ML20217E983

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Proposed Improved Tech Specs Sections 1.0,2.0,3.0 & 4.0
ML20217E983
Person / Time
Site: Catawba  Duke Energy icon.png
Issue date: 04/20/1998
From:
DUKE POWER CO.
To:
Shared Package
ML20217E978 List:
References
NUDOCS 9804280018
Download: ML20217E983 (48)


Text

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3 a

s Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions


NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

IRIE Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux (AFD) signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, display, and trip functions.

l Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessaent of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be perforned by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.

(continued)

Catawba Unit 1 1.1-1 Supplement 3 l

9804280018 900420 PDR ADOCK 05000413 P

PDR L

4 Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions


NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

IgIm Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state and the verification of the required logic output.. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux (AFD) signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, display, and trip functions.

l Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.

i (continued)

Catawba Unit 2 1.1-1 Supplement 3 l

j

Spec.% % l.0 SE AND : :~ ~

  1. ' # M DEFINIT 10NST f The defined terms of this section appear in capitalized ty and are applicable throughout these Technical Specifications ng l3gsg ACTION vt M

h ACTION [shall be that p%nd*r da<iaa=ted conditionart of arbm

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Qepedtfl museres)/bquired es&f)

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ACTUATION LOGIC TtST k

OA.I

@ An ACTUATION LOGIC TEST shall be the appilcation of various simulated3 fnput combinations in conjunction with each possible interlock logic state and

@ verification of the required loaic output. The ACR% TION LOGIC TESTV hall is includeacontinuitychecl(asaminimuTmof output devices. _ f M CHANNEL OPERATIONAL TEST 4,3NRtdii CllANNEL OPFJWWTONAtAST) be the injection of a simu e

a' into th* Channel as Close to lie sensor as practicable to verify 0)ERMILITY.oflalam, interlock and P trip functions. The(AllAt0E CBAN@M dDERMf0NAL4epshall include adju@stment[s as necessary, of theylars, inter-lock andypgetpoints@fhat the etpoints are within the greou, fred range and accuracy.

Q l

avd A.1.

AXIALFLUXDIFFERENCE[WDd (M4 MUX D shall be the difference in nomalized flux signals tretween the top ana potton alves of a two section excore neutron detector.

CHANNEL CALIBRATION CHANLEL CALIBRATION shall be the adjustment, as necessary, of the annel i)that it responds within the required range and accuracy to known ea w s input. The CHANNEL CALIBRATION sha capass the entire channel including theIsenso alam, inter an t p function y be l

m r ormed by ny ser s of sequential, lap g r total channe ps a

that the entire channel is calibrat g i,g,4,)

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o CHANNEL CHECK

@ing operationf,by observatinnT This determinaACHANNELCHFCrthal A.I dur on sha 1 ude, where pos-sible, c arison of the channel indication an status ther indica-tions r status derived from independent instrument channels measuring the same parameter.

6 I

CATAWBA - UNIT 1 1-1 Amendment No. 148 he3l / 0

E wo DEFT

[ M IfFINITIONSI w

-]

M*TE

{6 The fined tems of this section opear in capitalized type and are applicable thringhout these Teclanted Specificationgg ACTION f

ACTI shall be that part of a ITrMED Specification cribes as -murre9/4quiredyunder designated condition. A 9;g.

ACTUATION LOGIC TEST (or ock)

I An ACTUATION LOGIC TEST shall be the application of various simulatedl -

I i

g*g t combinations in conjunction with each possible interlock logic state and I

g ification of the required logic output. TheACTUATIONLOGICTESTYshall i

include a continuity checi as a minimum,7of output devices.

J t

Ad MCHAlelEL OPERATIONAL TEST (Cet)

,, 4 g,g 1.3 mrmau t OPssrATb -=v m TI I be the injection of a si dl signal inuo the channel as close tolbe sensor as practicable to verif OPERABILI'T ofdalars, interlock and(# trip functions. The GNAl arm 3tANN un*==ru= - m u shall include adjustment [s as necessary, of theglars, inter-i lock a

, Trip fetpoints hat the etpointsarewithinthe} required range accuracy.

AXIAL FLUX DIFFERENCE (AFD) -

@ GLXIMa nurFFERfhC be the difference in normalized flux signals

.g between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION

[

A CHAlelEL CALIBRATION shall be the adjustment, as necessary, of the

(!!B c 1 annnthat it respcnds within the required range and accuracy to known A.t f men input. The CHANNEL CALIBRATION siga11 encompass the entire channel

_ c neluding theysenso @ alars, inter an@ trip function @ pay be I

j performes pyyny seriYs of sequential, appin or total channel 1steps 6

t

@e that the lentire channel i ibr I

I TI l% 0 d %#

CHANNEL CHECK

@ A CHANNEL CHECK shall be{the qualitative assessment,ll include, w t f channel behavior during operationCby ooservattorg This determination sha gg

sible, arison of the channel indication an@ stat other indica-1 tions r status derived from independent instrument annels measuring the same parameter.

j CATAWBA - UNIT 2 1-1 Amendment No. 142 pyt !J 15

Discussien of Changes Secticn 1.0 - USE AND APPLICATION l

l ADMINISTRATIVE CHANGES information.

Removal of this limitation does not substantively change the requirements for operation in MODE 1 or MODE 2 since the reactivity threshold is unchanged. This change is considered administrative in nature.

This change is consistent with NUREG-1431.

1 A.22 The definitions of Hot Shutdown and Cold Shutdown in CTS Table 1.2 have been revised to provide clarity, completeness and avoid any potential misinterpretation. Specifically, a new footnote in ITS Table 1.1 stating "all reactor vessel head closure bolts fully tensioned" eliminates a potential overlap in defined MODES.

For example, when the vessel head is detensioned, both the definition of Refueling and Cold Shutdown could apply, dependent on temperature.

It is not the intent of the Technical Specification to allow an option of whether to apply Refueling applicable LCOs or to apply Cold Shutdown applicable LCOs. This change is editorial in nature since the intent of the existing specification is clarified to reflect actual industry practice. This change is consistent with NUREG-1431.

A.23 The definition of REFUELING in CTS Table 1.2 is changed to remove the limit on average reactor coolant temperature in MODE 6.

When the average coolant temperature exceeded 140*F, the CTS could be incorrectly interpreted as not requiring the application of the TS which are applicable when the reactor vessel head bolts are not i

fully tensioned.

By removing the temperature reference, the appropriate LCOs will be applied during this condition.

This change is editorial in nature since the intent of the existing specification is clarified to reflect actual industry practice.

This change is consistent with NUREG-1431.

A.24 The CTS 1.5 definition is revised to include required displays within the scope of a CHANNEL CALIBRATION.

The majority of CTS channels which require a calibration are those that perform trip or actuatton functions and do not require a colibratton of associated displays.

