ML17329A435

Proposed Tech Specs for Interim Plugging Criteria

ML17329A435
Person / Time
Site:	Cook
Issue date:	03/27/1992
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17329A434	List: ML17329A433 ML17329A435
References
NUDOCS 9204060377
Download: ML17329A435 (55)

Text

ATTACHMENT 2 to AEP:NRC:1166A PROPOSED REVISED TECHNICAL SPECIFICATIONS PAGES 92040b0377 920327 PDR ADOCK 05000315 P

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INDEX LIMITING CONDITIONS FOR OPE TIONS 6 SURVEILLANCE RE UIREMENTS SECTION PAGE 3 4.4 REACTOR COOLANT SYSTEM (Continued) 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (Continued)

Hot Standby N ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ N ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-2 SklutdowIl ~ ~ ~ ~ ~ ~ ~ N ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-3 Reactor Coolant Loops.................................. 3/4 4-3b 3/4.4.2 SAFETY VALVES SHUTDOWN .... . . . . ... ........NE

~ ~ 3/4 4-4 3/4.4.3 SAFETY VALVES - OPERATZNG.... ~ ~ ... E 3/4 4-5 3/4.4.4 PRESSURZZERo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-6 3/4.4.5 STEAM GENERATORS............,... N E . 3/4 4-7 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems............................. 3/4 "4-15 Operational Leakage................................... 3/4 4-16 3/4.4.7 CHEMISTRYN N ~ .. N N N N..... N . N .. N N N .N.N. ~ ~ . ~ ~ . ~ ~ ., ~,,

N ~ . ~ 3/4.4-18 3/4.4.8 SPECIFIC ACTIVITY .. ~ ~ NN ...NN 3/4 4-21 3/4.4.9 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System................................ 3/4 4-25 Pressurizer........................................... 3/4 4-30 Overpressure Protection Systems.. 3/4 4-31 3/4.4.10 STRUCTURAL INTEGRITY 3/4 4-33 3/4.4.11 RELIEF VALVES - OPERATING 3/4 4-35 3/4.4.12 REACTOR COOLANT VENT SYSTEM Reactor Vessel Head Vents............................. 3/4 4-37

Pressurizer Steam Space Vents..............;........... 3/4 4-39 COOK NUCLEAR PLANT UNIT 1 VZ AMENDMENT NO. 5&E +9S

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EACTOR COO T SYSTE STEAM GEN RA 0 S LIMITING CONDITIO FO OPERATION 3.4.5 Each steam generator shall be OPERABLE.

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

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

SURVEILLANCE RE UI EMENTS 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.

The .tubes selected for each inservice inspection shall. include at least 3% of

'he total number of'ubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a ~ 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 or repaired by sleeving in the affected area.

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

COOK NUCLEAR PLANT UNIT 1 3/4 4-7 AMENDMENT NO. 403

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

3. 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.

c. 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.

d. Implementation of the steam generator tube/tube support plate interim plugging criteria for one fuel cycle (Cycle 13) requires a 100% bobbin coil inspection for hot leg tube support plate intersections and cold leg intersections down to the lowest cold leg tube support plate with known outer diameter stress corrosion cracking (ODSCC) indications.

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

~Cate or Ins ection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective'-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5S 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 - UNIT 1 3/4 4-8 AMENDMENT NO. 08, 454

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REACTOR COO SYST SURVE LLANCE RE UI ENTS Cont ued 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 ect on Fre uencies - 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 within 24 calendar months of initial criticality. 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-l 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 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:

1. 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.

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.

d. Tubes left in service as a result of application of the tube support plate interim plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-9 AMENDMENT NO. 08

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EACTOR COO SYSTE SURVEILLANCE RE UIREHENTS Continued 4.4.5.4 Acce tance Crite ia

a. As used in this Specification:

contour of a tube or sleeve from that required by fabrication drawings or specifications. Eddy-current testing indications below 20$ of the nominal wall thickness, if detectable, may be considered as imperfections.

2.

general corrosion occurring on either inside or outside of a tube or sleeve.

