ML20205G878

Forwards Technical Justification & Analysis of Significant Hazards Consideration for Tech Spec Changes Requested in 850315,0731 & 0807 Ltrs

ML20205G878
Person / Time
Site:	Catawba
Issue date:	11/08/1985
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
TAC-65833, NUDOCS 8511130372
Download: ML20205G878 (8)

Text

Dunn POWER GOMPANY P.O. BOX 33180 CHARLOTTE. N.C. 28242 HALH. TUCKER Tut.zenoxx (yo4) an>4 sag m aensame=,

November 8, 1985

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Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Attention:

Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Proof and Review Technical Specifications

Dear Mr. Denton:

Please find attached the technical justification and analysis of significant hazards consideration for those Technical Specification changes requested in my March 15, 1985 and August 7, 1985 letters regarding the proposed Catawba Units 1 and 2 Technical Specifications. Also enclosed is additional justi-fication for the changes requested in attachment 4 of my July 31, 1985 letter.

The justifications and analyses are provided for those proposed changes which we wish to have included in the combined Units 1 and 2 Technical Specifications upon their issuance with the fuel load license.

Those changes which were requested but for which the justification and analysis have not yet been provided will be pursued at a' later date.

The proposed changes which we are providing the justification and analysis for are for those changes requested on:

The ventilation specifications, 3/4.6.1.8, 3/4.7.6, e

3/4.7.7. 3/4.9.4.2 and 3/4.9.11.

Operational Leakage Specification, Action Statement C.

e e Table 3.3-6 and 4.3-3, Radiation Monitoring Instrumentation for Plant Operations,. Item 4. - Auxiliary Building Ventilation High Gaseous Radioactivity e Ice Condenser Doors, Surveillance 4.6.5.3.1 Standby Shutdown System, 3/4.7.13 e

Administrative Controls, 6.5.1.8 and 6.8.2 e

Table 3.3-10, Accident Monitoring Instrumentation, Item 9 e

Auxiliary Feedwater Flowrate 8511130372 051100 g

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' Nk.1 Har:ld R. Denton, Dirsetcr.

November.8. 1985

.Page Two.

e Relief-Valves,JSurveillance 4.4.4.3 Table 3.2-1, DNB' Parameters e

Figure 3.2-3,. Reactor Coolant System Total Flow Rate Versus R e

Also enclosed are'several _dditional pages of proposed Technical Specification revisions.. The changes _ to _ the Standby Shutdown System (SSS) Specification are

-intended'to reflect Catawba specific equipment requirements.

The SSS speci-

'fication provided in my-letter of August 7, 1985 was derived from the McGuire proposed' specification and_ contained some McGuire specific information which has been found=to be not applicable to Catawba.

The proposed changes to the ventilation specifications, ice condenser doors (in part), administrative controls (in part), relief valves and Figure 3.2-3 have been incorporated in the Proof and Review Technical Specifications. The.

other requested changes have not.

The requested changes that have not yet

been incorporated into'the Proof and Review Technical Specifications are not identifying errors currently contained in the Specifications but are deemed necessary in order to ease operational restraints which are unnecessarily overly conservative.

If you have any questions regarding this submittal please contact Mr. Roger W.

Que11ette at (704)373-7530.

Very_truly1yours, 2/26T~A~

Hal B. Tucker RWO: sib Attacheent ec:

Dr. J. Nelson Grace, Regional Administrator.

U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta,_ Georgia 30323 p

Mr. Heyward Shealy, Chief Bureau of Radiologica Health

-S. C.' Department of Health & Environmental Control

-2600 Bull Street Columbus, South Carolina 29201 l

NRC Resident Inspector Catawba Nuclear Station

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. JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS' CONSIDERATION-s e

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The proposed changesfaffect the Technical Specifications for-the Annulus j

. Ventilation' System, Control Room Area Ventilation System, Auxiliary Building Filtered Exhaust. System -the Reactor Building Containment Purge

. System and'the Fuel Handling Ventilation Exhaust System.

.Several of;the proposed changes are administrative.in nature and are requested in order to cladfy wording or to correct nonapplicable state-

'ments.. Changing the wording from " charcoal adsorbers" to. " carbon hadsorbers" would make theIyqrding consistent.with the wording used at the

' station and would correctly' state the adsorber substance.

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Deleting the reference to ANSI N510-1980 in the fre-heater surveillances is necessary since,this. standard does not address heater power dissipation.

The surveillance will still be performed using acceptable methods.

The proposed. wording changes.to 4.7.6e.3) are, intended to clarify this

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Surveillance Requirement. The current wording is very confusing and possibly incorrect in its literal' interpretation M The.new wording will not change the-intdnt of the requirement or the way in,which compliance with ithe: specification is. demonstrated.

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The changes. requested to 4.7.7b.1), 3), d.1)..e.,

f., 4.9.11.2b.1), 3),

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'd.1), e..and fL are to clarify the flow rates required for the total. system

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'.asfopposed to those required for each train or each fan. There will be no change ~to the performance of the tests or' acceptance criteria.

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The' changes proposed 2to 4.6.1.8b.,

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4.7.6c., h., 4.7.7b., e.,-f.,

g., 4'.9.4.2b.,1g..and 4.9.11.2b., e.,

f. and g. would delete the requirement

~ to leak : test the. filter. units af ter painting, fire or a chemical release

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TheTe'chnicalSpecifications.and.ParagraphC.S.cohRegulatoryGuide1.52.

currently require thst in-place leak testing be-performed on a filter unit after thc. system has been exposed to paint,' fire or' chemicals.1 This type of-

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exposure would only.-affect the radioiodine efficiency of the carbon and not the structural integrity of.the filter media and its supports. Since the-in-place leak' tests are required to verify structural integrity, only a laboratory analysis.h>tsst-is required af ter:being'. exposed to these _ types of fumes A

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If particulates are considered froni spray painting, fire, etc.. they would

~ ffect the:HEPA= filters. However..only.an increase in the pressure drop' awould result due.to filter particulate loading, not leaks de9 eloping. Here, "

.again, structural integrity is not affected and leak tests a9e not warranted.

As for= smoke particulates, if hot particulates from a fire enter the system-ductwork. the system fans will shut down in response to a smoke detector signal. Therefore, they will.not impinge on the HEPA filters, (after having to pass through the' prefilters), leaks will not be caused due to burning l-

--of the media and therefore, should not result in leak testing.

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, The change to the footnote on each ventilation specification would delete the requirement to operate the system for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters on and operating. This is in accordance with Regulatory Guide 1.52, paragraph C.5.d.

This section of Regulatory Guide 1.52 was written and based on the premise that refrigerant would affect the radioiodine efficiency of the carbon. This has recently been proven incorrect as shown in the attached memos. The only other reason to purge the refrigerant from the carbon after the periodic test would be if the carbon bed was to become saturated and a retest had to be made.

This is accomodated in the periodic filter test by allowing a purge to be performed at the test coordinators discretion. Also, subsequent purging is accomplished by the Technical Specifications.

Since they require a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run, with the heaters energized, every 31 days, the refrigerant should be dissipated by the time the 18 month periodic test is due. This change will reduce the amount of run time on the filters which should extend their change-out time.

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10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The proposed amendments do not involve an increase in the probability or consequences of any previously evaluated accident. The accident analyses will be unaffected by these changes.

The proposed amendments do not create the possibility of a new or different kind of accident than any previously evaluated since the oparation of the plant will be unaffected.

The majority of the changes are administrativa in nature and will not affect safety in any way.

Several changes may reduce a margin of safety in an insignificant way by deleting unnecessary and overly conservative surveillances on the structural integrity of the filter housings and requiring a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run of the ventilation system with the heaters on.

The Commission has provided guidance concerning the application of standards of no significant hazards determination by providing certain examples (48 FR 14870).

Those changes involving clarifications are similar to example (i) and those changes to the surveillances for the filters housings and for the purging of residual refrigerant are similar to example (vi)

For the reasons stated above, it is concluded that the proposed amendments do not involve significant hazards considerations.

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' JUSTIFICATION AND ANALYSIS OF N0'SIGNIFICANT HAZARDS CONSIDERATION g

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',w The change requested to-ACTION c. on page 3/4 4-20 would require that

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action be taken with any Reactor Coolant System Pressure Isolation valve-

-leakage greater than specified limits in Modes 1 and 2 and would delete

~ he requirements'while'in any other Modes.

t Successful testing 1of the Pressure. Isolation Valves depends on obtaining

optimum system. conditions.. These valves seat better at.high. system,

pressure, therefore,'they leak less at full reactor coolant. system pressure.

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. Testing,~however, is usually_ conducted:at a pressure'of approximately 400 psig and.the resulting-leakrates adjusted up.to full-reactor coolant system pressure. This 1s-done since_during a normal unit st'artup,

testing at:a hf gher pressure ucually means system temperatuse would be higher.. This_can. lead to flashing problems when-performing the leak

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rate tiest.. Present test philcsophy is to test at the lower pressure andl "Vif the valve' exhibits a leak rate above allowable, then increase reactor.

coolant system pressure and attempt 3. testing again, since these-valves-s

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Ese'at better at the higher pressure.. The~ Technical Specification presently-requires'us to enter ACTION (c) should afvalve be found to leak in Modas'

3 or 4. _ This could.causeLa plant = shutdown prior to performing a retest -

at a higher pressure due to the 'large number of valves which are' required

' to be. tested..

I Sinc'e Surveillance ~ Requirement-4i4

.2.2 19 not required to be performed 1 L

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for_ entry into Modes 13 or 4 the Reactor Coolunt System-Pressure Isolation

. Valves'are not required'to be operable until entry into Mode 2.

Any' leak--

7 age due to these valves'does not impact the.sa'1 0 of the plant provided s - the leakage does.not impact. meeting 'lecho ted Sn cificatjons 3.4.6.2a., b.,

' c..' d..- and : e.' p Therefore, Technical-Speu i f at ' 2 3.4.6.2f. needs to

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[ reflect the testing requiremsnt'that the "i % q

<2. Valves must be tested

successfully prior to entering

Mode 2.: If.the valves are deterniined to

be leaking during'. Modes-3 or _4,- ACTION statement (c) should: not 'be

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capplicable.

10.CFR 50.92 states that a-proposed amendment involves no significant b,

hazards considerations;if. operation in accordance'with the proposed

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' amen'dmentswould not:

(1)" Involve a significant increase in the probability or

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consequences of an accident previously evaluated; or f'p,

(2). Create the possibility'of a new or different kind of

. accident'from any' accident previoualy evaluated; or

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(3) Involve.a significant reduction in a margin of safety..

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The proposed amendment does not involve any increase in the probability

, or consequences of any presviously evaluated accident. The accident analyses will be unaffected by these changes.

The proposed amendment does not create the possibility of a new or

' different kind of accident than any previously evaluated since the operation of the plant will be unaffected.

The proposed amendment does not involve a significant reduction in a

- margin of safety since the valves do not have to be declared operable until prior to entry into mode 2.

' The Commission has provided guidance concerning the application of standarde_of no significant hazards determination by providing certain examples (48.1RR 14870). This change is similar to example (vi).

' For the reasons-stated above, it is concluded that the proposed amendment

- does not-involve significant hazards considerations.

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I JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The change to Item 4 on pages 3/4 S-52 and 3/4 3-54 would change the applicable modes from "ALL" to "1, 2, 3, 4".

This change is necessary in order to provide consistency throughout the Technical Specifications. Technical Specification 3/4.7.7 Auxiliary Building Filtered Exhaust (VA) System calls for the VA system to be operable in modes 1, 2, 3 and 4.

Since the main function of

-radiation monitor EHF-41 is to realign the VA system to the filtered mode upon receipt of a high radiation signal, EMF-41 should only be required to be operable during those modes in which the VA system is required to be operable.

Therefore, requiring EMF-41 to be operable at all times is an operational burden which adds nothing to the safe operation of the station.

10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new of different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The proposed amendment does not involve an increase in the probability or consequences of any previously evaluated accident. The accident analyses will be unaffected by these changes.

The proposed amendment does'not create the possibility of a new or different kind of accident than any previously evaluated since the operation of the plant will be unaffected.

The proposed amendment does not involve a significant. reduction in a margin of safety since the VA system is not required to be operable in modes 5 or 6 anyway.

The Commission has provided guidance concerning the application of standards of no significant hazards determination by providing certain examples (48 FR 14870).

This change is similar to examples (i) and (vi).

For the reasons stated above, it is concluded that the proposed amendment does not. involve significant hazards considerations.

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- JUSTIFICATION'AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The proposed changes to Technical Specification 3/4.6.5.3 would decrease the-surveillance. frequency for testing the ice condenser inlet doors.

