the slowest gate valve to close, before water is injected into the valve seat. j Makeup water is provided from the Demineralized Water Storage Tank for testing and adding water to the surge l chamber during normal plant operation. Assured water is provided from the essential header of the Nuclear Service Water System (NSWS). This supply is assured for at least 30 g) v days following a postulated accident. If the water level in the surge chamber drops below the low-low level or if the surge chamber nitrogen pressure drops below the low-low pressure after a Phase A isolation signal, a solenoid valve in the supply line from the NSWS will automatically open and remains open, assuring makeup to the CVIWS at a pressure greater than 110% of peak Containment accident pressure. Overpressure protection is provided to relieve the pressure buildup caused by the heatup of a trapped volume of I incompressible fluid between two positively cl? sing valves (due to containment temperature transient) back into containment where an open relief path exists. APPLICABLE The CVIWS design basis is established by the consequences of SAFETY ANALYSES the limiting DBA, which is a LOCA. The accident analysis (Ref. 2) assumes that only one train of the CVIWS is functional due to a single failure that disables the other train. Makeup water can be assured from the NSWS for 30 days following a postulated LOCA. The CVIWS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 3). (continued) Catawba Unit 1 B 3.6-107 Supplement 1 l

CVIWS l B 3.6.17

 <-                                                                                                                   1

( BASES (c e tinued) LCO In the event of a DBA, one CVIWS train is required to provide the seal injection assumed in the safety analysis. Two trains of the CVIWS must be OPERABLE to ensure that at least one train will operate, assuming that the other train is disabled by a single active failure. APPLICABILITY In MODES 1, 2, 3 and 4, a DBA could require a containment isolation. The large break LOCA, on which this system's design is based, is a full power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and Reactor Coolant System > pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low. In MODES 5 and 6, the probability and consequences of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the CVIWS is not required to be OPERABLE. j . ACTIONS Ad With one CVIWS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of providing 100's of the valve injection needs after a DBA. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant CVIWS train and the low probability of a DBA occurring during this period. The Completion Time is adequate to make most repairs. B.1 and B.2 If the CVIWS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating O (continued) l Catawba Unit 1 B 3.6-108 Supplement 1

CVIWS B 3.6.17 O V BASES ACTIONS B.1 and B.2 (continued) experience, to reach the required plant conditions from full power conditions in an orderly manner and without c.hallenging plant systems. SURVEILLANCE SR 3.6.17.1 REQUIREMENTS Verifying each CVIWS train is pressurized to a 36.4 psig ensures the system can meet the design basis. Assured water is provided from the essential hender of the NSWS. The 31 day Frequency was developed in consideration of the known reliability of the system and the two train redundancy available. SR 3.6.17.2 This SR verifies that each CVIWS train can perform its required function when needed by measuring the existing n conditions for the valves being injected. Gate valves ( served by the CVIWS do not receive a conventional Type C leak rate test using air as a test medium. The containment isolation valves served by the CVIWS may be tested individually or simultaneously. Containment isolation valves are leak rate tested by this SR by injecting seal water from the CVIWS to the containment isolation valves. With the containment isolation valve closed, the leakage is determined by measuring flow rate of seal water out of the containment valve injection water surge chamber. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. Furthermore, the SR interval was developed considering that the CVIWS OPERABILITY is demonstrated at a 31 day Frequeniv by SR 3.6.17.1. l (continued) Catawba Unit 1 B 3.6-109 Supplement 1 l j

CVIWS B 3.6.17 BASES SURVEILLANCE SR 3.6.17.3 REQUIREMENTS (continued) This SR ensures that each CVIWS train responds properly to the appropriate actuation signal. The Surveillance verifies that the automatic valves actuate to their correct position. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50, Appendix A GDC 54.

2. UFSAR, Section 6.2.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

O i ) l l l l l Catawba Unit 1 B 3.6-110 Supplement 1

Containment Spray System -) B 3.6.6 ! BASES-(continued) SURVEILLANCE SR 3.6.6.3 and SR 3.6.6.4 (continued)

    . REQUIREMENTS' Frequency was concluded to be acceptable from a reliability standpoint.

The surveillance of containment sump isolation valves is also required by SR 3.6.6.3. A single surveillance may be , used to satisfy both requirements. SR 3.6.6.5 and SR 3.6.6.6 1' These SRs require verification that each containment spray pump discharge valve opens or is prevented from opening and each containment spray pump starts or is de-energized and prevented from starting upon receipt of Containment Pressure Contol System start and terminate signals. The CPCS is described in the Bases for LCO 3.3.2, "ESFAS." The 18 month Frequency is based on the need to perform these Surve111ances under the conditions that apply during a plant l outage. SR 3.6.6.7 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. The spray nozzles can also be periodically tested using a vacuum blower to induce air flow through each nozzle to verify unobstructed flow. This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded. Because of the passive design of the nozzle, a test at 10 year intervals is considered adequate to detect obstruction of the spray l nozzles. REFERENCES 1. 10 CFR 50, Appendix A. GDC 38, GDC 39, GDC 40, GDC 41, l l GDC 42, and GDC 43. ! l

2. UFSAR, Section 6.2.

3. 10 CFR 50.49.

l

4. 10 CFR 50, Appendix K.

l (continued) l Catawba Unit 2 B 3.6-44 Supplement 1 I

Containment Spray System B 3.6.6

   ' BASES (continued)
                                                                                     )

REFERENCES 5. 10 CFR 50.36, Technical Specifications, (c)(2)(11).

      .(continued)

6. ASME, Boiler. and Pressure Vessel Code, Section XI.

l I l \ i L j l O l Catawba Unit 2 B 3.6-45 Supplement 1 l

Hydrogen Recombiners B 3.6.7 h'v B 3.6 CONTAINMENT SYSTEMS B 3.6.7 Hydrogen Recombiners I BASES BACKGROUND The function of the hydrogen recombiners is to eliminate the potential breach of containment due to a hydrogen oxygen j reaction. I Per 10 CFR 50.44, " Standards for Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref.1), and GDC 41, " Containment Atmosphere Cleanup" (Ref. 2), hydrogen recombiners are required to reduce the hydrogen l concentration in the containment following a loss of coolant acciaent (LOCA). The recombiners accomplish this by recombining hydrogen and oxygen to form water vapor. The vapor remains in containment, thus eliminating any discharge to the environment. The hydrogen recombiners are manually initiated since flammable limits would not be reached until ; several days after a Design Basis Accident (DBA). l Two 100% capacity independent hydrogen recombiner systems are provided. Each consists of controls located outside pd containment in an area not exposed to the Post-Loss of l Coolant Accident environment, a power supply and a l recombiner. Recombination is accomplished by heating a hydrogen air mixture above 1150*F. The resulting water vapor and discharge gases are cooled prior to discharge from the recombiner. A single recombiner is capable of maintaining the hydrogen concentration in containment below the 4.0 volume percent (v/o) flambility limit. Two recombiners are provided to meet the requirement for i redundancy and independence. Each recombiner is powered ; from a separate Engineered Safety Features bus, and is provided with a separate power panel and control panel, c APPLICABLE The hydrogen recombiners provide for the capability of SAFETY ANALYSES controlling the bulk hydrogen concentration in containment to less than the lower flammable concentration of 4.0 v/o following a DBA. This control would prevent a containment wide hydroger burn, thus ensuring the pressure and temperature assumed in the analyses are not exceeded. The limiting DBA relative to hydrogen generation is a LOCA. f) V (continued) l Catawba Unit 2 B 3.6-46 Supplement 1

Hydrogen Recombiners , B 3.6.7

   )   BASES APPLICABLE      Hydrogen may accumulate in containment following a LOCA as a SAFETY ANALYSES result of:

(continued)

a. A metal steam reaction between the zirconium fuel rod cladding and the reactor coolant;

b. Radiolytic decomposition of water in the Reactor Coolant System (RCS) and the containment sump;

c. Hydrogen in the RCS at the time of the LOCA (i.e.,

hydrogen dissolved in the reactor coolant and hydrogen gas in the pressurizer vapor space); or

d. Corrosion of metals exposed to containment spray and Emergency Core Cooling System solutions.

To evaluate the potential for hydrogen accumulation in containment following a LOCA, the hydrogen generation as a 1 function of time following the initiation of the accident is calculated. Conservative assumptions reconmended by Reference 3 are used to maximize the amount of hydrogen calculated. . O d Based on the conservative assumptions used to calculate the hydrogen concentration versus time after a LOCA, the hydrogen concentration increases at different rates depending on the region of the containment being measured. The initiation of the Air Return System and Hydrogen Skimmer . System along with the hydrogen recombiners will maintain the I hydrogen concentration in the primary containment below l flanrnability limits. i The hydrogen recombiners are designed such tnat, wi'.h the conservatively calculated hydrogen generation rates, a single recombiner is capable of limiting the peak hydrogen concentration in containment to less than 4.0 v/o (Ref. 3). The hydrogen recombiners satisfy criterion 3 of 10 CFR 50.36 (Ref.4). (continued) Catawba Unit 2 B 3.6-47 Supplement 1 l

1 l Hydrogen Recombiners B 3.6.7 BASES (continued) i LCO Two hydrogen recombiners must be OPERABLE. This ensures operation of at least one hydrogen recombiner in the event of a worst case single active failure. Operation with at least one hydrogen recombiner ensures that the post LOCA hydrogen concentration can be prevented from exceeding the flannability limit. APPLICABILITY In MODES 1 and 2, two hydrogen recombiners are required to i control the hydrogen concentration within containment below its flannability limit of 4.0 v/o following a LOCA, assuming a worst case single failure. In MODES 3 and 4, both the hydrogen production rate and the ; total hydrogen produced after a LOCA would be less than that 1 calculated for the DBA LOCA. Also, because of the limited time in these MODES, the probability of an accident I requiring the hydrogen recombiners is low. Therefore, the ] hydrogen recombiners are not required in MODE 3 or 4. j 1 p., In MODES 5 and 6, the probability and consequences of a LOCA Q are low, due to the pressure and temperature limitations in these MODES. Therefore, hydrogen recombiners are not required in these MODES. ACTIONS L.1 With one containment hydrogen recombiner inoperable, the inoperable recombiner must be restored to OPERABLE status within 30 days. In this condition, the remaining OPERARLE hydrogen recombiner is adequate to perform the hydrogen i control function. However, the overall reliability is 1 reduced because a single failure in the OPERABLE recombiner 1 could result in reduced hydrogen control capability. The 30 day. Completion Time is based on the availability of the : other hydrogen recombiner, the small probability of a LOCA occurring (that would generate an amount of hydrogen that exceeds the flaninability limit), and the amount of time available after a LOCA (should one occur) for operator action te prevent hydrogen accumulation from exceeding the flammability limit. , (continued) l l Catawba Unit 2 B 3.6-48 Supplement 1

[: Hydrogen Recombiners B 3.6.7. I , D, BASES L I

    . ACTIONS        M (continued)

Required Action A.1 has been modified by a Note that states ~ the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when one recombiner is-inoperable._ This allowance is based on the availability of the other hydrogen recombiner, the small- probability of a 4 LOCA occurring (that would generate an amount cf hydrogen ! that exceeds the flammability limit), and the amount of time available after a LOCA (should one occur) for operator L action to prevent hydrogen accumulation from exceeding the flannability limit. i l M 1 l 1 If the inoperable hydrogen recombiner(s) cannot be restored l to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not l ' apply. To achieve this' status, the plant must be brought to l at least MODE 3 within 6 hours. The Completion Time of l 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner - and without challenging plant systems. ( SURVEILLANCE $R 3.6.7.1 , l REQUIREMENTS , Performance of a system functional test for each hyd, ogen l recombiner ensures the recombiners are operational and can attain and sustain the temperature necessary for hydragen recombination. In particular, this SR verifies that the minimum heater sheath temperature increases to a 700'F in s 90 minutes. After reaching 700*F, the power is increased to maximum power for approximately 2 minutes and power is l- verified to be a 60 kW. l Industry operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. (continued) Catawba Unit 2 B 3.6-49 Supplement 1 !

1 Hydrogen Recombiners B 3.6.7 BASES SURVEILLANCE SR 3.6.7.2 REQUIREMENTS (continued) This SR ensures there are no physical problems that could affect recombiner operation. Since the recombiners are mechanically passive, they are not subject to mechanical failure. The only credible failure involves loose wiring or structural connections, deposits of foreign materials, etc. A visual inspection is sufficient to determine abnormal conditions that could cause such failures. The 18 month Frequency for this SR was developed considering the incidence of hydrogen recombiners failing the SR 1.1 the past is low. SR 3.6.7.3 : This SR requires performance of a resistance to ground test for each heater phase to ensure that there are no detectable grounds in any heater phase. This SR should be performed following SR 3.6.7.1. This is accomplished by verifying that the resistance to ground for any heater phase is

t 10,000 ohms.

U The 18 month Frequency for this Surveillance was developed considering the incidence of hydrogen recombiners failing the SR in the past is low. i REFERENCES 1. 10 CFR 50.44. j

2. 10 CFR 50, Appendix A GDC 41.

3. UFSAR Section 6.2.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

O l Catawbr Unit 2 B 3.6-50 Supplement 1

HSS-B 3.6.8 4

~T B 3.6 CONTAINMENT SYSTEMS (G  B 3.6.8 Hydrogen Skinsner System (HSS)

BASES BACKGROUND The HSS reduces the potential for breach of containment due to a hydrogen oxygen reaction by providing a uniformly mixed post accident containment atmosphere, thereby minimizing the potential for local hydrogen burns due to a pocket of hydrogen above the flammable concentration. Maintaining a uniformly mixed containment atmosphere also ensures that the hydrogen monitors will give an accurate measure of the bulk hydrogen concentration and give the operator the capability of preventing the occurrence of a bulk hydrogen burn inside containment per 10 CFR 50.44, " Standards for Combustible Gas Control Systems in Light-Lter-Cooled Reactors" (Ref.1), and 10 CFR 50, GDC 41, " Containment Atmosphere Cleanup" (Ref. 2). ThepostaccidentHSSisanEngineeredSafetyFeature(ESF) and is designed to withstand a loss of coolant accident (LOCA) without loss of function. The System has two independent trains, each consisting of to fans with their O own motors and controls. Each train is sized for 4260 cfm. There is a normally closed, motor-operated valve on the hydrogen skinrner suction line to prevent ice condenser bypass during initial blowdown. The two trains are initiated automatically on i *ainment pressure high-high signal. The automatic actio, is to open the motor-operated valve on the hydrogen skinener suction line after a 9

1 minute delay. Once the valve has fully opened, the hydrogen ;

skimmer fan will start. Each train is powered from a i separate emergency power supply. Since each train fan can l provide 100% of the mixing requirements, the System will ! provide its design function with a limiting single active i failure. ' Air is drawn from the dead ended compartments by the mixing fans and is discharged toward the upper regions of the containment. This complements the air patterns established by the containment air return fans, which take suction from the operating floor level and discharge to the lower regions of the containment, and the containment spray, which cools the air and causes it to drop to lower elevations. The systems work together such that potentially stagnant areas where hydrogen pockets could develop are eliminated. (continued) Catawba Unit 2 B 3.6-51 Supplement 1 l

HSS B 3.6.8 BASES (continued) APPLICABLE The HSS provides the capability for reducing the local SAFETY ANALYSES hydrogen concentration to approximately the bulk average concentration. The limiting DBA relative to hydrogen concentration is a LOCA. Hydrogen may accumulate in containment following a LOCA as a result of:

a. A metal steam reaction between the zirconium fuel rod cladding and the reactor coolant;

b. Radioiytic decomposition of water in the Reactor Coolant System (RCS) and the containment sump;

c. Hydrogen in the RCS at the time of the LOCA (,e o hydrogen d.ssolved in the reactor coolant and hydrogen gas in the pressurizer vapor space); or

d. Corrosion of metals exposed to containment spray and Emergency Core Cooling System solutions.