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

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

5. Defect means an imperfection of such severity that it exceeds the repair limit.

6. e a r P u n Limit means the imperfection depth at or beyond which the tube or sleeved tube shall be repaired or 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 or repaired prior to returning the steam generator to service. Any sleeve which, upon inspection, exhibits wall degradation of 29 percent or more of the nominal wall thickness shall be plugged prior to returning the steam generator to service. In addition, any sleeve exhibiting any measurable wall loss in sleeve expansion transition or weld zones shall be plugged. This definition does not apply for tubes experiencing outer diameter stress corrosion cracking confirmed by bobbin probe inspection to be within the thickness of the tube support plates. See 4.4.5.4.a.10 for the plugging limit for use within the thickness of the tube support plate.

7. Unserviceable describes the condition of a tube or sleeve if it leaks or contains a defect large enough 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.

8.

or sleeve from the point of entry (hot leg side) completely COOK NUCLEAR PLANT - UNIT 1 3/4 4-10 AMENDMENT NO. QS,

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REACTOR COOLANT SYSTE S S VEILLANCE RE UIREMENTS Continued around the U-bend to the top support of the cold leg. For a tube in which the tube support plate elevation interim plugging limit has been applied, the inspection will include all the hot leg intersections and all cold leg intersections down to, at least, the level of the last crack indication.

9. ~sleevin a tube is peraitted only in areas where the sleeve spans the tubesheet area and whose lower joint is at the primary fluid tubesheet face.

10. The Tube Su ort Plate Interim Plu in Criteria is used for disposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates. For application of the tube support plate interim plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude. The plant-specific guidelines used for all inspections shall be amended as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the above voltage/depth parameters. Pending incorporation of the voltage verification requirement in ASME standard verifications, an ASME standard calibrated against. the laboratory standard will be utilized in the Donald C. Cook Nuclear Plant Unit 1 steam generator inspections for consistent voltage normalization.

1. A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.75 volts, regardless of the depth of tube wall penetration, if, as a result, the projected.end-of-cycle distribution of crack

'ndications is verified to result in primary-to-secondary leakage less than 120 gpm in the faulted loop during a postulated steam line break event. The methodology for calculating expected leak rates from the projected crack distribution must be consistent with WCAP-13187, Rev. 0.

2. A tube should be plugged or repaired if the signal amplitude of the crack indication is greater than 1.75.

volts.

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

C. Steam generator tube repairs may be made in accordance with the methods described in either WCAP-12623 or CEN-313-P.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-11 AMENDMENT NO. 98) 454

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REACTOR COOLANT SYSTEMS SURVEILLANCE RE UIREMENTS Conti ued 4.4.5.5 R~e orts

a. Following each inservice inspection of steam generator tubes, if there are any tubes requiring plugging or sleeving, the number of tubes plugged or sleeved 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 period 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 or sleeved.

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.

The results of inspections performed under 4.4.5.2 for all tubes in which the tube support plate interim plugging criteria has been applied shall be reported to the Commission within 15 days following the inspection. The report shall include:

1. Listing of applicable tubes.

2. Location (applicable intersections per tube) and extent of degradation (voltage).

COOK NUCLEAR PLANT - UNIT 1 3/4 4-12 AMENDMENT NO. 98, 454

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TABLE 4 4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection Yes No. of Steam Generators per Unit Four First Inservice Inspection Second & Subsequent Inservice Inspections One~

Table Notation:

1. The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N%

of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.

2. The third and fourth steam generators not inspected during the first inservice inspection shall be inspected during the second and third inspections, respectively. The fourth and subsequent inspections shall follow the instructions described in 1 above.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-13 AMENDMENT NO. ~

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TABLE 4.4-2 TEAM GENERATOR TUBE INSPECTIO 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size 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"or sleeve C-1 None N/A N/A defective tubes and inspect C-2 Plug or sleeve C-1 None additional 2S defective tubes C-2 Plug or tubes in this and inspect sleeve 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/A N/A tubes in this S.G.s are S.G., plug or C-1 sleeve 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 additional 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 or sleeve C-3 defectiv'e tubes.

Prompt notific-ation to NRC pursuant to specification 6.9.1 n S ere xs t e num er o steam generators n t e unxt, an n xs t e num er o steam generators inspected during an inspection COOK NUCLEAR PLANT - UNIT 1 3/4 4-14 AMENDMENT NO. 60,

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EACTOR COOLANT YST 3 4.4 6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a. One of the containment atmosphere particulate radioactivity monitoring channels (ERS-1301 or ERS-1401),

b. The containment sump level and flow monitoring system, and

c. Either the containment humidity monitor or one of the containment atmosphere gaseous radioactivity monitoring channels (ERS-1305 or ERS-1405).