This exact same' change was requested for'the McGuire Nuclear Station

- Technical. Specifications by Mr. Hal B. Tucker's' July.12, 1985 letter to Mr. Harold R. Denton (attached). The justification and Significant Hazards Considerations are applicable to this request for Catawba.

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F.O. mox 33189 m - arrE. N.C. 28949 RALB. TUCKER gaggy,yg (To4) 373-453 July 12, 1985 Mr. Harold R. Denton, Director Office of Nuclear Laactor Regulation

,U. S. Nuclear Regulatory. Commission Washington, D. C. 20555 Attention:- Ms. E. G. Adensam, Chief

. Licensing Branch No.-4

Subject:

, McGuire Nuclear Station Docket Nos. 50-369 and 50-370

. Technical Specification Amendment for Containment Ice Condenser

Dear Mr..'Denton:

Attached are proposed license amendments to Facility Operating Licenses NPF-9 and NPF-17 for McGuire Nuclear Station Units 1 and 2 respectively.

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.The proposed' amendments would extend the surveillance interval for the Ice Condenser Doors to at least once every.18 months. The current Technical Specification (4.6.5.3.1) requires the surveillance of the Ice Condenser Doors at least once

-every 9 months.

The' Technical Basis (3/4.6.5.3.) for the surveillance requirements is'to provide o

i reasonable. assurance that the doors should remain closed during normal operation and open during a LOCA or'a high energy line break. Duke Power Company's review

' of the Ice Condenser Doors has demonstrated that the closure of the doors during normal. operation is adequately addressed.by the Continuous Door Position Monitoring System. The opening of the ice condenser doors during a high energy line break is also'not in doubt as a minor increase in pressure (of 0.1 psig or more) exerts La force of about.10 times that employed in surveillance and testing of these doors.-

In the case of a real' accident,'the ice condenser doors are certain to open since the doors do not have the structural strength to remain or be restrained shut

.'against a pressure rise of 2 or 3 psig. The ice condenser doors meet'their functional requirements as envisaged in the Technical Specification Basis even if no periodic surveillance and ten. ting is performed on these doors. The. doors

.do not contain any components, failure of which, would prevent the doors from performing their intended function.- 'The door hinges, for instance are designed.

i to preclude galling or self welding, the door springs only hold the door in neutral position but otherwise exert a very'small torque, and the door seals a

prevent formation of condensation which could freeze the door.

Duke Power Company has determined.that extension.of the surveillance interval for the ice condenser doors from 9 to'18 months is not a safety concern. Attachment I 4

contains the proposed changes to the Technical Specification. Attachment II provides

" Justification'and Safety Analysis in support of the proposed changes. Attachment III provides an Analysis of Significant Hazards Consideration for the proposed changes

as. required by 10CFR50.91. The proposed changes have no safety or environmental

-impact..

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3 Mr. Harold R. Denton, Director July 12,1985 Page This request involves one application for amendment to McGuire's Technical Specifications. Accordingly, pursuant to 10CFR 170.21 a check for $150.00 is enclosed.

Please feel free to contact us if you require any additional information.

-Very truly yours, s/Hal B. Tucker Hal B. Tucker RLG/mj f Attachments cc:

Dr. J. Nelson Grace, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323 W. T. Orders Senior Resident Inspector McGuire Nuclear Station Mr. Dayne Brown, Chief Radiation Protection Branch Division of Facility Services Department of Human Resources P. O. Box 12200 Raleigh, North Carolina 27605

Mr. Harold R. Dinten, Diractor July 12,1985 Page

• HAL B. TUCKER, being duly sworn, states that he is Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this revision to the McGuire Nuclear Station License Nos. NPF-9 and NPF-17 and that all statements and matters set forth therein are true and correct to the best of his knowledge.

s/Hal B. Tucker Hal B. Tucker, Vice President Subscribed and sworn to before me this 12th day of July i

s/ Sue C. Sherrill Notary Public l-My Commission Expires:

September 20, 1989 l

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Mr.,Harhld R. Dinten, Dirsctor

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July 12s 1985

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bec:

P. M. Abraham K. S. Canady.

N. A.-. Rutherford

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Gill.

S.LA. Gewehr P.'B. Nardoci W. H. McDowell

'J.~ B. Day, Jr.

,:1R. T IOuellette;;

.E.-O. McCraw-(MNS)

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C. Futrell-

-T. L. McConnell

>C. McNeilly

.V. M.-Kapila File:-MC-801.01 i

p ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION REVISION CONTAINMENT ICE CONDENSER DOORS 5

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CONTAINMENT SYSTEMS' ICE CONDENSER DOORS LIMITING CONDITION FOR OPERATION 3.6.5.3 The ice condenser inlet doors, intermediate deck doors, and top deck-doors shall be closed and OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

2, ACTION:

With one or more ice condenser doors open or otherwise inoperable, POWER OPERATION may continue for up to 14 days provided the ice bed temperature is monitored at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the maximum ice bed temperature is maintained less than or equal to 27'F; otherwise, restore the doors to their closed positions or OPERABLE status (as applicable) within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at-least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS O

4.6.5.3.1 Inlet Doors - Ice condenser inlet doors shall be:

Continuously monitored and determined closed by the inlet door

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position monitoring system, and b.

Demonstrated OPERABLE during shutdown at least once per months by:

1)

Verifying that the torque required to initially open each door is less than or equal to 675 inch pounds; 2)

Verifying that opening of each door is not impaired by ice, frost or debris; 3)

Testing a sample of at least 50% of the doors and verifying that the torque required to open each door is less than 195 inch pounds when the door is 40 degrees open. This torque is defined as the " door opening torque" and is equal to the nominal door torque plus a frictional torque component. The doors selected for determination of the " door opening torque" shall be selected to ensure that all doors are tested at least once during two test intervals; O

Amendment No.

(Unit 2)

McGUIRE - UNITS 1 and 2 3/4 6-37 Amendment No.

(Unit 1)

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D ATTACHMENT II JUSTIFICATION AND SAFETY ANALYSIS O

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9 JUSTIFICATION AND SAFETY ANALYSIS The current Technical Specification 4.6.5.3.1 requires that the ice condenser inlet doors be demonstrated operable during shutdown at least once every 9 months. The testing and surveillance required to demonstrate operability is time consuming as well as requiring a power plant shutdown. In the past the ice condenser door testing was scheduled concurrently with the ice weighing operations in the containment. The ice weighing surveillance for the containment is also required once every 9 months. Weighing of the ice baskets of the ice condenser has t.a carried out with the reactor "0N-LINE".

Therefore,uan extension of the duration between two ice condenser door inspections to 18 months would avoid a plant shutdown to carry out the ice condenser door surveillance.

The proposed Technical Specification changes would require that the sur-veillance on the ice condenser doors be performed at least once every 18 months rather than every 9 months as currently required. The extension of the surveillance interval to 18 months would not be a safety concern due to high reliability and forgiving design features of the ice condenser doors as discussed below:

High Reliability of the Ice Condenser Doors Due Power has reviewed the surveillance records of the ice. condenser door to 2

evaluate the door reliability. Unit 1 and Unit 2 doors have been inspected periodically since 1981 and 1983 respectively. There have been 416 individual inspections of the doors at Unit 1 and 216 door inspections at Unit 2.

There has not been any failure to meet the acceptance criteria for door testing in all the inspections / tests carried on to date. One reason for the excellent surveillance history of the ice condenser doors is a design change made to the McGuire door seals to prevent the doors from freezing closed. The efficacy of this modification has been commended in the NRC Safety Evaluation Report (Amendment 36, License NPF-9, 1984) concerning the door surveillance requirements (Ref. 1).

In view of the fact that the ice condenser doors.have a history of very high reliability, extension of the surveillance interval to 18 months is not a safety concern.

Forgiving Design Features of the Ice Condenser Doors inte acceptance criteria for the ice condenser door passing the inspection is that the door should open with relatively low torque of 675 inch-pounds. This translates into a force of 17 pounds at the door handle. A very small change in the differential pressure across the door exerts a large force on the door.

During a LOCA or a high energy line break within the containment, the internal pressure of the containment will rise to between 0 to 15 psig. A rise of internal pressure of about 0.1 psia will open all ice condenser doors by exerting a force of at least 160 lbs.

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A rise of pressure of 1 or 2 psig following an accident would open the ice condenser doors even if these were obstructed by debris or were restrained shut by a Is" thick iron bar (Ref. 2).

The ice condenser doors are made of 0.5 inch thick Fiber Reinforced Polyester (FRP) reinforced by 6 steel ribs.

The doors are expected to rupture during accident if the doors fail to open, thus venting steam to the ice condenser (Ref. 3). The ice condenser doors are designed so that it is virtually impossible to keep the doors shut during any small increase of containment pressure. Surveillance or testing of the doors does not make the doors more or less reliable as far as accident uitigatien is concerned.

The doors pe. form two distinct functions. While the doors are shut, the doors insulate the lower containment from the ice condenser to prevent the degradation of the ice condenser by sublimation, etc. This function is monitored continuously by the inlet door position monitoring system. The second function for the doors is to open during a LOCA or high energy line break.- Our analysis has demonstrated that the doors are virtually impossible to keep shut in event of a minor rise of containment pressure. The extension of the. ice condenser door surveillance interval to 18 months would not have any adverse impact on the proper functioning of the ice condenser doors.

Conclusion The ice condenser doors at McGuire have a history of very reliable operation for several years. The doors are designed to make a door-failure virtually impossible during a LOCA or a large high energy line break; the proper functioning of the door during an accident is a passive design feature and it does not depend upon performance of periodic surveillance and testing of the doors. The current Technical Specifications for the ice condenser doors are excessive as well as disruptive. The functional readiness of the ice condenser doors as deemed necessary in the Bases 3/4.6.5.3 (pg. B 3/4 6-5) is adequately addressed by operability of the Door Position Monitoring System and door surveillance once every 18 months.

The extension of the door surveillance interval to 18 months would not have any significant adverse safety consequences.

REFERENCES:

1.

NRC Safety Evaluation Report related to Amendment No. 36 to Facility Operating License NPF-9 and Amendment No. 17 to Facility Operating License to Facility Operating License NPF-17.

2.

Letter frem H. B. Tucker to Dr. J. Nelson Grace (NRC, Region II) dated Feb. 8, 1985. Sub,,9ct : LER 413/85 Concerning the Blockage of Ice Condenser Lower Inlet Doors (Catawba Nuclear Station).

3.

McGuire FSAR Section 6.2.2.8.

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e ATT/.CIDfENT III SIGNIFICANT HAZARDS EVALUATION

ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION The proposed changes to the Technical Specification 4.6.5.3.1 have been reviewed pursuant to the requirements of 10CFR50.91. The following analysis provides a determination that the proposed amendment of the Technical Specifications does not involve any significant hazards considerat. ion, as defined by 10CFR50.92.

The proposed amendment seeks to relax an excessive surveillance requirement.

The Duke Power review of.the ice condenser door surveillance records concludes that higher frequency of door testing does not contribute to improving the reliability of the doors. S?nilarly reduction of routine surveillance of the ice condenser doors does not reduce their reliability in fulfilling their protective function during an accident. The protective function of the doors is semi-passive and the doors do not require much maintenance or testing.

Therefore, the extension of the surveillance interval to 18 months is not a safety concern.

The proposed amendments would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) Involve a significant reduction in a margin of safety.

Based upon the preceding analysis, Duke Power Company concludes that the proposed amendments do not involve a significant hazards consideration.

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JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION

.The proposed changes to Technical Specification 3/4.7.13 are.taken from the-proposed STANDBY SHUTDOWN SYSTEM (SSS) Technical Specification for the'McGuire Nuclear Station. By letters dated September 7, 1984 and April 9, 1985 Duke proposed a SSS specification. for McGuire and included the technical justification and a Significant Hazards Consideration Analysis (attached). Those analyses are applicable to this amendment raquest.

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o-Duxe POWER GOMPANY

,J P.O. BOX 33189 CHARLOTTE, N.C. 28242 i HAL B. TUCKER '

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-, 2.7E September 7.. 1984

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P Mr. Harold RL Denton,' Director L

^

Office of-Nuclear Reactor Regulation L

U.iS. Nuclear Regulatory Commission' Washington,1D.-C. 20555L

-Attention:-.Ms. ET Gi Adensam, Chief Licensing Branch No. 4

Subject:

'McGuire Nuclear Station

Docket.Nos. 50-369 and 50-370 Proposed Technical Specification

. Standby Shutdown System

=

Dear Mr.;Denton:

i Attached are~ proposed license amendments to Facility Operating Licenses NPF-9

.and NPF-17 for McGuira Nuclear-Station Units 1 and 2, respectively.L

[

yT,hese proposed Technical Specifications provide the operability and sur -

veillance requirements for the Standby. Shutdown System.