However, CTS 4.3.3.6 requires a calibration of the post accident monitoring channels.

These channels are display only, therefore, the inclusion of required displays within the ITS 1.0 definition of CHANNEL CALIBRATION is on administrative clarification consistent with current requirements and practices.

1 Catawba Units 1 & 2 Page A-77 Supplement 35/20/97l

+

1 4

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions

.....................................N0TE The defined terms of this section a ar in capitalized type and are f.!

.0. !!!!................

Igtm Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations j

in conjunction with each possible interlock logic

{

state and the verification of the required logic j

output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output

]

devices.

AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux (AFD) signalsbetweenthe,dopandbottomhalvesofa I

twosectionexcoreneutrondetectory CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the ired range and accuracy to known input. The L CALIBRATION shall encompass the entire channel. including

  • functions. dred sensor,_ g r

alarm, interlock.Fdis6129.7snd trip Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and. normal calibration of the remaining adjustable devices in the channel.

Whenever a sensing element is replaced, the next l

required CHANNEL CALIBRATION shall include an I

inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be l

performed by means of any series of sequential,

{

overlapping calibrations or total channel steps so that the entire channel is calibrated.

(continued) i HTS 1.1 1 Rev 1, 04/07/95 ca,A

r 8

Definitions 1.1 l

1.1 Definitions (continued)

CHANNEL CECK A CHANNEL CHECK shall be the qualitative assessment, by observation. of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrtment channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (C0T) actual signal into the channel as close to the sensor as practicable to veri he OPERABILITY of required alam, interlock, and trip functions. The COT shall luce adjustments, as necessary, of the required alam. interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel.

sources, or reactivity control w.eiits, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LINITS The COLR is the unit specific' document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be detemined for each reload cycle in accordance with Specification 5.6.5.

ation within these limits is addressed dual Specifications.

DOSE EWIVALENT I 131 DOSE EWI 131 shall be that concentration of I 131 (microcuries/ gram) that alone would produce the same thyroid dose as t.% quantity and isotopic mixture of I-131. I 132. I 133.1 134, and I 135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed irW LSDie II of TID 1 T

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estReaf. orsi sesKr t istlhA t

ormatory mme 1100-

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. or K 30. Supple % to Part 1 (continued) 1.1 2 Rev 1. 04/07/95 ca6 L

S Definitions 1.1 1.1 Definitions SHlfiDOWN MARGIN (SOH)

a. All rod cluster control assemblies (RCCAs) are (continued)

' fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDH; and b.

In MODES I and 2. the fuel and moderator teg eratures are changed to the p inal zero

[

power design level F SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay. The SLAVE RELAY TEST shall include. as a minimum, a continuity check of associated testable actuation devices.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during a Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERNAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip OPERATIONAL TEST actuating device and verifyino the OPERABILITY of (TADOT) required alarm, interlock, i ay and trip functions. The TADOT shall inc adjustment,as necessary, of the trip actuating device so that it m

actuates at the required setpoint within the required accuracy.

1.1 6 Rev 1. 04/07/95 w

o Justification for Deviations Section 1.0 - Use and Application TECHNICAL SPECIFICATIONS NOTE:

The first five justifications for these changes from NUREG-1431 were generically used throughout the individual LC0 section markups. Not all generic justifications are used in each section.

1.

The brackets have been removed and the proper plant specific infonnation or value has been provided, i

l 2.

Editorial change for clarity or for consistency with the Improved Technical Specifications (ITS) Writer's Guide.

I 3.

The requirement / statement has been deleted since it is not applicable to this facility.

The following requirements have been renumbered, where applicable, to raflect this deletion.

1 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.

This change reflects the current licensing basis / technical specifications.

4 6.

This change reflects the NRC model for implementation of option B to 10 CFR 50, Appendix J, enclosed in a letter from C.I. Grimes, NRC, to D.J.

Modeen, NEI, dated November 2, 1995.

7.

Duke Power has not proposed to use a pressure and temperature limits report at this time. The proposed specifications retain the current pressure and temperature limits in a format consistent with NUREG-1431.

8.

The NUREG definitions for Channel Operational Test (C0T) and Trip Actuating Device Operational Test (TADOT) include " displays" as part of the scope cf required testing.

This requirement is not included in the respective ITS definitions, consistent with the CTS.

These tests demonstrate the functional ability of devices which change state in response to a change in a monitored parameter, e.g. interlocks, bistables, and alarms.

A display provides indication only information and performs no " actuation" function, therefore, their inclusion within these tests is inappropriate and inconsistent with current practice. Displays which are required operable by the Technical Specifications (e.g. post accident monitoring indicators) are colibrated to ensure their functional ability to display required information.

Catawba Units 1 and 2 14 Supplement 35/20/97l

Catawba and McGuire Improved TS Review Comments Section 2.0, Safety Limits 2.0-01 JFD 4 Bases Background discussion for Safety Limits STS Bases markup page B 2.0-1 The Bases Background discussion omits the words " steady state" which are used in the STS Bases to describe the peak linear heat generation rate. Comment:

JFD 4 does not adequately justify this omission. Revise the Bases to read

" steady state and transient peak linear heat generation rates." This is more inforrnative than omitting " steady state." Also add a JFD that explains that McGuire and Catawba have both steady state and transient limits.

DEC Response:

The word was deleted because normal operation (condition 1 events) cannot result in center line fuel melt. This discussion is only appropriate for transient operation (condition 2,3, and 4 events). The Bases have been revised to replace the word steady state with transient and JFD 7 has been added to justify the change.

mc3_cr_2.0 J

April 20,1998

(

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref.1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (A00s). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence 1

level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the transient peak linear heat rate (LHR) below l

the level at which fuel centerline melting occurs.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

J Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form.

This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

(continued)

Catawba Unit 1 B 2.0-1 Supplement 3 l

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref.1) requires that specified acceptable fuel I

design limits are not exceeded during steady state

{

operation, normal operational transients, and anticipated operational occurrences (A00s). This is accom having a departure from nucleate boiling (DNB)plished by design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the transient peak linear heat rate (LHR) below l

the level at which fuel centerline melting occurs.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding j

temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction i

results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

(continued)

' Catawba Unit 2 B 2.0-1 Supplement 3 l

)

(

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUW GDC 10 (Ref.1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (A00s). This is accomplished by having a departure from nucleate boiling (DW) design basis, which corresponds to a 95% probability at a 95% conf ldence level (the 95/95 DNB criterion) that DW will not occur and by requiring that fuel cedterline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel h =.

and cladding, as well as possible cladding perforation, that

+

would result in the release of fission products to the

{hswa4 reactor coolant. 0verheating of the fuel is prevented by 1

ma' ntainina the emnrmte peak linear heat rate (L}R) be'ow the level at which fuel centerline melting occurs.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Fuel centerline melting occurs when 'the loifal LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DW and the resultant sharp reduction in heat transfer coefficient, Inside the steam film high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker fom may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

(continued) j!DMrl$

B 2.0 1 Rev 1, 04/07/95 CO S

Justificatien fer Dzviatiens Secticn 2.0 - Safety Limits BASES 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 Specification (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.