I ACTION'ith only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous and/or particulate radioactivity monitoring channels are inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:

a.

Containment atmosphere particulate and gaseous (if being used) monitoring system-perfonnance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3,

b. Containment sump level and flow monitoring system-performance of CHANNEL CALIBRATION at least once per 18 months,

c. Containment humidity monitor (if being used) - performance of CHANNEL CALIBRATION at least once per 18 months.

COOK NUCLEAR PLANT - UNIT 1 3/4 4-15 AMENDMENT NO. 400,~

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EACTOR COOLAN SYSTE OPERATIONAL LEAKAGE LI ITING CONDIT ON 0 0 E T ON 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 600 gallons per day total primary-to-secondary leakage through all steam generators and 150 gallons per day through any one steam generator for Fuel Cycle 13,

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

e. Seal line resistance greater than or equal to 2.27E-l ft/gpm2 and,

,f. . 1 GPM leakage from any reactor coolant system pressure isolation valve specified in Table 3.4-0.

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

A~OI ON:

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 'hours.

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, 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 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. 'ollowing C. With any reactor coolant system pressure isolation valve(s) leakage greater than the above limit, except when:

1. The leakage is less than or equal to 5.0 gpm, and

2. The most recent measured leakage does not exceed the previous measured leakage* by an amount that reduces the To satisfy ALARA requirements; measured leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

Specification 3.4.6.2.e is applicable with average pressure within 20 psi of the. nominal full pressure value.

COOK NUCLEAR PLANT.- UNIT 1 3/4 4-16 AMENDMENT NO.

Order dated April 20, 1981

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REACTOR COOLANT SYST BASES 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 second-ary 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 for one fuel cycle (Cycle 13). Axial or circumferentially oriented 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 or repaired. 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 or sleeving will be required for all tubes with imperfections exceeding the repair limit which is defined in 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.

Tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates are plugged"or repaired by the criteria of 4.4.5.4.a.10.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-2a AMENDMENT NO. 403-, 454

REACTOR COOLANT SYSTE BASES 3 4 4 5 STEAM GE ERA 0 S TUBE 'I TEGRITY (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-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.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-2b AMENDMENT NO.

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REACTOR COOLANT SYSTE BASES Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) for Fuel Cycle 13 will minimize the potential for a large leakage event during steam line break under LOCA conditions. Based on the NDE,uncertainties, bobbin coil voltage distribution.and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 120 gpm in the faulted loop and 150 gpd per steam generator in the intact loops, which will limit 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-to-secondary leakage greater than 120 gpm 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 120 gpm.

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.

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

3 4 4 7 CHEMISTRY The- limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces'he 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 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 the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-4 AMENDMENT NO.

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REACTOR COOLANT SYSTE BASES The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3 4.4 8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator

'ube 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.

Offsite doses following a main steam line break are limited to 10 percent of the 10 CFR 100 guideline. The restriction is based on a Cook Nuclear Plant site-specific radiological evaluation that assumes a post-accident primary-to-secondary leak rate of 120 gpm in the faulted loop and a primary coolant specific activity concentration corresponding to 1% fuel defects (approximately 4.6 microCuries/gram dose equivalent 1-131), rather than a specific activity of 1.0 microCuries dose equivalent I-131.

Reducing T,z 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'ctivity, 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 if justified by the data obtained.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 4-5 AMENDMENT NO.

ATTAQBKNT 3 to AEP:NRC:1166A CURRENT PAGES MARKED-UP TO REFLECT PROPOSED CHANGES

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REACTOR COOIAHT SYSTEM ttEILLAÃCE REOUIREME~S Continued

l. All tubes chat previously had detectable vali penetzacions (greater than or equal to 20t) that have not been plugged or z epaired by sleeving in the affected area.

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

3. A tube inspection (pursuant to Specification 4.4,5.4.a.8) shall be performed on each selecced tube. If any selected cube does noc permit the passage of the eddy cuzzent probe foz a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a cube inspection.

