ML20217E653

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Proposed Improved Tech Specs Re Section 1.0,2.0,3.0 & 4.0
ML20217E653
Person / Time
Site: McGuire, Mcguire  Duke Energy icon.png
Issue date: 04/20/1998
From:
DUKE POWER CO.
To:
Shared Package
ML20217E632 List:
References
NUDOCS 9804270370
Download: ML20217E653 (41)


Text

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Definitions 1.1 1.0 USE AND APPLICATION l

l 1.1 Definitions 1

________________________----------NOTE-------------------------------------

)

l The defined terms of this section appear in capitalized type arid are l

applicable throughout these Technical Specifications and Bases.

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

Times.

i 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 i

devices.

l 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 l

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, l

overlapping calibrations or total channel steps so l

that the entire channel is calibrated.

l (continued) l McGuire Unit 1 1.1-1 Supplement 3 l

9804270370 980420 PDR ADOCK 05000369 i

l P

PDR l

I

Definitions 1.1 1.0 USE AND APPLICATION l

1.1 Definitions


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

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

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

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 4

the entire channel, including the required sensor, i

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 perfonned by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.

t I

(continued)

McGuire Unit 2 1.1-1 Supplement 3 l

O uSE Ano Ocfnwr10 MT (IEFINITIONS)

The defined terms of this section appear in ca italized type and are applicable throughout these Technical Specification. W g 3cs ACTION 4(of ACT10hhall be that part of a 6cEMD Specification @ prescribes

-dgi measursi require n r designated condition ggg g.

(ktekn ACTUATION LOGIC TEST r

@ An ACTUATION LOGIC TEST shall be the application of various simulated 4 input combinations in conjunction with each possible interlock logic state and

$ verification of the required logic output. The ACTUATION LOGIC TESTyshall include a continuity checir, as a minismof output devices.

J 6E% CHANNEL OPERATIONAL TEST GoT)

@ MF ANAI K ynum I NPF RAFIONA) hall be the injection of a simulated d TM g

b.

sional into tb thannel as close to the sensor as practicable to verifyGID requwee) OPERABILITY ofialarw, interlock an@ trip functions. The dRAnFTHANK dWrmerrNA17 TNT shall include adjustments, as necessary, of th larm, inter-lockanpnJripgetpoints hatthegetpointsarewithintherequ_ ired range a accuracy.

AXIAL FLUX DIFFERENCE (AFD)

@ a PTA# FurtoI+n n be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION

.a f

@ A CHANNEL CALIBRATION shall be e adjustment, as necessary, of the channetrGD that it responds wit the required range and accuracy to known i

)

bc/ aim input. The CHANNEL CALIBRATION shall encompass the entire channel b

includino thessenso s of sequential., p n @ trip function @ l steps alarm, interl may be 50 9"

/ performed by y ser overlappinmor total channe l Q that the ent channel is calibrated. L -

I CHANNEL CHECK

@ A CHANNEL CHECK shall belthe qualitative assessment)f channel behavi.or during opttration[by observattog This determination shall include, where pos-sible, comparison of tne cnannel indication an(9 status other indica-tions @or status derived from independent instrument channe measuring the same parameter.

McGUIRE - UNIT 1 1-1 Amendment No. 166 o

Rt lY 0 l

Go uSC /WD der /NITIO

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lel L F DEFINITIONSI M E*. The defined te of this section appear in capitalized type and are applicable throughout these Technical Specification Q

ACTION ACTIObhall

"@ N that part of a @ignated condition Specification prescribes quire under des 8

'O A4 Q 7Td 3

ACTUATION LOGIC TEST eum J

@ An ACTUATION LOGIC TEST shall be the appitcation of various simulated /

4,g input combinations in conjunction with each possible interlock logic state and

@ include a continuity checiqras a minim 7umerification of the required logic ou of output devices.

g M CHANNEL OPERATIONAL TEST (cat)

@,,tQ r"

1 be the injection of 'a simulated 2 A.i sinnM into the channel as close to the sensor as practicable to verify ERA 8.LITY _oflalars, interlock an@ trip functions. The - a r--

- shall include adjustment as necessary, of the 41ars, inter-lock and@d ac/ripgetpoints that the tpoints are within the trequired range an curacy.