To evaluate the potential for hydrogen accumulation in g containment following a LOCA, the hydrogen generation.as a function of time following the initiation of the accident is calculated. Conservative assumptions recommended by , Reference 3 are used to maximize the amount of hydrogen l calculated. l 1 The HSS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4). ] LCO Two HSS trains must be OPERABLE, with power to each from an independent, safety related power supply. Each train consists of one fan with its own motor and controls and is automatically initiated by a containment pressure high-high signal. 1 Operation with at least one HSS train provides the mixing necessary to ensure uniform hydrogen concentration throughout containment. ! l (continued) l Catawba Unit 2 B 3.6-52 Supplement I l

HSS I B 3.6.8 i BASES (continued) APPLICABILITY In MODES 1 and 2, the two HSS trains ensure the capability 1 to prevent localized hydrogen concentrations above the ! flamability limit of 4.0 volume percent in containment I assuming a worst case single active failure. l 1 In MODE 3 or 4, both the hydrogen production rate and the total hydrogen produced after a LOCA would be less than that calculated for the DBA LOCA. Also, because of the limited , time in these MODES, the probability of an accident J requiring the HSS is low. Therefore, the HSS is not required in MODE 3 or 4. In MODES 5 and 6, the probability and consequences of a LOCA or steam line break (SLB) are reduced due to the pressure and temperature limitations in these MODES. Therefore, the HSS is not required in these MODES. I ACTIONS Ad With one HSS train inoperable, the inoperable train must be restored to OPERABLE status within 30 days. In this O Condition, the remaining OPERABLE HSS train is adequate to perform the hydrogen mixing function. However, the overall reliability is reduced because a single failure in the OPERABLE train could result in reduced hydrogen mixing capability. The 30 day Completion Time is based on the availability of the other HSS train, the small probability of a LOCA or SLB occurring (that would generate an amount of i hydrogen that exceeds the flamability limit), the amoun6 of ' time available after a LOCA or SLB (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flamability limit, and the availability of the hydrogen recombiners and hydrogen ignitors. Required Action A.1 has been modified by a Note that states the provisions of LC0 3.0.4 are not applicable. As a result, a MODE change is allowed when one HSS train is inoperable. This allowance is based on'the availability of the other HSS train, the small probability of a LOCA or SLB i occurring (that would generate an amount of hydrogen that exceeds the flamability limit), and the amount of time available after a LOCA or SLB (should one occur) for operator action to prevent hydrogen accumulation from exceeding the flammability limit. (continued) ( Catawba Unit 2 B 3.6-53 Supplement 1 l

HSS

                 ,                                                            B 3.6.8 DD    BASES ACTIONS        H.d (continued)

If an inoperable HSS train'cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least l MODE 3 within 6 hours. The allowed Completion Time of. 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.8.1-l REQUIREMENTS L Operating each HSS train for a 15 minutes ensures that each i train is OPERABLE and that all associated controls are functioning. properly. It also ensures that blockage, fan and/or motor failure, or excessive vibration can be detected j for corrective action. The 92 day Frequency is consistent l with Inservice Testing Program Surveillance Frequencies. l operating experience, the known reliability of the. fan motors and controls, and the two train redundancy available. ,O SR 3.6.8.2

                     -Verifying HSS fan motor current at rated speed with the i                      motor operated suction valves closed is indicative of l                      overall fan motor performance and system flow. Such

! inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating l abnormal performance. : The Frequency of 92 days was based on L operating experience which has shown this Frequency to be acceptable. SR 3.6.8.3 This SR verifies the operation of the motor operated suction j valves and HSS fans in response to a start permissive from ' the Containment Pressure Control System (CPCS). The CPCS is described in the Bases for LC0 3.3.2, "ESFAS.' The 1 (continued) l Catawba Unit 2 B 3.6-54 Supplement 1 I I

i j HSS l B 3.6.8 BASES SURVEILLANCE SR 3.6.8.3 (continued) REQUIREMENTS Frequency of 92 days was based on operating experience which has shown this Frequency to be acceptable. SR 3.6.8.4 This SR ensures that each HSS train responds properly to a i containment pressure high-high actuation signal. The Surveillance verifies that each fan starts after a delay of a 8 minutes and s 10 minutes. The Frequency of 92 days conforms with the testing requirements for similar ESF equipment and considers the known reliability of fan motors l and controls and the two train redundancy available, i Therefore, the Frequency was concluded to be accer' sie from j a reliability standpoint. 1 REFERENCES 1. 10 CFR 50.44. i

2. 10 CFR 50, Appendix A. GDC 41.

O lD

3. Regulatory Guide 1,7, Revision 1.

i l 4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). l t O l Catawba Unit 2 B 3.6-55 Supplement 1 l l l_

HIS B 3.6.9 B 3.6 CONTAINMENT SYSTEMS B 3.6.9 Hydrogen Ignition System (HIS) BASES BACKGROUND The HIS reduces the potential for breach of primary containment due to a hydrogen oxygen reaction in post accident environments. The HIS is required by 10 CFR 50.44,

                       " Standards for Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref. 1), and Appendix A, GDC 41, " Containment Atmosphere Cleanup" (Ref. 2), to reduce the hydrogen concentration in the primary containment following a degraded core accident. The HIS must be capable     i of handling an amount of hydrogen equivalent to that             {

generated from a metal water reaction involving 75% of the fuel cladding surrounding the active fuel region (excluding , the plenum volume). 10 CFR 50.44 (Ref.1) requires units with ice conW1ser containments to install suitable hydrogen contr * .jstems , that would acconinodate an amount of hydrogen equivalent to l that generated from the reaction of 75% of the fuel cladding I g) g v with water. The HIS provides this required capability. This requirement was placed on ice condenser units because of their small containment volume and low design pressure (compared with pressurized water reactor dry containments). Calculations indicate that if hydrogen equivalent to that generated from the reaction of 75% of the fuel cladding with water were to collect in the primary containment, the resulting hydrogen concentration would be far above the lower flammability limit such that, if ignited from a random ignition source, the resulting hydrogen burn would seriously challenge the containment and safety systems in the containment. The HIS is based on the concept of controlled ignition using ! thermal ignitors, designed to be capable of functioning in a l post accident environment, seismically supported, and capable of actuation from the control room. A total of 70 ignitors are distributed throughout the various regions j of containment in which hydrogen could be released or to which it could flow in significant quantities. The ignitors are arranged in two independent trains such that each containment region has at least two ignitors, one from each train, controlled and powered redundantly so that ignition (continued) l Catawba Unit 2 B 3.6-56 Supplement 1

HIS-B 3.6.9 BASES' BACKGROUND would occur in each region even if one train failed to (continued) energize. f When the HIS is initiated, the ignitor elements are energized and heat up to a surface temperature a 1700'F. At

,                 this temperature, they ignite the hydrogen gas that is                l present in the airspace in the vicinity of the ignitor. The         l HIS depends on the dispersed location of the ignitors so that local pockets of hydrogen at increased concentrations would burn before reaching a hydrogen concentration significantly higher than the lower flammability limit.

Hydrogen ignition in the vicinity of the ignitors is assumed to occur when the local hydrogen concentration reaches-8.5 volume percent (v/o) and results in 100% of the hydrogen present being consumed. APPLICABLE The HIS causes hydrogen in containment to burn in a , SAFETY ANALYSES controlled manner as it accumulates following a degraded l core accident (Ref. 3). Burning occurs at the lower ! flamability concentration, where the resulting temperatures ! and pressures are rela.tively benign. Without the system, ! f_ N

           .      hydrogen could build up to higher concentrations that could result in a violent reaction if ignited by a random ignition j

source after such a buildup. The hydrogen ignitors are not included for mitigation of a Design Basis Accident (DBA) because an amount of hydrogen , equivalent to that generated from the reaction of 75% of the ; fuel cladding with water is far in excess of the hydrogen' j calculated for the limiting DBA loss of coolant accident ' (LOCA). The hydrogen concentration resulting from a DBA can be maintained less than the flammability limit using the hydrogen recombiners. The hydrogen ignitors, however, have been shown by probabilistic risk analysis to be a significant contributor to limiting the severity of accident sequences that are comonly found to dominate risk for units with ice condenser containments. As such, the hydrogen ignitors satisfy Criterion 4 of 10 CFR 50.36 (Ref. 4). LCO Two HIS trains must be OPERABLE with power from two independent, safety related power supplies. (continued) Catawba Unit 2 B 3.6-57 Supplement 1 l

HIS 8 3.6.9 BASES LCO For this unit, an OPERABLE HIS train consists of 34 of 35 (continued) ignitors energized on the train. Operation with at least one HIS train ensures that the hydrogen in containment can'be burned in a controlled manner. Unavailability of both HIS trains could lead to hydrogen buildup to higher concentrations, which could result in a violent reaction if ignited. The reaction could take place fast enough to lead to high temperatures and overpressurization of containment and, as a result, breach containment or cause containment leakage rates above those assumed in the safety analyses. Damage to safety related equipment located in containment could also occur. APPLICABILITY Requiring OPERABILITY in MODES 1 and 2 for the HIS ensures its imediate availability after safety injection and scram actuated on a LOCA. initiation. In the post accident environment, the two HIS subsystems are required to control the hydrogen concentration within containment to near its flamability limit of 4.0 v/o assuming a worst case single failure. This prevents overpressurization of containment O and damage to safety related equipment and instruments

                     -located within containment.

In MODES 3 and 4, both the hydrogen production rate and the total hydrogen production after a LOCA would be significantly less than that calculated for the DBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the HIS is low. Therefore, the HIS is not required in MODES 3 and 4. In MODES 5 and 6, the probability and consequences of a LOCA are reduced due to the pressure and temperature limitations of these MODES. Therefore, the HIS is not required to be OPERABLE in MODES 5 and 6. ACTIONS A.1 and A.2 With one HIS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days or the OPERABLE train must be verified OPERABLE frequently by performance of SR 3.6.9.1. The 7_ day Completion Time is based on the low (continued) l Catawba Unit 2 B 3.6-58 Supplement 1 l

{ HIS B 3.6.9 l (] BASES G' ! ACTIONS A.1 and A.2 (continued) '

probability of the occurrence of a degraded core event that l l would generate hydrogen in amounts equivalent to a metal water reaction of 75% of the core cladding, the length of time after the event that operator action would be required to prevent hydrogen accumulation from exceeding this limit, I and the low probability of failure of the OPERABLE HIS train. Alternative Required Action A.2, by frequent surveillances, provides assurance that the OPERABLE trairi , continues to be OPERABLE. 1 IL1 l Condition B is one containment region with no OPERABLE hydrogen ignitor. Thus, while in Condition B, or in Conditions A and B simultaneously, there would always be ! ignition capability in the adjacent containment regions that I would provide redundant capability by flame propagation to l the region with no OPERABLE ignitors. Required Action B.1 calls for the restoration of one - hydrogen ignitor in each .egion to OPERABLE status within L 7 days. The 7 day Completion Time is based on the same reasons given under Required Action A.1. fu.1 . The unit must be placed in a MODE in which the LC0 does not I apply if the HIS subsystem (s) cannot be restored to OPERABLE status within the associated Completion Time. This is done by placing the unit in at least MODE 3 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power l conditions in an orderly manner and without challenging l plant systems. l SURVEILLANCE SR 3.6.9.1 i REQUIREMENTS This SR confirms that 2 34 of 35 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements. Therefore, energizing provides [ (continued) l

  \                                                                                     i Catawba Unit 2                    B 3.6-59                         Supplement 1  l l

l I HIS B 3.6.9 BASES SURVEILLANCE SR 3.6.9.1 (continued) I REQUIREMENTS assurance of OPERABILITY. The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise l redundancy in that region, the containment regions are interconnected so that ignition in one region would cause burning to progress to the others (i.e., there is overlap in each hydrogen ignitor's effectiveness between regions). The Frequency of 92 days has been shown to be acceptable through operating experience. 1 SR 3.6.9.2 This SR confirms that the two inoperable hydrogen ignitors ; allowed by SR 3.6.9.1 (i.e., one in each train) are not in l the same containment region. The Frequency of 92 days is acceptable based on the Frequency of SR 3.6.9.1, which provides the information for performing this SR. l SR 3.6.9.3 A more detailed functional test is performed every 18 months to verify system OPERABILITY. Each glow plug is visually examined to ensure that it is clean and that the electrical circuitry is energized. All ignitors (glow plugs), including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized. Additionally, the surface temperature of each glow plug is j measured to be it 1700*F to demonstrate that a temperature i sufficient for ignition is achieved. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. () (continued) l Catawba Unit 2 B 3.6-60 Supplement 1

I HIS B 3.6.9 l BASES (continued) REFERENCES 1. 10 CFR 50.44.

2. 10 CFR 50, Appendix A, GDC 41..

l 3. UFSAR, Section 6.2.

4. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

l i l O l Catawba Unit 2 B 3.6-61 Supplement 1 l

AVS B 3.6.10 p) B 3.6~ CONTAINMENT SYSTEMS B 3.6.10 Annulus Ventilation System (AVS) BASES BACKGROUND The AVS is requited by 10 CFR 50, Appendix A, GDC 41,

                        " Containment Atmosphere Cleanup" (Ref. 1), to ensure that radioactive materials that leak from the primary containment into the reactor building (secondary containment) following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.