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

~ Table 4.4-2) during each fnservice inspection may be subjected to a par" ial tube inspection provided:

l. The tubes selected for the samples include che tubes from those areas of the cubi sheet array vhere tubes vith imperfections vere previously found.

j~ + 2. The inspections inc lude those porc ions of the tubes vhere

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The results of each sample inspection shall be classified. into one of the, folloving chree categories:

Cacezozv Inspection Results C-1 Less than St of che total tubes inspected are degraded tubes and none of che inspected tubes aze defective.

C-2 One or more tubes, buc not more than lt of the tocal tubes inspected are defeccive or becveen 5t and 10t of

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the total tubes inspected are degzaded tubes.

C-3 More than 10t of the cotal tubes inspected are degraded tubes or more than lt of the inspected cubes aze defective.

Note: In all inspections, previously degraded tubes must exhibi significant (greater than or equal to 10t) further vali penetracions co be included in the above percentage calculations.

COOK NUCLEAR PLAIT - UNIT 1 3/4 4-8 ANEMDNEVZ NO. gg, f51

t pg h pp

RQ~iR Cga~~gT gz~

4,4.S.l fntpec-..'on ahull ~ ecue.. iet - "e above reouire! intervice ins-ec-;on.

~i Of Suan qene. a r .uoet snail be per/orated ae Ne fol1cwfnq irecuenc:et:

ae ~<..e fir t <nse. vice int ec. On th411 bt perfo~ af e, 5 K!!ec.fve Pomr Wn...s bu wf-nfn 24 calencar cgn nt in'..al cri:ical i:y. $ uhsequenz intervfce insoec~fons tnal! be per.-~ <. ineervalt of no. less t~n 12 nor mr>> Can 2a calendar en:.",t afar cne previous fns0ec fon. Lf leo consezu-

~ fvt int ec.fons follo&nq tervfce under AYT cond) fons, not incluCfnq cne prese. rica inspec-.fon, result fn al'l inspec:fon rtsul a fall inq inC" ine C-1 cauqory or if bo cons~Cf ve insole fons dmns aa t!laC pzvfo~sly observed eeqradaefcn tat no: concfnued and no acdf:fonal deqradatfon has

<<t .nt".ection in.erval may be extended 0 a olxfcLc of once octa, 4v z)nuns ~

..e resul:t cf  :.".e intervice '.ns-ec fon of a thea genea~r ccrcuc-.ec in ac:cr<rcc <<ie Table a.'-4 ae 44 ~n:n in:a~alt fall in Qegory'C-3, tne fnsyec:fon fluency sruti "e 'r"reate" " a: ia t: orle:e. 29 Antes. ~he inc-e te '".

ins:ec:.On frecuency tnat; app! y un@f1 ale su0squenc,,ins."e.:..:n fa t'fy One 1Cerla o Qec'ciC on < <eSe3.ao c"e in~trvhl

~ ~

My ...en be ex ence- c a ."4z.mac Of once e. 4" ccn..4.

C ~ Acof:fonaI, unscheduled inse~fcs fnspec.fons tnall be pe~:r..

on eacn t:ean qenerac"r in accor.ance aft'ae ffrt: ta~le fns"ec:fon cpcfffed in Table 4.4-4 ourfnq cM thutcee tuotecue.

40 any of cne follerfnq ccnofefons;

~

&flag c4 sec&ary.gk4j.).eag. QQ$ includfnq lQk4 orfqfnacfnq fm ~-u-cuoe theet alai) in excist.

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, <w 1/affix of Specfffca<on 3.4.5.Z.

I te<snfc oc.credence qreatar tMn C~w Operatfaq Satft Qr ".quake.

A lottwf~lane enqfneer eo tafequards.

accfdent ~fr fnq acmic.on of a".e

4. 4 e4in t.etc 1'.ne or ff44MAr line brea~.

d. Tubes left in service as a result of application of the tube support plate interim plugging criteria shall be inspected by bobbin coil probe during all future refueling outages.

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REACTOR COOLANT S.g-,~

Stan EILLA.tCX REOLyIR~~~Vi 5 (Continued) 4.4. j. 4 iccoo can<<C i teria f

chi5 Speci ication:

I-"orfeccion means an exception co the dimensions, finish or contour of a cube or sleeve froa chat required by fabrication dravings or speci ficaciola. Eddy-cu rent tasting indications be L ov 20i o f the noainaL vaLL chicknes s if da tac ah 1 e, oay cora idcrod as iaperfoccf.ons.

