ML20216C592

Proposed Tech Specs Changing Surveillance Intervals to Accommodate 24-month Fuel Cycle Per GL 91-04

ML20216C592
Person / Time
Site:	Seabrook
Issue date:	04/08/1998
From:	NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
To:
Shared Package
ML20216C563	List: ML20216C558 ML20216C575 ML20216C592
References
GL-91-04, GL-91-4, NUDOCS 9804140495
Download: ML20216C592 (19)

Text

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Section II Markup of Proposed Changes The attached markup reflects the currently issued revision of the Technical Specifications listed below.

Pending Technical Specifications or Technical Specification changes issued subsequent to this submittal are not reflected in the enclosed markup.

The following Technical Specifications are included in the attached markup:

Technical Specification Title Page(s) 4.4.5.3 Steam Generators - Inspection Frequencies 3/4 4-15 3.4.6.2c Reactor Coolant System Leakage 3/4 4-21 3/4.4.5 Steam Generators Bases B 3/4 4-2a 3/4.4.6.2 Operational Leakage Bases B 3/4 4-4 i

i age 7 I 9804140495 980408 l PDR ADOCK 05000443 p PDR

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l REACTOR COOLANT SYSTEM f -

l STEAM GENERATORS l

l SURVEILLANCE REQUIREMENTS 1

l 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed At the following frequencies:

a. The first inservice inspection shall be performed after 6 Effective Full-Power Months but no later than restart after first refueling.

Subsequent inservice inspections shall be performed at intervals of d not less than 12 nor more than 24 calendar months after the previous inspection. &If two consecutive inspections, not including the pre-service inspection, result in all inspection results falling in Cate-gory C-1 or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at l 1esst once per 20 months. The increase in inspection frequency I

shall apply until the subsequent inspections satisfy the criteria of I Specification 4.4.5.3a.; the interval may then be extended to a i

maximum of once er 40 monthsY %$1 As am4 con 6' j 30 OR

c. Additional, unschedu ed inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of l the following conditions:

1) Primary-to-secondary tubes leak (not including leaks originating i from tube-to-tubesheet welds) in excess of the limits of Specification 3.4.6.2, or t

2) A seismic occurrence greater than the Operating Basis Earthquake, or l j

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3) A loss-of-coolant accident requiring actuation of the Engineered j Safety Features, or '

4) A main steam line or feedwater line break; (Lb l

<-/ . IMe itT i SEABROOK - UNIT 1 3/4 4-15 L

REACTORCOOLANT'SYSTEh REACTOR COOLANT SYSTEM LEAKAGE 1 i

f' OPERATIONAL LEAKAGE J l

LIMITING CONDITION FOR OPERATION l

1 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE.

l I c. 1 gpm total reactor-to-secondary leakage through all steam y @y generators and 500 gallons per day through any one steam generator 3 .

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 psig 20 psig, and i f. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 l

gpm at a Reactor Coolant System pressure of 2235 t 20 psig from any Reactor Coolant System Pressure Isolation Valve.* l l

APPLICABILITY: MODES 1, 2, 3, and 4.

l ACTION: )

a, With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 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 />.

l b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 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 '

hours.

l l c. With any Reactor Coolant System Pressure Isolation Valve leakage i greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, 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 l SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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• Test pressures less than 2235 psig but greater than 150 psig are allowed.

Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure dif-ferential to the one-half power. '

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SEABROOK - UNIT 1 3/4 4-21 Amendment No.

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, Inserts for Proposed Wording Changes to Technical Specification Requirements

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T/S 3.4.6.2.e,4.4.5.3, Hases 3/4.4.5 & 3/4.4.6.2 l

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If 20 percent of the tubes were inspected and the results were in be C-1 Category or if 40 percent of the l tubes were inspected and were in the C-2 Category during the previeus inspection, the next inspection l may be extended up to a maximum of 30 months in order to correspond svith the next refueling outage if l the results of the two previous inspections were not in the C-3 Category. Ilowever, if the results of either of the previous two inspections were in C-2 Category, an engineering assessment shall be performed before operation beyond 24 months and shall provide assurance that all tubes will retain adequate structural margins against burst throughout normal operating, transient, and accident conditions until the end of the fuel cycle or 30 months, which ever occurs Orst.