Changes have been made to reflect those changes made to the Specification. The following requirements have been renumbered, where applicable, to reflect this change.

6.

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-5.

7.

The Bases Background for NUREG 2.1.1 indicates that overheating of the fuel is prevented by maintaining steady state peak iinear heat rates below centeriine fueI meIt temperature.

The ITS Bases has changed this to the transient peak linear heat rates.

Condition 1 events (steady state operation) can not cause centerline fuel melt.

A more occurate description is that the transient linear heat rates are limited such that a transient event does not result in centerline fuel melt.

Catawba Units 1 and 2 14 of 1 Supplement 35/20/97l

Catawba and McGuire Improved TS Review Comments Section 3.0, LCO and SR Applicability 3.0-01 JFD 9 (Catawba only)

ITS LCO 3.0.4 JFD 9 states that the ITS for Catawba incorporates TSTF-lO4, which moved the discussion of exceptions to LCO 3.0.4 to the Bases. Catawba's STS markup, however, falls to show this change. This appears to be just a markup error because the smooth version of the proposed Catawba ITS does not have the discussion of exceptions and is consistent with TSTF-104. Comment: Correct the STS markup of LCO 3.0.4 in the Catawba submittal.

DEC Response:

The STS markup for Cataba has been revised to show this information deleted in accordance with JFD 9, consistent with the STS markup for McGuire.

Additionally, it was discovered that ITS page 3.0-3 for Catawba was inadvertently omitted from the revised pages in ITS Supplement 1, dated March 9,1998, and has been included in this response. This page was repaginated due to changes on the previous page. There are no changes to the content on this page other than a carryover of information previously located on page 3.0-2. This page was already included in the McGuire supplement 1 package.

3.0-02 DOC L1 CTS 4.0.3 ITS SR 3.0.3 1

DOC L1 does not state why basing the missed surveillance performance allowance on the specified Frequency rather than on the applicable allowed outage time is less restrictive. This change is both more and less restrictive.

DOC M1 adequately explains why it is more restrictive. Comment: Revise DOC L1 with the missing statement.

DEC Response:

This phrase has been added to DOC L.1 and the no significant hazard consideration associated with DOC L.1.

i i

mc3_cr_3.0 1

April 20,1998

LCO Applicability 3.0 3.0 LCO APPLICABILITY (continued)

LCO 3.0.7 unchanged. Compliance with Test Exception LCOs is optional.

(continued)

When a Test Exception LCu is desired to be met but is not met, the ACTIONS of the Test Exception LC0 shall be met.

When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications.

)

l Catawba Unit 1 3.0-3 Supplement 1 l

l LCO Applicability 3.0 3.0 LC0 APPLICABILITY (continued)

LCO 3.0.7 unchanged. Compliance with Test Exception LCOs is optional.

(continued)

When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LC0 shall be met.

When a Test Exception LC0 is not desired to be met, entry I

into a MODE or other specified condition in the i

Applicability shall be made in accordance with the other applicable Specifications.

l I

l l

l l

l l

1 l

Catawba Unit.2 3.0-3 Supplement 1 l

l LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 Specification shall not prevent changes in H0 DES or other (continued) specified conditions in the Applicability that are required ly with ACTIONS @nat arypart ui e nuwown of tnh g

~

Exceptions to this Specification am emw e tM J ndividual Specifications.] These except allow entry

]

' into Hum or ner spectrTed conditi in the r

Applicabilit when the associated ACT to be entered allow unit ration in the MODE or her specified condition n the Applicability on1 for a limited period of time.

i C

i LC0 3.0.4 is ly applicable for entr into a H00E or others specified ition in the Applicabi ty in H00ES 1, 2, 3, and 4.

Revi s*s Note: LCO 3.0.4 ha n revised so that chan in H0 DES or other spec fed conditions in the d'

App cability that are part a shutdown of the unit shall be prevented. In addit

n. LCO 3.0.4 has been revised that it is only applic e for entry into a H00E or ot r specified condition in t Applicability in N0 DES 1. 2.

and 4.

The MODE cha restrictions in LCO 3.0.4 wer

/

previously applicable n all N0 DES. Before this ver on of LCO 3.0.4 can be i esented on a plant specific is, tre licensee must rev the existing technical s 1 cations to determine e specific restrictions on changes or Required Acti should be included in indiv ual LCOs to justify this ange: such an evaluation s d be summarized in a matri of all existing LCOs to faci ate NRC staff review of a conversion to the STS.

t

-J 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 testing required to demonstrate OPERABILITY.

(continued)

W9G4Ti 3.0 2 Rev 1, 04/07/95

& U.n-

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Discussien of Chang s 52cticn 3.0 - LCO and SR Applicability TECHNICAL CHANGES - LESS RESTRICTIVE applicable if the specified Frequency is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

1 l

The second and third paragraphs of ITS SR 3.0.3 are added to clearly state the actions to take if the Surveillance is not performed within the delay period or the surveillance fails when performed.

This clarification will help avoid confusion as to when the Completion Time (s) of the Required Action (s) begin in various situations.

This change is less restrictive since it allows a grace period to perform a missed survetIlance before entering the required actions regardless of the completion time.

This change is cons' stent with NUREG-1431.

L.2 ITS LC0 3.0.5 was added to establish the allowance of restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS.

The purpose of this Specification is to provide an exception to LC0 3.0.2 to allow the performance of Surveillance Requirements to demonstrate the OPERABILITY of the equipment being returned to service or to demonstrate the OPERABILITY of other equipment that otherwise could not be performed without returning the equipment to service. This LC0 is necessary to establish a concept that although utilized, is not formally recognized in the present Technical Specifications. Without this concept many Surveillance Requirements in Technical Specifications could not be performed and various equipment would not be able to be restored to OPERABLE status, and still other equipment would not be able to be maintained OPERABLE. This change is consistent with NUREG-1431.

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Catawba Units 1 and 2 Page L - 23 Supplement 35/20/97l l

A Na Significant Htzards Csnsid2ratien Szction 3.0 - LC0 and SR Applicability ACTIONS that have more than one Completion Time. The confusion associated with the application of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> deferral to the Completion Times of this example's Required Actions illustrates the potential for misapplication throughout the Technical Specifications.

In addition the limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not applicable if the specified Frequency is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The second and third paragraphs of ITS SR 3.0.3 are added to clearly state the actions to take if the Surveillance is not performed within the delay period or the surveillance fails when performed. This clarification will help avoid confusion as to when the Completion Time (s) of the Required Action (s) begin in various situations.