2. De a Rc'A Qn 1eans a service induced crack'g, va5 ci je vol 0.

general corrosion occurrinj on either inside or outside of 5 cube or sleeve.

Dot".adtd Tube or Sleeve cleans an inperfoction jroacar than or equal co 20'f the no!5inaL vali thickness caused by degradation. P

4. Peccant Degradation aeans che aaount of the valL chic&ass affected oz removed by do jradacion.

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e ~ !ac: =rane an Lvpa'r!aaeian a! auch aavaricy char ic cxeaaea eha ".apace iinic.

Re" air/Flu in Laic means the i=perfection depth ac or beyond vhich che cube or sLeeved tube shall be repaired or removed froa servf,ca. Any cube vhich, upon inspection,

~ xhibics tube vill degradation of 40 percent or moro of tho noninal tube vali thickness shall be plugged or repaired prior co rocurnUlg che statiR jeneracor t4 service, AXPy sleeve vhich, upon inspection, exhibits vali de jradation of 29 percent or iora of the aeeinaL vali thickness shall be plujjod prior co returning tho steam jenoracor co service. In additi,on, any aLaevo exhibiting any aeasurable vali loss in sleeve expansion transition or veld cones shill be plug jod.

7, alUta '

defect Larjo enough co affect ita i'c Leaks or contains a structural into jricy in the event of in Operacinj Sasis Earthquaka; a <oaa.of ceo%'m "acci'danr,'5'r a scoaa L(no or J !aaavaear iina braah aa apace!iaa in 4.a.5.3.a, abave.

(ce! tube or sleeve froa the point of entry (hot Leg aide) completely around the U-bend to the top support of the coLd Leg.

~eiaavin a cube ia parniceaa anly in arear vbara cha alaava 5pana the tubesheec area ind vhoae Lover ]oint ia at the priory fluid tubeaheet fice.

COOK NUCLEhR PLABT - VHIT 1 Sl4 4,-10 mmmm so. 98, /de

Implementation of the steam generator tube/tube support plate interim plugging criteria for one fuel cycle (Cycle 13) requires a 100% bobbin coil inspection for hot leg tube support plate intersections and cold leg intersections down to the lowest cold leg tube support plate with known outer diameter stress corrosion cracking (ODSCC) indications.

This definition does not apply for tubes experiencing outer diameter stress corrosion cracking confirmed by bobbin probe inspection to be within the thickness of the tube support plates. See 4.4.5.4.a.10 for the plugging limit for use within the thickness of the tube support plate.

For a tube in which the tube support plate elevation interim plugging limit has been applied, the inspection will include all the hot leg intersections and all cold leg intersections down to, at least, the level of the last crack indication.

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~ ace~ generacor shaLL be decaained OPQAILX after co~lecin 44rroaponcing ac i4ns (plugging 4r 5 Loeving all cubes exceed

. ~ repair Llmi and alL cubes concainlng through.vaLL cracks) toluitod bv Tabi o 4.4 2. ~

ca~ gon'racor c'-'" ~ rop<<rs cay be ~de Ln accor~ca vie/

Ie clods das bed in o i her VCQ L2523 or CU.3L3.P.

g r,.S,f RIsar=i ao Folloving <<ach '.neer-if,ca 1".speccion of acean genetacot cubes,

.aro a ~ any " ~ 5 rogui 5g plugging or sleeving, che ff ivies p Lugged ar 5 Loe>ed in each scan generacot shall be number of rapot:ad

..o ccn ~ .5 res"'

'..-.5pe" .=n s.-.a.'. bo 4 ha 5 aam genetacot c'+e inse&ce

'.ncL'od in 5e Agwull Opetacing Rapotc Eor i ~

pa 44 '". v.".' ..' 'lpec ion val couple cad ~ This taport shaLL and oxcon>> af cubes L:Qpeccado

2. Laclc.cn and percanc of vali chic~ass penectacioa f4t f gall s oi e

4. an iape r fac ion.

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oeaa eoaawaoec(>> otal n 4 F~ cubes

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Results of scan generator cube inspec iona dich

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and require ptoapc nocificacion of che CoaaSlaton sha.

