ML20091P104

Proposed Tech Specs Re Steam Generator Tube Sleeving

ML20091P104
Person / Time
Site:	Kewaunee
Issue date:	01/27/1992
From:	WISCONSIN PUBLIC SERVICE CORP.
To:
Shared Package
ML111650750	List: ML20091P104
References
NUDOCS 9201310228
Download: ML20091P104 (28)

Text

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A'lTACHMENT 2 To Letter from C. R. Steinhardt (WPSC) to Document Control Desk (NRC)

Dated Y

January 27, 1992 Proposed Arnendment No.106 Affected Th Pages 9201310220'920127 -

PDR ADOCK 05000305

-p PDR l

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TABLE OF CONTENTS l TECHNICAL i')ECIFICATIONS APPENDIX A

_Section Title Page 1.0 Definitions ............ . .. . .. 1.1-1 1.0.a Quadrant-to Average Power Tilt Ratio .......... 1.1-1 1.0.b Safety limits ............ ... . .. 1.1-1 1.0.c Limiting Safety System Settiegs ............. 1.1-1 1.0.d Limiting Conditions for Operation ...... .... .. 1.1-1 1.0.e Operable - Operability ................. 1.1-2 1.0.f Operating ................ .. 1.1-2 1.0 9 Containment System Integrity . . . . . . . . . . . . . 1.1-2a 1.0.h Protective Instrumentation Logic ..... . ... . . 1.1-2a 1.0.1 Instrumentation Surveillance .............. 1.1-3 1.0.j Operating Modes ............ . ... .. 1.1-4 1.0.k Reactor Critical .. .......... ... . .. 1.1-4 1.0.1 Refueling Operation ..... ............. 1.1-5 1.0.m Rated Power .. ................ 1.1-5 1.0.n Reportable Event ............ ...... 1.1-5 1.0.0 Radiological Effluents ................. 1.1-5 1.0.p Standard Shutdown Sequence ............... 1.1-7 2.0 Safety Limits and Limiting Safety System Settings ....... 2.1-1 2.1 Safety Limits, Reactor Core ............... 2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure .. . ... 2.2-1 2.3 Limiting Safety Systems Settings, Protective Instrumentation ........... ..... .. 2.3-1 2.3.a Reactor Trip Settings ............. 2.3-1 2.3.a.] Nuclear flux ..... ....... 2.3-1 2.3.a.2 Pressurizer ............ 2.3-1 2.3.a.3 Reactor Coolant Temperature . . .. 2.3-1 2.3.a.4 Reactor Coolant Flow ........ 2.3-3 2.3.a.5 Steam Generators .......... 2.3-3 2.3.a.6 Reactor Trip Interlocks ..... . 2.3-3 2.3.a.7 Other Trips ............ 2.3-3 3.0 Limiting Conditions for Operation ............... 3.1-1 3.1 Reactor Coolant System ................. 3.1-1 3.1.a Operational Components . . . .. ....... 3,1-1 3.1.a.1 Reactor Coolan+ Pumps ....... 3.1-1 3.1.a.2 Decay Heat Removal Capability . . 3.1-la 3.1.a.3 Pressurizer Safety Valves .. . .. 3.1-2 l 3.1.a.4 Pressure Isolation Valves .. .. . 3.1-2 3.1.a.5 Pressurizer PORV and Block Valves . 3.1-2a 3.1.a.6 Pressurizer Heaters ..... .. 3.1-2a 3.1.a.7 Reactor Coolant Vent System . .. . 3.1-2a

3.1.b Heat-up & Coel down Limit Curves for Normal Operation . ................ 3.1-3 l 3.1.c M

vimum Coolant Activi N ............ 3.1-8 3.1.d leakage of Rea'cor Coo; ant . . . . . . . . . . 3.1-10 3.1.e Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration . . . . . . . . . . . . 3.1-14 3.1.f Minimum Conditions f or Criticality . . . . . . 3.1-16 3.2 Chemical and Volume Control System .. ......... 3.2-1 Proposed Amendment No. 106 TS i 1/2//92 1

Section Title Eagg 3.3 Engineered Safety features and Auxiliary Systems .... 3.3-1 3.3.a Accumulators . . . . . . . . . . . . . . . . . . 3.3-1 ,

3.3.b Safety injection and Residual Heat Removal Systems .................. 3.3-2 3.3.c Containment Cooling Systems .......... 3.3-4 3.3.d Component Cooling System . . . . . . . . . . . . 3.3 6 13.3.e Service Water System . . . . . . . . . . . . . . 3.3 6 3.4 Steam and Power Conversion System ............ 3.4-1 3.5 Instrumentation System ................. 3.5-1 3.6 Containment System .................. 3.6-1 3.7 Auxiliary Electrical Systems .............. 3.7-1 3.8 Refueling ................. 3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits . . . . . . . 3.10-1 3.10.a Shutdown Reactivity . . . . . . . . . . . . . 3.10-1 3.10.b Power Distribution Limits . . . . . . . . . . 3.10-1 3.10.c Quadrant Power Tilt Limits . . . . . . . . . . 3.10-5 3.10.d Rod Insertion Limits . . . . . . . . . . . . 3.10-5 3.10.e Rod Misalignment Limitations . . . . . . . . . 3.10-6 3.10.f Inoperable Rod Position Indica +sr Channels . . 3.10-6a 3.10.g Inoperable Rod Limitations . . . . . . . . . . 3.10-6a 3.10.h Rod Drop Time _. . . . . . . . . . . . . . . . 3.10-7 3.10.1- Rnd Position Deviation Monitor . . . . . . . . 3.10-7 3.10.j Quadrant Power Tilt Monitor . . . . . . . . . 3.10-7 3.10.k Inlet Temperature .............. 3.10-7a 3.10.1 Operating Pressure . . . . . . . . . . . . . . 3.10-7a 3.10.m Coolant flow Rate .............. 3.10-7a 3.11 Core Surveillance Instrumentation . . . . . . . . . . . 3.11-1 3.12 Control Room Postaccident Recirculation System . . . . 3.12-1 3.14 Shock Suppressors (Snubbers) . . . . . . . . . . . . . 3.14-1 l 3.15 Deleted 4.0 Surveillance Requirements .................. 4.1-1 4.1 Operational Safety Review ................ 4.1-1 4.2 ASME Code Class In-service Inspection and Testing . .. 4.2-1 4.2.a ASME Code Class 1, 2, and 3 Components and Supports .................. 4.2-1 4.2.b Steam Generator Tubes ............. 4.2-2 4.2.b.1 Steam Generator Sample Selection and Inspection