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d. The provisions of specification 4.0.2 do not apply for extending the fnquency for perforir ing inservice inspections as speci6ed in Specifications 4.4.5.3a. and b.

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1 For plant operation beyond 24 months from the previous steam generator tube inspection when the  !

results of either of the two previous inspections are in the C-2 Category as defined by Specification 4.4.5.2, the leakage through any one steam generator not isolated from the Reactor Coolant System shall not exceed 100 gallons per day, I

REACTOR COOLANT SYSTEM q BASES 3/4.4.4 RELIEF VALVES (Continued)

(2) No Surveillance Requirement (ACOT or TAD 0T) exists for verifying automatic operation.

(3) The required ACTION for an inoperable PORV(s) (closing the block valve) conflicts with any presumed requirement for automatic f i

actuation.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes l ensure that the structural integrity of this portion of the RCS will be main-l l- tained. The program for inservice inspection of steam generator tubes is based

!' on a modification of Regulatory Guide 1.83. Revision 1. Inservice inspection

! of steam generator tubing is essential in order to maintain surveillance of the

! conditions of the tubes in the event that there is evidence of mechanical l damage or progressive degradation due to design, manufacturing errors, or in-l service conditions that lead to corrosion. Inservice inspection of steam

! generator tubing also provides a means of characterizing the nature and cause of any tube degradation, so that corrective measures can be taken.

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.w SEABROOK - UNIT 1 B 3/4 4-2a Amendment No.

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Inserts for Proposed Wording Changes to Technical Specification Requirements

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T/S 3.4.6.2.c,4.4.5.3, Hases 3/4.4.5 & 3/4.4.6.2 (continued) l INSERT An engineering assessment of steam generator tube integrity will confirm that no undue risk is associated ;

with plant operation beyond 24 months of the previous steam generator tube inspection. To provide this confirmation, the assessment would demonstrate that all tubes will retain adequate structural margins I against burst during all normal operating, transient, and accident conditions until the end of the fuel l cycle. This evaluation would include the following elements:

1. An assessment of the Haws found during the previous inspection of each steam l

generator. '

2. An assessment of the inaximum Daw size that can be expected before the end of the current fuel cycle or 30 months, wbichever comes Grst, and the corresponding structural margins relative to the criteria of Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes."

3. An update of the assessment model, as appropriate, based on comparison of the predicted results of the steam generator tube integrity assessment with actual inspection results from previous inspections.

INSERT

@ l For plant operation beyond 24 months from the previous steam generator tube inspection when the results of either of the two previous inspections are in the C-2 Category as denned by Speci0 cation 4.4.5.2, the more restrictive leakage through any one steam generator not isolated from the Reactor l Coolant System of 100 gallons per day is intended to provide additional margin to accommodate a tube Haw which might grow at a greater than expected rate. The more restrictive limit provides additional assurance that should a signiGeant leak he experienced in service the plant will be shut down in a timely j manner.

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REACTOR COOLANT SYSTEM n BASES REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

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 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 I in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the eventofamainsteamlineruptureorunderLOCAconditions.j The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited Q amount of leakage 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 safety analyses.

The specified allowed leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure. It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required.

Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure.

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 valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

I SEABROOK - UNIT 1 8 3/4 4-4 QCd t

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l SECTION III l

Retype of Proposed Changes 1

i l The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification changes or Technical Specification changes issued subsequent to this submittal ,

are not reflected in the enclosed retype. The enclosed retype should be checked for continuity with  !

, Technical Specifications prior to issuance.

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' I REACTOR COOLANT SYSIEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS '

4.4.5.3 .Insgection Frecuencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

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a. The first inservice inspection shall be performed after 6 Effective 1 Full-Power Months but no later than restart after first refueling.

Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If 20 percent of the tubes were inspected and the results were in the C-1 Category or if 40 percent of the tubes were inspected and were in the C-2 Category during the previous inspection, the next I inspection may be extended up to a maximum of 30 months in order to correspond with the next refueling outage if the results of the two previous inspections were not in the C-3 Category. However, if the results of either of the previous two inspections were in C-2 Category, an engineering assessment shall be performed before operation beyond 24 months and shall provide assurance that all tubes will retain adequate structural margins against burst throughout normal operating, transient, and accident conditions until the end of the fuel cycle or 30 months, whichever occurs fi rst . If two consecutive inspections, not including tne preservice inspection, result in all inspection results falling in Category C-1 or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a mavier of once per 40 months:

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a. : the interval may then be extended to a maximum of once per 30 or 40 months, as applicable: l

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1) Primary-to-secondary tubes leak (not including leaks  !

originating from tube-to-tubesheet welds) in excess of the l limits of Specification 3.4.6.2. or l 2) A seismic occurrence greater than the Operating Basis

! Earthquake, or

3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or

4) A main steam line or feedwater line break; and l SEABROOK - UNIT 1 3/4 4-15 Amendment No.

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REACTOR COOLANT SYSTEM STEAM GENERATORS I l

l l SURVEILLANCE REQUIREMENTS l

, 4.4.5.3 Insoection Frecuencies (continued)

! d. The provisions of specification 4.0.2 do not apply for extending the frequency for performing inservice inspections as specified in l Specifications 4.4.5.3a. and b. I l

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SEABROOK - UNIT 1 3/4 4-15A Ameadment No.

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REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE l DPERATIONAL LEAKAGE l

LIMITING CONDITION FOR OPERATION t

3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE.

l b. 1 gpm UNIDENTIFIED LEAKAGE,

! c. 1 gpm total reactor-to-secondary leakage through all steam generators l

and 500 gallons per day through any one steam generator. For plant operation beyond 24 months from the previous steam generator tube inspection when the results of either of the two previous inspections are in the C-2 Category as defined by Specification 4.4.5.2, the leakage through any one steam generator not isolated from the Reactor Coolant System shall not exceed 100 gallons per day. ,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of ,

l 2235 psig i 20 psig. and I 1

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f. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve.*

APPLICABILITY: MODES 1, 2. 3. and 4.  !

l ACTION: I l l

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within i 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 />. i

b. With any Reactor Coolant System leakage greater than any one of the I above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 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 />. .

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c. With any Reactor Coolant System Pressure Isolation Valve leakage i l greater than the above limit, isolate the high pressure portion of the affected system from the 104 pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, 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 />.

• Test pressures less than 2235 psig but greater than 150 psig are allowed.

Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure dif-ferential to the one-half power.

SEABROOK - UNIT 1 3/4 4-21 Amendment No.

REACTOR COOLANT SYSTEM i

BASES

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3/4.4.4 RELIEF VALVES (Continued)

(2) No Surveillance Requirement (ACOT or TADOT) exists for verifying automatic operation.

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(3) The requirec ACTION for an inoperable PORV(s) (closing the block  !

valve) conflicts with any presumed requirement for automatic ,

actuation.  !

3/4.4.5 STEAM GENERATORS f

The Surveillance Requirements for inspection of the steam generator tubes l ensure that the structural integrity of this portion of the RCS will be main-  !

tained. The program for inservice inspection of steam generator tubes is based '

on a modification of Regulatory Guide 1.83. Revision 1. Inservice inspection of <

steam generator tubing is essential in order to maintain surveillance of the I conditions of the tubes in the event that there is evidence of mechanical damage  !

or progressive degradation due to design, manufacturing errors, or inservice  ;

conditions that lead to corrosion. Inservice inspection of steam generator tubing i

also 3rovides a means of characterizing the nature and cause of any tube i degracation, so that corrective measures can be taken.

An engineering assessment of steam generator tube integrity will confirm

! that no undue risk is associated with plant operation beyond 24 months of the  !

l previous steam generator tube inspection. To provide this confirmation, the

! assessment would demonstrate that all tubes will retain adequate structural margins against burst during all normal operating, transient, and accident conditions until the end of the fuel cycle. This evaluation would include the following elements:

1. An' assessment of the flaws found during the previous inspection of each steam generator.

2. An assessment of the maximum flaw size that can be expected before t

the end of the current fuel cycle or 30 months, whichever comes first, and the corresponding structural margins relative to the l criteria of Regulatory Guide 1.121. " Bases for Plugging Degraded PWR Steam Generator Tubes."