Eall a iaco be zeporcad putsuanc co Specificacion plane operacion. The vtic:aa follows of'his 4.9.L ptiot co taauaycf,on 4E a dasctipcioa of invescigaciona tiyorc shall 'ptarida conchcce4 co 4eceraiaa cause of cha cube dagtadacion an4 cotteccive Ieatutes eaten co ptaveac racutzIoce ~

d The results of inspections performed under 4.4.5.2 for all tubes in which the tube support plate interim plugging criteria has been" applied shall be reported to the Commission within 15 days following the inspection. The report shall include:

1. Listing of applicable tubes.

2. Location (applicable intersections per tube) and extent of degradation (voltage).

COOK gL~~~ P~ ~ ~ZT 1 3/4 4 ll ~g~ ~, jl, f5f

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10, The Tube Su ort Plate Interim lu Cr teria is used for disposition of a steam generator tube for continued service that is experiencing outer diameter initiated stress corrosion cracking confined within the thickness of the tube support plates. For application of the tube support plate interim plugging limit, the tube's disposition for continued service will be based upon standard bobbin probe signal amplitude. The plant specific guidelines used for all inspections shall be amended as appropriate to accommodate the'additional information needed to evaluate tube support plate signals with respect to the above voltage/depth parameters. Pending incorporation of the voltage verification requirement in ASNE standard verifications, an ASME standard calibrated against the laboratory standard wi,ll be utilized in the Donald C. Cook Nuclear Plant Unit 1 steam generator inspections for consistent voltage normalization.

l. A tube can remain in service if the signal amplitude of a crack indication is less than or equal to 1.75 volts, regardless of the depth of tube wall penetration, if, as a result, the projected end-of-cycle distribution of crack indications is verified to result in primary-to-secondary leakage less than 120 gpm in the faulted loop during a postulated steam line break event. The methodology for calculating expected leak rates from the projected crack distribution must be consistent with VCAP-13187, Rev. 0.

2. A tube should be plugged or repaired if the signal amplitude of the crack indication is greater than 1.75 volts.

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REACTOR COOLANT SYSTEM PERATIOtQL LEAKAGE LIMITIN6 CONDITION FOR OPERATION 3.4.6s2 Reac.or Coolant System leakage shall be limited to:

a. No PRESSURE 3OUNOARY L:AKAGE,

b. I Gpla'UHIOENIIFI 0 LEAKAGE gO tora'I primary-to-secondary leakage through all steam gener-ators and ~gallons

/50 per day through any one steam generator~ ~/g

d. 10 GPM IOENTIFIEO LEAKAGE from the Reactor Coolant System, and

e. 52 GPM CONTROLLED LEAKAGE.

f, 1 GPM leakage from any reactor coolant system pressure isolation valve specif ed in Table 3.4-0.

APPLICABILITY: MOOES 1, 2, 3 and 4 ACTION:

a. With any PRESSURE BOUNOARY LEAKAGE, be in at least HOT STANGBY within 6 hours and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE 30UNOARY LEAKAGE, 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 STANCBY 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 COLO SHUTCOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

C. With any reactor coolant system pressure isolation valve(s) leak-age greater than the above limit, except when:

The leakage is less than or equal to 5.0 gpm, and

2. The most recent measured leakage does not exceed the previous measured leakage~ by an amount that reduces the

~ o satisfy LARA requirenents, measured leakage may be measured indirectly (as from the perfo~anc of pressure. indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capaole of demonstrating valve compliance with the leakage criteria.

0. C. CQQK - UNIT 1 3/4 4-16

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M5ra 3iC..4 5 5-~~ C-NDr T~RS -i.SZ:iZZCRr~i.

Tht Su~4lanoa Raqufzs tncs foz inspection of ='lt seta@ gtntracor c bts tnsuzt c'.-.ac cne sc.=crural fncagzfcy of ".fs portion of che RC5 vfJ>

be maintained. The program for fnstrvfcs fnsptc 'on of seta! genera or cubes fs based on a aodfffcacfon of ReguLacory Cuide l.d3. Ravfsfon l.

Inservf ca fnsptc on o f 5 et an gent a'co r 4fng is essenc'l in order co mafncafn surveillance of che cond'cfons of eha c&es in the event chat evidence of mechanfcal danagt or progress fve dagzadacfon due to design, nanu securing trzozs. Or fnservfcs condfcfons cblac lead co corrosion:

Instzvice fnspecc ion of sctu generator cubing also provides a scans of charac c 4 rid fng coat na cu 4 and cause of any cube dagzadacf on so chat corztccfvt ntasuzts can be taken.

