ML17326A094

Proposed Tech Specs,Removing voltage-based Repair Criteria, F* Repair Criteria & Sleeving Methodologies from Unit 1 TS & Clarifying Bases Section

ML17326A094
Person / Time
Site:	Cook
Issue date:	08/17/1999
From:	AMERICAN ELECTRIC POWER CO., INC.
To:
Shared Package
ML17326A093	List: ML17326A092 ML17326A094
References
NUDOCS 9908250195
Download: ML17326A094 (47)

Text

Attachment 4 to C0799-05 Page 1 ATTACHMENT 2 TO C0799"05 CURRENT PAGES MARKED-UP TO REFLECT PROPOSED CHANGES TO UNIT 1 TECHNICAL SPECIFICATIONS Page Numbers 3/4 4-7 3/4 4-8 3/4 4-9 3/4 4-10 3/4 4-11 3/4 4-lla 3/4 4-lib 3/4 4-12 3/4 4-14 B 3/4 4-2a B 3/4 4-2b B 3/4 4-2c B 3/4 4-3 B 3/4 4-5 9908250i95 990817 PDR ADQCK 0500035,5 i P PDR )

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~ i 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM STEAM GENERATORS LIMITINGCONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 a!ld 4.

ACTION:

With one or more steam generators inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing Tavg above 200'F.

SURVEILLANCE RE UIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirement of Specification 4.0.5.

4.4.5.1 Steam Generator Sam le Selection and Ins ection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sam le Selection and Ins ection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.

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. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

b. The first sample of tubes selected for, each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

1. All tubes that previously had detectable wall penetrations (greater than or equal to 20%) that have not been plugged.

This Specification does not apply in Mode 4 while performing crevice flushing as long as Limiting Conditions for Operation for Specification 3.4.1.3 are maintained.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-7 AMENDMENT4M, 446, %0, 205

0 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (continued)

Tubes in those areas where experience has indicated potential problems.

A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does 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.

The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

The tubes selected for the samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

The inspections include those portions of the tubes where imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

Cat'egoO Ins ection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

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

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1%

of the inspected tubes are defective.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-8 AMENDMENT48k, 446, kV7, 4-'78,

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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

.3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (continued)

Note: In all inspections, previously degraded tubes er-8Ieevee must exhibit significant (greater than or equal to 10%) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Ins tion Fre uencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

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Thc first inservice inspection shall erformed after 6 Effective Full Power Months ut within 24 calendar months of initial criticalit Subsequent inservicc inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVTconditions, not including the preservice inspection, result in all inspection results falling into the C-1 category 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 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 4.4.5.3.a; the interval may then be extended to a maximum'of once pcr 40 months.

I 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

'ubsequent to any of the following conditions:

1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube shcct welds) in excess of the limits of Specification 3 4.6.2.

2. A seismic occurrence greater than the Operating Basis Earthquake.

3. A loss-of-coolant accident requiring actuation of the engineered safeguards.

4. A main steam line or feedwater line break.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-9 AMENDME<NT, 98, 466, 200, 205

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I 3/4 LIMI'HNGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (continued) 4.4.5.4 Acce tance Criteria As used in this Specification e

I'lm erfeclion means an exception to the dimensions, finish or contour of a tubeetaetaevtv from that required by fabrication diawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

~De radation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube,er-slaevs,

3. De raded Tube eMleeve means an imperfection greater than or equal to 20% of the nominal wall thickness caused by degradation.

Percent De radation means the amount of the tube wall thickness affected or removed by degradation.

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5. Defect means an imperfection t'oe+cu~hng gt.

of such severity that it exceeds hda0cc-+ LS 6ckpe+ibd~.

the~ limit. A ~dt.

Plu in Limit means the imperfection depth at or beyond which the tube~

4ceved-tubs shall be ~alret~ removed from service. Any tube which, upon inspection, exhibits tube wall degradation of 40 percent or more of thc nominal tube wall thickness shall be plugged~iaired prior to returning the steam generator to service.

Unserviceable describes the condition of a tube er-sleeve if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a losswf-coolant accident, or a stcam linc or fecdwater line brcak as specified in 4.4.5.3.c, above.