. Attachment 2

' Attachment 1 contains the proposed Technical Specification; changes.

discusses.the Justification and Safety Analysis to support the proposed changes, and provides an_ analysis performed in accordance with the standards contained in 10 CFR 50.92 as required by 10 CFR 50.91.

This analysis concludes.that the pro-posed amendments.have been reviewed and have been' determined to have no. adverse

safety;or' environmental impact.

This request involves one application for amendment to McGuire's Technical a

Specifications. Accordingly.-pursuant to 10 CFR 170.21, a check for $150.00

-is enclosed.

Please feel free'to contact us if you require any additional information.

Very truly yours,

!?

Js/Ilal B. Tucker Hal B. Tucker l

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'Mr._ Harold;R. Denton, Director-

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~ September ~7.

1984-

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Page 2~

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L HAL B;' TUCKER, being duly sworn, states that he is'Vice President of-j' Duke PowerLCompany; that_he is authorized on the part of said. Company to L

sign'and. file with the Nuclear Regulatory Commission this revision to the McGuire Nuclear Station Technical Specifications, Appendix A to License No.zNPF-9 and NPF-17;-and that all statements and matters set forth g

i therein are true and correct.to the best of his knowledge.

i.-

t s/Hal B. Tucker

.Hal B. Tucker,.Vice President c.

. Subscribed and sworn to before me this 7th day of September 1984.

l Es/ Sue C. Sherrill Notary P3blic

.My Commission Expires:

September 20, 1984 3'

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I., *

.,Mr; Harold-R.-Denton,-Director

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. fSep'tember 7, 1984-

Page 3.

W cc: 1Mr.fJames.P.=0'Reilly Regional Administrator

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U.-!S. Nuclear Regulatory Commission

Region II.:

1011Marietta' Street, Suite 3100

' Atlanta, Georgia 30323-Mr. Ralph Birkel.

Project Manager-Office of Nuclear-Reactor Regulation U. S.l Nuclear Regulatory Commission'-

1 Washington, D. C.

20555 Mr. W. T. Orders

' Senior-Resident Inspector 4-

-McGuire Nuclear Station ibec:

R.. L'. Gill N.~A.- Rutherford K. S. Canady H. B.-Tucker G. E. Vaughn R. 0.1Sharpe.

..RP WN0uellette D..Mendezoff.

J. R. Hendricks-J. A.' Oldham R. C.:Futrell T.

A.~ Ledford

-H. D. Brandes-T.'C.'McNeekin

' B. F. Caldwell R.

W.' Revels D. W. Murdock

-H. L. Davenport Fil'e: MC-802.02 r

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s PLANT SYSTEMS 3/4.7.14 STANDBY SHUTDOWN SYSTEM LIMITING CONDITION FOR OPERATION 3.7.14 The Standby Shutdown System (SSS) shall be OPERABLE.

. APPLICABILITY:

MODES 1, 2, and 3.

. ACTION:

a.

With the Standby Shutdown System inoperable for more than 7 days, prepare and submit a Special Report to the Commis-sion pursuant to Specification 6.9.2 within the next 30 days outlining the cause of the inoperability, corrective actions being taken, and plans for restoring the system to OPERABLE status; and b.

With one or more SSS components inoperable, as determined by the performance of surveillance requirements of Specification 4.7.14, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> verify the OPERABILITY of fire detection and suppression systems (if installed) in areas as identified in Table 4.7-2 and, as appropriate, perform the actions identified in notes 3 and 4 of the table.

c.

With the total leakage from UNIDENTIFIED LEAKAGE, IDENTI-FIED LEAKAGE and reactor coolant pump seal leakoff greater than 26 gpm, declare the Standby Makeup Pump inoperable and take ACTIONS a. and b., above.

Ifequipmentwhichconstitutesf$noperablecomponentofSSS d.

is located inside containment, then repairs shall be effected at the first outage which permits containment access.

e.

If thr: diesel generator is inoperable, repairs shall be effected within 60 days, or a report shall be prepared within the next 14 days which:

1) details the extent of repairs required, 2) outlines the schedule for completion of thase repairs, and 3) provides a basis for continued operation.

f.

The provisions of Specifications 3.0.3 and 3.0.4 do not-APP y-etce ne f op bblic l

SURVEILLANCE REQUIREMENTS 4.7.14.1-The Standby Shutdown System diesel generator shall be demon-strated OPERABLE:

a.- At least once per 31 days by verifying:

McGuire - Units 1 and 2 3/4 7-47 o is,,s,i -, e -, ^ rs y g af/ /# #"

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1): The fuel level in the-fuel storage tank is greater than or equal to 4.0 feet, and

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2).The diesel starts from ambient conditions and operates y

for at lecst 30 minutes at greater than or equal to 700 kW.

b.

By sampling new fuel oil in accordance with ASTM D4057-81 prior to addition to the storage tanks and:

1) -By verifying in accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has:

a)- An API Gravity of within 0.3 degrees at 60 F or a specific gravity of within 0.0016 at 60/60*F, when compared to the supplier's certificate or an absolute specific gravity at 60/60 F of greater than or equal to 0.83 but less than or equal to 0.89 or an API gravity at 60 F of greater than or equal to 27 degrees but less than or equal to 39 degrees.

b)

A kinematic viscosity at 40*C of greater than or equal to 1.9 centistokes but less than or equal to 4.1 centistokes (on a Saybolt Universal Viscosity at 100 F of greater than or equal to 32.6 SUS but less than or equal to 40.1 SUS).

c) A flash point equal to or greater than 125 F, and d) A clear and bright appearance with proper ccior when tested in accordance with ASTM D4176-82.

2) _ By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM 0975-81 are met when tested in accordance with ASTM 0975-81 except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 or ASTM D2622-82.

c.' At least once per,18 months by subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

4.7.14.2 The Standby. Shutdown System diesel starting 24-volt Nickel-Cadmium battery bank and charger shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that the overall battery voltage is greater than or equal to 24 volts.

b.

At least once per 18 mo.iths by verifying that:

McGuire - Units 1 and 2 3/4 7-48

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1). The batteries and battery racks show no visual indi-cation of physical damage or abnormal deterioration, and

2) The battery-to-battery and terminal connections are clean, tight, and free of corrosion.

4.7.14.3 The Standby Makeup Pump water supply shall be demonstrated OPERABLE by:

a.

Verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that IDENTIFIED LEAKAGE, UNIDEPTIFIED LEAKAGE, and Reactor Coolant Pump Seal leakoff do not exceed a total of 26 gpm.

b. ' Verifying at least once per 7 days:

1) That the requirements of Specification 3.9.10 are met and the boron concentration in the storage pool is greater than or equal to 2000 ppm, or

2) That the refueling water' storage tank is capable of being aligned to the Standby Makeup Pump.

c.

Verifying at least once per 92 days that the Standby Make-up Pump develops a flow of greater than or equal to 26 gpm at a pressure greater than or equal to 2485 psig.

4.7.14.4 The Standby Shutdown System 250/125-Volt Battery Bank and its associated charger shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying:

1) That the electrolyte level of each battery is above the plates, and

2) The total battery terminal voltage is greater than or equal to 258/129 volts on float charge.

b.

At least once per 92 days by verifying that the average specific gravity is greater than or equal to 1.200.

c.

At least once per 18 months by verifying that:

1) The batteries, cell plates, and battery racks show no visual indications of physical damage or abnormal deterioration, and

2) The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

4.7.14.5 The Steam Turbine Driven Auxiliary Feedwater Pump and associ-ated components shall be demonstrated OPERABLE at least once McGuire - Units 1 and 2 3/4 7-49

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.per 18 months by verifying that.the "C" solenoid is capable of being deenergized to open valve SA48ABC to provide steam supply to the turbine driven auxiliary feedwater pump.

!4.7.14.6

' Standby Shutdown System instrumentation shall be demonstrated OPERABLE by performance of surveillance requirements listed in Table 4.7-3.

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McGuire - Units 1 and 2 3/4 7-50

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i TABLE 4.7-3 4

MINIMUM CHANNEL CHANNEL' READOUT CHANNELS Instrument CHECK CALIBRATION LOCATION OPERABLE Reactor Coolant Pressure M

R SSF Control 1

Panel

' Pressurizer Level M

R SSF Control 1

Panel St'eam Generator Level M

R SSF Control 1/S.G.

Panel Incore Temperature M

R SSF Control 1

Panel Standby Makeup Pump Flow N/A R

SSF Control 1

Panel l

McGuire - Units 1 and 2 3/4 7-52 L'

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.t.ut ama. *,ys t.w Cong.onent Aci!GN Statements EI With one or more of the following Verify operabilityI ) of fire detectors and t

((

555 components inoperable suppression systems (if installed) in the following areas:

M" EL 716 EL 733 EL 750 Control Battery Cable Turbine Driven Motor Driven Containment Other EE-KK EE-KK EE-KK Room Room Rooms AFW Pump AfW Pump Actions A.

Diesel Generator X

X X

X X

X X

X (2)

(3),(4) 8.

Diesel starting 24-Volt battery X

X X

X X

X X

X (2)

(3),(4) bank and charger b

C.

Standby makeup pump and water supply X

X X

}l D.

250/125 V battery bank and associated X

X X

(2)

(4) charger E.

Steam Turbine Driven Auxiliary Feedwater X

Pump Solenoid "C" F.

Instrumentation Reactor Coolant Pressure X

X X

(2)

Pressurizer Level X

X X

(2)

Steam Generator Level X

X X

(2)

Incore Temperature X

X X

(2)

NOTES: (1) If fire detection and/or suppression systems are inoperable, then the ACTION statement (s) of the appilcable fire detection and/or suppression technical specification (s) shall be complied with.

(2) Monitor containment air temperature at least once per hour at the locations specified in Specification 4.6.1.5.1 or 4.6.1.5.2, in lieu of verification of operability of systems inside containment.

(3) With this component inoperable, verify operability of off-site power and one blackout diesel generator.

(4) With this component inoperable, then denoted areas of both units are affected.

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ATTACHMENT II A.

Technical Justification The Standby Shutdown System (SSS) provides an additional degree of redund to the multiple systems which are available, under emergency conditions achieve and maintain the reactor in the Hot Standby Mode.

to The surveillance requirements and action statements presented in the Draft Technical Specification of this redundant system.are reasonable considering the safety significance of a By design, the SSS is intended to respond to those low-probability fire and/or sabotage events which render both the control room and automatic safety systems inoperable.

Because of the low probability of occurrence of these events, the action statements in the proposed specifica-tion rely on compensatory action, timely repair or return to operability and if necessary, a justification for continued operation.

Because the SSS performs a redundant fire protection function, compensatory action relies largely on assurance of the cperability of fire detection and suppression systems.

Table 4.7.2 of the proposed specification establishes requirements for operability of fire detection end suppression systems.

The surveillance requirements presented in Attachment I are intended to provide assurance that the various components of the Standby Shutdown System (SSS) per-form their functions.

The surveillance requirements are based largely on exist-ing SSS Technical Specification for the Catawba Nucicar Station, which was ap-proved prior to the issuance of the fuc1 load license for Unit 1 of that plant.

Also considered in the formulation of the surveillance requirements were 2xist-ing McGuire Technical Specifications, such as those for the 1E Diesel Gcnerators, Refueling Water Storage Tank, Fire Protection & Detection Systems, and other Tech Specs which are related to the safe operation and/or shutdown of the plant.

Following is an item-by-item justification of the proposed Technical Specification.

3.7.14 - Action Statements The SSS Technical Specification for Catawba requires that with the SSS inoperable, least hot standby in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. restore the inoperable equipment to operable Considering that'the SSS is a third line of defense for a low-probability fire event, it is not reasonable to require a plant shutdown for tlose occasions when a component of the system is inoperable.

For exampic, specification 4.7.14.1.C requires that every 18 months, the diesel generator be subjected to an inspection which involves a substantial disassembly.

It is not reasonable to expect that this inspection can be performed within 7 days. Therefore, at a multi-unit plant such as McGuire, both units would have to be shut down every 18 months for an extended outage,1.resumably in addition to at least one unit's refueling outage.

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ACTIONSTATEMENhC, Regarding total allowed leakage, is virtually the same as Catawba's..The quantities of identified and unidentified leakage' allowed 7 5 are doverned by' specification 3.4.6.2, " Operational Leakage." The allowance

.for Reactor Coolant Pump Seal Leakoff (15 GPM) is~a conservative amount which isla1mest (3.75 GPM/ Pump vs.-4.3) equal to-the degraded' seal performance value used as'the design basis. The.55 GPM/ Pump difference is not a safety Concern.