,g AXIAL FLUX DIFFERENCE (App)

{ @ MIIT mx Drumm.o be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION

. O A CHANN LIBRATION shall be he adjustment, as necessary, of the channel hat it responds wit the required range and accuracy to known h

A'l

  1. 13 FEN 5 input. The CHANNEL CALIBRATION shall encompass the entire channel

. ;c including thefsenso alarm, interlogn(B trip function may be performed Dy an ser es of s uential, overlapp<n or total channel steps that the ent re channel is ca ibrated.,

CHANNEL CHECK Q

@during operati K shall be(the qualitative assessmenMof channel behavior A CHANNEL arison of sne cnanne@l indication anGD status esp other indica.This determ y observatto.

sible tions

'r status derived from independent instrument channel measuring the same parameter.

(

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McGUIRE - UNIT 2 1-1 Amendment No. 148 e

p.e 1.413

Discussicn of Ch:nges Sscticn 1.0 - Use and Applicaticn 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.

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 l

Table 1.1 stating "all reactor vessel head closure bolts fully l

tensioned" eliminates a potential overlap in defined MODES.

For l

example, when the vessel head is detensioned, both the definition l

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 l

or to apply Cold Shutdown applicable LCOs. This change is l

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

This change is consistent with NUREG-1431.

l A.23 The definition of REFUELING in CTS Table 1.2 is changed to remove l

the limit on average reactor coolant temperature in MODE 6.

When j

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 l

fully tensioned. By removing the temperature reference, the l

appropriate LCOs will be applied during this condition.

This l

change is editorial in nature since the intent of the existing i

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 actuation functions and do not require a calibration of l

ossociated displays.

However, CTS 4.3.3.6 requires a calibration

\\

of the posi occident 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.

McGuire Units 1 and 2 Page A-77 Supplement 35/20/97l l

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.

IREE 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 l

of various simulated or actual input combinations in conjunction with each possible interlock logic l

state and the verification of the required logic l

output. The ACTUATION LOGIC TEST, as a minimum,

!$all include a continuity check of output devices.

i 1

l AXIAL FLUX DIFFERENCE AFD shall be the diffprence in normalized flux (AFD) signals between the4 top and bottom halves of a l

twosectionexcoreneutrondetectorg CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as l

necessary, of the channel so that it responds within the r ired range and accuracy to known input. The CALIBRATION shall encompass l

the entire channel ludjng the reouired sensor._Y

/

alarm. interlock, dis vr ana trip runcuans.

Calibration of instrument channels with resistance f

temperature detector (RTD) or thermocouple sensors l

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 res the other sensing elements with the recenti installed l

sensing element. The CHANNEL CAL BRATION may be l

performed by means of any series of sequential.

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

l (continued) 1.11 Rev 1. 04/07/95 m

Definitions 1.1 1.1 Definitions (conti.ned)

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative l

assessment, by observation, of channel behavior during operation. This determination shall l

include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument l

channels measuring the same parameter.

~

l J

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or i

TEST (C0T) actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock,(dispray) and trip i

functions. The COT shall include adjustments, as 1

necessary, of the required alarm interlock, and trip setpoints so that the setpoints are within i

the required range and accuracy.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control cumgients, vithin 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 determined for each j

reload c 1 in accordance with Specification 5.6.5.

operation within these limits is addressed in idual Specifications.

und DOSE EQUIVALENT I 131 DOSE EQUIV 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 132,1 133,1 134, and I 135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed inJITable II of TID 14844, /

, 1962, JGalculation of istance Factors for 4

PowerandfestReactorSi t

n in~

yau t-m wiaun i y nie 1.109, __R 4

plRC,

. orJ1 30, wppiemqt to ParM. pa j (continued) 1.1 2 Rev 1, 04/07/95 k.w l

Definitions 1.1 1.1 Definitions SHlfTDOWN MARGIN (SDH)

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

(continued) fully inserted except for the single RCCA of l

highest reactivity worth, which is assumed to be fully withdrawn. Witn any RCCA not capable of being fully inserted, the reactivity worth j

of the RCCA must be accounted for in the determination of SDN: and b.

In H00ES 1 and 2. the fuel and moderator temperatures are changed to the p inal zero i

power design levelf I

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizin,J l

each slave relay and verifying the OPERABILITY of l

each slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of.

l associated testable actuation devices.

l 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 n Surveillance Frequency intervals, where n is the total ntaber of systems, subsystems, channels, or other designated components in the associated function.