The plane fs expected co bt operated fn a aannez such thac the second.

azy coolanc vill bt "a .".cafntd vf chin chose chtafsczy limits found eo resul=

fn tltgLfgfble cozzos.'on of cht sceas generator cubes. If the secondary cooisnc chaaiscp is .-.o: cain:sL dev.ichin chase yaranecar iinics, iocalisad ~

corrosion csay lfka'y ". ~ suit fn scztss cor.osioa cracking. The extent, of ~

crash'd -dnrinr plan: or.eacionvocid be iiaiced by cha iinicacion of acean genetacor cube leaks 4 btcvten <<~e prf~ coolant coolant .5- .-,r.co ~ se da~

stai aad che secoada c /

s scam a 1 ype sc ax

). r .av"". p'.n

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'c gene da a 4 1 s 4 e p

x ce gi of saf t.& d t.

d o p4 L b s la 4 ci at . rati~ plac ha 4 n ct gc<<ee/ -st da a o 5 al tray sc u o 4

or ncf d

'co etc(en.

aquas' r d tc ex ect'.

0 s ia on oni ors vi ze ataas generator vhila undergoing crevice f

iz pl aaa 1 4 A

flushing in Mode 4 is available for decay heat recsoval and is operable/operaci."g upon refnstatacsent of auxiliary or aafn feed flov coaczol and seeaa coaczol.

l:aseage-type defects t ara unlikely vieh the all volatile treatment (A~)

of secondary coelanc. Hovavar. even H ~ derecc or sinilar cype shonid daveion in service, ic vill ba focnd dnrini schedclad insarvicc scssn rsnaracoc robe..

'txas3.nation. Sluing'in'r sleevfng vill be iicpxized for all tubes vith fsptz-factions exceeding the repair lLaft vhich fs defined fn Speoiffcacfoa

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gyrator. cube lsgpectinns of eperacfsg.'pbncs &e ...- h 4emonseracad ehe capability to reliably detect dagzadatfoa chac has penetzacad 20% of the original c&e vali ehfckness.

l henever che results of any steaa generator cubing inservice inspection fall fneo Cacagory C 3, chest results vilL be proaycly reported to the

~

Cocsslfssfon pursuant to Specification 6.9.1 prior co rescription of plant operation. Such cases vill be considered by the Coaefssioa oa a casa-byecsse besfs and may result in a requfzeaent for analysis, laboratory examinations tests. additional eddy.current fnspection, and revision of the Technical Specifications, ff necessary.

'CMK NUCZXAL PLhHT UNIT 1 I 3'.2a ~meiT So. EHN g~~

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The allowable primary-to-secondary leak rate is 150 gallons per day per steam generator for one fuel cycle (Cycle 13). Axial or circumferentially oriented cracks having a primary-to-secondary leakage less than this limit during operation vill have an adequate margin of safety to .

vithstand 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 or repaired.

QnS%r F Tubes experiencing outer diameter stress corrosion cracking vithin the thickness of the tube support plates are plugged or repaired by the criteria of 4.4.5.4.a.10.

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ator to s not fs eaka a

atad vill "

enertt f" e the lfeita use to akage ensure a s li

'ut fraction

'f l tne d GPH age Part c:

a11 0 lfef aao ibutfon n

'event f ~ ft." a st gener zr ~e pture or .aaa lf break.

lfm fs c sfstan ~f a t. asset's used f Lee an sis of ese a dents.

!00 d leak e lfaf per st generator nsures at sta gener tu fntagr fs ntafne fn the nt of a n sta line ure o under CA fans.

PR~'SSURK SQLNOARY L=4XPQc of any oagnf tude is unacceptable since it say be .fndfcatfve of an i.-.cencfrg gross failure of the pressure boundary. Should PRESSURE 80UHOARY L.".<CG"= occur through a ccayonent vhfch can be isolated froa the balance of the Raactor C:olant Systaa, plant operation may contfnue provice<

the leakfng component is promptly fsolatad fm the Reactor Coolant Systaa since fsolatfon removes ",e source. of. potent(al failure.