ACTION STATEMENT D,-Recognize $ that equipment inside containment is inaccessible during operation. As not.ed previously, it is not reasonable to require'a shut-down.to repair an inoperal.le component of the SSS. However, any' inoperable equipment will be repairea at the first available opportunity (i.e., Planned or forced outige.)

ACTION STATEMENT E, Provides an adequate length of time to repair'or insp'ect the diesel generator. During this time, at least one emergency diesel and off-site power.will be operable.

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The verification of operability of. fire detection and suppression systems in,the locations specified in Table 4.7.2 will provide adequate acs'aranca that.thohe systems for which.the SSS provides a backup will not be damaged by fire.'

4.7.14 - Surveillance Requirements 4.7.14.1,

s a.

1) A fuel leN 1 of 4.0 feet in the fuel storage tank is sufficient-to operate the diesel for at least 48, hours.

Ip.is assumed that, uithin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />,off-shs pWor,rkbe rest 5 red or additional fuel can be added to ths' M t"ahe Cank.

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' 2) This specifiestioii'ifidentical to the requirement

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for Catawba. This requirement is loosely based on s

Specification 4.8.1.1.'2a.4 (IE Diesels), but considers that the'SSF Diesels are manual start only, and not subject to a test ft'equency based on successful starts,

b. -Requirements for new fuel being added to the tank are virtually identical t

to those of the 1E Diesels.

c.

Manufactn,rer-recommended 18-month

• inspection requirement is similar to that for 1E Dioscis, except that in view of the redundant nature of the SSS function, plant shutdown is not required.

/t 4.7.14.2

?*

3 requirementsforSSSdiese)startingbatteriesaresimilarto Surveillan those for 1E Diesels, but reflect the fact that the batteries are nickel-l cadmium nr.d. therefore, critoria for specific gravity and eletrolyte icyc1 are not applicable.

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A 4.7.14.3 The specification to verify operability of the Standby Makeup Pump a.

water supply is based on the Catawba specification, except that-the alternate water supply of 112.320 gallons has been replaced by the Refueling Water Storage Tank (RWST). The RWST is verified weekly (per T/S. 3/4.5.5) to contain at least 372,100 gallons of water borated to a concentration between 2,000 and 2,100 PPM.

b.

As noted previously, the flow rate of 26 GPM is based upon the total of allowed leakage (Identified and unidentified) and reactor coolant pump seal leakoff. The required pressure is equal to the pressurizer safety set point of 2485 psig. According to the pump manufacturer, the pump can deliver 26 GPM at 2735 psig.

4.7.14.4 The specifications for surveillance of 250/125 Volt Battery Bank are based on those for the 1E Diesel batteries, although the surveillance is performed less frequently. This relaxed frequency is consistent with the probability that the batteries will be required.

4.7.14.5 The Steam Turbine Auxiliary Feedwater pump is determined operable per Tech Spec 4.7.1.2a.

Specification 4.7.14.5 assures that a source of steam is available through valve SA48ABC, which fails open on deenergization of the "C" solenoid.

4.7.14.6 The instrumentation surveillance requirements are consistent with speci-fications 3.3.3.5 and 3.3.3.6.

B.

Significant Hazards Considerations Analysis This analysis is provided to determine whether the proposed amendment involves a significant hazards consideration, as' defined by 10CFR50.92.

The proposed Technical Specification will not increase the probability of an accident occurring, either one previously evaluated or not, because the SSS is used to mitigate events which are already in progress, and because the equip-ment upon which surveillance is performed is not directly associated with equipment whose failure would cause an accident.

Inoperability of the SSS could reduce the capability to mitigate the conse-quences of an event which disables the normal and emergency shutdown systems, and renders the control room uninhabitable. Therefore, controls on the in-operability of the SSS, and formalized surveillance procedures, are expected to contribute to a not decrease in potential probability of an accident.

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's The Federal Register Notice which published the interim' final rule contained examples of amendments that are not li..ely to involve significant hszards considoratione. One example'was a char.ge that constitutes additional limita-tions, restrictions, or controls not presently included in the Technical Specifications.

The proposed Technical Specification does not reduce or affect any margin of safety, except as noted above with respect to a potential decrease in conse-quences to the public.

Based on the above Duke Power Company concludes _that the proposed amendment does not involve a significant hazards consideration.

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. / 4' (R UJ 0 DUKE POWER GoxPAxy P.O. Box 33189 CHARLO1TE. N.C. 28242 HAL H. TUCKER vna e e a$

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April 9, 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation

.U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Attention:

Ms. E. G. Adensam, Chief Licensing Branch No. 4

Subject:

McGuire Nuclear Station Docket Nos. 50-369 and 50-370 Proposed Technical Specification (Supplement 1)

Standby Shutdown System

Dear Mr. Denton:

By letter dated September 7, 1984 Duke initially submitted a proposed Technical Specification to address operability and surveillance requirements for the

i.

McGuire Stendby Shutdown System.

In recent months, discussions have taken place which indicate that the pro-4 posed Technical Specification as previously submitted should be supplemented.

Accordingly, please find attached a proposed Technical Specification address-

^

ing the SSS which supercedes the previous proposed Technical Specification in its entirety. However, please note that the previously provided Technical

}

Justification and Significant Hazards Consideration Analysis remain valid.

Duke requests that the effective date of the amendment be 30 days from the approval date in order to allow adequate time to fully implement the new procedural requirements contained in this Technical Specification.

Inasmuch as this proposal supplements a previous submittal still under NRC review, no application fee is considered necessary.

Please feel free to contact us if you require any additional information.

Very truly yours, d

Hal B. Tucker RLG:smh-Attachment f (A24 ? A n / /f

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L.

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s Mr. Harold R. Denton, Director April 9, 1985:

Page 2 Dr. J. Nelson Crace, Regional Administrator cc:-

U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100

' Atlanta, Georgia 30323 Mr. Darl Hood Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington,'D.C. 20555 Mr. Dayne Brown, Chief Radiation Protection Branch Division of Facility Services Department of Human Resources P.. O. Box 12200 Raleigh, North Carolina 27605

-Mr. W. T. Orders-Senior Resident Inspector McGuire Nuclear Station n-,,

' ?.:s

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,o Mr. Harold R.'-Denton, Director-April 9, 1985

,Page 3-bec:

R.-L. Gul.

N. A.:. Rutherford '

K._-'S. Canady H. B. Tucker G. E. Vaughn R. O.'Sharpe II:.R.'N.' 'Ouelletta'

..E. O. McCraw

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T. L. McConnell

-J. R. Hendricks

.-J. A.-Oldham R. C. Futrell

'T. A. Ledford

~

H.-D. Brandes R. W. Revels' D. W. Murdock H. L. Davenport File: MC-802.02

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t PLANT SYSTEMS 3/4.7.14 STANDBY SHOTDOWN SYSTEM LIMITING CONDITION FOR OPERATION 3.7.14 The Standby Shutdown System (SSS) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With less than the minimum SSS equipment in Table 3.7-8 OPERABLE, restore the inoperable equipment to OPERABLE within 7 days, or provide equivalent capability N.

to achieve HOT STANDBY and restore the inuperable equipment to OPERABLE within 60 days, or be in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and HOT SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

t b.

The provisions nf Specifications 3.0.3 g

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3.0.4 M '.0.4

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g SURVEILLANCE REQUIREMENTS 4.7.14./

The Standby Shutdown System diesel generatcr shall be h

t demonstrated OPERABLE:

k g-a.

At least once per 31 days by verifying:

1) The fuel level in the fuel storage tank is greater than or equal to 4.0 feet, and

2) The diesel starts from ambient conditions and operates for at least 30 minutes at greater than or equal to 700 kW.

b.

By sampling new fuel oil in accordance with ASTM 04057-81 prior to addition to the storage tanks and:

1) By verifying ' fn accordance with the tests specified in ASTM D975-81 prior to addition to the storage tanks that the sample has:

a) Aa API Gravity of within 0.3 degrees at 60 F or a i

specific gravity of within 0.0016 at 60/60 F, when compared to the supplier's certificate or an absolute specific gravity at 60/60 F of greater than or equal to 0.83 but less than or equal to 0.89 or an API gravity m.

60 F of greater than or equal to 27 degrees but less than or equal to 39 degrees.

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McGuire - Units 1 and 2 3/4 7-47 u

b) A kinematic viscosity at 40 C of greater than or equal to 1.9 centistokes but less than or equal to 4.1 centistokes (on a Saybolt Universal Viscommeter at 100 F of greater than or equal to 32.6 SUS but less than or equal to 40.1 SUS).

c) A flash point equal to or greater than 125 F, and d) A clear and bright appearance with proper color when tested in accordance with ASTM 04176-82.

2) By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are met when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D1552-79 or ASTM D2622-82.

At least once per 18 months by subjecting the diesel to an c.

inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

4.7.14.[)

The Standby Shutdown System diesel starting 24-volt Nickel-Cadmium battery bank and charger shall be demonstrated OPERABLE:

At least once per 31 days by verifying that the overall a.

battery voltage is greater than or equal to 24 volts, b.

At least once per 18 months by verifying that:

1) The batteries and battery racks show no visual indication of physical damage or abnormai i

deterioration, and

2) The battery-to-battery and terminal connections are y

clean, tight, and free of corrosion.

4.7.14./

The Standby Makeup Pump water supply shall be demonstrated l

OPE.9ABLE by:

Verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that IDENTIFIED a.

LEAKAGE, UNIDENTIFIED LEAKAGE, and Reactor Coolant Pump Seal leakoff do not exceed a total of 26 gpm.

b.

Verifying at least once per 7 days:

1) That the requirements of Specification 3.9.10 are met and the boron concentration in the storage pool is

_ greater than or equal to 2000 ppm, or

2) That the refueling water storage tank is capable t' being aligned to the Standby Makeup Pump.

McGuire - Units 1 and 2 3/4 7-48 i

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h Verifying at each Cold Shutdown, but not more than once c.

every 92 days, that the Standby Makeup Pump develops a flow of greater than or equal to 26 gpm at a pressure greater than or equal to 2485 psig.

4.7.14.A The Standby Shutdown System 250/125-Volt Battery Bank and its associated charger shall be demonstrated OPERABLE:

At least once per 31 days by verifying:

a.

1) That the electrolyte level of each battery is above the plates, and

2) The total battery terminal voltage is greater than or equal to 258/129 volts on float charge.

b.

At least once per 92 days by verifying that the average specific gravity is greater than or equal to 1.200.

At least once per 18 months by verifying that:

c.

1) The batteries, cell plates, and bl.ttery racks show no visual indications of physical damage or abnormal deterioration, and

2) The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

4.7.14.

The Steam Turbine Driven Auxiliary Feedwater Pump and associated components shall be demonstrated OPERABLE in accordance with portions of-Specification 3/4.7.1.2 applicable to SSS.

4.7.14./

The "C" solenoid shall be demonstrated OPERABLE and capable of being deenergized to open valve SA48ABC to provide steam supply to the turbine driven auxiliary feedwater pump at least once per 18 months.

9 4.7.14./

Standby Shutdown System instrumentation shall be demonstrated OPERABLE by performance of surveillance requirements listed in Table 4.7-2.

McGuire - Units 1 and 2 3/4 7-49 L

[

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Table 3.7-8 Standby Shutdown System.

Minimum Equipment.

INSTRUMENT READOUT-LOCATION' MINIMUM CHANNELS OPERABLE Reactor' Coolant Pressure SSF Control Panel 1

Pressurizer Level SSF Control Panel-1 Steam Generator Level SSF Control Panel

-1/S.G.

Incore Temperature SSF Control Panel 1*

Standby Makeup Pump Flow SSF Control Panel 1

• • • • • • • • • • • • • • • • • • • • • -a******************-***********

EQUIPMENT LOCATION Diesel Generator and associated switchgear SSF Diesel starting 24-Volt battery. bank and charger SSF-Standby makeup pump and water supply Containment / Spent Fuel Pool 250/125 V battery bank, associated charger, and associated switchgear SSF Steam Turbine Driven -Auxiliary Feedwater Pump Aux. Bldg.

Solenoid "C" to valve SA 48 ABC Interior Doghouse

• Capable of being' connected at SSF Control Panel.

U, f

.a Table 4.7 Standby. Shutdown System Instrumentation Surveillance Requirements INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION Reactor Coolant Pressure M

R Pressurizer Level M

R Steam Generator Level M

R Incore Temperature M

R Standby Makeup Pump Flow N/A R

C.

y 's,

PLANT SYSTEMS BASES 3/4.7.13 GROUNDWATER LEVEL This specification is provided to ensure that groundwater levels will be monitored and prevented from rising to excessively high levels which could cause unacceptable. structural stresses in the reactor containment and/or auxiliary building due to uplift forces.