THERMAL 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 verifying _Phe OPERABILITY of j

(TADOT) required alarm interlock. Csphay and trip functions. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the required accuracy.

1.1 6 Rev 1. 04/07/95 w

Justificaticn fcr Daviaticns 52ction 1.0 - Use and Applicaticn 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 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/orchangestotheNUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.

5.

This change reflects the current licensing basis / technical specifications.

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 Tcst (TA00T) include " displays" as part of the scope of 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 calibrated to ensure their functional ability to display required information.

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

1,.

Catawba and McGuire Improved TS Review Comments Section 2.0, Safety Limits l

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 Informative than omitting " steady state." Also add a JFD that explains that McGuire and Catawba have both steady state and transient limits.

l DEC Response:

l The word was deleted because normal operation (condition 1 events) cannot l

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 l

the word steady state with transient and JFD 7 has been added to justify the l

change.

l i

mc3_cr_2.0 1

April 20,1998

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 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 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 hept rate (LHR) below l

l the level at which fuel centerline melting vcurs.

Overheating of the fuel cladding is preventti by restricting l

fuel operation to within the nucleate bciling regime, where i

the heat transfer coefficient is large and tim cladding surface temperature is slightly above the coo? ant saturation temperature.

Fuel centerline melting occurs when the local LHR, or power l

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.

I (continued)

McGuire Unit 1 B 2.0-1 Supplement 3 l

l..

Reactor Core SLs i

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 j

design limits are not exceeded during steady state i

operation, normal operational transients, and anticipated operational occurrences (A00s). This is accomplished by l

having a departure from nucleate boiling (DNB) 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 l

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, whare the heat transfer coefficient is large and the cladding j

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 l

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 coef ficient.

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)

McGuire Unit 2 B 2.0-1 Supplement 3 l

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

B 2.1.1 Reactor Core SLs l

BASES BACKGROUND GDC 10 (Ref.1) requires that specified acceptaole fuel 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 (DE)plished by design basis, which corresponds to a 95% probability at a 95% confidence j

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

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, that m

would result in the release of fission products to the L

\\JJ reactor coolant. _0verheating of the fuel is prevented by l

"[ y4p L ma' ntaining the Eteadtr t1Eh peak linear heat rate (LHR) f s

J be'.ow the revel at which fuel centerline melting occurs.

l Overheating of the fuel cladding is prevented by restricting l

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 lo' al LHR. or power' c

peaking. in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of I

the fuel. Expansion of the pellet upon centerline melting I

may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

l Operation above the boundary of the nucleate boiling regime l

could result in excessive cladding temperature because of l

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

l l

(continued) 1 B 2.0 1 Rev 1. 04/07/95 MChwh

)

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1 Justificaticn fer Deviaticns S:cticn 2.0 - Safety Linits 1

1 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.

Therequirement/statementhasbeendeletedsinceitisnot 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.

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 linear heat rates below centerline fuel melt temperature.

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

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

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

McGuire Units 1 and 2 11 Supplement 35/20/97l

Catawba and McGuire Improved TS Review Comments Section 3.0, LCO and SR Applicability i

3.0-01 JFD 9 (Catawba only)

ITS LCO 3.0.4 JFD 9 states that the ITS for Catawba incorporates TSTF-104, which moved the discussion of exceptions to LCO 3.0.4 to the Bases. Catawba's STS markup, however, fails 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 l

the STS markup of LCO 3.0.4 in the Catawba submittal.

l DEC Response:

I 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.

l 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 ofinformation previously located on page 3.0-2. This page was already included in the McGuire suppaement 1 package.

3.0-02 DOC L1 CTS 4.0.3 ITS SR 3.0.3 DOC L1 does not state why basing the missed surveillance performance allowance on the specified Frequency rather than on the applicable allowed l

outage time is less restrictive. This change is both more and less restrictive.

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

l l

DEC Response:

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

I l

i mc3 cr 3.0 J

April 20,1998

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Discussicn of Changas Szcticn 3.0 - LC0 and SR Applicability l

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 />.

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 l

performed within the delay period or the surveillance fails when l

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 consistent with NUREG-1431.

l 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.

McGuire Units 1 and 2 Page L - 2B Supplement 35/20/97l

j..