The containment has a secondary containment called the reactor building, which is a concrete structure that surrounds the steel primary containment vessel. Between the containment vessel and the reactor building inner wall is an annulus that collects any containment leakage that may occur following a loss of coolant acciccnt (LOCA) or rod ejection accident. This space also allows for periodic inspection of tne outer surface of the steel containment vessel. The AVS establishes a negative pressure in the annulus between the reactor bu.ilding and the steel containment A vessel. Filters in the system then control the release of .U radioactive contaminants to the environment. Reactor building OPERABILITY is required to ensure retention of primary containment leakage and proper operation of the AVS. The AVS consists of two separate and redundant trains. Each train includes a heater, a prefilter/ moisture separator, upstream and downstream high efficiency particulate air (HEPA) filters, an activated charcoal adsorber section for removal of radioiodines, and a fan. Ductwork, valves and/or dampers, and instrumentation also form part of the system. The moisture separators function to reduce the moisture l content of the airstream. A second bank of HEPA filters ' follows the adsorber section to collect carbon fines and provide backup in case of failure of the main HEPA filter bank. Only the upstream HEPA filter and the charcoal i adsorber section are credited in the analysis. The system initiates and maintains a negative air pressure in the reactor building annulus.by means of filtered exhaust ventilation of the reactor building annulus following receipt of a safety injection (SI) signal. The system is , described in Reference 2. l (continued) l Catawba Unit 2 B 3.6-62 Supplement 1

AVS B 3.6.10 l O BASES { BACKGROUND The pref 11ters remove large particles in the air, and the (continued) moisture separators remove entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal absorbers. Heaters are included to reduce the relative humidity of the airstream. Continuous operation of each train, for at least 10 hours per month, with heaters . on, reduces moisture buildup on their HEPA filters and adsorbers. The AVS reduces the radioactive content in the annulus atmosphere following a DBA. Loss of the AVS could cause

                   -site boundary doses, in the event of a DBA, to exceed the values given in the licensing basis.

1 APPLICABLE The AVS design basis is established by the consequences of ) SAFETY ANALYSES the limiting DBA, which is a LOCA. The accident analysis 4 (Ref. 3) assumes that only one train of the AVS is functional due to a single failure that disables the other train. The accident analysis accounts for the reduction in airborne radioactive material provided by the remaining one ), train of this filtration system. The 'unt of fission l products available for release from con 6ainment is ( determined for a LOCA. The modeled AVS actuation in the safety analyses is based upon a worst case response time following an SI initiated at the limiting setpoint. The total response time, from exceeding the signal setpoint to attaining the negative pressure of 0.5 inch water gauge in the reactor building I annulus, is 1 minute. This response time is composed of i signal delay, diesel generator startup and sequencing time, i system startup time, and time for the system to attain the required pressure after starting. l The AVS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4). LC0 In the event of a DBA, one AVS train is required to provide the minimum particulate iodine removal assumed in the safety analysis. Two trains of the AVS must be OPERABLE to ensure l that at least one train will operate, assuming that the l other train is disabled by a single active failure. (continued) Catawba Unit 2 B 3.6-63 Supplement 1 l L

AVS B 3.6.10 p BASES (continued) , V APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could lead to fission product release to containment that leaks to the reactor building. The large break LOCA, on which this system's design is ! based, is a full power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and Reactor Coolant System pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low. In MODES 5 and 6, the probability and consequences of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the AVS is not required to be OPERABLE. I ACTIONS Ad j With one AVS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days. The 7 day 1 Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant AVS train and the Q C low probability of a DBA occurring during this period. The Completion Time is adequate to make most repairs. B.1 and B.2 ) i With one or more AVS heaters inoperable, the heater must be restored to OPERABLE status within 7 days. Alternatively, a report must be initiated within 7 days per Specification 5.6.6, which details the reason for the heater's I inoperability and the corrective action required to return ! the heater to OPERABLE status. l l The heaters do not affect OPERABILITY of the AVS filter trains because charcoal adsorber efficiency testing is performed at 30*C and 95% relative humidity. The accident analysis shows that site boundary radiation doses are within 10 CFR 100 limits during a DBA LOCA under these conditions. I l (continued) l Catawba Unit 2 B 3.6-64 Supplement 1 l

l AVS B 3.6.10 BASES ACTIONS C.1 and C.2- i q < (continued) i If the AVS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.10.1 REQUIREMENTS Operating each AVS train from the control room with flow through the HEPA filters and charcoal adsorbers ensures that all trains are OPERABLE and that all associated controls are j functioning properly. It also ensures that blockage, fan or i motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for j lt 10 continuous hours eliminates moisture on the adsorbers O and HEPA filters. Experience from filter testing at operating units indicates that the 10 hour period is 1 j i adequate for moisture elimination on the adsorbers and HEPA ' filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls, the two train redundancy available, and the iodine removal capability of the Containment Spray System and Ice Condenser. SR 3.6.10.2 This SR verifies that the required AVS filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The AVS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 5). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. (continued) Catawba Unit 2 B 3.6-65 Supplement 1 l

l AVS B 3.6.10

   /]    BASES V

SURVEILLANCE SR 3.6.10.3 , REQUIREMENTS } (continued) The automatic startup on a safety injection signal ensures that each AVS train responds properly. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating expefience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore the Frequency was concluded to be acceptable from a reliability standpoint. Furthermore, the SR interval was developed considering that the AVS equipment OPERABILITY is , demonstrated at a 3* day Frequency by SR 3.6.10.1. SR 3.6.10.4 , The AVS filter cooling electric motor-operated bypass valves are tested to verify 0PERABILITY. The valves are nonna11y closed and may need to be opened to initiate miniflow cooling through a filter unit that has been shutdown following a DBA LOCA. Miniflow cooling may be necessary to D [d limit temperature increases in the idle filter train due to decay heat from captured fission products. The 18 month Frequency is considered to be acceptable based on valve reliability and design, and the fact that operating experience has shown that the valves usually pass the Surveillance when performed at the 18 month Frequency. REFERENCES 1. 10 CFR 50, Appendix A, GDC 41,

2. UFSAR, Section 9.4.

3. UFSAR, Chapter 15.

4. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

5. Regulatory Guide 1.52, Revision 2.

l Catawba Unit 2 B 3.6-66 Supplement 1

ARS B 3.6.11 D B 3.6 CONTAINMENT SYSTEMS [Q B 3.6.11 Air Return System (ARS) BASES BACKGROUND The ARS is designed to assure the rapid return of air from the upper to the lower containment compartment after the initial blowdown following a Design Basis Accident (DBA). The return of this air to the lower compartment and subsequent recirculation back up through the ice condenser assists in cooling the containment atmosphere and limiting post accident pressure and temperature in containment to less than design values. Limiting pressure and temperature reduces the. release of fission product radioactivity from containment to the environment in the event of a DBA. The ARS also promotes hydrogen dilution by mixing the hydrogen with containment atmosphere and distributing throughout the containment. The ARS consists of two separate trains of equal capacity, each capable of meeting the design bases. Each train includes a 100% capacity air return fan, and associated motor operated damper in the fan discharge line to the containment lower compartment. The damper acts as a barrier between the upper and lower compartments to prevent reverse flow which would bypass the ice condenser. The damper is normally closed and remains closed throughout the initial blowdown following a postulated high energy line break. The damper motor is actuated several seconds after the Containment High-High pressure setpoint is reached and a start permissive from the Containment Pressure Control System is present. A backdraft damper is also provided at the discharge of each fan to serve as a check valve. Each train is powered from a separate Engineered Safety Features (ESF) bus. The ARS fans are automatically started by the containment pressure High-High signal 911 minutes after the containment pressure reaches the pressure setpoint and a start permissive from the Containment Pressure Control System is present. The time delay ensures that no energy released during the initial phase of a DBA will bypass the ice bed through the ARS fans. (continued) C Catawba Unit 2 B 3.6-67 Supplement 1 l

ARS B 3.6.11 BASES BACKGROUND After starting, the fans displace air from the upper (continued) compartment to the lower compartment, thereby returning the air that was displaced by the high energy line break blowdown from the lower compartment and equalizing pressures throughout containment. After discharge into the lower compartment, air flas with steam produced by residual heat through the ice condenser doors into the ice condenser compartment where the steam portion of the flow is condensed. The air flow returns to the upper compartment through the top deck doors in the upper portion of the ice condenser compartment. The ARS fans operate continuously after actuation, circulating air through the containment volume. When the containment pressure falls below a predetermined value, the ARS fans are automatically de-energized. Thereafter, the fans are automatically cycled on and off if necessary to control any additional containment pressure transients. The ARS also functions, after'all the ice has melted, to circulate any steam still entering the lower compartment to the upper compartment where the Containment Spray System can Cool it. The ARS is an ESF system. It is designed to ensure that the

heat removal capability required during the post accident period can be attained. The operation of the ARS, in conjunction with the ice bed, the Containment Spray System, and the Residual Heat Removal (RHR) System spray, provides the required heat removal capability to limit post accident conditions to less than the containment design values.

APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES temperature and pressure are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. DBAs are assumed not to occur simultaneously or consecutively. The postulated DBAs are analyzed, in regard to ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the Containment Spray System, RHR System, and t (continued) I l Catawba Unit 2 B 3.6-68 Supplement 1

ARS B 3.6.11 BASES APPLICABLE ARS being inoperable (P,ef. 1). The DBA analyses show that SAFETY ANALYSES the maximum peak containment pressure results from the LOCA (continued) analysis and is calculated to be less than the containment design pressure. For certain aspects of transient accident analyses, traximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures, in accordance with 10 CFR 50, Appendix K (Ref. 2). The analysis for minimum internal containment pressure (i.e., maximum external differential containment pressure) assumes inadvertent simultaneous actuation of both the ARS and the Containment Spray System. The modeled ARS actuation from the containment analysis is p based upon a response time associated with exceeding the i containment pressure High-High signal setpoint to achieving full ARS air flow. A delayed response time initiation ' provides conservative analyses of peak calculated containment temperature and pressure responses. The ARS total response time of 600 seconds includes signal delays. The ARS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 3). , 1 LCO In the event of a DBA, one train of the ARS is required to provide the minimum air recirculation for heat removal assumed in the safety analyses. To ensure this requirement ! is met, two trains of the ARS must be OPERABLE. This will ) ensure that at least one train will operate, assuming the l worst case single failure occurs, which is in the ESF power i supply. : l (continued) Catawba Unit 2 B 3.6-69 Supplement 1 l

T ARS B 3.6.11 f BASES (continued) 'V[] APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ARS. Therefore, the LC0 is applicable in MODES 1, 2, 3, and 4. In MODES 5 and 6 the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ARS is not required to te OPERABLE in these MODES. l ACTIONS Ad If one of the required trains of the ARS is inoperable, it must be restored to OPERABLE status within 72 hours. The 72 hour Completion Time was developed taking into account the redundant flow of the OPERABLE ARS train and the low probability of a DBA occurring in this period. B.1 and 8.2 O If the ARS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SB 3.6.11.1 REQUIREMENTS Verifying that each ARS fan starts on an actual or simulated actuation signal, after a delay 2 8 minutes and s 10 minutes, and operates for 215 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly. It also ensures that blockage, far, and/or motor failure, or

p (continued) g l Catawba Unit 2 B 3.6-70 Supplement 1 1

ARS B 3.6.11 BASES SURVEILLANCE SR 3.6.11.1 (continued) REQUIREMENTS excessive vibration can be detected for corrective action. The 92 day Frequency was developed considering the known reliability of fan motors and controls and the two train redundancy available. SR 3.6.11 2 Verifying ARS fan motor current to be at rated speed with the return air dampers closed confirms one operating condition of the fan. This test is indicative of overall fan motor performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of 92 days conforms with the testing requirements for similar ESF equipment and considers the knnyn i reliability of fan motors and controls and the two train I redundancy available. 1 SR 3.6.11.3 d \ - Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan J is started. This Surveillance also tests the circuitry, including time delays to ensure the system operates properly. The Frequency of 92 days was developed considering the importance of the dampers, their location, physical environment, and probability of failure. Operating experience has also shown this Frequency to be acceptable. SR 3.6.11.4 and SR 3.6.11.5 j Verifying the OPERABILITY of the check damper in the air return fan discharge line to the containment lower l compartment provides assurance th6t the proper flow path will exist when the fan is start" and that reverse flow can not occur when the fan is not 0;. Iting. The Frequency of 92 days was developed consideri < the importance of the dampers, their location, physicc.: snvironment, and probability of failure. Operating experience has also shown this Frequency to be acceptable. /N (continued) U Catawba Unit 2 B 3.6-71 Supplement 1 l

ARS B 3.6.11 p J BASES (continued). SURVEILLANCE SR 3.6.11.6 and SR 3.6.11.7 REQUIREMENTS (continued) These SRs require verification that each ARS motor operated damper opens or is prevented from opening and each ARS fan is prevented from starting upon ,eceipt of Containment Pressure Contol System start and terminate signals. The CPCS is described in the Bases for LCO 3.3.2, "ESFAS." The 18 month Frequency is based on operating experience which has shown it to be acceptable. REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50, Appendix K.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

O O l Catawba Unit 2 8 3.6-72 Supplement 1

Ice Bed B 3.6.12

B 3.6 CONTAINMENT SYSTEMS B 3.6.12 Ice Bed BASES BACKGROUND The ice bed consists of over 2,475,252 lb of ice stored in 1944 baskets within the ice condenser. Its primary purpose is to provide a large heat sink in the event of a release of energy from a Design Basis Accident (DBA) in containment.

The ice would absorb energy and limit containment peak pressure and temperature during the accident transient. Limiting the pressure and temperature reduces the release of fission product radioactivity from containment to the environment in the event of a DBA. The ice condenser is an annular compartment enclosing approximately 300* of the perimeter of the upper containment compartment, but penetrating the operating deck so that a portion extends into the lower containment compartment. The lower portion has a series of hinged doors exposed to the atmosphere of the lower containment compartment, which, for normal unit operation, are designed to remain closed. At the top of the ice condenser is another set of doors exposed O to the atmosphere of the upper compartment, which also remain closed du.-ing normal unit operation. Intermediate deck doors, located below the top deck doors, form the floor of a plenum at the opper part of the ice condenser. These doors also remain closed during normal unit operation. The upper plenum area is used to facilitate surveillance and maintenance of the ice bed. The ice baskets held in the ice bed within the ice condenser are arranged to promote heat transfer from steam to ice. This arrangement enhances the ice condenser's primary function of condensing steam and absorbing heat energy released to the containment during a DBA. In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment. This allows air and steam to flow from the lower compartment into the ice condenser. The resulting pressure increase within the ice condenser causes the intermediate deck doors and the top deck doors to open, which allows the air to flow out of the ice condenser into the upper compartment. Steam condensation within the ice condenser limits the pressure and temperature buildup in (continued) Catawba Unit 2 B 3.6-73 Supplem:..u 1 l

Ice Bed B 3.6.12 O BASES U BACKGROUND containment. A divider barrier separates the upper and (continued) lower compartments and ensures that the steam is directed into the ice condenser. 1 The ice, together with the containment spray, is adequate to absorb the initial blowdown of steam and water from a DBA and the additional heat loads that would enter containment during several hours following the initial blowdown. The additional heat loads would come from the residual heat in the reactor core, the hot piping and components, and the secondary system, including the steam generators. During i the post blowdown period, the Air Return System (ARS) ' returns upper compartment air through the divider barrier to the lower compartment. Thic serves to equalize pressures in , containment and to continue circulating heated air and steam from the lower compartment through the ire condenser where the heat is removed by the remaining ice. As ice melts, the water passes through the ice condenser floor drains into the lower compartment. Thus, a second function of the ice bed is to be a large source of borated water (via the containment sump) for long term Emergency , Core Cooling System (ECCS) and Containment Spray System heat

   )                  removal functions in the recirculation mode.