........... 4.2-3 4.2.b.2 Steam Generator Tube Sample Selection and Inspection ........... 4.E-3 4.2.b.3 Inspection frequencies ....... 4.2-4 4.2.b.4 Plugging Iimit Criteria ...... 4.2-5 4.2.b.5 Hot Leg Tubesheet Crevice Plugging Limit Criteria ........... 4.2-6 l 4.2.b.6 Reports .............. 4.2-7 4.3 Deleted 4.4 Containment Tests .................. 4.4-1 4.4.a Integrated Leak Rate Tests (Type A) ...... 4.4-1 4.4.b Local Leak Rate Tests (Type B and C) . . . . . . 4.4-3 4.4.c Shield Building Ventilation System . . . . . . . 4.4-6 4.4.d Auxiliary Building Special Ventilation System . 4.4-7 4.4.e Containment Vacuum Breaker System ....... 4.4-7 Proposed Amendment No. 106 TS ii 1/27/92

Section lille .PaE 4.5 Emergency Core Cooling System and Containment Air Cooling System Tests .................. 4.5-1 4.5.a System Tests . . . . . , . . . . . . . . . . . . 4.5-1 4.5.a.1 Safety injection System ...... 4.5 1 4.5.a.2 Containment Vessel Internal Spray System ............... 4.5-2 4.5.a.3 Containment fan Coil Units ..... 4.5-2 4.5.b Component Tests ................ 4.5-2 ,

4.5.b,1 Pumps ............... 4.5-2 1 4.5.b.2 Valves ............... 4.5-3 1 4.6 Periodic Testing of Emergency Power System ....... 4.6 1 4.6.a Diesel Generators ............... 4.0-1 4.6.b Station Batteries ............... L6-2 4.7 Main Steam Isolation Valves ............... 4.7-1 4.8 Auxiliary feedwater System .............. 4.8-1 4.9 Reactivity Anomalies .................. 4.9-1 4.10 Deleted 4.11 Deleted 4.12 Spent Fuel Pool Sweep System . . . . . . . . . . . . . 4.12-1 ,

4.13 Radioactive Materials Sources . . . . . . . . . . . . . 4.13-1 +

4.14 Testing and Surveillance of Shack Suppressors (Snubbers) 4.14-1 4.15 Deleted l 4.16 Reactor Coolant Vent System Tests . . . . . . . . . . . 4.16.~

4.17 Control Room Postaccident Recirculation System . . . . 4.17-1 5.0 Design Features .................. 5.1-1 5.1 Site .................. 5.1-1 5.2 Containment ..... ,........... 5.2-1 5.2.a Containment System . . . . . . . . . . . . . . . 5.2-1 5.2.b Reactor Containment Vessel .......... 5.2-2 5.2.c Shield Building ................ 5.2-2 5.2.d Shield Building Ventilation System . . . . ... 5.2-2 5.2.e Auxiliary Building Special Ventilation Zone and Special Ventilation System . . . . . . . . . . . 5.2-3 5.3 Reactor .................. 5.3-1 5.3.a Reactor Core . . . . . . . . . . . . . . . . . . 5.3-1 5.3.b- Reactor Cool ant System . . . . . . . . . . . . . 5.3-2 5.4 Fuel Storage .................. 5.4-1 ,

6.0 Administrative Controls ..............,.... 6-1 6.1 Responsibility ................... 61 6.2 Organization ................... 6-1 6.2.1 Off-Site ................... 6-1 6.2.2 Facility Staff . . . , , . . . . . . . . . . . . . 6-1 6.2.3 Organizational Changes . . . . . . . . . . . . . . 2

.6.3 Plant Staff Qualifications ................ 6-2 6.4 Training . . . . . . . . . , . . . . . . . . 6-2a Proposed Amendment No. 106 TS iii 1/27/92

Section Title Page 6.5 Review and Audit . . . . . . . . . . . . . . . . . 6-2a 6.5.1 Plant Operations Review Committee (PORC) . . . . 6-2a