N3 Significant Hazards C nsid:raticn l

S;cticn 3.0 - LCO and SR Applicability I

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 l

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 survetIlance 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 McGuire 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 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 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 operable. When a Surveillance Requirement is not performed within 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 evaluated accidents are not increased since accident mitigating McGuire Units 1 and 2 Page 66 of 11M Supplement 35/M/97 l l

l No Significint H:zards Crnsidtraticn 5:cticn 3.0 - LC0 cnd SR Applicability l

l equipment will continue to perform their intended safety I

functions.

I 2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated'l The possibility of a new or different kind of accident from any accident previously evaluated is not created because the proposed l

change does not introduce a new mode of plant operation and does i

not involve physical modification to the plant.

3.

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

l 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 l

low probability that the surveillance will not be completed l

satisfactorily.

The requested allowance will provide sufficient l

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 j

and shutdowns.

l i

I l

McGuire Units 1 and 2 Page 76 of 1144 Supplement 35/24/97l

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 CTS 5.3.1 l

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 5.0 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 l

the language in the CTS 3/4.9.13. " Spent Fuel Assembly Storage." Thus use of

" nominal" in ITS 4.3.1.1.a is acceptable. The removal of the information i

explaining the tolerance associated with the allowed maximum nominal i

enrichment, however, should be justified by an IA-type DOC. It was suggested I

that the Bases for ITS 3.7.16 for Catawba (3.7.15 for McGuire) would be an appropriate location for this information. It was noted that the paragraph at the end of CTS Table 3.9-1, for both sites, omits the word " nominal," although it ought to be included for consistency. This table for both units is retained in ITS Section 3.7, and ought to be made consistent. Comment: Revise the Catawba submittal with an LA-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 IA.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 CTS 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.

mc3_cr 4.0 1

April 20,1998

Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (page 1 of 1)

Minimum Qualifying Burnup Versus Initial Enrichment for Unrestricted Region 1 Storage Initial Nominal l

Enrichment Assembly Burnup I

(Weicht% U-235)

(GWD/MTU) l 4.19(orless) 0 4.20 0.04 4.50 1.92 4.75 3.40 i

1 5

S l

g4 l

k l

8

/

3 7

ACCEPTABLE For Unrestricted Storage c5 2-UNACCEPTABLE M

For Unrestricted Storage 0

4.00 4.25 4.50 4.75 Initial Nominal Enrichment (Weight % U-235)

NOTES:

l Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for Unrestricted Region 1 storage by means of an analysis using NRC approved methodology to assure that k,ff is less than or i

equal to 0.95.

Likewise, previously unanalyzed fuel up to a nominal 4.75 l

weight % U-235 may be qualified for Restricted Region 1 storage by means of an is less than or l

analysis using NRC approved methodology to assure that keff l

equal to 0.95.

l McGuire Unit 1 3.7-33 Supplement 3 l

l

Spent Fuel Assembly Storage 3.7.15 Table 3.7.15-1 (page 1 of 1)

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

(GWD/MTV) 4.19(or less) 0 l

4.20 0.04 4.50 1.92 4.75 3.40 5-S EiN k3' ACCEPTABLE y

For Unrestricted Storage 5 2 h

f UNACCEPTABLE 3

4 For Unrestricted Storage 0

4.00 4.25 4.50 4.75 initial Nominal Enrichment (Weight % U-235)

NOTES:

Fuel which differs from those designs used to determine the requirements of Table 3.7.15-1 may be qualified for Unrestricted Region 1 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 4.75 l

weight % U-235 may be qualified for Restricted Region 1 storage by means of an analysis using NRC approved methodology to assure that k,ff is less than or l

equal to 0.95.

McGuire Unit 2 3.7-33 Supplement 3 l

i l

a Spent Fuel Pool Boron Concentration B 3.7.14 8 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Boron Concentration BASES BACKGROUND In the two region poison fuel storage rack (Refs. 1 and 2) design, the spent fuel pool is divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. Region 1, with 286 storage positions, is designed to accommodate new fuel with a maximum nominal enrichment of 4.75 wt% U-235 l

(maximum tolerance of 10.05 wt%), which have accumulated minimum burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1.

Fuel assemblies not meeting the criteria of Table 3.7.15-1 shall be stored in accordance with Figures 3.7.15-1 through 3.7.15-3.

