ML20093E108

Amend 214 to License DPR-77,revising TS SR & Bases to Incorporate Alternate SG Tube Plugging Criteria at Tube Support Plate Intersections

ML20093E108
Person / Time
Site:	Sequoyah
Issue date:	10/11/1995
From:	Hebdon F NRC (Affiliation Not Assigned)
To:
Shared Package
ML20093E112	List: ML20093E108 ML20093E119
References
GL-95-05, GL-95-5, NUDOCS 9510160079
Download: ML20093E108 (12)

Text

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o ye UNITED STATES

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NUCLEAR REGULATORY COMMISSION s*,

WASHINGTON, D.C. 20065-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO 50-327 SE0V0YAH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 214 License No. DPR-77 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated July 19, 1995, which was u.+,

WI by letter dated September 7,1995, and supplemented by letta.., dateo September 15 and 26, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

)

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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9510160079 951011 PDR ADOCK 05000327 P

PDR

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4 :

1 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:

1 (2) Technical Specifications The Technical Specifications contained in Appendices A and 8, as d

revised through Amendment No. 214, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, to be implemented within 45 days.

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FOR THE NUCLEAR REGULATORY COMMISSION r~

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f i

7 Frederick J. Hebdon, Director i

Project Directorate 11-3 1

Division of Reactor Projects - I/II l

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical i

Specifications Date of Issuance:

October 11, 1995 l

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i ATTACHMENT TO LICENSE AMENDMENT NO. 214 FACILITY OPERATING LICENSE NO. OPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT 3/4 4-7 3/4 4-7 3/4 4-9 3/4 4-9 3/4 4-9a 3/4 4-9b 3/4 4-10 3/4 4-10 3/4 4-14 3/4 4-14 83/4 4-3 B3/4 4-3 B3/4 4-4 B3/4 4-4 B3/4 4-4a 83/4 4-4b a

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1..

1 REACTOR COOLANT SYSTEM f'

SURVEILLANCE REQUIREMENTS (Continued) i 3.

A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall j

be performed on each selected tube.

If any selected tube does i

not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

j 4.

Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by j

bobbin coil probe during all future refueling outages.

)

The tubes selected as the second and third samples (if required by c.

J Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

jl 1.

The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with' imperfections were j

previously found.

1 i

2.

The inspections include those portions of the tubes where j

imperfections were previously found.

NOTE:

Tube degradation identified in the portion of the tube that is not a reactor coolant pressure boundary (tube end up to the start of the tube-to-tubesheet weld) is excluded free the Result and Action Required in Table 4.4-2.

f d.

Implementation of the steam generator tube / tube support plate repair l

criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg-tube support plate with known outside diameter stress corrosion cracking (00 SCC) indications. The determination of the lowest cold-leg l

tube support plate intersections having 00 SCC indications shall be i

based on the performance of at least a 20 percent random sampling of

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tubes inspected over their full length.

The results of each sample inspection shall be classified into one of the j

following three categories:

(

Cateoory Insnection Results l

l C-1 Less than 5% of the total tubes inspected are degraded j

tubes and none of the inspected tubes are defect' ve.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of l

the total tubes inspected are degraded tubes.

i C-3 More than 10% of the total tubes inspected are degraded l

tubes or more than 1% of the inspected tubes are defective.

i I

i Note:

In all inspections, previously degraded tubes must exhibit i

significant (greater than 10%) further wall penetrations to i

be included in the above percentage calculations.

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'The indicated changes to this page are applicable to Cycle 8 operation only.

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SEQUOYAH - UNIT 1 3/4 4-7 Amendment No. 189, 214

REACTOR COOLANT SYSTEM f*

SURVEILLANCE REQUIREMENTS (Continued)

J 4.4.5.4 Accentance Criteria j

a.

As used in this Specification:

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1.

Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or j

specifications.

Eddy-current testing indications below 205 of i

the nominal tube wall thickness, if detectable, may be con-j sidered as imperfections.

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2.

Dearadation means a service-induced cracking,

wastage, wear or i

general corrosion occuring or either inside or outside of a tube.

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3.

Dearaded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation.

4.

% Dearadation means the percentage of the tube wall thickness i

affected or removed by degradation.

i 5.

Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

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6.

Pluaaina Limit means the imperfection depth at or beyond which the i

tube shall be removed from service and is equal to 40% of the nominal tube wall thickness.