.6.5.1.1 Function . . . . . . . . . . . . . . . G-2a 6.5.1.2 Composition .... .. . . . . . . 6-2a 6.5.1.3 Alternates ........ . . . . . 6-2a 6.5.1.4 Meeting Frequency ......... 6-3 6.5.1.5 Quorum ,............... 6-3 6.5.1.6 Responsibilities .... ....... 6-4 6.5.1.7 Authority .............. 6-5 6.5.1.8 Records ............... 6-3 6.5.2 Corporate Support Staff .....,....... 6-5 6.5.2.1 Function ............... 6-5 6.5.2.2 Organization ............. 6-6 6.5.2.3 Activities .............. 6-6 6.5.2.4 Deleted 6.5.3 Nuclear Safety Review and Audit Committee .... 6-7 6.5.3.1 Function ............... 6-7 6.5.3.2- Composition ............. 6-7 6.5 3.3 Alternates .. ............ 6-8 6.5.3.4 Consultants ............. 6-8 6.5.3.5 Meeting Frequency ...... ... 6-8 6.5.3.6 Quorum ................ 6-9 6.5.3.7 Review ................ 6-9 6.5.3.8 Audits .. . . . . . . . . . . . . 6-10 6.5.3.9 Authority . . . . . . . . . . . . . 6-11 6.5.3.10 -Records . , . . . . . . . . . . . . 6-11 6.6 Reportable Events .................. 6-lla 6.6.1 Actions ,,................ 6-11a 6.7 Safety Limit Violation . . . . . . . . . . . . . . . . . 6-12 6.8 Procedures . . . . . . . . . . . . . . . . . . 6-12 6.9 Reporting Requirements . . . . . . . . . . . . . . . . . 6-13 6.9.1 Routine Reports . . . . . . . . . . . . . . . . 6-13 6.9.1.e "Mrtup Report ........ . , . 6-13 6.9.1.b innual Reporting 'tequirements . . . 6-14 6.9.1.c Monthly Operating Report . . . . . . 6-15 6.9.2 Deleted 6.9.3 Unique Reporting Requirements ... . . . . . 6-16 6.9.3.a Annual Radiological Environmental Monitoring Report ... . . . . . . 6-16 6.9.3.b Semiannual Radioactive Ef luent Release Report .......... . 6-17 6.9.3.c Special Reports .......... . 6-19 I

6.10 Record Retention . . . . . . . . . . . . . . . . . . 6-20 6.11 Radiation Protection Program . . . . . . . . . . . . . . 6-21 6.12 System Integrity . . . . . . . . . . . . . . . . . . 6-21 6.13 High Radiation Area ...................C42 6.14 Post-Accident Sampling and Monitoring .... . . . . . . 6-22 6.15 Secondary Water Chemistry . . . . . . . . . . . . . . . . 6-23 6.16 Rcdiological Effluents . . . . . . . . . . . . . . . . . 6-23 6.17 - Process Control Program (PCP) . . . . . . . . . . . . . . 6-23 6.18 Offsite Dose Calculatior Manual (0DCM) . . . ...... 6-24 l

6.19 Major Changes to Radioactive Liquid, Gaseous and Solid Waste Treatment Systems . . . . . . . . . . . . 6-25 Prooosed Amendment No. 106 TS iv 1/27/92 i

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Osc M.00 Title Page 7/8.0 Radiological Effluent Technical Specifications and Surveillance Requirements ........ . ..... . . 7/8-1 7/8.1 Radioactive Lignid Effluent Monitoring i instrumentation . ........ . ...... . . 7/8-2 )

7/8.2 Radioactive Gaseous Effluent Monitoring l Instrumentation . ......., ........ 7/8-3 i 7/8.3 Liquid Eifluents . ......... ..... . . 7/8-4 l 7/8.3.1 Concentration ......,,......... 7/8-4 '

7/8.3 ? Dose . ................. 7/8-5 7/8.3.3 Liquid Paawaste Treatment System . . . . . . . . 7/8-6 ,

7/8.4 Gaseous Effluents . ................. 7/8-7 l 7/8.4.1 Dose Rate . . .... ......... .. 7/8-7 7/8.4.2 Dose _- Noble Gases . . . ..... . .. 7/8-8 7/8.4.3 Dose - lodine-131, lodi +-111 and '

Radionuci ides in Parti - . Form . .. . . .. 7/8-9 7/8.4.4 Gaseous Rad.._.te Treatment System ... .. . . 7/8-10 7/8.5 Solid Radioactive Waste . . . . . . . . . , . . . . . . 7/8-11 7/8.6 Total Dose . ............... .

7/8-12 7/8.7 Radiological Environmental Monitoring . . .. .. . . . 7/8-14 7/8.7.1 Monitoring Program . . . . . . . . . . . . . . 7/8-14 7/8.7.2 Land Use Census . . . . . . . . . . . . . , , . 7/8-16 7/8.7.3 Interlaboratory Comparison Program . . . . . . 7/8-18 7/8.8 Basis . . , . . . . . . . . . . . . . . 7/8-19 e

Proposeo Amendment No. 106 TS v 1/27/92

LIST OF FIGURES FIGURE TITLE l 2.1-1 . . Safety Limits Reactor Core, Thermal and Hydraulic 3.1-1 . . . Coolant Heatup Limitation Curves Applicable for Periods Up to 15 Effective Full Power Years 3.1-2 . . . Coolant Cooldown Limitations Applicable Fr Periods Up to 15 Effective full Power Years 3.1-3 . . . Deleted I

3.10-1 .. Required Shutdown Reactivity vs. Reactor Boron ConcentratNon 3.10-2 .. Hot Channel Factor Normali?.ed Operating Envelope 3.10-3 .. Control Bank Insertion Limits 3.10-4 .. Permissible Operating Bank on Indicated Flux Difference as a Function of Burnup (Typical) 3.10-5 .. Target Band on Indicated Flux Difference as a function of Operating Power Level (Typical) 3.10-6 .. V(Z) as a function of Core Height l