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3. An update of the assessment model, as appropriate, based on comparison of the predicted results of the steam generator tube integrity assessment with actual inspection results from previous l inspections.

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'SEABROOK - UNIT 1 B 3/4 4-2a Amendment No. M l

REACTOR COOLANT SYSTEM BASES REACTOR COOLANT SYSTEM LEAKAGE I 3/4.4.6.2 OPERATIONAL LEAKAGE (Continued)

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 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 ppd leakage limit per steam generator ensures that steam l generator tube integrity is maintained in the event of a main steam line rupture l or under LOCA conditions. For plant operation beyond 24 months from the previous steam generator tube inspection when the results of either of the two previous inspections are in the C-2 Category as defined by Specification 4.4.5.2, the more restrictive leakage through any one steam generator not isolated from the Reactor Coolant System of 100 gallons per day is intended to provide additional margin to accommodate a tube flaw which might grow at a greater than expected rate. The more l restrictive limit provides additional assurance that should a significant leak be experienced in service the plant will be shut down in a timely manner. i The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited l amount of leakage from known sources whose presence will not interfere with the i detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems. I l

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating i 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 safety analyses. l The specified allowed leakage from any RCS pressure isolation valve is l sufficiently low to ensure early detection of possible in-series check valve  !

failure. It is apparent that when pressure isolation is provided by two in- I series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of value integrity is required.

Since these valves are important in preventing overpressurization and rupture of I the ECCS low pressure piping which could result in a LOCA that by) asses containment, these valves should be tested periodically to ensure low proba)ility of gross failure.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consecuent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

l SEABROOK - UNIl 1 B 3/4 4-4 Amendment No.

4 Section IV I

Deterinination of Significant IIazards for Proposed Change

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IV. DETERMINATION OF SIGNIFICANT IIAZARDS FOR PROPOSED CHANGES License Amendment Request (LAR) 98-03 is the second submittal in a planned series of License Amendment Requests which propose changes to the Seabrook Station Technical Specifications to accommodate fuel cycles of up to 24 months. The proposed changes are associated with steam generator tube inspection surveillance requirements that are currently performed at each 18-month or other outage interval. The License Amendment Request has been prepared in accordance with the generic guidance contained in NRC Gerric Letter (GL) 91-04, " Changes In Technical Specification Surveillance Intervals To Accommodate A 24-Month Fuel Cycle."

The Technical Specifications proposed to be amended are:

4.4.5.3 Steam Generators - Inspection Frequencies 3.4.6.2c Reactor Coolant System Leakage 3/4.4.5 Steam Generators 13ases 1 3/4.4.6.2 Operational Leakage Bases I I

I In accordance with 10 CFR 50.92, North Atlantic has reviewed the attached proposed changes and has concluded that they do not involve a significant hazards consideration (SilC). The basis for the conclusion that the proposed changes do not involve a SliC is as follows:

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1. The proposed changes do not involve a significant increase in the probability or  !

consequences of an accident previously evaluated.

I Extending Surveillance Requirement (SR) 4.4.5.3 to accommodate a 24 month cycle for I inspection of steam generator tubes structural integrity, as well as, imposing a more restrictive Limiting Condition for Operation (TS 3.4.6.2.c) for reactor coolart system leakage through Category C-2 steam generators, will neither exacerbate nor significantly increase the probability 1 or consequences of an accident previously evaluated in the Seabrook Station UFSAR. l The proposed changes to SR 4.4.5.3 do not alier the intent or method by which the surveillances I are conducted, do not involve physical changes to the plant, do not alter the way structures, systems or components (SSCs) function, and do not modify the manner in which the plant is operated.

The proposed change to TS 3.4.6.2.c imposes more restrictive limits on plant operations due to RCS leakage through steam generators. The proposed change does not involve physical changes ,

to the plant or alter the way a SSC functions.  !

The proposed changes to SR 4.4.5.3 and TS 3.4.6.2.c, and their associated Bases, will not adversely affect the ability of the steam generators to perform their intended safety function.