~lns ection determines the condition of the steam generator tube~eve from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-10 AMENDMENT 98, 45I; 466, AVE; 240, 205

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. 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-11 AMENDMENT484, 446, 477, ~,

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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM

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I h/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (continued)

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plugging e~eeviag all tubes exceeding the limit and all tubes containing through-wall cracks) required by Table 4.4-2. ~ "Q 5 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-11b AMENDMENT ~ 208215

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~ /I I 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (continued) 4.4.5.5 ~Re ons Following each inservice inspection of steam generator tubes, to the Commission within 15 days. 'eported

b. The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report shall include:

1. Number and extent of tubes eacLsleeces. inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged,~leeved-.

C. Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-12 AMENDMENTQ6, 446, 477, %$ ,

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STEAM GENERATOR TUBE INSPECTI T S INSPECT ON ND S SE ON SP ECTION Sample S i.ze Result Action Required Result Action Required Result Action Required A C-1 None N/A N/A N/A N/A minimum of S Tubes per S.G.

C-2 Plug ~&cue. C-1 None N/A N/A defective tubes and inspect C-2 Plug a~lao~ C-1 None additional 2S defect,ive tubes C-2 Plug~

tubes in thi.s and inspect ~ve-S.G. additional 4S defective tubes in this tubes S.G. C-3 Perform action for C-3 result of first sample C-3 Perform action N/A N/A for C-3 result of first sample C-3 Inspect all All other None N/f tubes in this S.G., plug

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~ S.G.s are C-1 defective tubes, and inspect 2S tubes in each other S.G.

Some S.G.s Perform action N/A N/A C-2 but no for C-2 result addi,tional of second sample Prompt noti- S.G. are fication to NRC C-3 pursuant to specification 6.9.1 Inspect all tubes N/A N/A Additional in each S.G. and S.G. is plug ~~lee~

C-3 defective tubes.

Prompt notifi.c-ation to NRC pursuant to specification 6.9.1 M n ere s t e num er or steam generators t e un t, an n s t e num er oz steam generators inspected during an inspection COOK NUCLEAR PLANT - UNIT 1 3/4 4-14 AMENDMENT NO. 40, 454 1(

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. 3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS TUBE INTEGRITY The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inseivice 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 damage or progressive degradation due to design, manufacturing errors, or inseivice 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.

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 negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system. The allowable primaiy-to-secondary leak rate is 150 gallons per day per steam generator. acks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Leakage in excess of this limit will require plant shutdown and an inspection, during which the leaking tubes will be located and pluggedw~aked. A steam generator while undergoing crevice flushing in Mode 4 is available for decay heat removal and is operable/

operating upon reinstatement of auxiliary or main feed flow control and steam control.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging er-6Ieeving; will be required for all tubes with imperfections exceeding the 'imit which is defined m Specification 4.4.5.4.a. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

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3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM NDE COOK NUCLEAR PLANT-UNIT I Page B 3/4 4-2b AMENDMENT44', 2I5

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3/4.4 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS TUBE INTEGRITY Continued Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-bywase 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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3/4 BASES REACTOR COOLANT SYSTEM s's 3/4.4 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 10 GPM IDENTIFIED LEAKAGE limitations provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The limitation on seal line resistance ensures that the seal line resistance is greater than or equal to the resistance assumed in the minimum safeguards LOCA analysis. This analysis assumes that all of the flow that is diverted from ihe boron injection line to the seal injection line is unavailable for core coolin ts o ernie iat o.

Maintaining anoperating leakage limit oi i50 gpd persteam generator (600 gpd tet will mtntmtm t e potenn for a large leakage event during steam line break under LOCA conditions.

will ensure the calculated offsite doses will remain within 10 percent-of the 10 CFR 100 requirements and that control room habitability continues to meet GDC-19. Leakage in the intact loops is limited to 150 gpd.

Also, the 150 gpd leakage limit incorporated into this specification is more restrictive than the standard operating leakage limit and is intended to provide an additional margi Hence, the reduced leakage limit, when combined. with an effective leak rate monitoring program, provides additional assurance that should a significant leak be experienced in service, it will be detected and the plant shut down in a timely manner.