4.2-1 . . . Application of Plugging Limit for a Westinghouse Mechanical Sleeve L

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Proposed Amendment No. 106 TS vii 1/27/92

(4) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved "Kewaunee Nuclear Power Plant Security Manual",

Rev.1, approved by the NRC on December 15,1989; the "Kewaunee Nuclear Power Plant Security Force Training and Qualification Manual", Rev. 7, approved by the NRC on November 17, 1987; and the "Kewaunee Nuclear Power Plant Security Contingency Plan", Rev.1, approved by the NRC on September 1,1983. These manuals include amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

(5) Deleted l (6) Steam Gengtator Upper lateral Supports The design of the steam generator upper lateral supports may be modified by reducing the number of snubbers from four (4) to one (1) per steam generator.

D .- The licensees shall comply with applicable effluent limitations and other limitations and monitoring requirements, if any, specified pursuant to Section 401(d) of the Federal Water Pollution Control Act Amendments of 1972.

E. This license is effective as of the date of issuance, and shall expire at midnight on December 21, 2013.

l l Proposed Amendment No.106 5 1/27/92

4.2 ASME CODE CLASS IN-SERVICE INSPECTION AND TESTING APPLICABILITl Applies to in-service structural surveillance of the ASME Code Class components and supports and functional testing of pumps and valves.

OBJECTIVE To assure the continued integrity and operational readiness of ASME Code Class 1, 2 and 3 components.

SPECIFICATION

a. ASME Code Class 1, 2 and 3 Componer.ts and Supports

1. In-service inspection of ASME Code Class 1, Class 2 and Class 3 components and supports shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g),

except where relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i). The testing and surveillance of shock suppressors (snubbers) is detailed in Technical Specification Sections 3.14 and 4.14,

2. In-service' testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except where relief I has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).

3. Surveillance testing of pressure isolation valves:

a. Per iodic leakage testing ul on each valve listed in Table TS 3,1-2 shall be accomplished prior to entering the operating mode after every time the plant is placed in the cold -

shutdown condition for refueling, af ter each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service af ter maintenance,- repair, or l

replacement work is performed.

l UI To satisfy ALARA requirements, leakage may be measured indirectly (as l

from the performance of pressure indicators) if accomplished in accordance with l approved procedures and supported by computations showing that the method is L capable of demonstrating valve compliance with the leakage criteria.

l Proposed Amendment No. 106 TS 4.2-1 1/27/92 i

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b.-Whenever integrity of a pressure isolation valve listed in Table TS 3.1-2_ cannot be demonstrated, the integrity of the remaining pressure isolation viive in each high pressure line having a leaking valve shall be determined and recorded daily.

In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily.

b. Steam Generator Tubes Examinations of the steam generator tubes shall be in accordance with the in-service inspection program described herein. The following terms are defined to clarify the requirements of the inspectica program.

Imperfection is an exception to the dimension, finish, or contour required by drawing or specification. '

Dearadation means a service-induced cracking,

wastage, wear or general corrosion occurring on either inside or outside of a tube.

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% Degradation is an estimated 4 of the tube wall thickness affected or removed by degradation.

Degraded Tube means a tube contains an imperfection 220% of the nominal wall thickness caused by degradation.

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

Distorted Indication is a possible tube wall loss condition that is unquantifiable with a = numeric percent call due to the existing signal characteristics.

Tube Insnection means an inspection of the steam generator tube from the point of entry (e.g., hot leg side) completely around the U-bend to the top support of the opposite leg (cold leg).

! Tube is the Reactor Coolant System pressure boundary past the hot leg I L ,

side of the tubesheet and before the cold-leg _ side of the tubesheet.

!~ Tubesheet Crevice Reaion is, for the purposes of applying the in-service

. inspection program plug and repair criteria, the area from the tube end to 5 inches below the top of the tubesheet.

Pluaced Tube is a tube intentionally removed from service by plugging in the hot and cold legs because it is defective, or because its continued -

integrity could-not be assured.

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l Proposed Amendment Ne 106 l TS 4.2-2 1/27f92 l

Repaired Tutm is a tube that has been modified to allow continued service '

consistent _ with plant Technical Specifications regarding allowable tube wall- degradation, or to prevent further tube wall degradation. A tube without repairs is a nonrepaired tube.

Souirrel Indications are generally multiple stress corrosion cracks in the roll transition area and mid span of the tubesheet.

1. Steam Generator Sample Selection and insnection The in-service inspection may be limited to one steam generator on ,

a rotating schedule encompassing the number of tubes determined in TS 4.2.b.2.a provided the previous inspections-indicated that the two steam generators are performing in a like manner.

2. Steam Generator Tube Sample Selection and Inspection  !

The tubes selected for each in-service inspection shall: ,

a. Include at least 3% of the total number of nonrepaired tubes, in both steam generators, and 3% of the total number of repaired -

tubes in both steam generators. The tubes selected for these inspections shall be selected on a random basis except as noted i in 4.2.b.2.b.

b. ',oncentrate the inspection by selection of at least 50% of the-tubes to be inspected from critical areas where experience in l similar plants with similar water chemistry indicates higher  ;

potential for degradation,

c. -Include the inspection of all non-plugged tubes which previous inspections revealed in excess of 20% degradation. - The previously degraded tubes need only be inspected about the area of previous degradation -indication if their inspection is not employed- to satisfy 4.2.b.2.a -and 4.2,b 2.b above.

d. The second and third sample inspections during each in-service inspection may be less than the full length of each tube by concentrating the inspection on those areas of the -tubesheet array and on those' portions of - the tubes where tubes with imperfections were-previously found.  ;

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! e. If a tube - does not permit the passage of the eddy current L inspection probe the entire length and through the U-bend, this shall bel recorded and an adjacent tube shall be inspected. The tube which did not ~ allow passage of th~e eddy current probe shall-be considered degraded.