A third function of the ice bed and melted ice is to remove I fission product iodine that may be released from the core during a DBA. Iodine removal occurs during the ice melt phase of the accident and continues as the melted ice is sprayed into the containment atmosphere by the Containment Spray System. The ice is adjusted to an alkaline pH that facilitates removal of radioactive iodine from the containment atmosphere. The alkaline pH also minimizes the occurrence of the chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment Spray System fluids in the recirculation mode of operation. J It is important for the ice to be uniformly distributed around the 24 ice condenser bays and for open flow paths to exist around ice baskets. This is especially important ; during the initial blowdown so that the steam and water mixture entering the lower compartment do not pass through only part of the ice condenser, depleting the ice there while bypassing the ice in other bays. l (continued) l Catawba Unit 2 B 3.6-74 Supplement I r

Ice Bed B 3.6.12 BASES BACKGROUND Two phenomena that can degrade the ice bed during the long

    ,    (continued)   service period are:

a. Loss of ice by melting or sublimation; and

b. Obstruction of flow passages through the ice bed due to buildup of frost or ice. Both of these degrading phenomena are reduced by minimizing air leakage into and out of the ice condenser.

The ice bed limits the temperature and pressure that could be expected following a DBA, thus limiting leakage of fission product radioactivity from containment to the environment. APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES temperature and pressure are the loss of coolant accident l (LOCA) and the steam line break (SLB). The LOCA and SLB are l analyzed using computer codes designed to predict the l resultant containment pressure and temperature transients. I DBAs are not assumed to occur simultaneously or I consecutively. Although the ice condenser is a passive system that requires no electrical power to perform its function, the Containment ) Spray System RHR Spray System, and the ARS also fMetion to l assist the ice bed in limiting pressures and temperatures. Therefore, the postulated DBAs are analyzed in regards to I containment Engineered Safety Feature (ESF) systems, ; assuming the loss of one ESF bus, which is the worst case ' single active failure and results in one train each of the Containment Spray System, RHR Spray System, and ARS being inoperable. The limiting DBA analyses (Ref.1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. For certain aspects of the transient accident analyses, maximizing ths calculated containment pressure is not conservative. In particular, the cooling effectiveness of the ECCS during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For (] U (continued) Catawba Unit 2 B 3.6-75 Supplement 1 l t'

Ice Bed B 3.6.12 BASES APPLICABLE these calculations, the containment backpressure is SAFETY ANALYSES calculated in a manner designed to conservatively minimize, (continued) rather than maximize, the calculated transient containment pressures, in accordance with 10 CFR 50, Appendix K (Ref. 2). The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed in the Bases for LC0 3.6.5, " Containment Air Temperature." In addition to calculating the overall peak containment pressures, the DBA analyses include calculation of the transient differential pressures that occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand these local transient pressure differentials for the limiting DBAs. The ice bed satisfies Criterion 3 of 10 CFR 50.36 (Ref. 3). c LC0 The ice bed LC0 requires the existence of the required (' quantity of stored ice, appropriate distribution of the ice and the ice bed, open flow paths through the ice bed, and appropriate chemical content and pH of the stored ice. The stored ice functions to absorb heat during a DBA, thereby limiting containment air temperature and pressure. The chemical content and pH of the ice provide core SDr4 (boron content) and remove radioactive iodine from the containment atmosphere when the melted ice is recirculated through the ECCS and the Containment Spray System, respectively. APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed. Therefore, the LCO is applicable in MODES 1, 2, 3, and 4. In MODES 5 and 6, the probability and consequene n of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ice bed is not required to be OPERABLE in these MODES. (continued) l l Catawba Unit 2 B 3.6-76 Supplement 1 i i l l

Ice Bcd B 3.6.12 BASES (continued) ACTIONS 6.d If the ice bed is inoperable, it must be restored to OPERABLE status within 48 hours. The Completion Time was developed . based on operating experience, which confirms that due to the very large mass of stored ice, the parameters comprising OPERABILI.TY do not change appreciably in this time period. Because of this fact, the Surveillance Frequencies are long (months), except for the ice bed temperature, which is checked every 12 hours. If a degraded condition is identified, even for temperature, with such a-large mass of ice it is not possible for the degraded condition to significantly degrade further in a 48 hour period. Therefore, 48 hours is a reasonable amount of time to correct a degraded condition before initiating a shutdown. B.1 and B.2 If the ice bed cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this O- status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from. full power conditions in an r <derly manner and without challenging plant syste as. SURVEILLANCE SR 3.6.12.1 REQUIREMENTS Verifying that the maximum temperature of the ice bed is s 27'F ensures that the ice is kept well below the melting point. The 12 hour Frequency was based on operating experience, which confirmed that, due to the large mass of stored ice, it is not possible for the ice bed temperature to degrade significantly within a 12 hour period and was also based on assessing the proximity of the LC0 limit to the melting temperature. Furthermore, the 12 hour Frequency is considered adequate in view of indications in the control room, including the alarm, to alert the operator to an abnormal ice bed (continued) Catawba Unit 2 B 3.6-77 Supplement 1 l

E l-l Ice Bed B 3.6.12

l Q) BASES SURVEILLANCE SR 3.6.12.1 (continued) REQUIREMENTS temperature condition. This SR may be satisfied by use of the Ice Bed Temperature Monitoring System. SR 3.6.12.2 1 This SR ensures that the flow channels through the ice ) l condenser have not accumulated an excessive amount of ice or i frost blockage. The visual inspection must be made for two . or more flow channels per ice condenser bay and must include the following specific locations along the flow channel: i a. Past the lower inlet plenum support strJCtures and turning vanes; ! b. Between ice baskets;

c. Past lattice frames;

d. Through the intermediate floor grating; and

e. Through the top deck floor grating.

The allowable 0.38 inch thick buildup of frost or ice is based on the analysis of containment response to a DBA with partial blockage of the ice condenser flow passages. If a ! flow channel in a given bay is found to have an accumulation I of frost or ice > 0.38 inch thick, a representative sample l of 20 additional flow channels from the same bay must be visually inspected. If these additional flow channels are all found to be ! acceptable, the discrepant flow channel may be considered

single, unique, and acceptable deficiency. More than one discrepant flow channel in a bay is not acceptable, however.

These requirements are based on the sensitivity of the ! partial blockage analysis to additional blockage. The l Frequency of 9 months was based on ice storage tests and the allowance built into the required ice mass over and above the mass assumed in the safety analyses. p/ (continued) l Catawba Unit 2 8 3.6-78 Supplement 1

Ice Bed B 3.6.12 l BASES SURVEILLANCE SR 3.6.12.3 j REQUIREMENTS i (continued) Verifying the chemical composition of the stored ice ensures l that the stored ice has a boron concentration of at least 1800 ppm as sodium tetraborate and a high pH, a 9.0 and s 9.5, in order to meet the requirement for borated water l when the melted ice is used in the ECCS recirculation mode of operation. Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine. The high pH is required i to enhar.ce the effectiveness of the ice and the melted ice l in removing iodine from the containment atmosphere. This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment Spray System fluids in the recirculation mode of operation. The Frequency of 9 months ( is based on operating experience. l SR 3.6.12.4 m The weighing program is designed to obtain a representative

sample of the ice baskets. The representative sample shall include 6 baskets from each of the 24 ice condenser bays and shall consist of one basket from radial rows 1, 2, 4, 6, 8, and 9. If no bar.ket from a designated row can be obtained for weighing, a basket from the same ro.4 of an adjacent bay shall be weighed.

The rows chosen include the rows nearest the inside and outside walls of the ice condenser (rows 1 and 2, and 8 , and 9, respectively), where heat transfer into the ice condenser is most likely to influence melting or sublimation. Verifying the total weight of ice ensures that there is adequate ice. to absorb the required amount of energy to mitigate the DBAs. ! If a basket is found to contain < 1273 lb of ice, a l representative sample of 20 additional baskets from the same l bay shall be weighed. The average weight of ice in these 21 baskets (the discrepant basket and the 20 additional baskets) shall be 21273 lb at a 95% confident.e level. G (continued) O Catawba Unit 2 B 3.6-79 Supplement 1 l

T Ice Bed B 3.6.12 i BASES SURVEILLANCE SR 3.6.12.4 (continued) REQUIREMENTS Weighing 20 additional baskets from the same bay in the event a Surveillance reveals that a single basket contains

                    < 1273 lb ensures that no local zone exists that is grossly deficient-in ice. Such a zone could experience early melt         {

out during a DBA transient, creating a path for steam to pas:; through the ice bed without being condensed. The ) ! Frequency of 18 months was based on ice. storage tests and the allowance built into the required ice mass over and above the mass assumed in the safety analyses. Operating experience has' verified that, with the 18 month Frequency, the weight requirements are maintained with no significant

degradation between surveillances.

1 SR 3.6.12.5 l This SR ensures that the azimuthal distribution of ice is reasonably uniform, by verifying that the average ice weight l in each of three azimuthal groups of ice condenser bays is l within the limit. The Frequency of 18 months was based on ice storage tests and the allowance built into the required - l ice mass over and above the mass assumed in the safety i analyses. Operating experience has verified that, with the ' l 18 month Frequency, the weight requirements are maintained l with no significant degradation between surveillances. SR 3.6.12.6 This SR ensures that a representative sampling of ice . baskets, which are relatively thin walled, perforated l cylinders, have not been degraded by wear, cracks, ! corrosion, or other damage. Each ice basket must be raised ' at least 12 feet for this irspection. The Frequency of , 40 months for a visual inspection of the structural soundness of the ice baskets is based on engineering judgment and considers such factors as the thickness of the , basket walls relative to corrosion rates expected in their service environment and the results of the long term ice l storage testing. l 1 1 ( (continued) l Catawba Unit 2 B 3.6-80 Supplement 1

1 l l- Ice Bed B 3.6.12

L:

BASES (continued) 1 REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50, Appendix K i l

l

3. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

l l l I I lO l l O Catawba Unit 2 B 3.6-81 Supplement 1 l

                                                                                  .l t

Ice Condenser Doors B 3.6.13 B 3.6 CONTAINMENT SYSTEMS B 3.6.13 Ice Condenser Doors BASES BACKGROUND The ice condenser doors consist of the inlet doors, the intermediate deck doors, and the top deck doors. The functions of the doors are to:

a. Seal the ice condenser from air leakage during the lifetime of the unit; and

b. Open in the event of a Design Basis Accident (DBA) to direct the hot steam air mixture from the DBA into the ice bed, where the ice would absorb energy and limit containment peak pressure and temperature during the accident transient.

Limiting the pressure and temperature following a DBA reduces the release of fission product radioactivity from containment to the environment.

   -                      The ice condenser is an annular compartment enclosing approximately 300* of the perimeter of the upper containment compartment, but penetrating the operating deck so that a portion extends into the lower containment compartment. The inlet doors separate the atmosphere of the lower compartment from the ice bed inside the ice condenser. The top deck doors are above the ice bed and exposed to the atmosphere of the upper compartment. The intermediate deck doors, located below the top deck doors, form the floor of a plenum at the upper part of the ice condenser. This plenum area is used to facilitate surveillance and maintenance of the ice bed.

The ice baskets held in the ice bed within the ice condenser are arranged to promote heat transfer from steam to ice. This arrangement enhances the ice c]ndenser's primary function of condensing steam and absorbing heat energy released to the containment during a DBA. In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment. This allows air and steam to flow from the lower compartment into the ice condenser. The resulting pressure increase within the ice condenser causes the intermediate deck doors and the top deck doors to open, p (continued) l Catawba Unit 2 B 3.6-82 Supplement 1

Ice Condenser Doors B 3.6.13 BASES BACKGROUND which allows the air to flow out of the ice condenser into (continued) the upper compartment. Steam condensation within the ice condensers limits the pressure and temperature buildup in containment. A divider barrier separates the upper and lower compartments and ensures that the steam is directed into the ice condenser. The ice, together with the containment spray, serves as a containment heat removal system and is adequate to absorb the initial blowdown of steam and water from a DBA as well as the additional heat loads that would enter containment during the several hours following the initial blowdown. i The additional heat loads would come from the residual heat ! in the reactor core, the hot piping and components, and the- l secondary system, including the steam generators. During the post blowdown period, the Air Return System (ARS) returns upper compartment air through the divider barrier to the lower compartment. This serves to equalize pressures in containment and to continue circulating heated air and steam from the lower compartment through the ice condenser, where the heat is removed by the remaining ice. The water from the melted ice drains into the lower compartment where it serves as a source of borated water (via the containment sump) for the Emergency Core Cooling System (ECCS) and the Containment Spray System heat removal ! functions in the recirculation mode. The ice (via tue l Containment Spray System) and the recirculated ice uelt also serve to clean up the containment atmosphere. l l The ice condenser doors ensure that the ice stored in the ice bed is preserved during normal operation (doors closed) i and that the ice condenser functions as designed if called upon to act as a' passive heat sink following a DBA. APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES pressure and temperature are the loss of coolant accident (LOCA) and the steam line break (SLB). _ The LOCA and SLB are analyzed using computer codes designed to preact the resultant containment pressure and temperature transients.

                  - DBAs are assumed not to occur simultaneously or consecutively.

(continued) O Catawba Unit 2 B 3.6-83 Supplement 1 l

Ice Condenser Doors B 3.6.13 BASES

    -APPLICABLE       Although tne ice condenser is a passive. system that requires SAFETY ANALYSES no electrical power to perform its function, the Containment (continued)   Spray System and ARS also function to assist the ice bed in limiting pressures and temperatures. .Therefore, the postulated DBAs are analyzed with respect to Engineered Safety Feature (ESF) systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the Containment Spray System and the ARS being rendered inoperable.