Structural stresses due to high

. groundwater levels will not exceed allowable stresses until groundwater rises to elevation 732 MSL.

3/4.7.13 STANDBY SHUTDOWN SYSTEM The OPERABILITY of the SSS-ensures that a fire will not preclude achieving safe shutdown.

The SSS equipment are independent of areas where a fire could damage systems normally used to shutdown the reactor.

This capability is consistent with GDC 3 and 10CFR50, Appendix R.

The equivalent shutdown capability provided when the SSS is inoperable depends on the specific equipment involved and, therefore, should be sufficient to assure that the intended shutdown actions can be accomplished, or that fires can be reasonably precluded during that time for which SSS equipment would otherwise be required, consi. Stent with the SSS design basis.

Any temporary procedures or special fire watch patrols established to provide this equivalent capability shall be reviewed and approved prior to implementation in accordance with Section 6.0 of Technical Specification.

McGuire - Units 1 and 2 8 3/4 7-8 a.:

r

' JUSTIFICATION.AND ANAYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION

'T1.a proposed change ~to Specification 6.5.1.8 and'6.8.2 would allow the Superintendent of Station Services to review and approve changes to the

' station security program.

The' proposed change to 6.10.2 is necessary to

. reflect Duke Power's commitment to Regulatory Guide 1.88 which adopts ANSI' N 45.2.9-1974~..

.These changes were requested in a letter from Mr. Hal B. Tucker to Mr.

Harold R.-Denton, dated April 25, 1985 for the McGuire Nuclear Station.

(attached). The technical-justification and Significant Hazards Consideration.

contained therein are applicable to this change request.

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- DUKE POWER GOMPANY

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- P.O. box 33180 y.y CHARLOTTE. N.C. 28242

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,HALB.TUCKEH >

Trtre w er vur enreso.? -

(704) 373 453a 3

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-mm-April _25, 1985 b

4 DMr. Harold'R. Denton, Director Office of Nuclear Reactor Regulation LU. S.(Nuclear. Regulatory Commission u

4 Washington,LD.jC. 20555-

Attention

: Ms. E. G. Adensam, Chief

' Licensing-Branch No. 4

.Subj ect:.McGuire. Nuclear Station Docket-Nos. 50-369 and 50-370 Technical Specification Amendment, Administrative-Controls L

Dear Mr. Denton:

SAttach'd are proposed license amendments to Facility Operating Licenses NPF-9 e

.q L'

.and NPF-17 for McGuire Nuclear Station Units 1 and 2, respectively.

L

-!The proposed changes to'the Technical Specifications are concerned primarily with

',~

iSection16.0, Administrative Controls, and reportability requirements associated with (the: revision of the regulatory requirements of 10 CFR 50.72 and.50.73 as presented

~

>in Generic Letter 83-43 (dated December 19, 1983).

<The-proposed' changes-to the Administrative Controls section of the Technical 0;

Specifications ^are intended to streamline.and, clarify the existing' controls. This

' proposal seeks to delete corporate _ organizational charts in Section 6.0 of the 7;f

- ( Technical 1 Specifications, as _ this ois ' redundant to Chapter 13 of the McGuire Nuclear.

Station' Final Safety Analysis _ Report, which is required (by 10 CFR 50.71) to be'

, updated-nnually.. The" proposal also seeks to include _the Superintendent of Integrateo Scheduling in the review and approval'~ process of procedures and events.

'Also included in this proposal is a change to base QA record retention on ANSI

,N45-2.9-1974, as. recommended by Regulatory Guide 1.88;'a rewording of Section 6.12,

~

RRigh Radiation. Areas, to clarify requirements; a change that would allow the

Station Services Superintendent to review and approve modifications to the Station

-Security' Program and associated procedures; to update the solid waste reporting irequirements-(Specification 6.9.1.7) to make these requirements consistent.with

'.10 CFR part 61;.and'to' correct an error in Table 4.11-1~where the word " inlet" was inadvertantly substituted.for'" outlet". -The proposed changes to Specification te

-6;9.1.7 and: Table.4.11-1.were recommended by an in-house QA audit. These proposed changes are purely administrative 1.1. nature.

' Also included in this submittal are several proposed changes to remove inconsistences ithat were introduced;when reporting requirements (per 10 CFR sections 50.72 and

~

50.73) were changed and Section 6.0 of the McGuire Technical Specifications was accordingly_ modified. At the time that.the Administrative Controls section of

,?T

~the: Technical Specifications was changed,-references elsewhere in the Specifications to'this section were not modified, causing redundancies and meaningless references.

This1 proposal seeks to. remedy these inconsistencies and is administrative in nature.

b O K,.- m a n W c r

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Mr.1 Harold R. Denton, Director April 25,1 1985 Page 2 contains the proposed changes to'the Administrative Controls Section and Table 4.11-1. 1 contains'the proposed changes to the Technical Specifications concerning reporting requirements. 'Each attachment contains a Justification and Safety Analysis in support of the proposed changes, and an

. analysis performed in accordance with the Standards contained in 10 CFR 50.92 as required by 10CFR50.91. These analyses conclude that no Significant Hazards Considerationsare involved.

This request involves one. application for. amendment to McGuire's Technical Specifications. Accordingly, pursuant to 10CFR170.21 a check for $150.00 is enclosed.

Please feel free to contact us if you require any additional information.

Very truly yours,.

s/Hal B. Tucker Hal B. Tucker

~JBD/mjf-Attachment-cc: -Dr. J. Nelson Grace, Regional Administrator 1.U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30323 Mr. W. T. Orders Senior Resident Inspector-

-McGuire Nuclear Station Mr. Dayne Brown Chief Radiation Protection Branch Division of Facility Services Department of Human Resources P. 0. Box 12200 LRaleigh, North Carolina 27605 g

L.

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Mr. Harold R. Denton,-Director-April:25. 1985

-Page 3 HIL B. TUCKER. being duly sworn, states that he is Vice President of-Duke Power Company; that he is authorized on the.part of said Company to. sign and file with the Nuclear Regulatory Cr-mission this revision to the McGuire Nuclear Station. License Nos. NPF-9 and NPF-17 and that all? statements and matters set forth therein are true and correct to the best of: his knowledge.

s/Hal B. Tucker Hal B. Tucker, Vice President Subscribed and sworn to before me this 25th day of April, 1985 s/ Sue C. Sherrill Notary Public My Commission Expires:

= September 20, 1989

2

~

LMrs Hirald R.'Danten Dirsctor'

'b April 25, 1985 Page 4

- bec:

-P. M. Abraham -

.R. L. Weber I

K.-S. Canady.

L. M. Coggins

'N. A. Rutherford E. M. Geddie R.~L.

Gill R. P. Ruth'(MNS)

-S. A. Gewehr.

J. W.'Hampton.(CNS)

P. B. Nardoci C. F. York (MNS)

W. H. McDowell J. W. Cox (CNS)

J G. W. Hallman p w

{.,,B. DayunR gW.70be11stte;-

R. D. Groux E. 0."McGiaE(MNS)

Tom Mooney R; C. Futrell-Tom Ryan (MNS)

D. Mendezoff (MNS)

R. T. Bond (ONS)

T.-L..McConnell (MNS)

A'. L. Snow M. L. Birch R. S. 3hatnager L. Lewis..

T. P. Harrall-V. M. Kapila R. S. Howard-(W)

J.-M. Stewart Section File: MC-801.01 J. Effinger Section File: -MC-801.02

.'J. Silver-(MNS)

Section File: MC-815.02 N

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JAttachm:nt I.a l

INDEX

ADMINISTRATIVE CONTROLS J

SECTION PAGE 6.1 RESPONSIBILITY.................................................

6-1 6.2 ORGANIZATION 6.2.ls0FFSITE....'..................................................

6-1 6.2.2 UNIT STAFF.~..................................................

6-1 FIGURE 6.2-1 [De1'eted]..............................................

6-3 FIGURE 6.2-2 [ Deleted]..............................................

6-4 TABLE.6.2-1

~ MINIMUM SHIFT CREW COMPOSITION.......................

6-5 6.2.3 -STATION SAFETY REVIEW GROUP (SSRG)

Function..................................................

6-7 Composition...............................................

6-7

-Responsibilities..........................................

6-7 Authority.................................................

6-7 Records...................................................

6-7 J

6.2.4 SHIFT TECHNICAL ADVIS0R......................................

6-7 6.3 -UNIT STAFF QUALIFICATIONS......................................

6-7 6.4 TRAINING.......................................................

6-7 6.5' REVIEW AND AUDIT 6.5.1. TECHNICAL REVIEW AND CONTROL Activities................................................

6-8 MCGUIRE - UNITS I and 2 XXI

Attachm2nt I.a TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT

~

MINIMUM OF DETECTION LIQUID RELEASE -SAMPLING-ANALYSIS TYPE OF ACTIVITY (LLD)

TYPE-

. FREQUENCY FREQUENCY ANALYSIS (mci /ml)(y)

1. Batch Waste P

P Release (4).

Each Batch Each Batch Principaf6gama 5x10 m Tanks Emitters

-6 I-131 lx10

'(Waste.

Monitor

-5 PL M

Dissolved and ix10 "ks and One Batch /M Entrained Gases cy e Monitor Tank)-

(Gama emitters)

-5 P

M H-3 lx10 Each Batch Composite (2)

-7 Gross Alpha-1x10

-8 P

Q Each Batch Composite (2) -.Sr-89,-Sr-90 5x10

-6 Fe-55 lx10

2. Continuous (5)-

Continuous (3) Composite (3)

Principaf6gama 5x10 Releases Emitters

-6 I-131 lx10 (Containment Ventilation

-5 M

M Dissolved and lx10 Grab Sample Entrained Gases o ensate_

. Drain Tank (Gama EmMers)

Discharge and

-5 Conventional M

H-3 lx10 Waste Water Continuous (3) Composite (3)

_y Treatment Gross Alpha lx10 1 System Outlet)

-8 l

Q Sr-89, Sr-90 5x10 Continuous (3) Composite (3)

-6 Fe-55 lx10 McGUIRE - UNITS'1 and 2 3/4 11-2

E Attachmsnt I.a 6.0 ' ADMINISTRATIVE CONTROLS 6:1 RESPONSIBILITY 6.1.1 The Station Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor (or during his absence from the control room, a designated individual) shall be responsible for the control room command function.

A management directive to this effect, signed by the Vice-President Nuclear Production shall be reissued to all Nuclear Production Department station personnel on an annual basis.

6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as described in the FSAR, Chapter 13.

l UNIT STAFF 6.2.2 The unit organization shall be as described in the FSAR, Chapter 13, and:

a.

Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1; b.

At least one licensed Operator for each unit shall be in the control room when fuel is in either reactor.

In addition, while either unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room; A Health Physics Technician # shall be on site when fuel is in either c.

reactor; d.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibil-ities during this operation; AsiteFireBrigadegfatleastfivemembersshallbemaintained e.

onsite at all times.

The Fire Brigade shall not include three members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and

1. The Health Physics Technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

McGUIRE - UNITS 1 and 2 6-1

r Attrchment I.0 (Figure Deleted)

McGUIRE - UNITS 1 ar.d 2 6-3

y-Attcchment I.a r

-(Figure Deleted) i McGUIRE - UNITS 1 and 2 6-4

Attachmtnt I.a' ADMINISTRATIVE CONTROLS 6.5 REVIEW AND AUDIT 6.5.1 TECHNICAL REVIEW AND CONTROL ACTIVITIES 6.5.1.1. Each procedure and program required by Specification 6.8 and other procedures which affect nuclear safety, and changes thereto, shall be prepared by a qualified individual / organization.

Each such procedure, and changes thereto, shall be reviewed by an. individual / group other than the individual /

group which prepared the procedure, or changes thereto, but who may be from the same organization as the individual / group which prepared the procedure, or changes thereto.

6.5.1.2 Proposed changes to the Appendix A Technical Specifications shall be prepared by a qualified individual / organization.

The preparation of each proposed Technical Specifications change shall be reviewed by an individual /

group other than the individual / group which prepared the proposed change, but who may be from the same organization as the individual / group which prepared the proposed change.

Proposed changes to the Technical Specifications shall be approved by the Station Manager.

6. 5.1. 3 Proposed modifications to unit nuclear safety-related structures, systems and components shall be designed by a qualified individual / organization.

Each such modification shall be reviewed by an individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which-designed the modification.

Proposed modifications to nuclear safety-related structures, systems, and components shall be approved prior to implementation by the Station Manager; or by the Operating Superintendent, the Technical Services Superintendent, the Superintendent of Integrated Scheduling, or the Maintenance Superintendent, as previously designated by the Station Manager.