This change is less restrictive since it allows a grace period to perform a missed surveillance before entering the required actions regardless of the completion time.

This change is consistent with NUREG-1431.

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?

1 The change does not result in any hardware or operating procedure changes. The Surveillance Frequencies are not assumed to be the initiator of any analyzed event. This change will allow delaying the entry into the Required Actions for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when a Surveillance Requirement has not been performed within the l

requirements of proposed SR 3.0.2.

It is overly conservative to assume that systems or components are inoperable when a Surveillance Requirement has not been performed.

In fact, the opposite is the case; the vast majority of Surveillance Requirements performed demonstrate that systems or components are operabl e.

When a Surveillance Requirement is not performed within i

the requirements of SR 3.0.2, it is primarily a question of operability that has not been verified by the performance of the Surveillance Requirement.

The probability of accidents previously evaluated is not significantly increased since the impact of the small increase in time that an inoperable component will go undetected is' minimal.

In addition the consequences of previously Catawba Units 1 and 2 Page 67 of 11M Supplement 35/20/97l l

N3 Significant Htzards Ccnsideration cticn 3.0 - LCO and SR Applicability evaluated accidents are not increased since accident mitigating equipment will continue to perform their intended safety function.

2.

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

The possibility of a new or different kind of accident from any accident previously evaluated is not created because the proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant.

3.

Does this change involve a significant reduction in the margin of safety?

The increased time allowed for the performance of a Surveillance Requirement discovered to have not been performed within the requirements of SR 3.0.2 is acceptable based on the small probability of an event requiring the associated component and the low probability that the surveillance will not be completed satisfactorily. The requested allowance will provide sufficient time to perform the missed Surveillance in an orderly manner.

Without the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> delay, it is possible that the missed Surveillance would force an unnecessary plant shutdown; imposing a transient on plant systems. As such, any reduction in the margin of safety will be insignificant and offset by the benefit gained in plant safety due to avoidance of unnecessary plant transients and shutdowns.

Catawba Units 1 and 2 Page 77 of 1144 Supplement 35/20/07l 2

Catawba and McGuire improved TS Review Comments Section 4.0, Design Features 4.0-01 Catawba McGuire DOC A3 DOC M1 JFD 5 JFD 5 CI'S 5.3.1 ITS 4.3.1.1.a ITS 4.3.1.1.a CTS Table 3.9-1 CTS Table 3.9-1 ITS Table 3.7.16-1 ITS Table 3.7.15-1 Catawba CTS 5.3.1 limits reload fuel to a maximum nominal enrichment of fe.O weight percent U-235 with a maximum tolerance of.05 weight percent U-235.

In a telcon on 3/12/98 with DEC, it was stated that the tolerance in this statement is what is implied by the word " nominal." It was pointed out that use of the word nominal is a deviation from the STS, but that it is consistent with the language in the CTS 3/4.9.13. " Spent Fuel Assembly Storage." Thus use of

" nomina. in ITS 4.3.1.1.a is acceptable. The removal of the information explaining the tolerance associated with the allowcd maximum nominal enrichment, however, should be justified by an IA-type DOC. It was suggested that the Bases for ITS 3.7.16 for Catawba (3.7.15 for McGuire) would be an j

approoriate location for this information. It was noted that the paragraph at the ent' f CTS Table 3.9-1, for both sites, omits the word " nominal," although it okpa to be included for consistency. This table for both units is retained in ITS l

Section 3.7, and ought to be made consistent. Comment: Revise the Catawba 4

submittal with an IA-type DOC to justify removal of the tolerance information to the Bases. Modify the paragraph in Catawba ITS Table 3.7.16-1 and McGuire ITS Table 3.7.15-1 by adding the word nominal and revise the ITS Bases to explain what nominal means in terms of tolerance.

DEC Response:

DOC LA.5 has been written to show the movement of the Catawba 5.3.1 maximum tolerance information to the Bases for ITS 3.7. The paragraph at the bottom of CI'S Table 3.9-1 and corresponding ITS 3.7 figure for McGuire and Catawba has also been revised to include the word nominal consistent with the labeling on the included figure. The Bases for ITS 3.7.15 and 3.7.16 for Catawba and 3.7.14 and 3.7.15 for McGuire have been revised to include the tolerance information.

j mc3_cr_4.0 J

April 20,1998

Spent Fuel Assembly Storage 3.7.16 Table 3.7.16-1 Minimum Qualifying Burnup Versus Initial Enrichment for Unrestricted Storage Initial Nominal Enrichment Assembly Burnup (Weicht% U-235)

(GWD/MTU) 4.05 (or less) 0 4.50 2.73 5.00 5.67 10-S Ei 8

D 6

UNRESTRICTED STORAGE c8 4

-0?h2 RESTRICTED e

STORAGE O

4.00 4.25 4.50 4.75 5.00 Initial Nominal Enrichment (Weight % U-23 NOTES:

Fuel which differs from those designs used to determine the requirements of Table 3.7.16-1 may be qualified for Unrestricted storage by means of an analysis using NRC approved methodology to assure that keff is less than or equal to 0.95.

Likewise, previously unanalyzed fuel up to a nominal 5.0 l

weight % U-235 may be qualified for Restricted storage by means of an analysis using NRC approved methodology to assure that k,ff is less than or equal to 0.95.

Catawba Unit 1 3.7-35 Supplement 3 l

Spent Fuel Assembly Storage 3.7.16 Table 3.7.16-1 Minimum Qualifying Burnup Versus Initial Enrichment for Unrestricted Storage Initial Nominal Enrichment Assembly Burnup (Weicht% U-235)

(GWD/MTU) 4.05 (or less) 0 4.50 2.73 5.00 5.67 10 Sg 8

h 6

UNRESTRICTED STORAGE c5 4

2$2 RESTRICTED 2

STORAGE O

4.00 4.25 4.50 4.75 5.00 Initial Nominal Enrichment (Weight % U-23 NOTES:

Fuel which differs from those designs used to determine the requirements of Table 3.7.16-1 may be qualified for Unrestricted storage by means of an analysis using NRC approved methodology to assure that k,ff is less than or equal to 0.95.

Likewise, previously unanalyzed fuel up to a nominal 5.0 l

weight % U-235 may be qualified for Restricted storage by means of an analysis f

using NRC approved methodology to assure that k,ff is less than or equal to 0.95.

I Catawba Unit 2 3.7-35 Supplement 3 l

Spent Fuel Pool Boron Concentration B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Pool Boron Concentration BASES BACKGROUND The spent fuel storage rack (Ref. 1) is limited to a capacity of 1418 fuel assemblies. The s>ent fuel storage rack is designed to accommodate fuel witi a maximum nominal enrichment of 5.0 wt% U-235 (maximum tolerance of 10.05 wt%) which have accumulated minimum burnups greater than or equal to the minimum qualifying burnups in Table 3.7.16-1.

Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Figure 3.7.16-1.

The water in the spent fuel pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting k,ff of 0.95 be evaluated in the absence of soluble boron. Hence, the design of the spent fuel storage racks is based on the use of unborated water, which maintains the spent fuel pool in a subcritical condition during nomal operation when fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 2) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For exam)1e, the most severe accident scenario is associated with t1e accidental misloading of a fuel assembly. This could potentially increase the reactivity of the spent fuel pool.

To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the spent fuel pool storage rack with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with LC0 3.7.16 " Spent Fuel Assembly Storage." Prior to movement of an assembly, it is necessary to perform SR 3.7.15.1.

APPLICABLE Most accident conditions do not result in an increase in SAFETY ANALYSES the reactivity of the spent fuel storage rack. An example of these accident conditions is the dropping of a fuel assembly on the top of the rack. However, accidents can be postulated that could increase the reactivity. This (continued)

Catawba Unit 1 B 3.7-75 Supplement 3 l

1

i Spent Fuel Assembly Storage B 3.7.16 j

B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage l

BASES BACKGROUND The spent fuel storage rack (Ref. 1) is limited to a capacity of 1418 fuel assemblies. The spent fuel storage rack is designed to accommodate fuel with a maximum nominal enrichment of 5.0 wt% U-235 (maximum tolerance of 0.05 wt%) which have accumulated minimum burnups greater than or equal to the minimum qualifying burnups in Table 3.7.16-1.

Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Figure 3.7.16-1.

The water in the spent fuel pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting k,ff of 0.95 be evaluated in the absence of soluble l

boron. Hence, the design of the spent fuel storage racks is based on the use of unborated water, which maintains the spent fuel pool in a subcritical condition during normal operation when fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 2) allows credit for soluble boron under other abnonnal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the accidental misloading of a fuel assembly. This could potentially increase the reactivity of the spent fuel pool.

To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the spent fuel pool storage rack with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with the accompanying LCO.

Prior to movement of an assembly, it is necessary to perfonn SR 3.7.15.1.

(continued) l Catawba Unit 1 B 3.7-78 Supplement 3

Spent Funi Pool Boron Concentration B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Pool Boron Concentration BASES BACKGROUND The spent fuel storage rack (Ref. 1) is limited to a capacity of 1418 fuel assemblies. The s:ent fuel storage rack is designed to accommodate fuel witi a maximum nominal enrichment of 5.0 wt% U-235 (maximum tolerance of 10.05 wt%) which have accumulated minimum burnups greater than or equal to the minimum qualifying burnups in Table 3.7.16-1.

fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Figure 3.7.16-1.

The water in the spent fuel pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting k,ff of 0.95 be evaluated in the absence of soluble i

baron. Hence, the design of the spent fuel storage racks is based on the use of unborated water, which maintains the spent fuel pool in a subcritical condition during normal operation when fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 2) allows credit for soluble boron under other abnomal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the i

accidental misloading of a fuel assembly. This could potentially increase the reactivity of the spent fuel pool.

To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the spent fuel pool storage rack with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with LC0 3.7.16, " Spent Fuel i

Assembly Storage." Prior to movement of an assembly, it is necessary to perfom SR 3.7.15.1.

i APPLICABLE Most accident conditions do not result in an increase in SAFETY ANALYSES the reactivity of the spent fuel storage rack. An example of these accident conditions is the dropping of a fuel assembly on the top of the rack. However, accidents can be postulated that could increase the reactivity. This (continued)

Catawba Unit 2 B 3.7-75 Supplement 3 l

Spent Fuel Assembly Storage B 3.7.16 B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage BASES BACKGROUND The spent fuel storage rack (Ref. 1) is limited to a capacity of 1418 fuel assemblies. The s)ent fuel storage rack is designed to accommodate fuel wit 1 a maximum nominal enrichment of 5.0 wt% U-235 (maximum tolerance of 0.05 wt%) which have accumulated minimum burnups greater than or equal to the minimum qualifying burnups in Table 3.7.16-1.

Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Figure 3.7.16-1.

The water in the spent fuel pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting k fr of 0.95 be evaluated in the absence of soluble boron. Hence, the design of the spent fuel storage racks is based on the use of unborated water, which maintains.the spent fuel pool in a subcritical condition during nonnal operation when fully loaded. The double contingency-principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 2) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the accidental misloading of a fuel. assembly. This could potentially increase the reactivity of the spent fuel pool.

To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the spent fuel pool storage rack with no movement of assemblies may therefore be achieved by controlling the location of each assembly in.accordance with the accompanying LCO.

Prior to movement of an assembly, it is necessary to perfonn SR 3.7.15.1.

i (continued) l Catawba Unit 2 B 3.7-78 Supplement 3 4

1 Specification 3.7.16 Minimum Oualifyina Burnuo Versus initial Enrichment for Unrestricted Storace Initial Nominal Enrichment Assembly Burnup (Weicht% U-235)

(GWO/NTU) 4.05 (or less) 0 4.50 2.73 5.00 5.67 10 -

h8^

lif 6-UNRESTRICTED E

STORAGE n

g4 9

$2 RESTRICTED 4

STORAGE O

-1 4.00 4.25 4.50 4.75 6.00 l

Initial Nominal Enrichment (Weight % U-235) 1 c.

I h

7.14 J

Fuel which differs from those designs used to determine the requirements of Table 3

-1 may be qualified for Unrestricted storage by means of an analysis I

using NRC approved methodology to assure that k.,, is les han or equal to l

0.95.

Ca mdw Likewise, previously unanalyzed fuel up to45.0 weight % U-235 may be qualified for Restricted storage by means of an analysis using NRC approved methodology to assure that k,g is less than or equal to 0.95.

CATAWBA - UNIT 1 3/4 9-19 Amendment No. 148 l

Pagelof1/-

l l

l

h 3pd%

3.7.16.