The limiting DBA analyses (Ref.1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. For certain aspects of transient accident analyses,. maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the ECCS during the core reflood phase of a LOCA analysis , increases with increasing containment backpressure. For ! these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures, in accordance with 10 CFR 50. Appendix K j (Ref. 2). , O- The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed in the Bases for LCO 3.6.5, " Containment Air Temperature." An additional design requirement was imposed on the ice condenser door design for a small break accident in which the flow of heated air and steam is not sufficient to fully open the doors. For this situation, the doors are designed so that all of the doors would partially open by approximately the same amount. Thus, the partially opened doors would modulate the flow so that each ice bay would receive an approximately equal fraction of the total flow. This design feature ensures that the heated air and steam will not flow preferentially to some ice bays and deplete 1 the ice there without utilizing the ice in the other bays. )

In addition to calculating the overall peak containment pressures, the DBA analyses include the calculation of the transient differential pressures that would occur across , l (continued) O ] l l - Catawba Unit 2 B 3.6-84 Supplement 1 j

Ice Condenser Doors B 3.6.13 /O BASES b APPLICABLE subcompartment walls during the initial blowdown phase of SAFETY ANALYSES the accident transient. The internal containment walls and (continued) structures are designed to withstand the local transient pressure differentials for the limiting DBAs. The ice condenser doors satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3). LCO This LC0 establishes the minimum equipment requirements to assure that the ice condenser doors perform their safety function. The ice condenser inlet doors, intermediate deck doors, and top deck doors must be closed to minimize air leakage into and out of the ice condenser, with its attendant leakage of heat into the ice condenser and loss of ice through melting and sublimation. The doors must be OPERABLE to ensure the proper opening of the ice condenser in the event of a DBA. OPERABILITY includes being free of any obstructions that would limit their opening, and for the inlet doors, being adjusted such that the opening and closing torques are within limits. The ice condenser doors function with the ice condenser to limit the pressure and temperature that could be expected following a DBA. APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice condenser doors. Therefore, the LCO is applicable in MODES 1, 2, 3, and 4. The probability and consequences of these events in MODES 5 and 6 are reduced due to the pressure and temperature limitations of these MODES ~ Therefore, the ice condenser doors are not required to be OPERABLE in these MODES. ACTIONS A Note provides clarification that, for this LCO, separate Condition entry is allowed for each ice condenser door. (continued) O Catawba Unit 2 B 3.6-85 Supplement 1 l

Ice Condenser Doors B 3.6.13 BASES ACTIONS L.1 (continued) If one or more ice condenser inlet doors are inoperable due to being physically restrained from opening, the door (s) must be restored to OPERABLE status within l' hour. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1,

                   " Containment," which requires containment to be restored to OPERABLE status within 1 hour.

B.1 and B.2 if one or more ice condenser doors are determined to be partially open or otherwise inoperable for reasons other than Condition A or if a door is found that is not closed, it is acceptable to continue unit operation for up to 14 days, provided the ice bed temperature instrumentation is monitored once per 4 hours to ensure.that the open or inoperable door is not allowing enough air leakage to cause the maximum ice bed temperature to approach the melting point. The Frequency of 4 hours is based on the fact that O temperature changes cannot occur rapidly in the. ice bed because of the large mass of ice involved. The 14 day Completion Time is based on long term ice storage tests that indicate that if the temperature is' maintained below 27'F, there would not be a significant-loss of ice from sublimation. If the maximum ice bed temperature is > 27'F at any time or if the doors are not closed and restored to OPERABLE status within 14 days, the situation reverts to Condition C and a Completion Time of 48 hours is allowed to restore the inoperable door to OPERABLE status or enter into Required Actions D.1 and D.2. Ice bed temperature must be verified to be within the specified Frequency as augmented by the provisions of SR 3.0.2. fu.1 If Required Actions B.1 or B.2 are not met, the doors must be restored to OPERABLE status and closed positions within 48 hours. The 48 hour Completion Time is based on the fact that, with the very large mass of ice involved, it would not be possible for the temperature to increase to the melting (continued) l- Catawba Unit 2 B 3.6-86 Supplement 1 1

Ice Condenser Doors B 3.6.13 (l' BASES \_/ ACTIONS [.d (continued) point and a significant amoun't of ice to melt in a 48 hour period. D.1 and D.2 If the ice condenser doors cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.13.1 REQUIREMENTS Verifying, by means of the Inlet Door Position Monitoring System, that the inlet doors are in their closed positions makes the operator aware of an inadvertent opening of one or more doors. The Frequency of 12 hours ensures that operators on each shift are aware of the status of the doors. SR 3.6.13.2 Verifying, by, visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed. In determining if a door is impaired by ice, the frost accumulation on the doors, joints, and hinges are to be considered in conjunction with the lifting force limits of SR 3.6.13.7. The Frequency of 7 days is based on engineering judgment and takes into consideration such factors as the frequency of entry into the intermediate ice condenser deck, the time required for significant frost buildup, and the probability that a DBA will occur. (continued) k Catawba Unit 2 B 3.6-87 Supplement 1 l l

Ice Condenser D: ors B 3.6.13 O b BASES SURVEILLANCE SR 3.6.13.3 REQUIREMENTS (continued) Verifying, by visual inspection, that the ice condenser l inlet doors are not impaired by ice, frost, or debris provides assurance that the doors are free to open in the event of a DBA. For this unit, the Frequency of 18 months I is based on door design, which does not allow water I condensation to freeze, and operating experience, which indicates that the inlet doors very rarely fail to meet their SR acceptance criteria. Because of high radiation in the vicinity of the inlet doors during power operation, this Surveillance is normally performed during a shutdown. SR 3.6.13.4 Verifying the opening torque of the inlet doors provides assurance that no doors have become stuck in the closed position. The value of 675 in-lb is based on the design opening pressure on the doors of 1.0 lb/ft 2

                                                                   . For this unit, the Frequency of 18 months is based on the passive nature of the closing mechanism (i.e., once adjusted, there _ are no known factors that would change the setting, except possibly T             -

a buildup of ice; ice buildup is not likely, however, because of the door design, which does not allow water condensation to freeze). Operating experience indicates that the inlet doors usually meet their SR acceptance criteria. Because of high radiation in the vicinity of the ) inlet doors during power operation, this Surveillance is normally performed during a shutdown. SR 3.6.13.5 The torque test Surveillance ensures that the inlet doors have not developed excessive friction and that the return springs are producing a door return torque within limits. l The torque test consists of the following: 1

1. Verify that the torque, T(OPEN), required to cause opening motion at the 40* open position is s 195 in-lb; i

1' (continued) l Catawba Unit 2 B 3.6-88 Supplement 1

Ice Condenser Doors B 3.6.13 BASES SURVEILLANCE SR 3.6.13.5 (continued) REQUIREMENTS

2. Verify that the torque T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 40' open position is a 78 in-lb; and

3. Calculate the frictional torque, T(FRICT) = 0.5

{T(0 PEN) - T(CLOSE)}, and verify that the T(FRICT) is s 40 in-lb. The purpose of the friction and return torque Specifications is to ensure that, in the event of a small break LOCA or SLB, all of the 24 door pairs open uniformly. This assures that, during the initial blowdown phase, the steam and water i mixture entering the lower compartment does not pass through part of the ice condenser, depleting the ice there, while bypassing the ice in other bays. The Frequency of 18 months is based on the passive nature of the closing mechanism (i.e., once adjusted, there are no known factors that would change the setting, except possibly a buildup of ice; ice buildup is not likely, however, because of the door design, which does not allow water condensation to freeze). p Operating experience indicates that the inlet doors very , rarely fail to meet their SR acceptance criteria. Because of high radiation in the vicinity of the inlet doors during power operation, this Surveillance is normally performed i during a shutdown. SR 3.6.13.6 Verifying the OPERABILITY of the intermediate deck doors provides assurance that the intermediate deck doors are free to open in the event of a DBA. The verification consists of visually inspecting the intermediate doors for structural deterioration, verifying free movement of the vent assemblies, and ascertaining free movement cf each door when lifted with the applicable force shown below: p (continued) O Catawba Unit 2 B 3.6-89 Supplement 1 l

Ice Condenser Doors B 3.6.13 BASES l I l SURVEILLANCE SR 3.6.13.6 (continued) i REQUIREMENTS l Qgs.t Liftina Force

a. Adjacent to crane wall < 37.4 lb

b. Paired with door adjacent to crane wall s 33.8 lb

c. Adjacent to containment wall s 31.8 lb l d. Paired with door adjacent to containment s 31.0 lb wall The 18 month Frequency is based on the passive design of the intermediate deck doors, the frequency of personnel entry into the intermediate deck, and the fact that SR 3.6.13.2 confirms on a 7 day Frequency that the doors are not impaired by ice, frost, or debris, which are ways a door would fail the opening force test (i.e., by sticking or from i increaseddoorweight).

l SR 3.6.13.7 Verifying, by visual inspection, that the top deck doors are . l in place and not obstructed provides assurance that the

doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA.

The Frequency of 92 days is based on engineering judgment, which considered such factors as the following: l

a. The relative inaccessibility and lack of traffic in the vicinity of the doors make it unlikely that a door would be inadvertently left open;

b. Excessive air leakage would be detected by temperature monitoring in the ice condenser; and

c. The light construction of the doors would ensure that, in the event of a DBA, air and gases passing through the ice condenser would find a flow path, even if a door were obstructed.

(continued) l Catawba Unit 2 B 3.6-90 Supplement 1 l

Ice Condenser Doors B 3.6.13 O BASES (continued)- ! V l ! REFERENCES 1. UFSAR, Chapter 6.

2. 10 CFR 50, Appendix K.-

l l 3. 10 CFR 50.36, Technical Specifications, (c)(2)(11). l l l l l O l I O Catawba Unit 2 B 3.6-91 Supplement 1 l

Dividar Barrier Integrity B 3.6.14 B 3.6 CONTAINMENT SYSTEMS B 3.6.14 Divider Barrier Integrity BASES. BACKGROUND The divider barrier consists of. the operating deck and associated seals, personnel access doors, and equipment

hatches that separate the upper and lower containment compartments. - Divider barrier integrity is necessary to minimize bypassing of the ice condenser by the hot steam and air mixture released into the lower c(mpartment during a

                         ' Design Basis Accident (DBA). This ensures that most of the gases pass through the ice bed, which condenses the steam and limits pressure and temperature during the accident transient. Limiting the pressure and temperature reduces the release of fission product radioactivity from-containment to the environment in the event of a DBA.

In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment. This allows air and steam to flow from the lower compartment into the ice condenser. The resulting pressure increase within the ice condenser causes the intermediata deck doors and the door panels at the top

 $O                       of the condenser to open, which allows the air to flow out of the ice condenser into the upper compartment. The ice condenses the steam as it enters, thus limiting the pressure and temperature buildup in containment. The divider barrier separates the upper and lower compartments and ensures that the steam is directed into the ice condenser. The ice, together with the containment spray, is adequate to absorb the initial blowdown of steam and water from a DBA as well as the additional heat loads that would enter containment over several hours following the initial blowdown. The          i additional heat loads would come from the residual heat in      1 the reactor core, the hot piping and components, and the        ;

secondary system, including the steam generators. During the post blowdown period, the Air Return System (ARS) returns upper compartment air through the divider barrier to the lower compartment. This serves to equalize pressures in containment and to continue circulating heated air and steam - from the lower compartment through the ice condenser, where ; the heat is removed by the remaining ice. . 1 Divider barrier integrity ensures that the high energy 4 fluids released during a DBA would be directed through the (continued) l Catawba Unit 2 B 3.6-92 Supplement 1

Divider Barrier Integrity B 3.6.14 BASES BACKGROUND ice condenser and that the ice condenser would function as (continued) designed if called upon to act as a passive heat sink following a DBA. i APPLICABLE Divider barrier integrity ensures the functioning of SAFETY ANALYSES the ice condenser to the limiting containment pressure and temperature that could be experienced following a DBA. The limiting DBAs considered relative to containment tem and pressure are the loss of coolant accident (LOCA)perature and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. DBAs are assumed not to occur simultaneously or consecutively. Although the ice condenser is a passive system that requires no electrical power to perform its function, the Containment Spray System, RHR Spray System, and the ARS also function to assist the ice bed in limiting pressures and temperatures. Therefore, the postulated DBAs arc analyzed, with respect to containment Engineered Safety Feature (ESF) systems, f assuming the loss of one ESF bus, which is the worst case ,

     \                               single active failure and results in the inoperability of onetraininboththeContainmentSpraySystgm,RHRSpray System, and the ARS. Additionally, a 5.0 ft opening is conservatively assumed to exist in the divider barrier in the LOCA and SLB DBA analyses.

The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. The maximum peak containment temperature results from the SLB analysis and is discussed in the Bases for LCO 3.6.5, " Containment Air Temperature." In addition to calculating the overall peak containment pressures, the DBA analyses include calculation of the transient differential pressures that occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand these local transient pressure differentials for the limiting DBAs. The divider barrier satisfies Criterion 3 of 10 CFR 50.36 (Ref. 2). (continued) Catawba Unit 2 B 3.6-93 Supplement 1 l l 1

Dividar Barrier Inttgrity B 3.6.14 BASES (continued) LCO This LCO establishes the minimum equipment requirements to ensure that the divider barrier performs its safety function of ensuring that bypass leakage, in the event of a DBA, does r.ot exceed the bypass leakage assumed in the accident analysis. Included are the requirements that the personnel access doors and equipment hatches in the divider barrier are OPERABLE and closed and that the divider barrier seal is properly installed and has not degraded with time. An exception to the requirement that the doors be closed is made to allow personnel transit entry through the divider barrier. The basis of this exception is the assumption that, for personnel transit, the time during which a door is open will be short (i.e., shorter than the Completion Time of 1 hour for Condition A). The divider barrier functions with the ice condenser to limit the pressure and temperature that could be expected following a DBA. APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in contair" : nressure and temperature requiring the integrity of t'~ r, v m barrier. Therefore, the LC0 is applicable in MODES ' .? J nd 4. - O C T. - ano v e 2 cy and consequences of these events in MODES 5 sw due to the pressure and temperature limitat..ms of these MODES. As such, divider barrier integrity is not required in these MODES. ACTIONS M If one or more personnel access doors or eq ipment hatches (other than the pressurizer enclosure hatch are inoperable or open, except for personnel transit entry, I hour is allowed to restore the door (s) and equipment hatches to OPERABLE status and the closed position. The 1 hour Completion Time is sonsistent with LCO 3.6.1, " Containment," which requires that containment be restored to OPERABLE status within 1 h>ur. Condition A has been modified by a Note to provide clarification that, for this LCO, separate Condition entry is allowed for each personnel access door or equipment hatch. (continued) l l Catawba Unit 2 B 3.6-94 Supplement i 1

Divider Barrier Inttgrity , B 3.6.14 O V BASES ACTIONS B.d (continued) If the pressurizer enclosure hatch is inoperable or open, 6 hours are allowed to restore the hatch to OPERABLE status and the closed position. The 6 hour completion time is based on the need to perform inspections in the pressurizer compartment during power operatipn and analysis performed that shows an open hatch (7.5 ft bypass area) during a DBA does not impact the design pressure or temperature of the containment. L1 If the divider barrier seal is inoperable, I hour is allowed to restore the seal to OPERABLE status. The 1 hour Completion Time is consistent with LCO 3.6.1, which requires that containment be restored to OPERABLE status within 1 hour.- D.1 anu D.2