Region 2, with 1177 storage positions, is designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups in accordance with the accompanying LCO.

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 each region in a subcritical condition during normal operation l

with the spent fuel pool fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the i

April 1978 NRC letter (Ref. 3) allows credit for soluble boron under other abnomal or accident conditions, since l

only a single accident need be considered at one time. For example, the most severe accident scenario is associated with the movement of fuel from Region 1 to Region 2, and accidental misloading of a fuel assembly in Region 1 or Region 2.

This could potentially increase the reactivity of l

the spent fuel pool. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the two region poison fuel l

storage rack with no movement of assemblies may therefore be 1

achieved by controlling the location of each assembly in accordance with LC0 3.7.15. " Spent Fuel Assembly Storage."

Prior to movement of an assembly, it is necessary to perform SR 3.7.14.1.

i (continued) l McGuire Unit 1 B 3.7-70 Supplement 3 l

La 4

l l

Spent Fuel Assembly Storage B 3.7.15 l

B.3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage i

BASES BACKGROUND In the two region poison fuel storage rack (Refs. I and 2) l design, the spent fuel pool is divided into two separate and l

distinct regions which, for the purpose of criticality l

considerations, are considered as separate pools. Region 1, l

with 286 storage positions, is designed to accommodate new l

fuel with a maximum nominal enrichment of 4.75 wt% U-235 l

. l (maximum tolerance of 1 0.05 wt%), which have accumulated I

minimum burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1.

Fuel assemblies not meeting the criteria of Table 3.7.15-1 shall be stored in accordance with Figures 3.7.15-1 through 3.7.15-3.

Region 2, with 1177 storage positions, is designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups in accordance with the accompanying LCO.

The water in the spent fuel pool nomally 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 each region in a subcritical condition during normal operation with the spent fuel pool fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 3) 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 movement of fuel from Region 1 to Region 2, and accidental misloading of a fuel assembly in Region 1 or Region 2.

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 two region poison fuel 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 perform SR 3.7.14.1.

(continued) l McGuire Unit 1 B 3.7-74 Supplement 3

Spent Fuel Pool Boron Concentration B 3.7.14 8 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Boron Concentration BASES BACKGROUND In the two region poison fuel storage rack (Refs. I and 2) design, the spent fuel pool is divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. Region 1, with 286 storage positions, is designed to accommodate new fuel with a maximum nominal enrichment of 4.75 wt% U-235 l

(maximum tolerance of 10.05 wt%), which have accumulated minimum burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1.

Fuel assemblies not meeting the criteria of Table 3.7.15-1 shall be stored in accordance with Figures 3.7.15-1 through 3.7.15-3.

Region 2, with 1177 storage positions, is designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups in accordance with the accompanying LCO.

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 limitin boron. g k,ff of 0.95 be evaluated in the absence of soluble Hence, the design of the spent fuel storage racks is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the spent fuel pool fully loaded. The double contingency )rinciple discussed in ANSI N-16.1-1975 and the April 1978 NIC letter (Ref. 3) allows credit for soluble boron ude 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 movement of fuel from Region 1 to Region 2, and accidental misloading of a fuel assembly in Region 1 or Region 2.

This could potentially increase the reactivity of the spent fuel pool.

Io mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the two region poison fuel storage rack with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with LC0 3.7.15, " Spent Fuel Assembly Storage."

Prior to movement of an assembly, it is necessary to perform SR 3.7.14.1.

(continued) l McGuire Unit 2 B 3.7-70 Supplement 3

Spent Fuel Assembly Storage B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage BASES BACKGROUND In the two region poison fuel storage rack (Refs. 1 and 2) design, the spent fuel pool is divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. Region 1, with 286 storage positions, is designed to accommodate new fuel with a maximum nominal enrichment of 4.75 wt% U-235 l

(maximum tolerance of 10.05 wt%), which have accumulated minimum burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1.

Fuel assemblies not meeting the criteria of Table 3.7.15-1 shall be stored in accordance with Figures 3.7.15-1 through 3.7.15-3.

Region 2, with 1177 storage positions, is designed to accommodate fuel of varicus initial enrichments which have accumulated minimum burnups in accordance with the accompanying LCO.

The water in the spent fuel pool nomally 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 each region in a subcritical condition during normal operation with the spent fuel pool fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 3) 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 movement of fuel from Region 1 to Region 2, and accidental misloading of a fuel assembly in Region 1 or Region 2.