7 Table 3

__1 Minimum Oualifyina Burnuo Versus Initial Enrichment for Unrestricted Storace Initial Nominal Enrichment (Weicht% U-235)

Assembly Burnup (GWD/MTU) 4.05 (or less) 0 4.50 2.73 5.00 5.67 10 -

8-g 6-UNRESTRICTED STORAGE d..

dii 4 5

5 2

RESTRICTED E

STORAGE O

4.00 4.25 4.50 4.75 5.00 Initial Nominal Ervichment (Weight % U 235)

@f

,.w Fuel which differs from those designs used to determine the requirements of Table 3.5-may be qualified for Unrestricted storage by means of an analysis using NRC approved methodology to assure that k,,is_less n or equal to 0.95.

a whD Likewise, previously unanalyzed fuel up tod5.0 weight % U-235 may be qualified t

for Restricted storage by means of an analysis using NRC approved methodology j

to assure that k is less than or equal to 0.95.

g CATAWBA - UNIT 2 3/4 9-19 Amendment No. 142 Pay 2 J 4

l 1

INSERT Table 3.7.16-1 Minimum Qualifying Burnup Versus Initial Enrichment for Unrestricted Storage Initial Nominal Enrichment Assembly Burnup (Weicht% U-235)

(GWD/MTU) 3 4.05 (or less) 0 4.50 2.73 5.00 5.67 10-g 4

Ei 8

D 6

UNRESTRICTED STORAGE c8 4

2$2 RESTRICTED q

J2 STORAGE 1

0 4.00 4.25 4.50 4.75 5.00 Initial Nominal Enrichment (Weight % U-23 NOTES:

Fuel which differs from those designs used to determine the requirements of Table 3.7.16-1 may be qualified for Unrestricted storage by means of an analysis using NRC approved methodology to assure that k,ff is less than or l equal to 0.95.

Likewise, previously unanalyzed fuel up to a nominal 5.0 weight % U-235 may be qualified for Restricted storage by means of an analysis i

using NRC approved methodology to assure that k,ff is less than or equal to 0.95.

INSERT Page 3.7-39a Catawba

INSERT -

The s)ent fuel storage rack (Ref.1) is limited to a capacity of 1418 fuel assem) lies. The spent fuel storage rack is designed to accommodate fuel with a maximum nominal enrichment of 5.0 wt% U-235 (maximum tolerance of 10.05 wt%) which have accumulated minimum burnups greater than or equal to the minimum qualifying burnups in Table 3.7.16-1.

Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Figure 3.7.16-1.

i 1

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INSERT Page B 3.7-81 Catawba

INSERT i

The spent fuel storage rack (Ref.1) is limited to a capacity of 1418 fuel assemblies. The spent fuel storage rack is designed to accommodate fuel with a maximum nominal enrichment of 5.0 wt% U-235 (maximum tolerance of i 0.05 wt%) which have accumule.ted minimum burnups greater than or equal to the minimum qualifying burnups in Table 3.7.16-1.

Fuel assemblies not meeting the j

criteria of Table 3.7.16-1 shall be stored in accordance with Figure 3.7.16-1.

l l

INSERT Page B 3.7-85 l

. Catawba

Sce: lr:.amw %0 j

DESIGN FEATURES

[5.2 CONTAIletENT

~~C------=

=

CONFIGURATION 5.2.1 The c tainment struc re is comprised a steel containee vessel surrounded y a concrete co ainment having t following design eatures:

a.

Containment Vess

1) Nominal side diameter =

5 feet.

2,)

Nominal nside height = 1 feet.

3) Nomin thickness of ve el walls = 0.75 inc 4J Nomi i thickness of v sel done = 0.6875 i h.

[

51 1 thickness of essegbottom=0.25 ch.

6!)

free volume = 1 x 10 cubic feet.

b.

Rea or Buildino Nominal Annul space = 6 feet.

n Annulus nomi 1 volume = 484,090 e ic feet.

3 Nominal out deheight(topoffo dation base to to of done) = 17 feet.

4 ) Nominal i side diameter = 127 f t.

5 f Minimum ylinder wall thicknes = 3 feet.

6 ) Nini done thickness = 2.25 feet.

7J Dome side radius = 87 fee DESIGN PRESSURE TEMPERATURE 5.2.2 The re tor containment vess is designed and s 11 be maintained for (amaximumi ernal pressure of 15 ig and a temperat e of 328'F.

/Ch REACTOR CORE L ASSDSLIES S'

I 5.

1 The reactor shall contain 193 fuel ass lies. Each assembly hall i

consist of a matrix of gE3Midricabzircaloy fuel rods with nitial composition of natural or slightly enriched uranium dioxide s fuel material.

Limited substitutions of zirco alloy or stainless steel filler rods for fuel rods, in accordance with i

configurations, may be used. ( ue approved applications of fuel rod assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A Ilmited number of lead test assemblies that have not completed e

representative testing may be olaced in non-limitin't core regions.jRelinitial l;e d fuel l (sita11 De s g* 3,(a maxi- _ yer in physic 4rdesign togt aa*=av1J-23 1

_inal enrichgefit of 5.0 we 3

telera y or s.Un we y y m m u-)l557

_1

\\

CATAW8A - UNIT 1 5-6 Amendment No. 148 Pye. '7 of //

900' b CA kon t/, o DESIGN FEATURES f.2 CONTAINMENT 2 CONFIGURATI I

5.2.1 Th containment tructure is omprised of steel contain ntvessel]

surroun d by a coner e contain t having the ollowing desig features:

~

. Containmen Vessel 11 Nom' al inside f aaeter = 115 eet.

inal inst height = 171 et.

2)))

inal thi ness of vessel alls = 0.75 in 3

L g,p 4

ominal th'ckness of vess done = 0.6875 ' ch.

Sp Nominal ickness of ves gbottom=0.25 inch.

Met fre volume = 1.2 10 cubic feet.

b.

Reactor B 1dina j

2p) 1 inal Annular s ce = 6 feet.

nulus nominal olume = 484,090 ubic feet.

3) ominal outsid height (topof undation base top of done) = 177 f et.

Nominal ins e diameter = 127 feet.

i h Minimum cy inder wall thick ss = 3 feet.

6h Minimum thickness =

5 feet.

7) Dome i de radius = 87 et.

DE. GM PRESSURE TEMPERATURE 2.2 The tor containment v ssel is designe and shall be a ntained for a maximum i rnal pressure of 5 psig and a t rature of 328F.

h REACTOR CORE j

FUEL ASSENBLIES (llON A,/

J<t The reactor shall contain 193 fuel as tes. Each assembly shall co st of a matrix ofc15Matfcka )zircaloy uel ith an initial l

composition of natural or slightly enriched um dioxid as fuel material.

Limited substitutions of zircon alloy or stainless steel filler rods for fuel rods, in accordance with pproved applications of fuel rod corfigurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not comp 1_eitIL representative testine may be placed in non-limiting core regions.

f R ad fuel )

(shall be yslar in physi design to th Alttal co

_aG)ng an(

all have )

A.

[a nawi-nominal enri nt of 5.0 t per m * '

"*t tn r, maximum,

)

tolepe of *.05pght percent,W6ZJh.f CAT M - UNIT 2 5-6 Amendment No. 142 pge '7 o ! //

CNS ITS Ccnysrsien - Discussion of Changes Section 4.0 - Design Features TECHNICAL CHANGES - REMOVAL OF DETAIL LA.4 CTS 5.5 describing the location of the meteorological tower has been relocated to the UFSAR.