Furthermore, the proposed changes do not adversely affect the physical protective boundaries of the plant. The proposed changes do not affect accident initiators or precursors and do not alter the design assumptions, conditions, configuration of the facility or the manner in which the piant is operated. The proposed changes do not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an initiating event within the acceptance limits assumed in the Updated Final Safety Analysis Report (UFSAR). The proposed changes are administrative in nature and do not change the level of programmatic controls or the procedural Page 10

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, details associated with aforementioned surveillance requirements. While the proposed changes will lengthen the interval between surveillances, the increase in interval has been evaluated; and i based on the reviews of the steam generator tube eddy current test (ECT) inspections, it is

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concluded that the wear growth rate of the only active degradation mechanism (Anti-Vibration l Bar (AVB) wear) identified to date at Seabrook Station is such that sufficient margin exists between the plugging criteria and structural limit such that no tubes are predicted to exceed the structural limit even with the longer surveillance interval.

l Since there are no changes to previous accident analyses, the radiological consequences l associated with these analyses remain unchanged, therefore, the proposed changes do not involve l a significant increase in the probability or consequences of an accident previously evaluated.

Therefore, the proposed changes will not significantly increase the probability or consequences of any previously analyzed accident.

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2. The proposed changes do not create the possibility of a new or different kind of accident from any presiously analyzed.

l The proposed changes to TS 3.4.6.2 and SR 4.4.5.3, and associated Bases, do not alter the design l

assumptions, conditions, configuration of the facility or the manner in which the plant is operated. There are no changes to the source term, containment isolation or radiological release assumptions used in evaluating the radiological consequences in the Seabrook Station UFSAR.

i Existing system and component redundancy is not being changed by the proposed changes. The l proposed changes have no impact on component or system interactions. The proposed changes are administrative in nature and do not change the level of programmatic controls and procedural details associated with the aforementioned surveillance requirements. Therefore, since there are no changes to the design assumptions, conditions, configuration of the facility, or the manner in which the plant is operated and surveilled, the proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed. 1 l

3. The proposed changes do not involve a significant reduction in a margin of safety, i The proposed changes to the surveillance intervals for SR 4.4.5.3 is still consistent with the basis 1

for the interval. The intent or method of performing the surveillances remains unchanged. The  !

more restrictive limit for leakage through any one steam generator placed in Category C-2, as well as, the requirement to do an engineering assessment of steam generator tube integrity, provides additional margin of ensuring safe plant operation.  !

In addition, there : no adverse affect on equipment design or operation and there are no changes being made to the Technical Specification required safety limits or safety system settings that i would adversely affect plant safety. The proposed changes are administrative in nature and do not change th level of programmatic controls and procedural details associated with the j aforementioned surveillance requirements. While the proposed changes will lengthen the interval between surveillances, the increase in interval has been evaluated; and based on the reviews of the steam generator tube ECT inspections, it is concluded that the wear growth rate of the only active degradation mechanism (AVB wear) identified to date at Seabrook Station is such that sufficient margin exists between the plugging criteria and structural limit such that no tubes l

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l are predicted to exceed the structural limit even with the longer surveillance interval. Therefore, extension of the current surveillance intervals to' accommodate a 24 month cycle will not

- significantly degrade the ability, the availability or the reliability of the steam generators to perform their intended safety function, thus, it is concluded that there is no significant reduction in a margin of safety. "

Based on the above evaluation, North Atlantic concludes that the proposed changes do not constitute a significant hazard.

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Sections V & VI Proposed Schedule for License Amendinent Issuance and Effectiveness and ,

EnvironmentriInnpact Assessinent i

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V. PROPOSED SCHEDULE FOR LICENSE AMENDMENT ISSUANCE AND l EFFECTIVENESS 1

North Atlantic requests NRC review of License Amendment Request 98-03 and issuance of a license j amendment by October 10,1998, having immediate effectiveness and implementation required within 60 days.

i VI. ENVIRONMENTAL IMPACT ASSESSMENT North Atlantic has reviewed the proposed license amendment against the criteria of 10CFR51.22 for l environmental considerations. The proposed changes do not involve a significant hazards consideration, nor inciease the types and amounts of effluent that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, North Atlantic concludes that the proposed change meets the criteria delineated in 10CFR51.22(cX9) and l 10CFR51.22(cX10) for a categorical exclusion from the requirements for an Environmental Impact i Statement.

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