PRESSURE BOUNDARY LEAKAGEof any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Should PRESSURE BOUNDARY LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

COOK NUCLEAR PLANT-UNIT I Page B 3/4 4-3 AMENDMENT$ 3, N, 454., ~, %0, 215

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I4/JV I I h I kl 3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primaiy-to-secondary steam generator leakage rate of 1.0 GPM.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Cook Nuclear Plant site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.

Reducing T,~ to less than 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible ifjustified by the data obtained.

COOK NUCLEAR PLANT-UNIT I Page B 3/4 4-5 AMENDMENT445, 466, 200 to C0799-05 Page 1 ATTACHMENT 3 TO C0799-05 PROPOSED CHANGES TO UNIT 1 TECHNICAL SPECIFICATIONS Page Numbers 3/4 4-7 3/4 4-8 3/4 4-9 3/4 4-10 3/4 4-11 3/4 4-12 3/4 4-14 B 3/4 4-2a B 3/4 4-3 B 3/4 4-5

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I 3/4 LIMH'INGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS

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3/4.4 REACTOR COOLANT SYSTE<M STEAM GENERATORS LIMITINGCONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With oneor more stcam generators inoperable, restore the inoperable generator(s) to OPERABLE status prior to increasing T,~ above 200'F.

SURVEILLANCERE UIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirement of Specification 4.0.5.

4.4.5.1 Steam Generator Sam le Selection and Ins tion - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sam le Selection and Ins ction - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in.Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each I inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall bc from these critical areas.

b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

1. All tubes that previously had detectable wall penetrations (greater than or equal to 20%) that have not been plugged..

This Specification does not apply in Mode 4 while performing crevice flushing as long as Limiting Conditions for Operation for Specification 3.4.1.3 are maintained.

COOK'NUCLEAR PLANT-UNIT1 Page 3/4 4-7 AMENDMENT403, 466, $ $ , %5

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> 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (continued)

2. Tubes in those areas where experience has indicated potential problems.

A tube inspection (pursuant to Specification 4.4.5.4.a.8) shall be performed on each selected tube. If any selected tube does 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.

I The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:

1. The tubes selected for the samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.

2. The inspections include those portions of the tubes where, imperfections were previously found.

The results of each sample inspection shall be classified into one of the following three categories:

CategoO Ins ction Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

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

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1%

of the inspected tubes are defective.

COOK NUCLEAR PLANT-UNITI Page 3/4 4-8 AMENDMENT S, 466, %VV, 4V8,

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I 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS

.i 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued)

Note: In all inspections, previously degraded tubes must exhibit significant (greater than or equal to 10%) further wall penetrations to be included in the above percentage calculations.

4.4.5.3 Ins ction Fre uencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality or replacement of steam generators. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category 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. Ifthe results of 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 4.4.5.3.a; the interval may then be extended to a maximum of once per 40 months.

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:

Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2.

A seismic occurrence greater than the Operating Basis Earthquake.

A losswf~lant accident requiring actuation of the engineered safeguards.

A main steam line or feedwater line break.

COOK NUCLEAR PLANT-UNITI Page 3/4 4-9 AMENDMENT98, 466, 200i 285

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s 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCERE UIREMENTS (continued) 4.4.5.4 Acce tance Criteria ae As used in this Specification:

~lm rfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

~De radation masm a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube. I thickness caused by degradation.

4. Percent De radation means the amount of the tube wall thickness affected or removed by degradation.

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

removed from service. Any tube which, upon inspection, exhibits tube wall degradation of 40 percent or more of the nominal tube wall thickness shall be plugged prior to returning the steam generator to service.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough i to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

~tns ction determines the condition of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-10 AMENDMENT 98, 45k, 466k

3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS o 3/4.4 REACTOR COOLANT SYSTEM

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

COOK NUCLEAR PLANT-UNlT1 Page 3/4 4-11 AMENDMENT45k) 466) %VV) 4V8) 200) 205) K5

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> 3/4.4 REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (continued) 4.4.5.5 ~Re orts Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

The complete results of the steam generator tube inservice inspection shall be included in the Annual Operating Report for the perio'd in which this inspection was completed. This report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

Results of steam generator tube inspections which fall into Category C-3 and require prompt notification of the Commission shall be reported pursuant to Specification 6.9.1 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine, cause of the tube degradation and corrective measures taken to prevent recurrence.