6.5.1.4 Individuals responsible for reviews performed in accordance with Specifications 6.5.1.1, 6.5.1.2, and 6.5.1.3 shall be members of the station supervisory staff, previously designated by the Station Manager to perform such reviews.

Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary.

If deemed necessary, such review shall be performed by the appropriate designated station review personnel.

6.5.1.5 Proposed tests and experiments which affect station nuclear safety and are not addressed in the FSAR or Technical Specifications shall be reviewed by the Station Manager; or by the Operating Superintendent, the Technical Services Superintendent, the Maintenance Superintendent, or the Superintendent of Integrated Scheduling as previously designated by the Station Manager.

-McGUIRE - UNITS 1 and 2 6-8

1

-l Attachm:nt I.a ADMINISTRATIVE CONTROLS ACTIVITIES (Continued) 6.5.1.6 ALL REPORTABLE EVENTS and all violations of Technical Specifications

-shall be, investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence.

Such reports shall be Japproved by the Station Manager and transmitted to the Vice President, Nuclear Production,,andlto the Director of the Nuclear Safety Review Board.

6.5.1.7 The' Station Manager shall assure the performance of special reviews and in'estigations, and the preparation and submittal of reports thereon, as v

requested by the Vice President, Nuclear Production.

6.5.1.8--The station security program, and implementing procedures, shall be reviewed at least-once per 12 months.

Recommended changes shall be approved b

~

by.the Station Manager or Superintendent.of Station Services and transmitted to the Vice President, Nuclear Production, and to the Director of the Nuclear Safety Review Board.

.6.5.1.9 The~ station emergency plan, and implementing procedures, shall be reviewed at least once per 12 months.

Recommended changes shall be approved l

-by the Station Manager and transmitted to the Vice President, Nuclear Production, and to the Director of the Nuclear Safety Review Board.

6.5.1.10 - The Station Manager shall assure the performance of a review by a

- qualified individual / organization of every unplanned onsite release of radio-active material to the environs including the preparation and forwarding of reports covering evaluation, recommendations, and disposition'of the corrective

' ACTION to-prevent recurrence to the Vice President, Nuclear Production and to the Nuclear Safety Review Board.

6.5.1.-11 The Station Manager shall assure the performance of a review by a qualified individual / organization of changes to the PROCESS CONTROL PROGRAM,

~

0FFSITE DOSE CALCULATION MANUAL, and Radwaste Treatment Systems.

6.5.1.12. Reports documenting each of the activities performed under Specifi-cations 6.5.1.1 through 6.5.1.11 shall be maintained.

Copies shall be provided to the Vice President, Nuclear Production, and the Nuclear Safety' Review Board.

6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB)

FUNCTION-6.5.2.1 The NSRB shall function to provide independent review and audit of

. designated activities'in the areas of:

a.

2 Nuclear power plant operations, b.

Nuclear engineering, c.

Chemistry and radiochemistry, McGUIRE - UNITS 1 and 2 6-9

rc Attachm:nt I.a

-ADMINISTRATIVE CONTROLS RECORDS 6.5.2.11-Records of NSRB activities shall be prepared, approved, and distributed as indicated below:

.a.

Minutes of each NSRB meeting shall be prepared, approved, and forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Engineering, Construction, and Production, within 14 days l

following each meeting; b.

Reports of reviews encompassed by Specification 6.5.2.8 above, shall be prepared, approved and forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Engineering, Construction, and Production, within 14 days following completion of the review; and Audit reports encompassed by Specification 6.5.2.9 above, shall be c.

forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Engineering, Construction, and Production, and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

The Commission shall be notified and a report submitted pursuant to a.

the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the Station Manager; or by:

(1) the Operating Superintendent, (2) the Technical Services Superintendent, (3) the Maintenance Superintendent, or (4) the Superintendent of Integrated Scheduling, as previously designated by the Station Manager, and the results of the review shall be

-submitted to the NSRB and the Vice President, Nuclear Production.

6. 7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The Vice President, Nuclear Production, and the NSRB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; b.

A Safety Limit Violation Report shall be prepared.

The report shall be reviewed by the' Operating Superintendent and the Station Manager.

This report shall describe:

(1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence; McGUIRE - UNITS 1 and 2 6-13 Amendment No. (Unit 1)

Amendment No. (Unit 2)

1 Attachmant I.a ADMINISTRATIVE CONTROLS SAFETY LIMIT VIOLATION (Continued)

-c.

The Safety Limit Violation Report shall be submitted to the Commission, the NSRB and the Vice President, Nuclear Production, within 14 days of the violation; and d.

Critical operation of the unit shall not be resumed until authorized by the Commission.

6.8 PROCEDURES AND-PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978; b.

The applicable procedures required to implement the requirements of NUREG-0737; c.

Security Plan implementation;*

d.

Emergency Plan implementation; e.

PROCESS CONTROL PROGRAM implementation; f.

OFFSITE DOSE CALCULATION MANUAL implementation; and g.

Quality Assurance Program for effluent and environmental monitoring.

6.8.2 Each procedure of Specification 6.8.1 above, and changes thereto, shall be reviewed and approved by the Station Manager; or by:

(1) the Operating Superintendent, (2) the Technical Services Superintendent, (3) the Maintenance Superintendent, or (4) the Superintendent of Integrated Scheduling, as previously designated by the Station Manager; prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of Specification 6.8.1 above may be made provided:

a.

The intent of the original procedure is not altered; b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and McGUIRE - UNITS 1 and 2 6-14

• Review and approval may be performed by the Superintendent of Station Services.

Attachm nt I.a ADMINISTRATIVE CONTROLS

. PROCEDURES AND PROGRAMS (Continued')

c.

The change is documented, reviewed, and approved by the Station Manager; or by:

(1) the Operating Superintendent, (2) the Technical

' Services Superintendent, (3) the Maintenance Superintendent; or (4) the' Superintendent of Integrated Scheduling, as previously designated by the Stntion Manager, within 14 days of implementation.

6. 8.' 4 The following programs shall be established, implemented, and maintained:

a.

Reactor Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

The systems include RHR, Boron Recycle, Refueling Water, Liquid Waste, Waste Gas, Safety Injection, Chemical and Volume Control, Contain-ment Spray, and Nuclear Sampling.

The program shall include the following:

1)

Preventive maintenance and periodic visual inspection requirements, and 2)

Integrated leak test requirements for each system at refueling cycle intervals or less.

b.

In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

This program shall include the following:

1)

Training of personnel, 2)

Procedures for monitoring, and 3)

Provisions for maintenance of sampling and analysis equipment.

c.

Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation.

This program shall include:

1)

Identification of a sampling schedule for the critical variables and control points for these variables, 2)

Identification of the procedures used to measure the values of the critical variables, 3)

Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, McGUIRE - UNITS 1 and 2 6-15

1 Attachment I.a-ADMINISTRATIVE CONTROLS

-SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)

.i The Radioactrue Effluent Release Reports shall include the following information for eaca-type of solid waste shipped offsite during the report period:

-q.

j a.

Total con ~tainer volume, in cubic meters, b.

Total Curie quantity (determined by measurement or estimate),

c.

Principal' radionuelides (determined by measurement or estimate),

d.

Type of' Waste (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e.

Number of shipments, and f.

Solidification agent or absorbent (e.g., cement, or other approved 2 agents (media)).

The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from (ne site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (0DCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Manage '

ment, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy.

to the NRC Regional Office, no later thaa the '15th of each month following the calendar month covered by the report.

4 McGUIRE - UNITS 1 and 2 6-20 Amendment No.

(Unit 1)

Amendment Noi (Unit 2)

.x Attachmsnt I.a ADMINISTRATIVE' CONTROLS

-RECORD RETENTION (Continue'd) g.

-Records of; training and qualification'for current members of the unit staff; h.

Records.of inservice inspections performed pursuant to these Technical-Specifications; i..

Records of reviews performed for changes made to procedures or l

equipment or reviews of tests and experiments pursuant to 10 CFR 50.59; j.

. Records of meetings of the NSRB and reports required by l

Specification 6.5.1.12; k.

Records of the service lives of all snubbers listed in Tables 3.7-4a l

and 3.7-4b including the date at which the service life commences and associated installation and maintenance records; 1.

Records of secondary water sampling and water quality; and l

. Records of analyses required by the Radiological Environmental l

m.

' Monitoring Program that would permit evaluation of the accuracy of

.the analysis at a later date.

This should include procedures effective at specified times and QA records showing that these procedures were followed.~

6.10.3. Records of quality assurance activities required by the QA Manual shall be retained for a period of time required by ANSI N45.2.9-1974.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and

. adhered to for all operations involving personnel radiation exposure.

6;12 HIGil RADIATION AREA 6.12.1. In lieu of the " control device" or " alarm signal" required by 1paragraph.20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity lof. radiation is equal to or less than 1000 mres/hr at 45 CM (18 in.) from the radiation source or from any surface which the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation' Work Permit (RWP).

Individuals qualified in radiation protection procedures'(e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 arem/hr provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.

McGUIRE - UNITS 1 and 2 6-23

e Attachmznt I.a 3

ADMINISTRATIVE CONTROLS

\\

HIGH RADIATION AREA (Continued) 1 Any individual or group'of individuals permitted to enter such areas shall be i

i rprov ded w th or accompanied by one or,more of the following:

sr a.

A radiation monitoring device which continuously indicates the

{

radiation dose rate in the area; or v

b.

A radiation monitoring device which continuously integrates theb radiation dose rate in the area and alarms when a preset integrated dose is received.

Entry into such areas with this moriitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of'them; or j

c.

An individual qualified in radiation protection proc dures with a radiation dose. rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the. frequency specified by the Station Health Physicist in the RWP.

6.12.2 In ddition to the requirements' of Specification 6.12.1, areas accessible to. personnel with radiation levels greater than-1000 mrem /hr at 45 CM (18 in.) from'th'e radiation source or from ariVsurface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry,

.and the keys shall be maintained under the administrative control of th,e Shift Foreman on duty and/or health physics supervision.

Doors shall remain

~

locked except during' periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work ~ area and the -

maximum allowable stay time for individuals in that area.

In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance'may be made by personnel qual _ified in radiation protection procedures _to provide positive exposure control over.the activities being performed within the area. '

O For individual areas accessible.to. personnel with radiation levels greater.

.than 1000 mrem /hr that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual arsas, that area shall be barric'aded, conspicuously pos^ted, and a flashing light shall be activated as

-a warning device.

s 6.13 PROCESS CONTROL PROGRAM (PCP) s

-6.13.1 The PCP'shall be appr Oed by the Commission prior to implementation.

• Measurement made at 18 inches from source of radioactivity.

l McGUIRE - UNITS 1 and 2 6-24 Amendment No.

(Unit 1)

Amendment No.,(Unit 2)

'i 7

h

Attechment I.a ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (PCP) (Continued) 6.13.2: Licensee-initiated changes to the PCP:

Shall be submitted to the' Commission in the Semiannual Radioactive a.

~

Effluent Release Report for the period in which the change (s) was made.

This submittal shall cor.tain:

1)

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental informatinn; 2)

A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and 3)

Documentation of the fact that the change has been reviewed and found acceptable by the Station Manager.

b.

Shall become effective upon review and acceptance by a qualified individual / organization.

6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall he approved by the Commission prior to-implementation.

6.14.2 Licensee-initiated changes to the ODCM:

Shall be f.ubmitted to the Commission in the Semiannual Radioactive a.

Effluent f.elease Report for the period in which the change (s) was made effective.

This submittal shall contain:

1)

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental informction.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered, dated and containing the revision number

-together with appropriate analyses or evaluations justifying the change (s);

2)

A determination that the charg will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and 3)

Documentation of the fact that the change has been reviewed and found acceptable by the Station Manager.

b.

Shall become effective upon review and acceptance by a qualified individual / organization.

McGUIRE --UNITS 1 and 2 6-25 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

L

n ww :

i n

v

?'

I',.a (Delete Page)

-f l

Y

$. ~

McGUIRE - UNITS 1 and 2 6-27

'i

a Attachment I.a (Delete Page)

McGUIRE - CNITS 1 and 2 6-28

- = -

Attaa' :-f. 6.b

' Justification and Safety Analysis The proposed changes to the McGuire Technical Specifications address a range of administrative matters. This proposal seeks to streamline, update, correct, and clarify the administrative controls at McGuire Nuclear Station, as specified in the Technical Specifications.

This proposal also includes the correction of a typographical error in Table 4.11-3.

'In Table 4.11-1 the word " inlet" has been used inadvertantly instead of " outlet".