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c Proposed Amendment No. 106 l TS 4.2-3 1/27/92

r The results of each sample inspection shall be classified into one of the following three categories, and actions taken as described in Table 4.2-2.

Category inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes, and none of the inspectcd 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.

NOTE: In all inspections, previously degraded tubes must exhibit significant (>10%) further wall penetrations to be included in the above percentage calculations.

3. Inspection Frecuencies The above required in-service inspections of steam ger.erator tubes shall be performed at the following frequencies:

a. In-service inspections shall be performed at refueling intervals not more than 24 calendar months after the previous inspection.

If two consecutive inspections following service under AVT conditions, not including the pre-service inspection, result in all inspection results falling into the C-1 category; or if two consecutive inspections dcmonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months.

b. If the results of the in-service inspection of a steam generator conducted in accordance with Table 4.2-2 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 a subsequent inspection meets the conditions _specified in 4.2.b.3.a and the interval- can be extended to a 40-month period.

Proposed Amendment No. 106 TS 4.2-4 1/27/92

c, Additional, unscheduled in-service. inspections shall be performed i on _ each steam generator in accordance with the first sample inspection specified in Table 4.2-2 during the shutdowr subsequent to any_of-the following' conditions:

1. Primary-to-secondary tube leaks (not including leaks originating from tube-to-tubesheet welds) in excess of the limits of Specifications 3.1.d and 3.4.a.4, or

-2. A seismic occurrence greater than the Operating Basis Earthquake, or

3. A loss-of-coolant accident requiring actuation of the engineering safeguards, where the cooldown rate of the Reactor

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Coolant System exceeded 100*F/hr, or

4. - A main steam line or feedwater line break, where the cooldown rt te of the Reactor Coolant System exceeded 100*F/hr. j

d. If the type of steam generator chemistry treatment is changed significantly, the steam generators shall be inspected at the next c tage of sufficient duration following 3 months of power operation since the change.

4. Plugaina limit Criteria _

The -following criteria apply independently to tute m and sleeve wall degradation:

a. Any tube m hich, w upon inspection, exhibits tube wall degradation 4 of 50% or more shall be plugged or repaired prior to returning the steam generator to service. If significant- gencral tube thinning occurs, this criterion will be ' reduced to 40% wall F digradation. Tube repair shall be in accordance with the methods described- in ; WCAP-Il643, '"Kewaunee Steam Generator Sleeving-Report (Mechanical Sleeves)" or CEN-413-P, "Kewaunee Steam Generator Tube Repair Using Leak Tight Sleeves."

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__WFor the' 1991-1992 operating cycle only, Specification 4.2.b.4 applies to the tube excluding the hot leg tubesheet crevice region. Refer to Specification 4.2.b.5 for the hot leg tubesheet crevice region criteria.

Proposed Amendment No. 106

. TS 4.?-5 1/27/92 1

l b. Any Westinghouse mechanical sleeve which, upon inspection, l exhibits wall degradation of 31% or mcre shall be plugged prior

-l to returning the steam generator to service. Figure TS 4.2-1 l illustrates the application of tube, sleeve, and tube / sleeve

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j joint plugging limit criteria,

c. Any Combustion Engineering leak tight sleeve which, upon inspection, exhibits wall degradation of 40% or more shall be plugged prior to returning the steam generator to service. This plugging limit applies to the sleeve up to and including the weld '

region.

5. Hot tea Tubesheet Crevice Plunginq limit CriteriaI3)

The following criteria applies to indications in the hot leg tubesheet crevice region:

a. Any tubesheet crevice indication which:

1. Exhibits tube wall degradstion of 50% or more with the bobbin coil exam, or

2. Is ident ified as a multiple circumferential indication or single circumferential' indication with the motorized rotating pancake coil (MRPC) exam, or

3. Is identified as a multiple axial indict. tion (MAI) or single -

axial-indication (SAI) with the MRPC exam and is repairable by sleeving within the 27-inch sleeving boundary, oi

4. Is identified as a MAI or SAI with HRPC exam and the corresponding bobbin call was either a distorted roll indication, distorted crevice indication or squirrel, shall. be plugged or repaired prior to returning the steam generator to service. If significant general tube thinning l occurs, this criterion will be reduced to 40% wall degradation.

l Repair methods will be submitted under 10 CFR 50.90 to be incorporated as an amendment in the facility license. The l

Commission will review the repair method, issue a significant hazards determination, and amend the facility license.

b. Any tubesheet crevice indication which is not categorized in Specification 4.2.b.5.a may be left in service provided that the number of crevice indications left in service does riot exceed-a total of 388 tubes per steam generator.

(3I Specification 4.2.b.5 is applicable for the 1991-1992 operating cycle only. _

Proposed Amendment No. 106 TS 4.2-6 1/27/92 P

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6s-' Reports

a. Following each in-service inspection of steam generator tubes, if-there. are any tubes requiring plugging or repairing, the number of tubes plugged or repaired shall be reported to the Commission within 30 days.

b. The' results. of the steam generator tube in-service 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 a degradation.