This could potentially increase the reactivity of the spent fuel pool. To mitigate these postulated criticality related accidents, baron is dissolved in the pool water. Safe operation of the two region poison fuel 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 perform SR 3.7.14.1.

(continued) l McGuire Unit 2 8 3.7-74 Supplement 3

Spet N=M Yl.15 Ninimum Oualifvino Burnuo Versus initial Enrichment for Unrestricted Recion 1 Storace Initial Nominal Enrichment Assembly Burnup (Weicht% U-235)

(GWD/MTU) 4.19(or less) 0 4.20 0.04 4.50 1.92 4.75 3.40 5-4--

8 3--

ACCEPTABe g

For Unrestricted Storage m 2-3!ir t

UNACCEPTABLE 4

For Unrestridad Storace 0

4.00 425 4.50 4.75 initial Nonitial Enrichment (Weight % U-235) e Fuel whict) differs from those designs used to determine the requirements of 3

Table 35-1 may be qualified. for Unrestricted Region 1 storage by means of an analysis using NRC approved methodology to assure that k a is less than or equal to 0.95.

@ nom *a3 Likewise, previously unanalyzed fuel up toJ4.75 wetght% U-5 may be qualified for Restricted Region 1 storage by means of an analysis using NRC approved methodology to essure that k,,, is less than or equal to 0.95.

McGUIRE - UNIT 1 3/4 9-18 Amendment No.

166 l

f9 pay 2 o

Sys'Scu 3.9. !$

I Table 3.9'-1 Minimum Oualifyino Burnuo Versus initial Enrichment for Unrestricted Recion 1 Storace Initial Nominal Enrichment Assembly Burnup

~

(Weicht% U-235)

(GWD/NTU1 4.19(or less) 0 4.20 0.04 4.50 1.92 4.75 3.40 5-4-

3-AOCEPT M E

For Unrestrknad Storage b1 N

\\

E UNACCEPTABl.E 4'

rw unr.setned stora00 0

--4 4.00 4.25 4.50 4.75 inilial Nonnial Enrkfiment (Weight % U-24 D

w fuel which differs from those designs used to determine the requirements of Table 3. - may be qualified for Unrestricted Region 1 storage by means of an analysis using NRC approved methodology to assure that s less than or equal to 0.95.

(a noeni e}

Likewise, previously unanalyzed fuel up tom.75 wel t% U-235 may be qualified for Restricted Region 1 storage y means of an anal sis using NRC approved methodology to assure that k i less than or equal to 0.95.

g McGUIRE - UNIT 2 3/4 9-18 Amendment No. 148 9

l INSERT Table 3.7.15-1 (page 1 of 1)

Minimum Qualifying Burnup Versus Initial Enrichment for Unrestricted Region 1 Storage Initial Nominal Enrichment Assembly Burnup 1

(Weicht% U-235)

(GWD/MTV) 4.19(or less) 0 4.20 0.04 4.50 1.92 4.75 3.40 5-S Ei l

a h3 ACCEPTABLE E

For Unrestricted Storage l

55 2 2

l UNACCEPTABLE 3

4 For Unrestricted Storage 0

4.00 4.25 4.50 4.75 Initial Nominal Enrichment (Weight % U-235) l NOTES:

Fuel which differs from those designs.used to determine the requirements of Table 3.7.15-1 may be qualified for Unrestricted Region 1 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 4.75 l

weight % U-235 may be qualified for Restricted Region 1 storage by means of an analysis using NRC approved methodology to assure that k,ff is less than or equal to 0.95.

INSERT Page 3.7-39 (page 1 of 8)

McGuire

INSERT l (maximum tolerance of 10.05 wt%), which have accumulated minimum burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1.

Fuel assemblies not meeting the criteria of Table 3.7.15-1 shall be stored in accordance with Figures 3.7.15-1 through 3.7.15-3.

l l

l INSERT Page B 3.7 - 81 McGuire

[

INSERT (maximumtoleranceof 0.05 wt%), which have accumulated minimum burnup l

greater than or equal to the minimum qualified burnups in Table 3.7.15-1.

l Fuel assemblies not meeting the criteria of Table 3.7.15-1 shall be stored in i

accordance with Figures 3.7.15-1 through 3.7.15-3.

t l

l l

l l

l l

INSERT Page B 3.7 - 85 McGuire l

O O

4 4

ENCLOSURE 2 WITHDRAWAL OF DEFINITION CHANGE AND REVISED ATTACHMENTS 5 AND 6 l

l l

l I

i l

i

Definitions e

1.1 1.1 Definitions (continued) l 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 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, l

1962, " Calculation of Distance Factors for Power and Test Reactor Sites."