                       .If divider barrier integrity cannot be restored to OPERABLE O'                    status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.14.1 REQUIREMENTS Verification, by visual inspection, that all personnel access doors and equipment hatches between the upper and lower containment compartments are closed provides assurance that divider barrier integrity is maintained prior to the reactor being taken from MODE 5 to MODE 4. This SR is necessary because many of the doors and hatches may have l- been opened for maintenance during the shutdown. (continued) Catawba Unit 2 B 3.6-95 Supplement 1 l

Divider Barrier Integrity B 3.6.14 BASES b(N ' i SURVEILLANCE SR 3.6.142 i REQUIREMENTS l (continued) Verification, by visual inspection, that the personnel access door and equipment hatch seals, sealing surfaces, and alignments are acceptable provides assurance that divider barrier integrity is maintained. This inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.14.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for long periods of time. Those that use resilient materials in the seals must be opened and i inspected at least once every 10 years to provide assurance ! that the seal material has not aged to the point of degraded j performance. The Frequency of 10 years is based on the i known resiliency of the materials used for seals, the fact I that the openings have not been opened (to cause wear), and l operating experience that confirms that the seals inspected ! at this Frequency have been found to be acceptable. i SR 3.6.14.3 Verification, by visual inspection, after each opening of a O V personnel access door or equipment hatch that it has been closed makes the operator aware of the importance of cPsing it and thereby provides additional assurance that diviaer barrier integrity is maintained while in applicable MODES. SR 3.6.14.4 Conducting periodic physical property tests on divider barrier seal test coupons provides assurance that the seal material has not degraded in the containment environment, including the effects of irradiation with the reactor at power. The required tests include a tensile strength test. The Frequency of 18 months was developed considering such . factors as the known resiliency of the seal material used, ! the inaccessibility of the seals and absence of traffic in their vicinity, and the unit conditions needed to perform the SR. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was ; concluded to be acceptable from a reliability standpoint. (continued) l Catawba Unit 2 B 3.6-96 Supplement 1

Divider Barrier Integrity B 3.6.14 r^ BASES (s) SURVEILLANCE SR 3.6.14.5 REQUIREMENTS (continued) Visual inspection of the seal around the perimeter provides assurance that the seal is properly secured in place. The Frequency of 18 months was developed considering such factors as the inaccessibility of the seals and absence of traffic in their vicinity, the strength of the bolts and mechanisms used to secure the seal, and the unit conditions needed to perform the SR. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. l REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

I O Catawba Unit 2 B 3.6-97 Supplement 1 l l

( Containment Recirculation Drains l B 3.6.15 ! I I l B 3.6 CONTAINMENT SYSTEMS B 3.6.15 Containment Recirculation Drains BASES BACKGROUND The containment recirculation drains consist of the ice , condenser drains and the refueling canal drains. The ice condenser is partitioned into 24 bays, each having a pair of inlet doors that open from the bottom plenum to allow the I hot steam-air mixture from a Design Basis Accident (DBA) to i enter the ice condenser. Twenty of the 24 bays have an ice l condenser floor drain at the bottom to drain the melted ice I into the lower compartment (in the 4 bays that do not have I drains, the water drains through the floor drains in the i adjacentbays). Each drain leads to a drain pipe that drops j l down several feet, then makes one or more 90* bends and j l exits into the lower compartment. A check (flapper) valve j l at the end of each pipe keeps warm air from entering during I normal operation, but when the water exerts pressure, it I opens to allow the water to spill into the lower compartment. This prevents water from backing up and interfering with the ice condenser inlet doors. The water delivered to the lower containment serves to cool the atmosphere as it falls through to the floor and provides a ' source of borated water at the containment sump for long term use by the Emergency Core Cooling System (ECCS) and the Containment Spray System during the recirculation mode of operation. The refueling canal drains are at low points in the refueling canal. During a refueling, valves are closed in the drains and the canal is flooded to facilitate the refueling process. The water acts to shield and cool the spent fuel as it is transferred from the reactor vessel to storage. After refueling, the canal is drained and the valves are locked open. In the event of a DBA, the refueling canal drains are the main return path to the lower compartment for Containment Spray' System water sprayed into the upper compartment. L The ice condenser drains and the refueling canal drains function with the ice bed, the Containment Spray System, and the ECCS to limit the pressure and temperature that could be l expected following a DBA. t

 \                                                                          (continued) l Catawba Unit 2                        B 3.6-98                        Supplement 1 l

L Containment Recirculation Drains B 3.6.15 BASES (continued). l APPLICABLE The limiting DBAs considered relative to containment SAFETY ANALYSES temperature and pressure are the loss of coolant accident l (LOCA) and the steam line break (SLB). The LOCA and SLB are l

                      ' analyzed using computer codes designed to predict the            !

resultant containment pressure and temperature transients. J DBAs are assumed not to occur simultaneously or , consecutively. Although the ice condenser is a passive I system that requires no electrical power to perform its function, the Containment Spray System and the Air Return System (ARS) also function to assist the ice bed in limiting pressures and temperatures. Therefore. the analysis of the l p(ostulated DBAs, with respect to Engineered Safety FeatureESF worst case single active failure and results in one. train of the Containment _ Spray System and one train of the ARS being rendered inoperable. The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is' calculated to be less than the containment design pressure. The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed ' in the Bases for LCO 3.6.5, " Containment Air Temperature." O In addition to calculating the overall. peak containment pressures, the DBA analyses include calculation of the transient differential pressures that occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand these local transient pressure differentials for the limiting DBAs. i The containment recirculation drains satisfy Criterion 3 of i 10 CFR 50.36 (Ref. 2). l LCO _ This LCO establishes the minimum requirements to ensure that . the containment recirculation drains perform their safety l functions. The ice condenser floor drain valve disks must be closed to minimize air leakage into and out of the ice condenser during normal operation and must open in the event ! of a DBA when water begins to drain out. The refueling canal drains must have their plugs removed and remain clear to ensure the return of Containment Spray System water to ! ' the lower containment in the event of a DBA. The i containment recirculation drains function with the ice 1 lO l (continued) l Catawba lait 2 B 3.6-99 Supplement 1 l

Containment Recirculation Drains l B 3.6.15 ) BASES LCO condenser, ECCS, and Containment Spray System to limit the (continued) pressure and temperature that could be expected following a DBA. APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature, which would require the operation of the containment recirculation drains. Therefore, the LC0 is applicable in MODES 1, 2, 3, and 4. The probability and consequences of these events in MODES 5 and 6 are low due to the pressure and temperature limitations of these MODES. As such, the containment recirculation drains are not required to be OPERABLE in these MODES. ACTIONS L1 If one ice condenser floor drain is inoperable,1 hour is allowed to restore the drain to OPERABLE status. The O Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, I

                     " Containment," which requires that containment be restored to OPERABLE status within'1 hour.

B.d If one refueling canal drain is inoperable, I hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the' containment analysis. The 1 hour > Completion Time is consistent with the ACTIONS of LCO 3.6.1, which requires that containment be restored to OPERABLE status in 1 hour. C.1 and C.2 l If the affected drain (s) cannot be restored to OPERABLE L status within the required Completion Time, the plant must p (continued)

 \d                                                                                  j l Catawba Unit 2                    B 3.6-100                       Supplement 1

Containment Recirculation Drains B 3.6.15 O BASES , V l ACTIONS C.1 and C.2 (continued) be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.6.15.1 and SR 3.6.15.2 REQUIREMENTS Verifying the OPERABILITY of the refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.15.1 confirms that the refueling canal drain valves have been locked open and that the drains i are clear of any obstructions that could impair their functioning. In addition to debris near the drains, SR 3.6.15.2 requires attention be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water ( is flowing to the drains. SR 3,6.15.1 must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the plugs have been removed and that no debris that could impair the drains was deposited during the time the canal was filled. SR 3.6.15.2 is performed every 92 days for the upper compartment and refuel canal areas. The 92 day Frequency was developed considering such factors as the inaccessibility of the drains, the absence of traffic in the vicinity of the drains, and the redundancy of the drains. l SR 3.6.15.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA. Inspecting the drain valve disk ensures , that the valve is performing its function of sealing the l drain line from warm air leakage into the ice condenser l during normal operation, yet will open if melted ice fills l the line following a DBA. Verifying that the drain lines ! are not obstructed ensures their readiness to drain water l from the ice condenser. The 18 month Frequency was p (continued) V l Catawba Unit 2 B 3.6-101 Supplement 1 l l

I Centainment Recirculation Drains B 3.6.15 1 BASES l SURVEILLANCE SR 3.6.15.3 (continued) REQUIREMENTS developed considering such factors as the inaccessibility of i the drains during power operation; the design of the ice ) condenser, which precludes melting and refreezing of the ice; and operating experience that has confirmed that the drains are found to be acceptable when the Surveillance is performed at an 18 month Frequency. Because of high i radiation in the vicinity of the drains during power l operation, this Surveillance is normally done during a i shutdown. l REFERENCES 1. UFSAR, Section 6.2.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

l 0 - J l 1 O l Catawba Unit 2 B 3.6-102 Supplement 1

f l Reactor Building B 3.6.16 lO l () B 3.6 CONTAINMENT SYSTEMS B 3.6.16 Reactor Building l BASES BACKGROUND The reactor building is a concrete structure that surrounds l the steel containment vessel. Between the containment vessel and the reactor building inner wall is an annular space that ccliects containment leakage that may occur following a loss of coolant accident (LOCA). This space also allows for periodic inspection of the outer surface of the steel containment vessel. 1 The Annulus Ventilation System (AVS) establishes a negative pressure in the annulus between the reactor building and the steel containment vessel under post-accident conditions. Filters in the system then control the release of radioactive contaminants to the environment. The reactor ' building is required to be OPERABLE to ensure retention of containment leakage and proper operation of the AVS. k Q

     ' APPLICABLE        The design basis for reactor building OPERABILITY is a SAFETY ANALYSES   LOCA. Maintaining reactor building OPERABILITY ensures that the release of radioactive material from the containment            '

atmosphere is restricted to those leakage paths and associated leakage rates assumed in the accident analyses. The reactor building satisfies Criterion 3 of 10 CFR 50.36 l (Ref. 1). l l LCO Reactor building OPERABILITY must be maintained to ensure proper operation of the AVS and to limit radioactive leakage from the containment to those paths and leakage rates assumed in the accident analyses. 1 i APPLICABILITY Maintaining reactor building OPERABILITY prevents leakage of I radioactive material from the reactor building. Radioactive l material may enter the reactor building from the containment following a LOCA. Therefore, reactor building OPERABILITY 4 p (continued) d Catawba Unit 2 B 3.6-103 Supplement 1 l

1 Reactor Building B 3.6.16 O BASES G APPLICABILITY is required in MODES 1, 2, 3, and 4 when a steam line break, l (continued) LOCA, or rod ejection accident could release radioactive material to the containment atmosphere. In MODES 5 and 6, the probability and consequences of these l events are low due to the Reactor Coolant System temperature i and pressure limitations in these MODES. Therefore, reactor I building OPERABILITY is not required in MODE 5 or 6. ACTIONS 6.d ; In the event reactor building OPERABILITY is not maintained, reactor building OPERABILITY must be restored within 24 hours. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and. 3 the low probability of a Design Basis Accident occurring during this time period. B.1 and B.2 ,

 /~'N If the reactor building cannot be restored to OPERABLE Q'                    status within the required Completion Time, the plant must I

be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full , power conditions in an orderly manner and without l challenging plant syster s. j l l SURVEILLANCE SR 3.6.16.1 REQUIREMENTS Maintaining reactor building OPERABILITY requires i maintaining each door in the access opening closed, except when the access opening is being used for normal transit entry and exit. The 31 day Frequency of this SR is based on engineering judgment and is considered adequate in view of the other indications of door status that are available to the operator. (continued) l Catawba Unit 2 B 3.C-104 Supplement 1

t Reactor Building l B 3.6.16 l l'O l BASES lv SURVEILLANCE SR 3.6.16.2 REQUIREMENTS i (continued) The ability of a AVS train to produce the required negative pressure 2 0.5 inch water gauge during the test operation within 1 minute provides assurance that the building is adequately sealed. The negative pressure prevents leakage

from the building, since outside air will be drawn in by the low pressure. The negative pressure must be established l within the time limit to ensure that no significant quantity
of radioactive material leaks from the reactor building l prior to developing the negative pressure.

The AVS traNs are teste every 18 months on a STAGGERED l TEST BASIS to ensure that in addition to the requirements of LC0 3.6.10, " Annulus Ventilation System," either AVS train will perform this test. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. SR 3.6.16.3 i ,g This SR would give advance indication of gross deterioration of the concrete structural integrity of the reactor 'rd building. The 40 month Frequency is based on the l requirement to perform two additional inspections at approximately equal intervals between the Type A tests required by SR 3.6.1.1 performed on a 10-year interval. The verification is done during shutdown. REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). l l

l O Catawba Unit 2 B 3.6-105 Supplement 1 l

f~ CVIWS B 3.6.17 B 3,6 CONTAINMENT SYSTEMS V B 3.6.17 Containment Valve Injection Water System (CVIWS) BASES r j , BACKGROUND The CVIWS is required by 10 CFR 50, Appendix A, GDC 54,

                          " Piping Systems Penetrating Containment" (Ref. 1), to ensure a water seal to a specific class of containment isolation valves (double disc gate valves) during a LOCA, to prevent leakage of containment atmosphere through the gate valves.

The CVIWS is designed to inject water between the two seating surfaces of double disc gate valves used for l Containment isolation. The injection pressure is higher than Containment design peak pressure during a LOCA. This will prevent leakage of the Containment atmosphere through the gate valves, thereby reducing potential offsite dose below the values specified by 10 CFR 100 limits following . the postulated accident. During normal power operation, the system is in a standby mode and does not perform any function. During accident situations the CVIWS is activated to perform its safety related function, thus limiting the release of containment O atmosphere past specific containment isolation valves, in order to mitigate the consequences of a LOCA. Containment isolation valves, for systems which are not used to mitigate the consequences of an accident, will be supplied with CVIWS seal water upon receipt of a Phase A isolation signal. Containment isolation valves, for accident mitigating systems which are supplied with seal water from the CVIWS, l have their seal water supplies actuated by a Containment 4 Pressure - High-High signal. j The system consists of two independent, redundant trains; one supplying gate valves that are powered by the A train diesel and the other supplying gate valves powered by the B train diesel. This separation of trains prevents the possibility of both containment isolation valves not sealing due to a single failure. Each train consists of a surge chamber which is filled with water and pressurized with nitrogen. One main header exits the chamber and splits into several headers. A solenoid valve is located in the main header before any of the branch headers which will open after a 60 second delay on a l (continued) l l Catawba Unit 2 B 3.6-106 Supplement 1

CVIWS B 3.6.17 I O BASES V BACKGROUND Phase A isolation signal. Each of the headers supply (continued) injection water to containment isolation valves located in the same general location, and close on the same engineered safety signal. A solenoid valve is located in each header which supplies seal water to valves closing on a Containment Pressure - High-High signal. These solenoid valves open i after a 60 second delay on a Containment Pressure - High- ! High signal. Since a Phase A isolation signal occurs before ! a Containment Pressure - High-High signal, the solenoid i valve located in the main header will already be injecting ! water to Containment isolation valves closing on a Phase A isolation signal. This leaves an open path to the headers supplying injection water on a Containment Pressure - High-High signal. The delay for the solenoid valves opening is to allow adequate time for the slowest gate valve to close, before water is injected into the valve seat. Makeup water is provided from the Demineralized Water Storage Tank for testing and adding water to the surge chamber during normal plant operation. Assured water is provided from the essential header of the Nuclear Service Water System (NSWS). This supply is assured for at least 30 days following a postulated accident. If the water level in f

the surge chamber drops below the low-low level or if the k: surge chamber nitrogen pressure drops below the low-low pressure after a Phase A isolation signal, a solenoid valve in the supply line from the NSWS will automatically open and remains open, assuring makeup to the CVIWS at a pressure greater than 110% of peak Containment accident pressure.