Plugging limit does not apply to that portion of the tube that is not within the pressure boundary of the reactor coolant system (tube end up to the start of the tube-to-tubesheet weld). This definition does not apply to tube support plate intersections if the voltage-based repair criteria t

are being applied. Refer to 4.4.5.4.a.10 for the repair limit l

applicable to these intersections.

l 7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a l

loss-of-coolant accident, or a steam line or feedwater line l

break as specified in 4.4.5.3.c, above.

i 8.

Tube Insnaction means an inspection of the steam generator tube i

from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

9.

Preservice Insnection means a tube inspection of the full length of each tube in each steam generator perfonned by addy current techniques prior to service establish a baseline con-i dition of the tubing. This inspection shall be perfoneed prior i

to initial POWER OPERATION using the equipment and techniques l

expected to be used during subsequent inservice inspections.

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10. Tube Suonort Plate Pluaaina Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented outside diameter i

stress corrosion cracking confined within the i

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• The indicated changes to this page are applicable to Cycle 8 operation only.

l SEQUOYAH - UNIT 1 3/4 4-9 Amendment No. 189, 214

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• REACTOR COOLANT SYSTEM j'

SURVEILLANCE REOUIREMENTS (Continued) l thickness of the tube support plates. At tube support plate I

intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described 4

below:

a.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the j

bounds of the tube support plate with bobbin voltages less than or equal to the lower voltage repair limit (Note 1),

will be allowed to remain in service.

b.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the I

bounds of the tube support plate with a bobbin voltage i

greater than the lower voltage repair limit (Note 1), will be repaired or plugged, except as noted in 4.4.5.4.a.10.c i

j below.

c.

Steam generator tubes, with indications of potential degradation attributed to outside diameter stress l

corrosion-cracking within the bounds of the tube support l

plate with a bobbin voltage greater than the lower voltage i

repair limit (Note 1), but less than or equal to upper voltage repair limit (Note 2), may remain in service if a i

rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indications of 2

outside diameter stress corrosion-cracking degradation I

with a bobbin coil voltage greater than the upper voltage

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repair limit (Note 2) will be plugged or repaired.

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d.

Not applicable to SQN.

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l e.

If an unscheduled mid-cycle inspection is performed, the i

j following mid-cycle repair limits apply instead of the limits l-identified in 4.4.5.4.a.10.a, 4,4.5.4.a.10.b, and 4.4.5.4.a.10.c.

The mid-cycle repair limits are deterair;ad from the following equations:

Y u

I 1.0 +NDK+Gr ("' A0 CL l

v,v,-(v -v f 'c','O a

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• The indicated changes to this page are applicable to Cycle 8 operation only.

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SEQUOYAH - UNIT 1 3/4 4-9a Amendment No. 189, 214

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Contineed) where:

V upper voltage repair limit m

V lower voltage repair limit tat V.

mid-cycle upper voltage repair limit based on time into cycle mid-cycle lower voltage repair limit based on V.g V

and time into cycle length of time since last scheduled inspection At during which V, and V, were imp 1peented cycle length (the time between two scheduled steam CL generator inspections) structural limit voltage V

n average growth rate per cycle length Gr.

95-percent cumulative probability allowance for NDE nondestructive examination uncertainty (i.e., a value of 20-percent has been approved by NRC)

Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.a.10.a. 4.4.5.4,a.10.b, and 4.4.5.4.a.10.c.

Note 1:

The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing or 2.0 volts for 7/8-inch diameter tubing.

i Note 2:

The upper voltage repair limit is' calculated according to the methodology i

in GL 95-05 as supplemented.

V, may differ at the TSPs and flow distribution baffle.

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• The indicated changes to this page are applicable to Cycie 8 operation only.

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SEQUOYAH - UNIT 1 3/4 4-9b Amendment No.18g 214 i

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j REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) b.

The steam generator shall be determined OPERA 8LE 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 Rann.th a.

Following each inservice inspection of steam generator tubes, the i

number of tubes plugged in esca steam generator shall be reported to the Commission within 15 days.

l 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.g.2 within 12 months following completion of the inspection. This Special Report shall < nclude:

1.

Number and extent of tubes inspected.

j 2.

Location and percent of wall-thickness penetration for each j

indication of an imperfection.

3.

Identification of tubes plugged.

i c.