3. Identification of tubes plugged.

4. Identification of tubes repaired.

c. Results : of a steam generator tube inspection which fall .into Category C-3' require prompt (within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) notification of the Commission' consistent with 10 CFR 50.72(b)(2)(i). A written follow up report shall be submitted to the Commission consistent with Specification 4.2.b.6.a using the Licensee Event Report System to satisfy the' intent of 10 CFR 50.73(a)(2)(ii).

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proposed Amendment No. 106 i- TS 4 2-7 1/27/92 i

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BASIS _

The plant was not specifically designed to meet the requirements of Section XI of the ASME Code; therefore, 100% compliance may not be feasible or practical. However, access for in-service inspection was considered during the design and modifications have been made where practical to make provisions for maximum access within the limits of the current plant design.

Where practical, the inspection of ASME Code Class 1, Class 2 and Class 3 components is performed in accordance with Section XI of the ASME Code. If a code required inspection is impractical, a request for a deviation from the requirement is submitted to the Commission for approval.

The basis for surveillance testing of the Reactor Coolant System pressure isolation valves identified in Table T.S. 3.1-2 is contained within " Order for Modification of License" dated April 20, 1981. _

Technical Specification 4.2.b These Technical Specifications provide the inspection and repair / plugging requirements for the steam generator tubes at the Kewaunee Nuclear Power Pl ant . Fulfilling these specifications will assure the KNPP steam gener: tor tubes are inspected and maintained in a manner consistent with current NRC regulations and guidelines including the General Design Criteria in 10 CFR Part 50, Appendix A.

General Design Criterion (GDC) 14 " Reactor Coolant Pressure Boundary," and GDC 31, " Fracture Prevcntion of Reactor Coolant Pressure Boundary," require that the reactor coolant pressure boundary have ao extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. Also, GDC 15, " Reactor Coolant System Design," requires that the Reactor Coolant System and associated auxiliary, control, and protection systems be designed with sufficient margin 'n ense e that the design conditions of the reactor coolant pressure boundary are not exceeded dJring any condition of normal operation, including anticipated operational -

occurrences. Furthermore, GDC 32 " Inspection of Reactor Coolant System Pressure Boundary," requires that components that are part of the reactor coolant pressure boundary be designed to permit periodic inspection and testing of critical areas to assess their structural and leak tight integrity.

The NRC has developed guidance for steam genera'or tube inspections and maintenance including Regulatory Guides 1.83 and 1.121. Regulatory Guide 1.83, "In-service Inspection of Pressurized Water Reactor Steam Generator 3 Tubes," forms the basis for many of the requirements in this section and should be consulted prior to any revisions. Regulatory Guide 1.121, " Bases for Pluaging Degraded PWR Steam Generator Tubes," defines the minimum wall thickness in a steam generator tube, and may be applied to tube sleeves in determining their minimum wall thickness.

Proposed Amendment No. 106 TS 4.2-8 1/27/92

l Technical-Specification 4.2.b.1 >

. If the steam generators are shown to be performing in a like manner, it_is '

appropriate to limit the inspection to one steem generator on a rotating schedule. Economic savings as well as reductions in personnel exposure and 4

outage. duration can be realized.

Technical Specification 4.2.b'g .

Periodic inspection of the steam generator tubes allows evaluation of their ,

service condition. As operational experience has become available it is evident that certain types of steam generators are susceptible to generic degradation mechanisms. Site specific steam gq1erator tube degradation has also occurred throughout the industry. The inspection program at Kewaunee is designed to identify both generic and site specific tube degradation

. mechanisms.

L Steam generator' tube surveillance at Kewaunee is generally performed using eddy current techniques. Various methods of eddy current (EC) testing are

used to inspect steam generator tubes for wall degradation. EC methods have improved considerably since Kewaunee began commercial operation in 1774.

Single frequency EC testing with a single probe and X-Y plotter have evoh .d

into multifrequency techniques with assorted probe types and sophisticated software to allow more accurate volumetric tube examinations. Profilometery

techniques are also being developed which detect imperfections in a tube's original geometry. WPSC is committed to utilize advancing EC testing technology, as appropriate, to assure accurate determination of the stean generator tubes' service condition.

Technical Specification 4.2.b.3 Steam generator tube itespections are generally scheduled during refueling outages at the Kewaunee Nuclear Power Plant. The tubes scheduled for a ,

given inspection are based upon their service condition determined during i- previous _ inspections, and operational experience from other plants with similar steam generators and water chemistry. Identification of degraded

steam generator tube conditions results in augmentation of the inspection effort as well as increasing the frequency _ of subsequent inspections. In d

this manner, steam generator tube surveillance is consistent with service i

conditions.

There_ are several operational c:currences or transients that will require

_ subsequent steam generator tube inspectier.s. These inspections are required i- as a yesult of excessive primary-to-secondary leakage or transients imposir.g large mechanical and thermal stresses on the tubes.

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l-E Proposed Amendment No. 106

TS 4.2-9 1/27/92 r

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Technical Specification 4.2,b.4 -

Steam-generator tubes N -found with less _ than the minimum walgickness criteria determined by analysis, as described in WCAP 7832 , must either be repaired to be kept in service or removed from service by plugging.

Steam generator tube plugging is a common method of preventing primary-to-secondary .stecm generator tube leakage and has been utilized since the inception of PWR nuclear reactor plants. This method is relat.vely uncomplicated.from a structural / mechanical standpoint as flow is cut off from the affected tube by plugging it in the hot and cold leg faces of the tubesheet.