(continued) l McGuire Unit 1 1.1-2 Supplement 3

1.-

1 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) actudi 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 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.

l 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, 1-133, 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."

l l

4 I

(continued) i l McGuire Unit 2 1.1-2 Supplement 3 i

Speb bo'w l.0 l

DEFINITIONS DOSE EOUIVALENT i-J3J A.

@ DOSE EQUIVALENT I-131' shall be that concentration of 1-131 (microc Q gram) SD alone would produce the same thyroid dose as the quantity and iso.

topic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present. The g g b thyroid da m raa-ard an factors used for this calculation shall be those listed-

" i a eI or i m- --- - _ - - a-mun_ancy Factor (far pawar and Ae srst

' M tJer c Cl, W b g.";l".c- % 6 %'s "5'

E - AVERAGE DISINTEGRATION ENERGY Q % 7*T A.I se,. c $

E shall.be the average 7 h @each radionuclide in theqEIb)

(wei ted in proportion to the concentration of f the sum of the average beta and gamma energies per disin ton V/d) for thepr ::a 6-s -

z__samois.

M n,*res..u, +$.a t.J %,;+i. w i o 7 a M6 ENGINEERED SAFETY FEATURE SPONSE TIME %k.g.,.+ k.a at 1.,.r. m 4.ht,s. 2:a %4l,,:g GTD@ Thel _ructurdrn ufrhr i su w.har

~

RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESFt(ctuationdfetpoint 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.

FREQUENCY NQ(AT h OA.l?

~

1.14 Th FREQUENCY NOTAT specified for t performance of S eillance

<Reouir nts shall corre nd to the inte s defined in Tabl 1.1.

EDENTE IEDlLEAKAGE 1

I LEAKAGE shall be:

a. ria utJa ww Liakpge7(_ xcept' CONTReii m wumrdQnteTIMy,stte) such aslpump A.

e f

(tvsu+ '* h seafs)or valve packinqJDER tfiat aptured anc conducted toya sump t-a-a6 t Or.W or collecting tanl@@

ebb. uw,4 +<)

i E

  • ss.\\

su s3es

'O>

kakagdintothecontainmentatmospherefromsourcesthatareboth g)f.

a specifically located and known either not to interfere with the operation of leakage detection systems or not to befKta) UKE BUUNDARQ LEAKAGQor p.

Reactor Coolant

%.akaoeithrough a steam generat o the SecondaryCl3 tut)Systeeg MASTER RELAY TEST

-_M.fe-en,,n. D A MASTER RELAY TEST shall me am ro & aa at each master relay and w.%

ri s cq w ne OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

4 McGUIRE - UNIT 2 1-3 Amendment No.

148 6

k

r 5905f'ed', l.O i

l l

DEFINITIONS DOSE EOUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (microcurth g @ topic mixture of I-131, 1-132, 1-133, 1-134, and I-135

,l gram), @ ZR alone would produce the same thyroid dose as the quantity and iso-The h hC,M@2 thyroid staca ennvere4^n factors used for this calculation shall be those listed le v.

. uv a o,, _ -moiamon Af Distance Factorvior rowerano/ies l

'qm;gmtW:nm=gy '

s=r

's E - AVERAGE DISINTEGRAI10N LNtRGY phrMab df 4u hm.f $=r 'a3

@ E shall be the average (weighted in proportion to the concentration of A1 each radionuclide in the of the sum of the average beta and gamma M

energies per disintegration eV/d) for the aaianuci se ri v se sampig 3

cf G s.,essue R..,Lau u titlives>10 iso ENGINEERED SAFETY FEATURES SE TIME m.-nes ett*.e a te.h 9r1 g ne.u.t

(

twi..Ji-s,ac4V6 w 44 wi.-4.

f A.