The proposed amendment corrects this error.

The proposed change to Specification 6.1.2 is to clarify the intent of the Speci-fication as to include only Nuclear Production Department Station personnel, not vendor, contractor, or other personnel.

This is the result of the resolution of a concern identified at Catawba Nuclear Station (Reference Inspection Report 50-413/84-29) and corresponds to the current Catawba Nuclear Station Technical Specifications, which reflect the resolution of this situation.

The proposed amendment of the Technical Specifications seeks to correct certain job titles; Manager of Nuclear Production, or Vice President, Steam Production, and Executive Vice President, Power Operations, are being changed to Vice President, Nuclear Production, and Executive Vice President, Engineering, Construction and Production, respectively. The proposed amendment also seeks to delete figures 6.2-1 and 6.2-2, Offsite organization and Station Organization, respectively. These two figures are presently in error in some espects, and are contained in Chapter 13 of the McGuire Nuclear Station Final Safety Analysis Report, and are maintained there in accordance with 10CFR50.71. Page XXI of the Index is modified to indicate these deletions.

Th-inclusion of the Superintendent of Integrated Scheduling in Specifications 6.5.1.2, 6.5.1.5, 6.6.lb, 6.8.2, and 6.8.3c is an administrative matter and involves no safety questions. The proposed changes would allow the Superintendent of Integrated Scheduling to review and/or approve modifications of safety-related

. structures, systems or components (6.5.1.3), proposed tests and experiments which affect nuclear safety and are not addressed in the FSAR or Technical Specifications (6.5.1.5), REPORTABLE. EVENTS (6.6.lb), and procedures specified under Specifi-cation 6.8.1 and changes thereto (6.8.2 and 6.8.3c), if so designated by the Station Manager.

'In each of tne above cases, the Operating Superintendent, the Technical Services Superintendent,and the Maintenance Superintendent 3each have the same authority as described above.

Since the Superintendent of Integrated Scheduling is required to meet the same qualifications as each of these Superintendents, no loss of Technical Review Capability can occur, therefore there can be no impact on safety.

The purpose of the proposed chaage to the Technical Specification 6.5.1.8 and 6.8.1 is to allow the Station Services Superintendent to review and approve modifications relating to the Station Security Program and associated procedures. Currently the responsibility is discharged by the Station Manager. The proposed change is purely administrative in nature and would facilitate efficient resolution of security related matters.

The modifications approved by the Station Services Superintendent will be transmitted to the Vice President, Nuclear Production and the Director of the Nuclear Safety Review Board. Also, two typographical errors (one each in 6.5.1.8 and 6.5.1.9) are corrected.

The Semiannual Radioactive Effluent Release requirements listed in the Technical Specification 6.9.1.7 (pg. 6-20) require minor changes of words to make these requirements consistent with 10CFR Part 61. A footnote on this page is outdated and is alto being deleted by the proposed amendment. This error and the error in Table 4.11-1 were detected by an in-house QA audit of the Technical Specificati;ns governing the radwaste systems.

The purpose of the proposed changes to Specification 6.10.2 is to eliminate an inconsistency between the Technical Specification 6.10.2 and the Duke Power Company record retaining procedures based upon ANSI N45.2.9-1974.

The proposed

. change would-require that the records of the quality assurance activities as described-in the QA Manual may be retained in accordance with the ANS1 N45.2.9-1974

. and not for the duration of the unit Operating License as currently required by

- the Teshnical Specifications.

McGuire Nuclear Station is committed to the directives in the Regulatory Guide 1.88.

The regulatory guide affirms the use of ANSI N45.29-1974 for administrative cc"trols governing the QA records.

The proposed change to Specification 6.12 provides consistency between McGuire and Catawba High Radiation Area Technical Specifications, and reflects the wording-of the current draft of Revision 5 to the Standard Technical Specifications. The change will not significantly affect current radiation protection practices at McGuire..but is intended to clarify the requirements relative to definition of and' access to high radiation areas.'The change is largely semantic and does not affect plant safety.

The issuance of Amendments 32 (Unit 1) and 13 (Unit 2) resulted in the renumbering of several pages in Section 6.

This resulted in duplication when pages 6-27 and 6-28 were not deleted.. Since page 6-27 duplicates.6-25, and 6-28 duplicates 6-26, we propose to delete pages 6-27 and 6-28 from the Technical Specifications.

These proposed changes to the McGuire Technical Specifications are intended to promote a more efficient resolution of administrative matters at McGuire Nuclear Station. Wording changes are proposed to clarify requirements and procedure review is expanded to improve efficiency with no loss of review capability. These changes are administrative in nature and will not impact plant safety.

L_

. s --

Attachment I.c A'nalysis of Significant Hazards Consideration As requ' ired by 10CFR50.91 this analysis provides a determination that the proposed change to the Technica1' Specifications does not involve any significant hazards consideration, as defined by 10CFR50.92.

The proposed changes to the Technical' Specifications are purely administrative J

in nature and do not affect plant safety. The changes simply update, clarify, and correct the Technical Specifications. The proposed' changes would not affect plant equipment or plant operational procedures and have no safety implications.

-The proposed amendment would not:

1) Involve a significant increase in the probability or consequences.of an accident previously evaluated; or

.2)

Create the possibility of a new of different kind of accident from any accident previously evaluated; or 3)

Involve a significant reduction in a margin of safety.

. Based upon the preceding analysis. Duke Power Company concludes that the proposed amendments do not involve a significant hazards consideration.

. N',

i l-ATTACIDIENT 11

Attachm:nt II.a INSTRUMENTATION RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels

.shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to

-ensure that the limits of Specification 3.11.1.1 are not exceeded.

The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

' APPLICABILITY:

At all times.

ACTION:

a.

With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above.epecification, immediately suspend the release of radio-active liquid effluents monitored by the affected channel, or

' declare the channel inoperable.

b.

With less than the minimum number of radioactive liquid effluent monitoring. instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12.

Restore-the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, in lieu of a Licenser. Event Report, explain in the next Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.8 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL-CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-8.

McGUIRE - UNITS 1 and 2 3/4 3-66

m Attechm:nt 11.a

-INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

' LIMITING CONDITION-FOR OPERATION 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY:

As shown in Table 3.3-13.

ACTION:

a.

With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.

b.

With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.

Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or, in lieu of a Licensee Event Report, explain in tne next Semiannual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-9.

McGUIRE - UNITS 1 and 2 3/4 3-71

Attechment II.s SURVEILLANCE REQUIREMENTS (Continued) 9)

Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.

This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during_

subsequent inservice inspections.

b.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports a;

Within 15 days following the completion of each inservice inspection of steam. generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.

The complete results of the steam generator tube inservice inspection shall be' submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection.

This Special Report shall include:

1)

Number and extent of tubes inspected, 2)

Location and percent of wall-thickness penetration for each indication of an imperfection, and 3)

Identification of tubes plugged.

McGUIRE - UNITS 1 and 2 3/4 4-15

q Attachment ILo TABLE 4.4-2.

STEAM GENERATOR TUBE INSPECTION 5

N 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION

-3RD SAMPLE INSPECTION h

Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per

,g S. G.

C-2 Plug defective tubes.

C-1 None N/A N/A and inspect additional Plug defective tubes C-1 None y

25 tubes in this S. G.

C-2 and inspect additional C-2 Plug defective tubes 45 tubes in this S. G.

Perform action for C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N/A N/A sample my C-3 Inspect all tubes in All other this S.G., plug de-S.G.s are None N/A N/A i

fective tubes and C-1 O

inspect 25 tubes in Some S.G.s Perform action for each other S.G.

C-2 but no C-2 result of N/A N/A additional second sample S.G. are C-3 Additional Inspect all tubes S.G. is in each S.G. and N/A N/A C-3 plug defective tubes.

S = 3 h Where N is the number of steam. generators in the unit, and n is the number of' steam generators inspected during an inspection 1

I i

Atts.chment II.a ACTION:

(Continued)

MODES'1, 2, 3, 4, and 5:

With the ' specific activity.of -the reactor coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/_E microcuries per gram of gross specific activity, perform the sampling and-analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.

In lieu of a. Licensee Event Report, for this ACTION statement within 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 with a copy to the Director, Nuclear Reactor Regulation, Attention:

Chief, Core Performance Branch, and Chief, Accident Evaluation Branch, U.S. Nuclear Regulatory Commission, Washington, D.C., 20555.

This report shall contain the

.results of the specific activity analyses toget.her with the following information:

1.

Reactor power history _ starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded;.

~

2.

Results of the last isotopic analysis for radioiodines performed prior to exceeding the limit, while limit was exceeded, and one analysis after the radioiodine activity was reduced to less than the limit, including for each isotopic analysis, the date and time of sampling and the radiofodine concentrations; 3.

Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; 4.

History of degassing operations, if any, starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; and 5.

The time duration when the specific activity of the reactor coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENT I-131.

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

McGUIRE - UNITS 1 and 2 3/4 4-26

r Attechm:nt II.a CONTAINMENT SYSTEMS

' CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained.

at a levol consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With the structural integrity of the containment vessel:not: conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200 F.

SURVEILLANCE REQUIREMENTS 4.6.1.6 The structural integrity of the containment vessel shall be determined during the shutdown for each Type A containment leakage rate test (reference

, Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the vessel.

This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance of the surfaces or other abnormal degradation.

Any abnormal degradation of the containment vessel detected during the above required inspections shall be reporte.d to the Commission pursuant to 10CFR Sections 50.72 and 50.73.

McGUIRE - UNITS 1 and 2 3/4 6-14

'Attcchment II.e f

CONTAINMENT SYSTEMS i

REACTOR BUILDING STRUCTURAL INTEGRITY t~

LIMITING CONDITION FOR OPERATION 3.6.1.7 The structural integrity of the reactor building shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.7.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

I Withthestructuralintegrityofthereactorbuildingnotconformingtfthe; above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200 F._,

o a

o SURVEILLANCE REQUIREMENTS e

^

4. 6.1. 7

.The structural integrity of the reactor _ building shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the exposed accessible interior and exterior surfaces of the reactor building and verifying no apparent changes in appearance of the concrete surfaces or other abnormal degradation.

Any abnormal degradation of the reactor building detected during the above' required inspections shall be reported to the Commission pursuant to 10CFR Sections 50.72 and 50.73.

f McGUIRE - UNITS 1 and 2 3/4 6-15 L.

Attachment II.a

' ELECTRICAL POWER SYSTEMS

,; 5

.;(

SURVEILLANCE REQUIREMENTS (Continued) 1)

Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution, and 2)

Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure.

4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2 within 30

- days.

Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

c' 4.8.1.:.4 Diesel Generator Batteries - Each diesel generator 125-volt battery bank aid charger shall be demonstrated OPERABLE:

c a.

At least once per 7 days by verifying that:

1)

The electrolyte level of each battery is above the plates, and 2)

The overall battery voltage is greater than or equal to 125 volts under a float charge.

b.

At least once per 18 months by verifying that:

1)

The batteries, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration; 2)

The battery-to-battery and terminal connections are clear, tight, free of corrosion and coated with anti-corrosion material; and 3)

The battery capacity is adequate to supply and maintain in OPERABLE status its emergency loads when subjected to a battery service test.

f McGUIRE - UNITS 1 and 2 3/4 8-7 en L

f-Atttchm:nt II.a

)

3/4.11 RADIOACTIVE EFFLUENTS t

3/4.11.1 LIQUID EFFLUENTS o

CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1-4) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B. Table II, Column 2 for

- radionuclides other than dissolved or entrained noble gases.

Fordigsolvedor entrained noble gases, the concentration shall be limited to 2 x 10 microcurie /ml total activity.

4 APPLICABILITY:

At all times.

).

ACTION:

L With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immedi-ately restore-the concentration to within the above limits.

I SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1.

L 4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.

h-L i

i l

McGUIRE - UNITS 1 and 2 3/4 11-1 l

L

C-Attcchment II.a 4ECid!U.t P0. M IDA'TIVE EFFLUENTS C

3/4..ll.2.GASE005 EFFLUENTS

g. W ! L t M.. -

1)

Orai LIMITING CONDITION FOR OPERATION 3.r'1',"2."lThe dose rate due to radioactive materials released in gaseous

.I effluents from_the site to areas at and beyond the SITE BOUNDARY (see Figure,5.1-3)shallbelimitedtothefollowing:

a.

For noble gases:

Less than or equal to 500 mrem /yr to the whole

, dJe[5h fN. ' body and less than or equal to 3000 mrem /yr to the skin, and ys ' M cets v gr "For Iodine-131 and 133, for tritium, and for all radioactive s m r.n d ~r a,c 9

g g_3; g

(,",,,,,

Less than or equal tc,1500 mrem /yr to any organ.