To determine the basis- for the sleeve plugging limit, the minimum sleeve wall thickness was calculated in accordance with Draft Regulatory Guide 1.121 (August 1976).

For the Westinghouse mechanical sleeves, the sleeve plugging limi. of 31%

is applied to the sleeve as shown on Figure TS 4.2-1. For the Combustion Engineering leak tight sleeves, a plugging limit of 40% is applied to the sleeve and we'd region. The sleeve plugging 1imits allow for eddy current testing inaccuracies and continued operational degradation per Draft Regulatory Guide 1.121 (August 1976).

Repair by sleeving, or other methods, has been recognized as a viable a'i ' native for isolating unacceptab'e tube degradation and preventing tube 1-noge. Sleeving isolates unacceptr.Me degradation and extends the service life of the ' tube, and the spam generator. T accordance with WCAP 11643 and CEN-413-P L gbe repair, by sleeving in has been evaluated and analyzed as acceptable. The Westinghouse mechanical sleeve spans the degraded area of the parent tube in the tubesheet region. The sleeves are either 36", 30" or _27" to allow access permitted by channel head bowl geometry. The sleeve is hydraulically expanded and hard rolled into the parent tubing.

NFor the 1991-1992~ operating cycle only, Specification 4.2.b.4 applies to the tube ' excluding the hot leg tubesheet crevice region. Refer to Specification 4.2.b.5 for the tubesheet crevice _ region criteria.

' N WCAP'7832, " Evaluation d Steam Generator Tube, Tube Sheet .d Divider

_ Plate Under Combined LOCA Plu~ > Cc7ditions."

N .E W. James, WPSC, to A. schwencer, NRC, dated September 6, 1977.

m WCAP - 11643, Kewaunee Steam Generator- Sleeving Report, Revision 1, November 1988 (Proprietary).

N CEN-413-P, "Kewaunee Steam Generator Tube hpair Using Leak Tight Sleeves," January 1992 (Prcprietary).

Proposed imer.dment No. 106 TS 4.2-10 1/27/92

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[ ihert are three types of Combustion Engineering leak tight sleeves.- The i first type, the straight tubesheet sleeve, spans the degraded area of_ the parent tube-in the tubesheet crevice region. The sleeve is welded to the parent tube n_ ear each end. The second type of sleeve is the peripheral ,

3 tubesheet -sleeve. The sleeve is initially curved as part of the 1 manufacturing process and straightened as part of_ the installation i process. The third type of sleeve, the tube support plate sleeve, spans the degraded area of the tube support plate and is installed up to the sixth support plate. This sleeve is welded tn the parent tube near each

end of the sleese.

The hydraulic equivslency ratios for the application of normal operating, 4 upset, and accident condition bounding analyses have been evaluated.

Design, installation, testing, and inspection of steam generator tube sleeves requires substantially more engineering than plugging, as the

. tube remains in service. Because of this, the NRC has iefined steam

generator tube repair _to be an Unreviewed Safety Question as described in 10 CFR 50.59(a)(2),

As such, other tube repair. methods will be submitted under 10 CFR 50.90; ard in accordance with 10 CFR S0.91 and 92, the Commission will review the method, issue a significant hazards determination, and amend the facility license accordingly. A 90-day time I frame for NRC review and approval is expected.

! Technical Specification 4.2.b.5 4

The purpose of Specification 4.2.b.5 is to clarify the repair criteria for ambiguous eddy current indications in the hot leg tubesheet crevice region and is applicable for the 1991-1992 operating cycle only. During the spring i 1990 refueling outage, eddy current inspections using a rotating pancake coil found axial indications in the tubesheet crevice region which were no?

discernible using the standard bobbin coil eddy current probe. A metallurgical exam of two tubes pulled from Steam - Generator B hot leg

revealed the presence of axial cracks within the tubesheet crevice area

which could not be reliably detected and sized with the standard bobbin coil i technology.

l An evaluation of tube integrity and associated radiological consequences was performed to show that continued operation of - the plant with these i indications in service provided an adequate margin of safety. This evaluation- was based the pulled tube -exam and leakage rate testing of n crevices restricted by top-of-tubesheet dents analogous to those present in l the Kewaunee steam gen The details - of- this evaluation are documented in WCAP 12790gators.

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W WCAP 12790 "Kewaunee Steam Generator Mid-Cycle Report," December 1990 k

j (Proprietary).

i Proposed Amendment No. 106 j TS 4.2-11 1/27/92

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The results of this evaluation conservatively demonstrate that with an operating -leak rate ' limit. of 200 gallons per day (administrative limit imposed for the 1991-1992 operating cycle), a total of 388 tubes per steam generator with through wall cracks in the tubesheet crevice region can be in service without exceeding 10% of the 10 CFR Part 100 guidelines during a postulated steam line break.

4' During the 1991 refueling outage 27-inch sleeves will be installed in addition to the 30- or 36-inch sleeves which were used in pn t lous outages.