@from when tie monitored parameter exceeds its ESF/ctuation getpoint at the l

The ENGINFFDGer SAFET7 FEATU NSE TIME shall be that time interval 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.v L-<suMAT 2 fTiiE00ENCY NOTATLON 1

The FR UENCY N0JAnuii specinea tor /tne performance ofj$urveillanc de.14 quirement shall co/ respond to the int (rvals defined in Ta61e 1.1.

iTDENTLPTED1 LEAKAGE l

(.

LEAKAGE shall be:

gg

a. AWtal teatair.

~ W.

ifeakaqd ExceoVLUNIR01i FD}'FMAGFD(McJefed sf{teinD such asdpump sea @or valve packing Qthat G9,ysaptured and conducted t a sump gy

c.. epm **

or collecting tang @

ae-.:g;,a...e se.t.[${, ggyf ht

@h.

((eakaatfintothecontainmentatmospherefrom$ourcesthatareboth specifically located and known either not to interfere with the operation of leakage detection systems or not to befRLhbUKtg i

LEAS (AG@r g

@c.

Reactor Coolant System agd throu % a steam generat o the l

(msco 3y Secondary (diR':iii) Systemq MASTER RELAY TEST g,

Q A MASTER RELAY TEST shallAG rnmnernizpum on each master relay and 4.

ver.

rnmond OPERABILITY of each relay. The HASTER RELAY TEST shall include a continuity check of each associated slave relay.

McGUIRE - UNIT 1 1-3 Amendment No. 166 q J 13

Discussien of Chang 2s Srctien 1.0 - Usa cnd Applicatien 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 ALTERATIONS as movement of only those components which can affect core reactivity. The basis for this is evident in that the Specifications applicable during CORE ALTERATIONS are those that i

protect from or mitigate a reactivity excursion event.

In keeping with this, ITS Specification 1.1 provides that movement of equipment other than fuel, sources, or reactivity control l

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 l

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.

L.2 Not used.r.T.c

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< 4.

4. +. 4..-. i..n. e - - 4 4. - _.

1.

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4.... T.., L.1. -

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" Calculation of Distance Facter; for P0wer and Tc;t "cactor e 4. +. -.. -

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equivalence calculation; to reficct =cre recent =cd-ling, ;; well

to en;ure c0n;i;tency between current Off,ite dO;c calculation practices, plant ;;mpling, and counting practice;, thi; change is pr0pc;cd to ; witch the standard u;cd for corvees 40n factor; from T. f n.
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eenversion facter; cf ICR" 30.

Thi; i; acceptable because Regulatory Cuide 1.4, which i; u;cd ;; ; guidance document for l

dc;c calculation; sugge:t; u;c Of de c conver:fon facter; from i

T. e. n n

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Thi; change 4s -. _.,., 4,, +._. 4.. 4.. u.

u. n. n. e r.

1,A 31 2

1 McGuire Units 1 and 2 Page L - 14 Supplement 35/20/97l l

i l

i Definitions 1.1 1.1 Definitions (continued) l 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 inte the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock,(dispray) and trip functions. The COT shall include adjustments, as i

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, vithin 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 n accordance with Specification 5.6.5.

ration within these limits is addressed in al Specifications, n.4 DOSE EQUIVALENT I 131 DOSE EQUIVALE 31 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 132,1-133. I 134, and I 135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed inf[ Table II /of TID 14844, /

C, 1962, 1 Calculation of stance Factors for 4

PowerandfestReactorSi t

en in av t-vi wiaun.y ty nie 1.109, R

, orJ1gJu, duppiemqt to Part% page (continued) 1.1 2 Rev 1, 04/07/95 a.w

Definitions 1.1 1.1 Definitions

+ __.

DOSE EQUIVALENT I 131 f 1922 Tabl ti ed.

  • ted Dose 1

(continued)

Equival in Targe Organs o Tissues per ak Lnf finit ivi t y* y E-AVERAGE Eshallbetheaverage(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 gamma energies per h

disintegration (in Me'g_far i etnnes. other than

/

iodines with half lives inutes, making up at least 95% of the total ine activity in the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time FEATURE (ESF) RESPONSE interval from when the monitored parameter TIME 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.

The maximum allowable p mary containment le age ~

rate. L,. shall be [

of primary contai t air weight per day at t calculated peak c inment peasgure (P.). f-u i

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE 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.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection (continued) 1.1 3 Rev 1. 04/07/95 l

h(c gum I

4 e

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