["'AP'Pi.IC'ABI' ITY:'

At all times.

L y

. ACTION:

8. t.1. 4 kn!' and cvor-With the dose rate (',) exceeding the above limits, immediately restore the r: lease rate to within the above limit (s).

3 qg

.y SURVEILLANCE REQUIREMENTS

r..

at %.

4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be li determined to be within the above limits in accordance with the methodology and parameters in the ODCM.

2 4.11i2.1;2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in,e'cordance with the methodology and parameters of the ODCM by obtaining c

representative samples and performing analyses in accordance with the sampling 3T and analysis program specified in Table 4.11-2.

Ct GM :.

"'" ~ # kcGb!$'E"- UNITS 1 and 2 3/4 11-9 L

AttEchment II.a RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.l' The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:

a.

For noble gases:

Less than or equal to 500 mrem /yr to the whole body and less than or equal to 3000 mrem /yr to the skin, and b.

For Iodine-131 and 133, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days:

Less than or equal to 1500 mrem /yr to any organ.

APPLICABILITY:

At all times.

ACTION:

With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.

4.11.2.1.2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.

McGUIRE - UNITS 1 and 2 3/4 11-9

7 Attachm2nt II.a REACTOR COOLANT SYSTEM BASES-STEAM GENERATORS (Continued)

Whenever the results of any steam generat u tubing inservice inspection fall into Category C-3, these results will be promptly reported to the

~ Commission pursuant to 10CFR Sections 50.72 and 50.73 prior to resumption of l

plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the

. Technical Specifications, if necessary.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm.

This threshold value is sufficiently low to ensure early detection of additional leakage.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount ofsleakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.

This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in the accident analyses.

The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break.

The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents.

The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

McGUIRE - UNITS I and 2 B 3/4 4-4 Y

u.

F-1.b Justification and Safet

,alysis

.The proposed' changes to the Technical Specifications involve the reportability requirements of.Section 6.9 of the McGuire Technical Specifications.

cAs. presented in Generic Letter No. 83-43 (dated December 19, 1983),

the regulations regarding reporting of events-(10CFR sections 50.72 and 50.73) were changed... Subsequently, section 6.9 of McGuire's Technical Specifications was amended to reflect the new reporting requirements. When this section was amended, however, the references in other parts of-the Specifications were not updated. This proposal is to appropriately update these references. References to Specification 6.9.1.11b have been deleted because Specification 6.9.1.11b has been deleted. Technical Specification 6.9.1 has been so changed that references to it are virtually meaningless. These references to Specification 6.9.1 have.been modified on an individual basis to preserve present reporting requirements, but to clarify these requirements and remove redundancies in.the requirements.

.These proposed changes are administrative in nature and will have no impact upon. plant safety.

a

4 N.-

Attachment II.c g y S OF SIGNIFICANT HAZARDS CONSIDERATION As required by 10CFR50.91 this analysis provides a determination that the proposed changes to the Technical Specifications do not involve any significant hazards consideration, as defined by 10CFR50.92.

The proposed changes in the Technical Specifications are purely administrative in nature to ;orrect inconsistencies arising from the modification of Section 6.0 of the Technical Specifications following the change of the reporting requirements et 10 CFR Sections 50.72 and 50.73.

The proposed change would not affect any equipment or plant operational procedures and has no safety implications.

The proposed amendment would not:

1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2) Create the possibility of a new of different kind of accident from any accident previously evaluated; or 3)

Involve a significant reduction in a margin of safety.

Based upon the preceding analysis, Duke Power Company concludes that the proposed amendments do not involve a significant hazards consideration.

- J

JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The proposed change to Table 3.3-10, item 9 - Auxiliary Feedwater Flow Rate would change the Total Number of Channels from 2/ steam generator to 1/ steam generator. This change is needed in order to maintain consistency between the FSAR and the Technical Specifications.

There are 2 flow transmitters in each steam generator however only one transmitter per steam generator is safety grade and relied upon to function during an accident.

SER section 7.3.2.6 reflects the fact that the Staff has found that having a single auxiliary feedwater flow transmitter in each steam generator is acceptable and thus the design meets the requirements outlined in TMI-2 Action Plan, item II.E.L.2.

The Technical Specifications need to be revised to reflect this.

10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The proposed amendment does not involve an increase in the probability or consequences of any previously evaluated accident.

There will be no changes made to the design or construction of the plant therefore, the accident analyses will be unaffected by this change.

The proposed amendment does not create the possibility of a new or different kind of accident than any previously evaluated since the operation of the plant will be unaffected.

The proposed amendment does not involve a significant reduction in a margin of safety since the Staff has reviewed the Catawba design and has found that it meets the criteria in the TMI Action Plan, item II.E.1.2.

The Commission has provided guidance concerning the application of standards of no significant hazards determination by providing certain examples (48 FR 14870).

This change is similar to example (vi).

For the reasons stated above, it is concluded that the proposed amendment does not involve significant hazards considerations.

i

JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The change to Surveillance Requirement 4,4.4.3 would add a surveillance step requiring isolation and venting of the normal air supply to the PORV's.

This change request is an additional restriction on a surveillance and is thus similar to example (ii) as outlined in the guidance provided concerning the application of standards of no significant hazards determination (48 FR 14870).

JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION Per the July 31, 1985 Duke request to revise Technical Specification

' Figure 3.2-3, the DNB LIMIT Line is to be removed. Thia part of the figure was added when rod bow concerns were addressed via burnup dependent penalties and the R2 Factor which allowed a tradeoff of RCS F1w and FAH based on the relationships determined by core thermal-hydraulic analyses. Upon the removal of burnup dependent rod bow penalties and the R2 factor from the Standard Technical Specifications, the DNB LIMIT Line no longer serves a useful purpose and therefore Duke requests its removal in order to prevent misinterpretation of the figure.

The rod bow penalty as a function of burnup is calculated with the methods described in WCAP-8691, Revision 1 " Fuel Rod Bow Evaluation" (July, 1979).

The maximum rod bow penalty is 2.7% DNBR. The safety analysis and deter-mination of core thermal limits are performed with DNBR limits of 1.49 and 1.47 compared to the design limits of 1.34 and 1.32 for typical and thimble cells, respectively.. The 410% thermal margin between safety analysis and design limit DNBR values is partially used to offset the 2.7% DNBR rod bow penalty.

The justification for the 2% Power /1% Flow tradeoff for RCS flows less than 396,100 gpm is based upon the following relationships:

3 Flow N 1%

~

3 DNBR 1.6%

B Power N 1%

3 DNBR 1.8%

3 Power N 1.6% N 0.88 a Flow 1.8%

Although the above sensitivities apparently justify a tradeoff of nearly 1% Power /1% Flow, the ratio of 2/1 is suggested in order to ensure con-servatism with respect to DNB protection.

The a Flow /3 Power relationship is based upon sensitivity studies performed by Westinghouse for designs and parameter ranges associated with the McGuire and Catawba Nuclear Stations. The conditions resulting in the maximum DNBR sensitivity to flow were 120% power, 2400 psia, and a temperature necessary to yield the minimum DNBR limit. The typical cell sensitivity was used since the typical cell is more sensitive to flow than is the thimble cell.

Statepoints representative of possible operating conditions (normal and accident) were analyzed using TilINC-IV. Each statepoint was analyzed at a nominal flow value for core limit conditions, and at 97.5% of nominal flow.

' The largest penalty determined for the 2.5% flow reduction was 4.0% DNBR which results in the sensitivity of 1% flow /1.6% DNBR. Design flow rates for McGuire and Catawba are nearly the same (core thermal limit lines are based on 388.880 gpm at McGuire and 387,600 gpm at Catawba) and therefore the sensitivities are valid for both stations.

L

E e value of 1.8 used for the sensitivity of power to DNBR was derived from wark performed in the area of rod bow penalty in 1976-1977.

Information dealing with the penalty and generic margins is discussed in " Revised Interim Evaluation Report on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors", NRC report (February, 1977). The thermal-hydraulic analyses performed to derive this value show that the conditions used are similar to those used to calculate the flow sensitivities.

Previous DNBR sensitivity studies have shown that the DNBR sensitivity to power is very similar to the sensitivity to FN Therefore, a value of AH.

1.8 is appropriate for power when considering the tradeoff between power and flow.

In the SER for Amendment No. 9 to the McGuire Unit 1 license (NPF-9), the NRC Staff referenced the Battelle Northwest Laboratory report FATE-79-101,

" Analysis of the Sensitivity of Calculated NDNBR to Eight Selected DNB Parameters" in which the power to flow sensitivity ratio is determined to be 40.86.

This value agrees with the Westinghouse sensitivity of 40.88 for McGuire and Catawba and confirms the conservatism of the proposed 2%

power /1% flow tradeoff requested for the Catawba Technical Specifications.

10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:

1.

Involve a significant increast in the probability or consequences of an accident previously evaluated; or 2.

Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3.

Involve a significant reduction in a margin of safety.

The proposed amendment does not increase the probability or consequences of an accident previously evaluated, it does not create the possibility of a new or different kind of accident and it does not involve a significant reduction in a margin of safety.

W e impact of operation at reduced power and reduced Thermal Design Flow upon the LOCA analysis was evaluated and described in the November 18, 1983 submittal for McGuire Nuclear Station.

Since RCS parameters eniculated for 95% flow will exhibit little change in Tcold, the RCS blowdown transient would not be significantly impacted. At the reduced core power level, the lower flow will not cause a greater initial fluid enthalpy rise in either the hot assembly or the core as a whole, so no penalty in calculated peaking factor will occur. The tradeoff of 2% power /1% flow is therefore sufficient to ensure the existing ECCS analysis for Catawba remains conservative.

The current best estimate flow at Catawba Unit 1 is 399,112 gpm.

The margin above the Technical Specification limit is therefore only 0.76%. Consider-ing an expected variance in flow measurement of N2.2%, Duke believes the proposed Technical Specification amendment is required based upon the

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unnecessary severity of the existing-Technical Specification Action

- Statement if any future flow measurement indicates a flow slightly less :than 396,100 gpm..

-Based upon'the above analysis, the proposed amendment is determined

.to.involvenosignificanthazat$1 considerations.

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PROOF & REVIEW kpy TA8LE 3.2-1 DN8 PARAMETERS

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PARAMETER LIMITS 1

Four Loops in Ooeration,

'Averace Temperature Meter average

- 4 channels: < 592*F

- 3 channels: 5592'F Computer average

- 4 channels: 1 593*F

- 3 channels: 1 593'F Pressurizer Pressure 7

Meter-average

- 4 channels: 122254 psig*

- 3 channels: 2 2230 psig*

2 Computer average

- 4 channels: 1 222f psig*

- 3 channels: 1 2224 psiga l

• Limit not appifcable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

OCT 7 1985 CATAWBA'- UNITS 1 AND 2 3/4 2-16 i

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PROOF & REVIEW COPY POWER DISTRIBUTION LIMITS BASES QUADRANT POWER TILT RATIO (Continued) l flux map or two sets of 'four symmetric thimbles.

The two sets of four symmetric thimbles is a unique set of eight detector locations. The normal locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

Alternate locations are available if any of the normal locations are unavailable.

3/4.2.5 DN8 PARAMETE.15

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The limits on the DN8-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in

.the. transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a design limit DNBR throughout each analyzed transient. The indicated T,yg value messzmusF and the indicated pressurizer pressure value6 corre-spond to analytical limits of 594.8'F and psig respectively, with allow-ance for measurement uncertainty.

zzos,3 The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. Measure-ment uncertainties must be accounted for during the periodic surveillance.

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CATAW8A - UNITS 1 AND 2 8 3/4 2-6 N

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JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The proposed changes to Technical Specification Table 3.2-1 "DNB Parameters" are based on a new analysis taking into account the source of the indications (meter or computer) and the number of available operable channels.

One additional change to the Proof and Review Table 3.2-1 is offered to maintain consistency and changes to the bases are also proposed to remove some discrepancies (attached).

10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operetion in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The proposed amendment does not involve an increase in the probability or consequences of any previously evaluated accident. There will be no changes made to the design or construction of the plant therefore, the accident analyses will be unaffected by this change.

The proposed amendment does not create the possiblity of a new or different kind of accident:than any previously evaluated since the operation of the plant will be unaffected.

The proposed amendment does not involve a significant reduction in a margin of safety, the proposed changes actually result in slightly more conservative numbers being used.

The Commission has provided guidance concerning the application of standards of no significant hazards determination by providing certain examples (48 FR 14870). This change is similar to example (ii).

For the reasons stated above, it is concluded that the proposed amendment does not involve significant hazards considerations.

E