The' 27-inch sleeves expand the current sleeving boundary to cover approximatel) 84% of the tube bundle. During the 1992 refueling outage, flexible., sleeving technology may be used which will extend the sleeving boundary ta all but the- otitermost tubes. Therefore, this specification is an . interim measure for the 1991-1992 operating cycle until the sleeving bcundary is extended, e

. Technical. Speci ficpti.g3 4.2.b.6 Category C-3 inspection- results are considered abnormal degradation to a principal safety barrier and are therefore reportabl.e under 10 CFR Part 50.72(b)(2)(1) and 10 CFR Part 50.73(a)(2)(ii),

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i Proposed Amendment No. 106 TS 4.2-12 1/27/92 l

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Plugging tialt

{ GPPile s to tube caly i

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> (l, Plugging limits UPPER ^d Sleeve Jann l

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IC l l}, N'N EXISTING l TUBE l-l SLEEVE I,

l TUBESHEET

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l sleeve only N

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opp tf te ve r n } -p q ond tube APPLICATION OF PLUGGING LIMIT FOR A WESTINGHGUSE MECHANICAL SLEEVE l Proposed Amendment No. 106 FIGURE TS 4.2-1 1/27/92

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- ATTACHMENT 3 To 9 Letter from C. R. Steinhardt (WPSC) to Document Control Desk (NRC) j i

Dated January 27,1992 b

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l Proposed Amendment No.106 .

Affidavit Pursuant to 10CFR2.790 CEN-413-P, "Kewaunce Steam Generator Tube Repair Using Leak Tight Sleeves" January,1992

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[ 'i?.Ilu f LT_.E U 61MllT TO ULMR_b192 l

Combustioni Engineering, Inc. )

State of Connecticut )

Cour.tr i SS :

g i of Ec"?fot%

[

I[ 1, S. A. Toclle, dept 49 M say that. I am the Manager, Operating y

a ctrr Licensing, of Comturtion Engins;cring, Inc., duly authorized

?%s by

.s bis affida91t, ;zr$c M ve rov bwed or caused to have reviewed 5 1:e inrarmatica which is identified as proprietary and referenced in

,e pcragraph immediately be

0w. . ?m submitting this affidavit in

'snformance with the orgvisions of *.0 CFR 2.790 of the Commission's regulations and in conjunction with the application of Wisconsin Pub'ic Servico engporation for M thholding this information.

The infortaation for wnich proprietary treatment is sought is contained in the following documents:

CEN-413-P, "Kewaunco Steam Generator Tube Repair Using Leak Tight _

Sleeves," Januarv 1992.

These documents have boet appropriately designated as proprietary.

I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged cr as confidential commercial or financial information.

Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the

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information' sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

F

1. The information sought to be withheld from public disclosure, which is owned 3nd has been held in confidence by Combustion Engineering, concerns the design, manuf acture, installation, and testing of the steam generator tube welded sloove for repairing degraded tubes.

2. The information' consists of test data or other similar data concerning a process, method or component, the application of which results in substantial competitive advantage to combustion Engineering.

3. The information is of a type customarily held in confidence by Combustion Engineering and not customarily disclosed to the public. Combustion Engineering has a rational basis for detarmining the~ types of information customarily held in confidence by.it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confiaence. The details of the aforementioned system were provided:to the Nuclear Regulatory Commission via letter DP-537

.from F. M. Stern to Frank Schroeder dated December 2, 1974.

This system was applied in determining that the subject document herein is proprietary.

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4. The information is being transmitted to the Commission in confidence ader the provisions of 10 CFR 2.790 with the ur.derstanding that it is to be received in confidence by the Commission.

5. The information, to the best of my knowledge and belief, is not available in public sources, and any disclosure to third parties has beeri made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

6. Public disclosure of the information is likely to causa substantial harm to the competitive position of Combustion Engineering because:

a. A similar product is manufactured and sold by major pressurized water reactor competitors of combustion Engineering.

b. Development of this information by C-E required tens of thousands of manhours and millions of dollars. To the best of my knowledge and belief, h competitor would have to undergo similar expense in generating equivalent information.

c. In order to acquire such information, a competitor would also require considerable time and inconvenience to develop

{ the methodology for steam generator tube repair using leak tight sleeves for degraded tubes.

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d. The information required significant ef fort and expense to obtain the licensing approvals necessary for application of the information. Avoidance of this expense would decrease L a competitor's cost in applying the information- and 4

marketing the product to which the information is applicable.

e. The informatior, consists of analyses of the methodology used to repair steam generator tubes using leak tight sleeves, the application of which provides a competitive economic advantage. The availability of such information to competitors would enable them to modify their product to better compete with Combustion Engineering, take marketing or other actions to improve their product's position or impair the position of Combustion Engineering's product, and avoid developing similar data and analyses in support of their processes, methods or apparatus, f, In pricing Combustion Engineering's products and services, significant research, development, engineering, analytical, manuf acturing, licensing, quality assurance and other costs and expenses must be included. The ability of Combustion Engineering's competitors to utilize such information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.

E

g. Use of the information by competitors in the international-marketplace would increase their ability to market nuclear steam supply systems by reducing the costs associated with 1

r their technology development. In addition, disclosure would have an adverse economic impact on combustion Engineering's potential for obtaining or maintaining foreign licensees.

Further the deponent sayeth not, n

s. A. W(

S. A. Toolle Manager Operating Reactor Licenr.ing Sworn to before me this R day of IrWLM N / , 1992 1

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ac s.. w~ lhtGbl.

Ifotary Public My commission expires: F/3/ 94 4

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NITACilMENT 4 l To l

1 etter from C. R. Steinhardt (WPSC) to Document Control Desk (NRC) l l

Dated January 27, 1992 l'roposed Aniendnient No.106 Combustion Engineering Report

" Test Report in Load Cycle Test of Axial Tube End Cracks in Welded Sleeve /T'ibe Samples" January,1992