10 CFR 50.36 (c)(4) states that Design Features are those features.such as material of construction and geometric arrangements which, if altered or modified, would have a significant effect on safety and are not covered in other TS sections. The location of the meteorological tower does not meet these requirements. Changes to the UFSAR are controlled by 10 CFR 50.59. The 10 CFR 50.59 evaluation ensures that changes to this requirement will not have any adverse impact on the safe operation of the plant. This change is consistent with NUREG-1431.

LA.5 CTS 5.3.1 requires that reload fuel have a maximum nominal enrichment of 5 wt% U-235 and defines the tolerance associated with the nominal enrichment as 1 0.05 wth U-235.

ITS 4.3.1.1.0 and 4.3.1.2.a retain the nominal enrichment requirement but does not include the tolerance.

The tolerance information is not necessary for inclusion within the technical specifications and is adequately controlled by industry and NRC accepted measurement standards.

The specific tolerance is moved to the Bases for ITS 3.7.15 and 3.7.16.

Changes to the Bases are controlled in accordance with the controls in ITS Chapter 5.0, " Administrative Controls," and 10 CFR 50.59 which ensure that any changes are appropriately reviewed.

1 Catawba Units 1 and 2 Page LA - 23 Supplement 35/20/97l

l l

l 1

i ENCLOSURE 2 WITHDRAWAL OF DEFINITION CHANGE AND REVISED ATTACHMENTS 5 AND 6 4

i

Definitions 1.1 1.1 Cefinitions (continued)

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel

{

indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (C0T) actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, and trip functions.

The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

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

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle.

These cycle specific l

parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5.

. Unit operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and j

isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, AEC, 1962, " Calculation of Distance Factors for Power and Test Reactor Sites."

J (continued) l Catawba. Unit 1 1.1-2 Supplement 3

Definitions 1.1 1.1 Definitions (continued)

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (C0T) actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, and trip functions.

The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, j

sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

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

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5.

Unit operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132. I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, AEC, 1962, " Calculation of Distance Factors for Power and Test Reactor Sites."

(continued) l l Catawba Unit 2 1.1-2 Supplement 3

  • 5 h be, l. o DEFINITIONS I P-30,Surp Fbh 1, p.

' D hT M" N %M DOSE EQUIVALENT I-131 pg

? ](ryjo,. s *Tm.

l A,/ (O DOSE EQUIVALENT I-131 shall be that concentration of a-ui

{ q(microcuriefgram) c alone would produce the same thyroid dose as the

}

uantity and isotopic mixture of I-131 I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calcula-l tion shall be those listedgjale III pr IID-leapg,fcalpGlation op ptstanc E t arraM Mt" Passer asM leSt Reattor SitaL "r

.I I

hcoche9, M(sap i - AVERAGE DISINTEGRATION ENERGY I

e

.,a

@ l shall be the averha e

[weighte in p,vvvrtion to the concentration of M

each radionuclide in the E cf the sum of the average beta and gamma energies per disinte ra on io. for 3:anionuanaes w sne sanom.

A.

i N'

, ENGINEERED SAFETY FEA RESPONSE TIME

+

I ab

( Q The h(ESFsf RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF,4ctuation getpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where appitcable.

h rFRE00ENCTIOTATION NT 1

(

FREQUENCYNOTApT specified fo/the performance A SurvetIIance) 1.14 s.

Jten aaad to the int /rvals defined in Aable 1.1 ramants shall carra fIDENTIFI5E LEAKAGE l.

N)ENTIEIED LEAKAGE shall be:

t h [ seal or valve oackina N h

@ that G u ear iexcasrr umilioti r1 LEAKAGFffgcMdstne]s e

suc l

captured ano conducted toya sump 4 F(.eace.gv <oic.h, wt.,y or collecting tang Q!)

Ce, w.,4.o.c3 em7 (sc.n v.1 wake p.g.*' q,(,q 8

@r. L Oka into the containment atmosphere from sources that are both specifically located apd known either not to interfere _ with_the

(

operation of geakage petectiongystems or not to be4 PRES 5URE 80UNDARj LEAKAG@or N

(pt. Reactor Coolant I

through a steam gener o the (am 3 Secondary m System

"^8 I

c4.k#egs3 i

A MASTER RELAY TEST shall ~ uw'eneramrtan/n1 each master relay and OA,1?

emf 1 cation iOOPERA81LITY of each relay. The MASTER RELAY TEST shall d

include a continuity check of each associated slave relay.

CATAWBA - UNIT 2 1-3 Amendment No. 142 9dO po7

1 i

6fus %

i.o n

i DEFINITIONS (C R P-Suff w & h hrVl, l

1 DOSE EOUIVALENT I-131

!j g

7-x 44 4 '

DOSE EQUIVAL

-131 shall be that concentration of I-131 1

s icrocurieYgram) m c alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133,1-134, and I-135 actually present. The thyroid dose conversion factors used far' this calcula-I-

tion shall be those If sted_ #1Ta 111 of JfD-14844.a*C#r"1=H= M stan (Facto pfor Powerprd Test Re or Sitesyj h 61C l%

r y

E - AVERAGE DISINTEGRATION ENERGY M

@ l shall be the avera e ' weighted in proportion to the concentration of rod each radionuclide in th of the sum of the average beta and aansa f7M'r

,3 energies per dis egra V/d) for#Gw r M anuel W < in t V sans B.

h p

e 'Mah'N[

af.d.

ENGINEERED SAFETY TEATURESIRESPONSE TINE y

OM @ interval from when the monitorea parameteY exceeds its ES The [N6 #EE M FETY F_f,NI MTJRFjhRESPONSETIMEshallbethattime at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

( FRE0VENCY NO (in*

1.14 e FREQUENCY NOTAT specified for the fonsance of Su 11ance 9

Re rements shall co pond to the intery defined in Table

.1

% IFfED' LEAKAGE IDhFIED) LEAKAGE shMai I

$.. w.~ -

G eaka_FrexceminGMfDLEAKAGE @ f M ed-e BAe St such as pump f$,g a sw or collect' ng tan tgy@that G ptured and conducted } sump

,gg gp seat $or valve Dac cinev(

T g.,, % g g 5 j g

,,sube

.' daka e into the containment atmosphere from sources that are both V3 spectf1cally located and known either not to interfere with the

~

I operation of geakagegetectiongystems or not to befPRESSURE BOUNDARY [

LEAKAGE, or

[akaaeIt!. rough a steam generator to the Reactor Coolant Sv

(/N537 3 Condary(CQS46fft/ System /g.

MASTER RELAY TEST y,p og 61 M A MASTER RELAY TEST shall a tha-en 'ra b M each master relay and e

fr1= =meriDOPERA81LITY of each relay. 'The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

CATAWBA - UNIT 1 1-3 Amendment No. 148 P

4 / I3 y

I i

\\

l.