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0 TABLE 4.4-2 O SrEAM GENERATOR TUBE INSPECrlON IST SAMPLE INSPECHON 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of S C-1 None N/A N/A N/A N/A Tubes per S.G.

C-2 Plug defective tubes and inspect C-1 None N/A N/A additional 2S tubes in this S.G.

C-2 Plug defective tubes and C-1 None inspect additional 4S tubes in this S.G.

C-2 Plug defective tubes C-3 Perform action for C-3 result of first sample C-3 Perform action for C-3 result of first sample N/A N/A C-3 Inspect all tubes in this S.G., plug All other S.G.s are defective tubes, and inspect 2S C-1 None N/A N/A tubes in each other S.G.

Prompt notification to NRC pursuant to specification 6.9.1 Some S.G.s C-2 Perform action for C-2 result but no additional of second sample N/A N/A S.G.s are C-3 Additional S.G. is Inspect all tubes in each S.G. N/A N/A C-3 and plug defective tubes.

Prompt notification to NRC pursuant to specification 6.9.1 S =3(N/n)% N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection.

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3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS TUBE INTEGRITY The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. 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 damage or progressive degradation due to design, manufacturing errors, or inservice 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.

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 negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation, of steam generator tube leakage between the primary coolant system and the secondary coolant system. The allowable primary-to-secondary leak rate is 150 gallons per day per steam generator. Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Leakage in excess of this limit will require plant shutdown and an inspection, during which the leaking tubes will be located and plugged. A steam generator while undergoing crevice flushing in Mode 4 is available for decay heat removal and is operable/operating upon reinstatement of auxiliary or main feed fiow control and steam control.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfections exceeding the plugging limit which is defined in Specification 4.4.5.4.a.6. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-blase basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddywurrent inspection, and revision of the Technical Specifications, if necessary.

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--t 3/4.4 REACTOR COOLANT SYSTEM 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 10 GPM IDENTIFIED LEAKAGE limitations provides:allowance for a limited amount'of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIEDLEAKAGE by the leakage detection systems.

The limitation on seal line resistance ensures that the seal line resistance is greater than or equal to.the resistance assumed in the minimum safeguards LOCA analysis. This analysis assumes that all of the flow that is diverted from the boron injection line to the seal injection line is unavailable for core cooling.

'Maintaining an operating leakage limit of 150 gpd.per steam generator (600 gpd total) will minimize the potential for a large leakage event during steam line break under LOCA conditions. This operating leakage limit will ensure the calculated offsite doses will remain within 10 percent of the 10 CFR 100 requirements and that control room habitability continues to meet GDC-19. Leakage in the intact loops is limited to 150 gpd.

Also, the 150 gpd leakage limit incorporated into this specification is more restrictive than the standard operating leakage limit and is intended to provide an additional margin. Hence, the reduced leakage limit,.when combined with an effective leak rate monitoring program, provides additional assurance that should a significant leak be experienced in service, it will be detected and the plant shut down in a timely manner.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Should PRESSURE BOUNDARY LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since. isolation. removes the source of potential failure.

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3/4 BASES 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the'Cook Nuclear, Plant site,-

such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finalizing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this

'site. This reevaluation may result in higher limits.

Reducing T,~ to less than 500'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive'specific activity levels in the primly coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will,be used to assess the parameters associated with spiking phenomena. A reduction:in frequency of isotopic analyses, following power changes may be permissible ifjustified by the data obtained.

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Al~g to C0799-05 Page 1 ATTACHMENT 4 TO C0799-05 ENVIRONMENTALASSESSMENT The Licensee has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. The Licensee has determined that this license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c) (9) . This change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following speci fic criteria.

(i) The amendment involves no significant hazards consideration.

As demonstrated in attachment 1, this proposed amendment does not involve any significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

As documented in attachment 1, there will be no significant change in the types or significant increase in the amounts of any effluents released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor is the proposal anticipated to result in any change in the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.