Results of steam generator tube inspections which fall into Category C-3 shall be reported pursuant to Specification 6.6.1 prior to resumption of plant operation. The written followup of 8

l this report shall provide a description of investigations conducted l

to determine cause of the tube degradation and corrective measures l

taken to prevent recurrence.

l d.

For implementation of the voltage-based repair criteria to tube i

support plate intersections, notify the staff prior to returning l

the steam generators to service should any of the following i

conditions arise:

l 1.

If estimated leakage based on the projected end-of-cycle (or if not practical using the actual measured end-of-l cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next j

operating cycle.

2.

If circumferent)al crack-like indications are detected at i

the tube support plate intersections.

3.

If indications art identified that extend beyond the i

confines of the tube support plate.

4.

If indications are identified at the tube support plate elevations that are attributablesto primary water stress corrosion cracking.

5.

If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual pasured end-of-cycle) voltage distribution exceeds 1 x 10', notify the NRC and provide an assessment of the safety significance of the occurrence.

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• The indicated changes to this page are applicable to Cycle 8 operation only.

SEQUOYAH - UNIT 1 3/4 4-10 Amendment No. 189, 214 I...

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REACTOR.C00LART SYSTEM f

OPERATIONAL LEAKAGE LIMITING C0EITION FOR OPERATION i

i 3.4.6.2 Reactor Coolant System leakage shall be limited to:

1 a.

No PRESSURE B0UNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

I GPM total primary-to-secondary leakage through all steam generators j

and 500 gallons per day through any one steam generator, 1

l cc.

150 gallons per day of primary-to-secondary leakage through any one-j steam generator, d.

10 GPN IDENTIFIED LEAKAGE from the Reactor Coolant System, and l

e.

40 GPN CONTROLLED LEAKAGE at a Reactor Coolant System j

pressure of 2235 i 20 psig.

f.

1 GPM leakage at a Reactor Coolant System pressure of 2235 t 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2, 3 and 4 l

l AGIl0N:

I a.

With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STAND 8Y 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 />.

b.

With any Reactor Coolant System leakage greater than any one of the l

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 H0T STAND 8Y 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 1

greater than the above limit, isolate the high pressure portion of 4

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 STAND 8Y 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

)

i 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 SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

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• Replacement of c. with cc. is applicable for Cycle 8 operation only.

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SEQUOYAH - UNIT 1 3/4 4-14 Amendment No. 12, 214 l

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REACTOR COOLANT SYSTEM

]

BASES i

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in i

negligible corrosion of the steam generator tubes.

If the secondary coolant i

chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the 1<mitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 150 gallons per day per steam generator).

Cracks leaving a primary-to-secondary leakage less than this limit during operation wi'l have an adecuate margin of safety to withstand the loads imposed l

during normal operation anc by postulated accidents.

Sequoyah has demonstrated that primary-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown or condenser off-gas.

Leakage in excess of this limit will require plant shutdown j

and an unscheduled inspection, during which the leaking tubes will be itcated and plugged.

The voltage-based repair limits of SR 4.4.5 implement the guidance in GL l

95-05 and are applicable only to Westinghouse-designed steam generators (S/Gs) j with outside diameter stress corrosion cracking (00 SCC) located at the tube-to-tube support plate intersections.

The voltage-based repair limits are not i

applicable to other forms of S/G tube degradation nor are they applicable to i

00 SCC that occurs at other locations within the S/G. Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial 00 SCC with no significant cracks extending outside the j

thickness of the support plate.

Refer to GL 95-05 for additional description of the degradation morphology, e.

l Implementation of SR 4.4.5 requires a derivation of.the voltage structural j

limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).

l l

The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation, at the 95 percent prediction interval curve reduced to account for the lower 95/95 mrcent tolerance bound for tubing material i

i.e., tse 95 percent LTL curve). The voltage structural properties at 650*F (d downward to account for potential flaw growth during an i

limit must be adjuste operating interval and to account for NDE uncertainty.

The upper voltage repair limit; V is determined from the structural voltage limit by applying i

the following e

, tion:

qua V

-Va - V. - V.,

where V represents the allowance for flaw growth between inspections and V l

represeEs the allowance for potential sources of error in the measurement oY j

the bobbin coil voltage.

Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.

i

• The indicated changes to this page are applicable to Cycle 8 operation only.