~ ~ &4 The Surveillance Rcquf resents for RC$ Pressure isolation'alves provfde added assurance of valve fntagrfty thereby reducfng the pea4tlflfty. of gross valve faf CHEMISTRY lure and <<onsecuent fntarsystaa LXA. Laakage fry tM RCS Pressure lsolatfon Valves is lCKNT~Fi.=3 L~~QK and ~f11 be considered as a portion of toe all~4 fait. 1 314. 4. 7 The lfaftatfons on Reactor Coolant Systee chesfstry ensure that corrosion .

of, the Reactor. C.olant Systaa fs afnfofzed and reduces the potantfal for Reac-tor Coolant Systaa leakage or failure due to stress carrosfon. Hafntafnfnq the chesfstry within the Steady Stata LfafM provfdes adequate corrosfon pro-tactfon to ensure the structural fntegrity of the Reactor Coolant Sysm over the 1 ffe of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride Ifafts are tiae and taaeerature dependent. Corrosion studies she that operation Nay he continued Wtn contaminant concentration levels fn excess

- K the Sbeady State l.fails,:uq: co the rsnefent Lfafta,.-',or tw tice fntariels ~ftMut having a sfgnificant effec. on tee structural fntagrf ay spiff.~ l,fai:aa of the Rer Coolant Systas. The tfse interval peraf tting contfnued operation "vithfrf'Mw'~&ac f'ons'of 'the.'7~$ evt." Lian&'~fdos"tfoe for'ak<aq'correc-:

tive ac-.fons to restore t1e contaminant concentrations to vfcafa ~ SCaacy 5 tlta Li ef ts.

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3]a.4.8 SPEC'."-~C PC--Ig~

Wi.e L&itations on che spec'c activity of c"e pr~~ry cooLant ensur ~ Chat che resulting 2 hou" doses ac che si=e boundary viLL not exceed an appropriaceLy saaLL f:action of Par: L00 Limits foLLoving a stean genera cube rupcure accident Ln confu.".c ion vich an assumed steady state pri>p-co-secondary sceaa generator leakage rate of L.0 CPA. The val es:->r Limni s on s'peel f ic activit/ represe'n'c f.ncerA LLLi s based upon para~et c evaluation by c"e BC of <<(p ical 5 ice Locac ons These val" es conservac've Ln c..a= speci "c si e parameters of che Cook Huclear P~ ant si:a,

~

such as si:e bou.".dary Location and zeceorologicaL conditions, vere noc considered Ln (h's eva-'uation. Yhe %C is f'."zli in'ite specific crica"'a which ri'LL be used as -he basis for ch ~ reevaluation of che specific acc'.ri=v L I 0 4a s of 7

., s s ! .

F aia4 ~ .. s eevalua ion Cay resuLc in higher Lairs .

Beduc'ng T co '.ass chan 500 .".

prevents ch ~ rel,ease of act'vicy shouLd a stean genera or = be -. p- re since che saturation pressure of -".e presa.

cooLant Ls belov c.-.e '.i = pressure of =".e exospheric sceaa relief vaLves.

~asA sue Nilllnce r(qu onencs prove,~A adequate assurance 4%ac excessive speci 'c act'vicy Lave's in . he pr&rf coolant viLL be detec ed fn su 'c'enc toe co "ake cor=ective ac 'on. j."~orsacion obtained on Eod.'.-.e spik'ng viLL be sed co.assess che parameters associated rich spiking phenoaena. A reduce'on in frequency of tsocopic anaLyses following pover changes say be pea'ssible Lf fusti 'ed by Che data obtained.

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Maintaining an operating leakage limit of 150 gpd per steam generator (600 gpd total) for Fuel Cycle 13 will minimize the potential for a large leakage event during steam line break under LOCA conditions. Based on the NDE uncertainties, bobbin coil voltage distribution and crack growth rate from the previous inspection, the expected leak rate following a steam line rupture is limited to below 120 gpm in the faulted loop and 150 gpd per steam generator in the intact loops, which will limit 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-to-secondary

'eakage greater than 120 gpm 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 120 gpm.

Offsite doses following a main steam line break are limited to 10 percent of the 10 CFR 100 guideline. The restriction is based on a Cook Nuclear Plant site-specific radiological'evaluation that assumes a post-accident primary-to-secondary leak rate of 120 gpm in the faulted loop and a primary coolant specific activity concentration corresponding to 1% fuel defects (approximately 4.6 microCuries/gram dose equivalent I-131), rather than a specific activity of 1.0 microCuries dose equivalent I-131.

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