Discussicn of Changes l

Section 1.0 - USE AND APPLICATION 1

TECHNICAL CHANGES - LESS RESTRICTIVE L.1 CTS 1.9 definition has been revised to remove the " manipulation of any component within the reactor pressure vessel" from consideration as a CORE ALTERATION.

This change maintains CORE I

ALTERATIONS as movement of only those components which can affect core reactivity.

The basis for this is evident in that the j

Specifications applicable during CORE ALTERATIONS are those that protect from or mitigate a reactivity excursion event.

In keeping with this, ITS Specification 1.1 provides that movement of j

equipment other than fuel, sources, or reactivity control components, are not considered CORE ALTERATIONS.

Since other equipment (e.g. cameras, thimble plugs, upper internals) will have negligible (if any) effect on core reactivity, any movement has essentially no impact on core reactivity. Therefore, the revised definition places no restrictions on movement of equipment other than fuel, sources, and reactivity control components. Source range instrumentation is available for monitoring core reactivity and boron concentration is maintained within COLR limits during MODE 6.

This change is less restrictive and is consistent with NUREG-1431.

1 L.2 Not used.P.T.C A n #. 4..n. 4. +. 4. m m.4,,,

r a a + J. a n.

f..a m. n.n.c. E Enk.

U..A. f E M. T.

i. S.i 1 11 T

-s Specific; th^ thyroid de:^ conversion facters, c cd for a.% 1. s. 1..s + 4. m.m. e.,, L..% 1 1, a.

+ L. a r. a 1 4., 4. a A 4.m.. T..s k i. a

.a f, T. i.n 1 A. 0 A,A,

La 777

" Calculation of Distance Factor; for "cwcr and Tc;t "cactor C 4. +. n,

Tn.

,,nAs+a

4. L. w m a 4. k. n A.a l a n s. La4.mn m e n A. +a. namf.a-m

+La Y 191 a

s.

.yuu.s inw

.vu svyj

-w siy u w yws i sii

.iiw equ: valence calculation; to reflect re recent mcdeling,

acil as to cr,;ure consistency betwc^n current Offsite dc ^ calculation practices, plant ;;mpling, and cour, ting practices, this change i; prcpo cd to switch the standard used for conversion facter; from T.Th-1.,A D A A.
4. a T P n.n Sn,
c. o m a l mm.a..n +

n.%

+

1.,

.%na i n. S T.., k l. a am

.a 010

.rr m.....

r3 m.m, titled "Ccmmitted Dc:c Equ!v;1cnt in Target Organ: cr Tissue: per 4.

+,ca a s. n _. 4. +.,n a. 4.,. 4. +.,...-

c a n. 4. a._.

1.

1.,

n a s. 4.

4. +. 4., _. -

nner.

,Te

<a_

....m.

m.

EQUI"ALENT I 131 ill not, require calculaticn: to use the dc c conversion factor; cf ICR" 30.

Thi; i; acceptable because Regulatory Cuide 1.4, ahich i; used : a guidance document for de:c calculation; suggest; u;c cf dose conversion factor; from onA%+ar m,, ~a + % k. a 1 4. e LP n,n O.

YPnn S n.

comam.e. n A a, T e n.n..

O-s n.. A

+ L. m a

m

..m.

..rm m,_,,

.r.

modeling and physiologic dat and thereferc cheuld be referred to

the ;0urcc-documcr.t for dose corversion; facters.

This change 4,,

.a_,4.,.+.a_+

-,, n, u.n. n E P

1.,A S 1 l

4+

mm l

t i

i Catawba Units 1 & 2 Page L - 11 Supplement 35/20/97l 1

e Definitions 1.1 1.1 Definitions (continued)

CHANNEL CECK A CHANNEL CHECK shall be the qualitative assessment, by observation, or channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (C0T) actual signal into the channel as close to the sensor as practicable to veri the OPERABILITY of required alarm, interlock, and trip functions. The COT shall luce adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

1 CORE ALTERATION CORE ALTERATION shall be the movement of any fuel.

sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of sovement of a component to a safe position.

CORE OPERATIIE LIMITS The COLR is the unit specific document that REPORT (C0tR) provides cycle specific parameter limits for the current reload cycle. These cycle specific i

parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. (EL ation within these limits is 1

4 addressed -

idual Specifications.

DOSE EQUIVALENT I 131 DOSE EQUI 131 shall be that concentration of I 131 (microcuries/ gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I 131. I 132. I 133,1134, and I 135 actually present. The thyroid dose conversion factors used for this calculation shall be_those listed irHti Die II of TID-1 T

SffT g (AEC.

, /Laicula on of istan

  • N + c fc r A P

and/ Test Rea or Si s,Mtir thos isth t

oTeequeatory nume 1 109.

v.

- /

C.

, or 30 Supple to Part I N age _

(continued)

W MT 1.1 2 Rev 1, 04/07/95 c6k

1 Definitions 1.1 1

1.1 Definitions

)

m -

ble b tle /. [ommitte_d foseT/

DOSE EQUIVALENT I 131 F192 212.

(continued)

Eaulval t in Taraaf Oraans or Wssues per Jdtakd Oi L a# uni Ar+4vitv F E-AVERAGE lshallbetheaverage(weightedinproportionto DISINTEGRATION ENERGY the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and energies per Q @

disintegration (in hew) opes, otner than iodines, with half lives minutes, making up i

at least 95% of the total ine activity in the coolant.

FNGINEERED SAFETY The ESF RESPONSE TIME shall be that time FEATURE (ESF) RESPONSE interval from when the monitored parameter TIE exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pum pressures reach their required values, p discharge etc.).

Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

L, The max' allowable imary contai leakage rate.

. shall be [

of primary ainment air wei per day at t calculated pe containment ore sure (P,). r LEAKAGE LEAKAGE shall be:

a.

Identified LEAQE 1.

LEAKAGE. such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff),

that is captured and conducted to collection systems or a sump or collecting tank:

2.

LIAKAGEintothecontai$mentatmosphere i

'from sources that are both specifically located and known either not to interfere with the operation of leakage detection (continued)

MJMlI 1.1 3 Rev 1. 04/07/95 cateA

4 4

4 N3 Significant Hazards Consideratien

+

Section 1.0 - Use and Application LESS RESTRICTIVE CHANGE L.2 Notused.','^".."..,'.'""-'.^.'.'*..'*.'.^.^.'.._~^.".^.'.*..^.^,*.^*.'..^'..,-.-"._^'

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w m 3 Catawba Units 1 and 2 Page 67 of 10M Supplement 35/20/97l

s o

ys

+

N3 Significant Hazards C:nsidIratien Section 1.0 - Use and Application L.,,., u..,. _ u, _n m..

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