4 SEQUOYAH - UNIT 1 B 3/4 4-3 Amendment No. 36,18S 214

i j

REACTOR. COOLANT _ SYSTEM j

BASES 4

The mid-cycle equation of SR 4.4.5.4.a.10.e should only be used during j

unplanned inspection in which addy current data is acquired for indications at j

the tube support plates.

SR 4.4.5.5 implements several reporting requirements recommended by GL 95-j 05 for situations which NRC wants to be nottfled prior to returning the S/Gs to q

service.

For SR 4.4.5.5.d. Items 3 and 4, indications are applicable only where alternate plugging criteria is being applied.

For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected i

end-of-cycle voltage distribution (refer to GL 95-05 for more information) when i

it is not practical to complete these calculations using the projected E0C voltage distributions prior to returning the S/Gs to service.' Note that if i

leakage and conditional burst probability were calculated using the measured i

EOC voltage distribution for the purposes of addressing GL Sections 6.t.1 and i

6.a.3 reporting criteria, then the results of the projected EOC voltage j

distribution should be provided per GL Section 6.b(c) criteria.

I Wastage-type defects are unlikely with proper chemistry treatment of the i

secondary coolant.

However, even if a def6ct should develop in service, it will be found during scheduled inservice steam generator tu>e examinations.

Plugging will be required for all tubes with imperfections exceeding the repair limit defined in Surveillance Requirement 4.4.5.4.a.

The portion of the tune I

that the plugging limit does not apply to is the portion of the tube that is i

not within the RCS pressure boundary (tube end up to the start of the tube-to-tubesheet weld). The tube end to tube-to-tubesheet weld portion of the tube does not affect structural integrity of the steam generator tubes and therefore j

indications found in this portion of the tube will be excluded from the Result l

and Action Required for tube inspections.

It is expected that any indications i

that extend from this region will be detected during the scheduled tube j

inspections.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated i

20% of the original tube wall thickness.

l l

Tubes experiencing outside diameter stress corrosion cracking within the j

thickness of the tube support plate are plugged or repaired by the criteria of 4.4.5.4.a.10.

1 i

Whenever the results of any steam generator tubing inservice inspection i

fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.6.1 prior to resumption of plant opera-tion.

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.

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• The indicated changes to this page are applicable to Cycle 8 operation only.

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SEQUOYAH - UNIT 1 8 3/4 4-4 Amendment No. 36, 189, 214 i

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l REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEN LEAKAGE j

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS i

The RCS leakage detection systems required by this specification are i

provided to monitor and detect leakage from the Reactor Coolant Pressure Soundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Preser

• Soundary Leakage Detection 3

Systems," May 1973.

t j

3/4.4.6.2 OPERATIONAL LEAKAGE l

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.

1 The surveillance requirements for RCS Pressure Isolation Valves provide added assurances of valve integrity thereby reducing the probability of gross valve failure and consequent intertystem LOCA.

Leakage from the RCS isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allound f

limit.

i The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the j

detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

l The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 GPM with the modulatin l

valve in the supply line fully open at a nominal RCS pressure of 2235 psig. g'his limitation ensures that in the event of a LOCA, the safety injection flow will

(

not be less than assumed in the accident analyses.

t i

The total steam generator tube leakage limit of 600 gallons per day for all l

steam generators and 150 gallons per day for any one steam generator will minimize the potential for a significant leakage event during steam line break.

Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate followinq a steam line rupture is limited to below 3.7 gpa in the faulted loop, which wil' limit the j

calculated offsite doses to within 10 percent of the 10 CFR 100 guidelines.

If the projected end of cycle distribution of crack indications results in primary-l to-secondary leakage greater than 3.7 gpa in the faulted loop during a postulated steam line break event, additional tubes must be removed from service in order to reduce the postulated primary-to-secondary steam line break leakage to below l

3.7 gps.

l PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, j

the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

i i

• The indicated changes to this page are applicable to Cycle 8 operation only.

b l

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SEQUOYAH - UNIT 1 B 3/4 4-4a Amendment No. 36, 189, 214 l

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4 REAGTOR COOLANT SYSTEM l

aASES 3/4.4.7 CHEMISTRY a

i The limitations on Reactor Coolant System chemistry ensure that corresten of l

the Reactor Coolant System is mir.imized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of 1

i the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within i

the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

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l SEQUOYAH - UNIT 1 8 3/4 4-4b Amendment No. 36, 189, 214

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