Overpressure protection is provided to relieve the pressure buildup caused by the heatup of a trapped volume of incompressible fluid between two positively closing valves (due to containment temperature transient) back into containment where an open relief path exists. APPLICABLE The CVIWS design basis is established by the consequences of SAFETY ANALYSES the limiting DBA, which is a LOCA. The accident analysis (Ref. 2) assumes that only one train of the CVIWS is functional due to a single failure that disables the other train. Makeup water can be assured from the NSWS for 30 days following a postulated LOCA. The CVIWS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 3). (continued) Catawba Unit 2 B 3.6-107 Supplement 1 l

CVIWS B 3.6.17 l (v BASES (continued) LCO In the event of a DBA, one CVIWS train is required to l provide the seal injection assumed in the safety analysis. 4 Two trains of the CVIWS must be OPERABLE to ensure that at ] least one train will operate, assuming that the other train j is disabled by a single active failure. J APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could require a containment i isolation. The large break LOCA, on which this system's 1 design is based, is a full power event. Less severe LOCAs i and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and Reactor Coolant System pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such i an accident is low. In MODES 5 and 6, the probabillty and consequences of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the CVIWS is not required to be OPERABLE. O O ACTIONS Ad With one CVIWS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of l providing 100% of the valve injection needs after a DBA. The 7 day Completion Time is based on consideration of such ! , factors as the availability of the OPERABLE redundant.CVIWS I train and the low probability of a DBA occurring during this period. The Completion Time is adequate to make most l repairs. l B.1 andjh2 i If the CVIWS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating (continued) l Catawba Unit 2 B 3.6-108 Supplement 1

                                                                                          ~l CVIWS B 3.6.17
      ' BASES O

ACTIONS B.1 and B.2 (continued) experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE- SR 3.6.17.1 - REQUIREMENTS Verifying each CVIWS train is pressurized to a 36.4 psig ensures the system can meet the design basis. Assured water is provided from the essential header of the NSWS. The 31 1 day Frequency was developed in consideration of.the known reliability of the system and the two train redundancy available. I SR- 3.6.17.2 This SR verifies that each CVIWS train can perform its required function when needed by measuring the existing conditions for the valves being injected. Gate valves A . served by the CVIWS do not receive a conventional. Type C leak rate test using air as a test _ medium. U The containment isolation valves served by the CVIWS may be j tested individually or simultaneously. Containment- ) isolation valves are leak rate tested by this SR by injecting seal water from the CVIWS to the containment isolation valves. With the containment isolation valve closed, the leakage is determined by measuring flow rate of seal water out of the containment valve injection water surge chamber. 1 The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown that these l components usually pass the Surveillance when performed at the 18 month frequency. Therefore, the Frequency was , concluded to be acceptable from a reliability standpoint. ' ! Furthermore, the SR interval was developed considering that the CVIWS OPERABILITY is demonstrated at a 31 day Frequency by SR 3.6.17.1. (continued) Catawba Unit 2 B 3.6-109 Supplement 1 l l' ; 1 t

CVIWS B 3.6.17 l lO%J BASES SURVEILLANCE SR 3.6.17.3 REQUIREMENTS (continued) This SR ensures that each CVIWS train responds properly to the appropriate actuation signal. The Surveillance verifies that the automatic valves actuate to their correct position. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore the Frequency was concluded to be acceptable from a reliability standpoint. REFERENCES 1. 10 CFR 50, Appendix A, GDC 54. ' l

2. UFSAR, Section 6.2.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

                                                                                     . I O                                                                                     !

l l l l I I U l l Catawba Unit 2 B 3.6-110 Supplement 1 l t

Contairveent Spray System fle*Eanda6s3D B 3.6.60 BASES SURVEILLANCE SR 3.6.60.2 (continued) REQUIREMENTS perfonnance, and detect incipient failures by indicating abnornel perfoneance. The Frequency of this SR is in accordance with the Inservice Testing Progree. SR 3.6.60.3 and SR 3.5.68.4 These SRs require verificat' ion that each automatic containment spray valve actuates to its correct position and each containment spray pimp starts upon receipt of an actual or simulated Aontainment ---- -- - signal. This Surveillance is not required for valves that are locked. sealed, or otherwise secured in the required p'osition under administrative controls. The (185 month Frequency is based on the need to perform these Surveillances under the @ conditions that apply during a plant outage and.the potential for an unplanned transient if the Surve111ances were performed with the reactor at power. Operating 1 experience has shown these components usually pass the Surveillances when performed at the f18& month Frequency. Therefore, the Fr was concluded to be acceptable from @ f a reliability s int. The surveillance of contairment sump isolation valves is also required by SR 3.6.6.3. A single surveillance may be

         %W                        used to satisfy both requirements.

SR 3.6.2.I h With the containment spray inlet valves closed and the spray 7%5frs4 vazW . . ca. headea drained of any solution, low pressure air or smoke alsokt peri.JttO) can be blown thr u tant connections. VThis SR ensures that Mked. t.dt' VEtaum each spray nozzle is unobstructed and that spray coverage of M f 4, y A-* j**' #k the containment during an accident is not degraded. Because of the passive desi of the nozzle, a test at (IETTF2D N 05 L 'd "'**( e, afDEPINranif'E11Iyear intervals is considered adequate @ hvA uno e m teJ Tl to detect obstruction of the spray nozzles. ' REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43. (continued) s WOG-STS- B 3.6-92 Rev 1, 04/07/95 Cala.wbo

issue Number ] 45i Affected Section l3.830_ Distr,1bution System-Shutdown l Affected Unas CNS: l Yes) MNS: l Nol Affected Pages ITS: 3.8-40 ITS: ITS Boeos: ITS Bases: l CTS: I CTS: DOCS: 1 DOCS: NRG: NRG: NRG Soses: NRG Bases: JFD: JFD: NSHC: - NSHC: Deectlption The logicot connector between Required Action A.2.5 ond A.2.6 of 3.8.10 is missng. Should be 'AND" consistent with CTS markup and NUREG markup. l i l i I l l l l i c 1 I l l

Distribution Systems-Shutdown 3.8.10 ACTIONS O CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.4 Initiate actions to Inunediately restore required AC, channels of DC, DC trains, and AC vital bus electrical power distribution subsystems to OPERABLE status. Afin A.2.5 Declare associated Inunediately required residual heat removal subsystem (s) inoperable and not in operation. AliQ Inynediately A.2.6 Declare affected Low O , Temperature Overpressure Protection feature (s) inoperable. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and 7 days voltage to required AC, DC channel, DC train, and AC vital bus electrical power distribution subsystems. O 3.8-40 Supplement 1 l Catawba Unit 1

1 Distributien Systems-Shutdown 3.8.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

i l A. (continued) A.2.4 Initiate actions to Inunediately restore required AC, channels of DC, DC trains, and AC vital bus electrical power 4 distribution subsystems to OPERABLE status. AliQ A.2.5 Declare associated Inunediately required residual heat removal subsystem (s) i inoperable and not in operation.- AfiQ Immediately A.2.6 Declare affected Low O Temperature Overpressure Protection feature (s) inoperable. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and 7 days voltage to required AC, DC channel, DC train, and AC vital bus electrical power distribution subsystems. O l Catawba Unit 2 3.8-40 Supplement 1

O leeue Number l 49] ANected Section 13.3.2 ESFAS. NSWS Suctkm Transfer . l Affected Unas CNS: l Yesl MNS: l Noj Affected Pogos ITS: ITS: ITS Saoes: B 3.3-100 ITS Bones:

                     ' CTS:                                                  CTS:

DOCS: DOCS: NRG: 'IRG:

NRG Bones: B 3.3-114 (insert) 4tG Bones: JFD: JFD: NSHC: NSHC: EIN N Ws for 3.3.2, Action Q hos two errors. Lost sentonce in first paragraph should read "one<xJt-of-two" configuration not "one-out-of-three* when one of the three channels is placed in trip. Second porograph should read 4 hours instood of I hour to place in trip consistent with the LCO. 6 i

ESFAS Instrumentation B 3.3.2 q BASES V ACTIONS 0.1. 0.2.1. and 0.2.2 (continued) With one channel of NSWS Suction Transfer - Low Pit Level inoperable in one or more NSWS pits, 4 hours are allowed to restore the channel to OPERABLE status or to place it in the tripped condition. The failure of one channel places the l Function in a two-out-of-two configuration. The failed channel must be tripped to place the Function in a one-out-l of-two configuration that satisfies redundancy requirements. Failure to restore the inoperable channel to OPERABLE status l or place it in the tripped condition within 4 hours requires the unit be placed in MODE 3 within the following 6 hours and MODE 5 within the next 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, this Function is no longer required OPERABLE. R.1. R.2.1. and R.2.2 /~ With two channels of NSWS Suction Transfer - Low Pit Level ( ' inoperable in one or more pits, one channel must be restored to OPERABLE status or the NSWS must be aligned to the Standy NSWS Pond within 4 hours. Failure to restore one channel or to accomplish the realignment within 4 hours requires the unit be placed in MODE 3 within the following 6 hours and MODE 5 within the next 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, this Function is no i longer required OPERABLE. l l SURVEILLANCE The SRs for each ESFAS Function are identified by the SRs REQUIREMENTS column of Table 3.3.2-1. A Note has been added to the SR Table to clarify that Table 3.3.2-1 determines which SRs apply to which ESFAS Functions. (continued) k. l Catawba Unit 1 B 3.3-100 Supplement 1

ESFAS Instrumentation B 3.3.2 BASES ACTIONS 0.1. 0.2.1. and 0.2.2 (continued) With one channel of NSWS Suction Transfer - Low Pit Level inoperable in one or more NSWS pits, 4 hours are allowed to restore the channel to OPERABLE status or to place it in the tripped condition. The failure of one channel places the

    -l                 Function in a two-out-of-two configuration. The failed channel must be tripped to place the Function in a one-out-l                 of-two configuration that satisfies redundancy requirements.

Failure to restore the inoperable channel to OPERABLE status l or place it in the tripped condition within 4 hours requires the unit be placed in MODE 3 within the following 6 hours and MODE 5 within the next 30 hours. ! The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, this Function is no longer required OPERABLE. R.I. R.2.1. and R.2.2 CJ V With two channels of NSWS Suction Transfer - Low Pit Level inoperable in one or more pits, one channel must be restored to OPERABLE status or the NSWS must be aligned to the Standy NSWS Pond within 4 hours. Failure to restore one channel or to accomplish the realignment within 4 hours requires the unit be placed in MODE 3 within the following 6 hours and MODE 5 within the next 30 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, this Function is no I longer required OPERABLE. l l SURVEILLANCE The SRs for each ESFAS Function are identified by the SRs REQUIREMENTS column of Table 3.3.2-1. A Note has been added to the SR Table to clarify that Table 3.3.2-1 determines which SRs apply to which ESFAS Functions. (continued) l Catawba Unit 2 B 3.3-100 Supplement 1

INSERT P.1. P.2.1 and P.2.2 Condition P applies to the Containment Pressure Control System Start and Terminate Permissives. With one or more channels inoperable, the affected containment spray, containment air return fans, and hydrogen skimer fans must be declared inoperable imediately. The supported system LCOs provide the appropriate Required Actions and Completion Times 'for the equipment made inoperable by the inoperable channel. The imediate Completion. Time is appropriate since the inoperable channel could prevent the supported equipment from starting when required. Additionally, protection from an inadvertent actuation may not be provided if the terminate function is not OPERABLE. l 1 0.1. 0.2.1. and 0.2.2 With one channel of NSWS Suction Transfer - Low Pit Level inoperable in one or more NSWS pits, 4 hours are allowed to restore the channel to OPERABLE status or to place it in the tripped condition. The failure of one channel places the Function in a two-out-of-two configuration. The failed channel must be tripped to place the Function in a one-out-of-two configuration that satisfies

redundancy requirements.

O Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 4 hours requires the unit be placed in MODE 3 within the following 6 hours and MODE 5 within the next 30 hours. l l I The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, this Function is no i longer required OPERABLE. R.1. R.2.1. and R.2.2 With two channels of NSWS Suction Transfer - Low Pit Level inoperable in one or more pits, one channel must be restored to OPERABLE status or the NSWS must be aligned to the Standby NSWS Pond within 4 hours. Failure to restore one channel or to accomplish the realignment within 4 hours requires the unit be , ( placed in MODE 3 within the following 6 hours and MODE 5 within the next ! 30 hours. ' The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. In MODE 5, this function is no longer required OPERABLE. INSERT Page B 3.3-114 O Catawba

5 leeue Number l 50j Affected Section 13.1.4 Rod Group Alignment Umits l Affected Unlis CNS: l Noj MNS: l Yesj

   ~ Affected Pogos     ITS:           3.1-7                                ITS:            3.1-7. 8 ITS Bones:                                          ITS Bases:

CTS: CTS: DOCS: DOCS: NItG: NRG: NRG Bones: NRG Bones: JFD: JFD: NSHC: NSHC: , l , Deecstplion [ l Completion times for Condmon A and B do not line up with the required octions. Revise consistent with writers j guide. Only applies to Condition A for Cotowbo. l l r l l i i I

l Rod Group Alignment Limits 3.1.4 Im 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits l l LCO 3.1.4 All shutdown and control rods shall be OPERABLE, with all 1 individual indicated rod positions within 12 stens of their l group step counter demand position. APPLICABILITY: MODES 1 and 2. ) l ) i l ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod (s) A.1.1 Verify SDM is within 1 hour l untrippable. the limit specified in the COLR. E t l . A.1.2 Initiate boration to I hour l [ restore SDM to within i limit. ANQ A.2 Be in MODE 3. 6 hours l l B. One rod not within B.1 Restore rod to within 1 hour alignment limits, alignment limits. E (continued) l O Catawba Unit 1 3.1-7 Supplement 1 1 l

Rod Group Alignment Limits 3.1.4 I f 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits LCO 3.1.4 All. shutdown and control rods shall be OPERABLE, with' all individual indicated rod positions within 12 steps of their group step counter demand position. APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod (s) A.1.1 Verify SDM is within 1 hour untrippable. the limit specified in the COLR. E ' l A.1.2 Initiate boration to I hour l O , restore SDM to within limit. AE A.2 Be in MODE 3. 6 hours l B. One rod not within B.1 Restore rod to within 1 hour l alignment limits. alignment limits. ] E I (continued) . l O Catawba Unit 2 3.1-7 Supplement 1 l

leets Number l 511 Affected Sectior. 13.3.4 Rod Group Ahonment l Affected Units CNS: l Yesl MNS: l Yetj Affected Pages ITS: 17 5: ITS Soses. B 3.1-26 ITS Bases: B 3.1-26 CTS: CTS: DOCS: DOCS: NRG: NIIG: NilG Bones: B 3.1-28 NRG Boess: B 3.1-28 JFD: 83 JFD: B3 NSHC' NSHC. Desctlpilon The Boses for Action A.I.1 and A.I.2 implies hat the boration to rectore SDM is on emergency boration. This is not consistent with the requirement of the LCO oction and should be veleted. TNs change was already justified by LA.1, JFD 17 is odded to the STS markup O 1 l l

Rod Grcup Alignment Limits B 3.1.4

 ,G    BASES (continued)

U ACTIONS A.1.1 and A.1.2 When one or more rods are untrippable, there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovereG. The Completion Time of I hour is adequate for determining SDM l and, if necessary, for initiating boration to restore SDM. In this situation, SDM veification must include the worth of the untrippable rod, as well as a rod of maximum worth. M If the untrippable rod (s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on O operating experience, .for reaching MODE 3 from full power b) conditions in an orderly manner and without challenging plant systems. IL1 When a rod becomes misaligned, " can usually be moved and is still trippable. If the rod <en be realigned within the Completion Time of I hour and the rod was not misaligned for a significant period of time before being discovered, local xenon redistribution during this short interval will not be significant, and operation may proceed without further restriction. An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LC0 3.1.5, " Shutdown Bank Insertion Limits," and LC0 3.1.6, " Control Bank Insertion Limits." The Completion Time of 1 hour gives the operator sufficient time to adjust the rod positions in an orderly manner. (continued) l Catawba Unit 1 8 3.1-26 Supplement 1

Rod Group Alignment Limits B 3.1.4 BASES (continued)

ACTIONS A.1.1 and A.1.2 When one or more rods are untrippable, there is a

! possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovered. The Completion Time of 1 hour is adequate for determining SDM l and, if necessary, for initiating boration to restore SDM. In this situation, SDM verification must include the worth o; the untrippable rod, as well as a rod of maximum worth. I U If the untrippable rod (s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are ret applicable. To achieve this status, the unit must be.' brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on

         .                      operating experience, for reaching MODE 3 from full power

t conditions in an orderly manrer and without challenging plant systems.

M When a rod becomes misaligned, it can usually be moved and is still trippable. If the rod can be realigned within the Completion Time of 1 hour and the rod was not misaligned for a significant period of time before being discovered, local xenon redistribution during this short interval will not be significant, and operation may proceed without further ; restriction. l An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LC0 3.1.5, " Shutdown Bank Insertion Limits," and LCO 3.1.6, " Control Bank Insertion Limi ts . " The Completion Time of 1 hour gives the operator l sufficient time to adjust the rod positions in an orderly manner. Q (continued) fV l Catawba Unit 2 B 3.1-26 Supplement 1

Rod Group Alignment Linits B 3.1 o V - LCO Failure to meet the requirements of this LCO may produce (continued) unacceptable power peaking factors and URs. or unacceptable 50Hs, all of which may constitute initial conditions inconsistent with the safety analysis. APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the ODERABILIlY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3. 4, 5. and 6. the ali limits do not apply because the control rods are, t+r=ad marf

rametor u m r

shut down and not producing fission power. In the shutdown ' y ,,,,l g W MODES, the OPERABILITY of the shutdown and control rods has - the potential to affect the required SON. but this effect can be cospensated for by an increase in the boron concentration of tte RCS. See LCO 3.1.1 *SitJTomel MARGIN ' (50MM-K V20VF/* for SOM in MODES 3. 4. and fi and LCO 3.9.1. noron concentration." for boron concentration i requirements during refueling. ACTIONS A.1.1 and A.1.2 When one or more rods are untr yiable. there is a O possibility that the re Under these conditions. itquired sDM may is important be adversely to determine 50r., and if it is less than the required value. initiate theaffected. boration until the required SOM is recovered. The SOM Completion and. if necessary,Time of 1 hour is adequate for initiatingfensegenc3 boration for detemining% h restorg50M. g In this situation. SOM verification must include the worth of the untrippe.ble rod, as well as a rod of maximum worth. I b.1 If the untrippable rod (s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve (continued) ' B 3.1 28 Rev 1. 04/07/95 e,AA O

l Jurtificatico f;r Deviations S:ctica 3.1 - ht:ctivity C:ntr:1 Systems I BASES control bank information was deleted and/or modified in ITS 3.1.5 to reflect the Bases specific to the shutdown bank insertion limits. The control bcnk discussion is not appropriate to this Bases for shutdown banks. The detailed information for control banks is described in the Bases for ITS 3.1.6 consistent with STS 3.1.7.

16. The STS 3.1.7 Bases provides an example figure which illustrates insertion limits. The specific insertion limit figures for unit operation are maintained in the COLR in accordance with the current licensing basis. The inclusion of " example" curves within the Bases of the Technical Specifications is not currently required and creates an unnecessary aaministrative burden for maintaining examples which remain consistent with any changes in COLR methodology. The inclusion of these examples is not necessary to maintain compliance with the LCO and could lead to inappropriate use of figures which do not reflect the current cycle core design limits. Therefore, the ITS 3.1.6 Bases does not contain any example insertton limit figures.

17. The STS 3.1.5 Bases for Actions A.1.1 and A.2.2 imply that the required boration to restore shutdown margin is emergency boration, however, this is not ccnsistent with the actual stated action. The operator wtll use the appropriate means for restoring SDM based on the amount that it is out of limits. ITS 3.1.4 Bases does not include " emergency. " This change is justified in Discussion of }

Change LA.1. Catawba Units 1 and 2 33 Supplement 15/20/07l

leeue Number l 52j Affected Section 13.3.1 RTS l Affected Units CNS: l Yes) MNS: l Yesl Affected Pages ITS: ITS: ITS Bases: B 3.3-49. 50 ITS Boess: B 3.3-50. 51 CTS: CTS: DOCS: DOCS: NRG: NRG: NRG Boess: B 3.3-55 NRG Bases: B 3.3-55 JFD: JFD: NSHC: NSHC: hription i

 '9 3.31.8 nbte requires vermcation of P6 and PIO interlocks. This suNeillonce oppus to more than one function. The    l Po interlock is only vermed during the Intermedicte Range COT and the P10 interlock is only verfied during the Power Range COT. This is consistent wtth the design and the CTS. The Boses for SR 3.3.1.8 ore clarlfled to Indicate that the interiocks are only verifled ossociated with their source function.

1

RTS Instrumentation B 3.3.1 O BASES U SURVILLANCE SR 3.3.1.6 REQUIREMENTS (continued) SR 3.3.1.6 is a calibration of the excore channels to the incore channels. If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are declared inoperable. This Surveillance is performed to verify the f(AI) input to the overtemperature AT Function and overpower AT Function. l A Note modifies SR 3.3.1.6. The Note states that this ; Surveillance is required only if reactor power is > 75% RTP and that 24 hours is allowed for completing the first surveillance after reaching 75% RTP. The Frequency of 92 EFPD is adequate. It is based on industry operating experience, considering instrument 1 reliability and operating history data for instrument drift. ' SR 3.3.1.7 SR 3.3.1.7 is the performance of a COT every 92 days. A COT is performed on each required channel to ensure the channel will perform the intended Function. l The tested portion of the loop must trip within the < Allowable Values specified in Table 3.3.1-1. , 1 l The setpoint shall be left set consistent with the assumptions of the setpoint methodology. SR 3.3.1.7 is modified by a Note that provides a 4 hour delay in the requirement to perform this Surveillance for l source range instrumentation when entering MODE 3 from MODE

2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short tima in MODE 3 l until the RTBs are open and SR 3.3.1.7 is no longer required 1

(continued) Catawba Unit 1 8 3.3-49 Supplement 1 l

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.7 (continued) REQUIREMENTS to be performed. If the unit is to be in MODE 3 with the RTBs closed for > 4 hours this Surveillance must be completed within 4 hours after entry into MODE 3. The Frequency of 92 days is justified in Reference 7. SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7, except it is modified by a Note that this test shall include verification that the P-6, during the Intermediate Range COT, and P-10, during the Power Range COT, interlocks are in their required state for the existing unit condition. The verification is perfonned by visual observation of the permissive status light in the unit control room. _The Frequency is modified by a Note that e allows this surveillance to be satisfied if it has been performed within 92 days of the Frequencies prior to reactor startup and four hours after reducing power below P-10 and (' P-6. The Frequency of " prior to startup" ensures this . surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels. The Frequency of "4 hours after reducing power below P-10" (applicable to intermediate and power range low channels) and "4 hours after reducing power below P-6" (applicable to' source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a delay to perform the testing required by this surveillance. The Frequency of every 92 days thereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P-10 or P-6. The MODE of Applicability for this surveillance is < P-10 for the power range low and intermediata range channels and < P-6 for the source range channels. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-10 or < P-6 for more than 4 hours, then the testing required by this surveillance must be performed prior to the expiration of the 4 hour limit. Four hours is a reasonable time to complete the required testing or place the unit in a MODE where this surveillance is no longer (continued) l Catawba Unit 1 B 3.3-50 Supplement 1

RTS Instrumentation B 3.3.1 (m Q BASES SURVILLANCE SR 3.3.1.6 REQUIREMENTS (continued) SR 3.3.1.6 is a calibration of the excere channels to the incore channels. If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are declared inoperable. This Surveillance is performed to verify the f(AI) input to the overtemperature AT Function and overpower AT Function. A Note modifies SR 3.3.1.6. The Note states that this Surveillance is required only if reactor power is > 75% RTP and that 24 hours is allowed for completing the first surveillance after reaching 75% RTP. The Frequency of 92 EFPD is adequate. It is based on industry operating experience, considering instrument reliability and operating history data for instrument drift. SR 3.3.1.7 C SR 3.3.1.7 is the performance of a COT every 92 days. A COT is performed on each required channel to ensure the channel will perform the intended Function. The tested portion of the loop must trip within the Allowable Values specified in Table 3.3.1-1. The setpoint shall be left set consistent with the assumptions of the setpoint methodology. SR 3.3.1.7 is modified by a Note that provides a 4 hour delay in the requirement to perform this Surveillrze for source range instrumentation when entering MODE T from MODE

2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the RTBs are open and SR 3.3.1.7 is no longer required (continued)

Catawba Unit 2 B 3.3-49 Supplement 1 l

RTS Instrumentation B 3.3.1 BASES 1

l. SURVEILLANCE SR 3.3.1.7 (continued)

REQUIREMENTS l to be perfonned. If the unit is to be in MODE 3 with the RTBs closed for > 4 hours this Surveillance must be completed within 4 hourt after entry into MODE 3. The Frequency of 92 days fs justified in Reference 7. L SR 3.3.1.8 SR 3.3.1.8 is-the performance of a COT as described in SR 3.3.1.7, except it is modified by a Note that this test shall include verification that the P-6, during the Intermediate Range COT, and P-10, during the Power Range COT, interlocks are in their required state for the existing unit condition. The verification is cerformed by visual observation of the permissive status light in the unit control room. The Frequency is modified by a Note that allows this surveillance to be satisfied if it has been performed within 92 days of the Frequencies prior to reactor startup and four hours after reducing pwer below P-10 and D P-6. The Frequency of " prior to startup' ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and pcrer range low instrument channels. The Frequency of "4 Nurs after reducing power below P-10" (applicable to intermediate and power range low channels) and "4 hours after reducing power below P-6" (applicable to source range channers) allows a normal shutdown to be completed and the unit ramoved from the MODE of Applicability for this surveillance without a delay to perform the testing required by this wrveillance. The Frequency of every 92 days thereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P-10 or P-6. The MODE of Applicability for this surveillance is < P-10 for the power range low and intermediate range channels and < P-6 for the source range channels. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-10 or < P-6 for more than 4 hours, then the testing required by this surveillance must be performed prior to the expiration of the 4 hour limit. Four hours is a reasonable time to complete the required testing or place the unit in a MODE where this surveillance is no longier (continued) l Catawba Unit 2 B 3.3-50 Supplement 1

RTS Instrumentation B 3.3.1 BASES r _ SURVEILLANCE SR 3.3.1.8 j ( A Inkm<. hah) [j ha M Pow) REQUIRE 6 5 % CoTj j (Ranp MW.I (continued) SR 3.3.1.8 is the performance of a COT as descri in l SR 3.3.1.7, except it is modified by a Jiote that shall include verification that the P 4,and P 10 nterlocks~ are Frequency in their required state for the s test Q I The is modified by a Note thatexisting allows thisunit condition.(,ryW4',Q surveillance to be satisfied if it has been performed within q Oi 92fdays of the Frequencies prior to reactor startig and PMM $ 1 our hours after reducing power below P 10 and P 6. The Vu,ul 6.74' F of

prior to startig" ansures this surveillance is ,

t g **W per ormed prior to critical operations and aDDlies to the snrce. intemodiate and power range low insfriament g h fa channels. & 4 * + c.4,1 The Frequency of '4 hours after reducing power

                                                                 ~        '

wt** % . J below P 10" (aglicable to intehiediate and power rang low channels) and 4 hours after reducing power below P 6 (applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MXIE of Applicability for this surveillance without a delay to perform the testing required by this surveillance. The Frequency of every 92 asys thereafter applies if the plant russins in the MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P 10 or P 6. The MXIE of' f Applicability for this surveillance is <.P 10 for the power range low and intermediate range channels and < P 6 for the source range channels. Once the unit is in MODE 3. this surveillance is no lon O maintained < P 10 or <ger required. If power is to be P 6 for more than 4 hours, then the testing required by this surveillance must be performed price to the expiration of the 4 hour limit. Four. hours is a reasonable time to maplete the required testing or place the unit in a M where this surveillance is no longer required. This test ensures that the NIS source. intermediate, and power range low channels are OlERABLE prior to taking the reactor critical and after reducing , power. into the applicable MXIE (< P-10 or < P 6) for periods-

                       > 4 hours.

SR 3.3.1.9 SR 3.3.1.9 the performance of a TA00T and is performed every 92 as justified in Reference 7. . (continued. B 3.3 55 Rev 1. 04/07/95 O \ .. .. i . ...i..-...-. .

                                                                                                                           ..}}

ML20216B769

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