ML20203A275

Amends 101 & 101 to Licenses NPF-37 & NPF-66,respectively, Reflect Forthcoming Replacement of Original Steam Generators

ML20203A275
Person / Time
Site:	Byron
Issue date:	02/03/1998
From:	Lynch M Office of Nuclear Reactor Regulation
To:	Commonwealth Edison Co
Shared Package
ML20203A279	List: ML20203A275 ML20203A282 ML20203B723
References
NPF-37-A-101, NPF-66-A-101 NUDOCS 9802230327
Download: ML20203A275 (19)

Text

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j NUCLEAR REGULATORY COMMISSION WAsHINPToN, D.C. 30066 4 001

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COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.101 License No NPF-37 1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated February 28,1997, as supplemented on November 13,1996, and March 20, June 24, August 19 and November 3,1997, complies with the standards and requirements of the Atomic Energy Ac* of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; 4.

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

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

The iscuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i

E.

The issuance of this amendmerit is in accordance with 10 CFR Part 51 of the Commission's regulations s1d all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as Indicated in the attachment to this license smendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:

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i 9802230327 980203

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PDR ADOCK 05000454 l

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2-(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 101 ano the Environmental Protection Plan contained in Appendix B, both of which are attached hertto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented in the first operating cycle after installation of the Babcock and Wilcox, Intemational replacement steam generators.

FOR THE NUCLEAR REGULATORY COMMISSION f

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M. David Lynch, Senior Project Manager Project Directorat3111-2 Division of Reac'cr Projects -!!!/IV Office of Nuclecr Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: February 3, 1998

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t UNITED STATES j

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30666-0001

,o COMMONWEALTH EDISON COMPAt[(

DOCKET NO. STN 50455 BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENE Amendment No.101 License No. NPF 66 1.

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

A.

7.ne application for amendment by Commonwealth Edison Company (the

,;censee) dated February 28,1997, as supplemented on November 13,1996, and March 20, June 24, August 19 and November 3,1997, complias with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility wie merate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No NPF-66 :s hereby amended to read as follows:

E

2-(2)

Technical Specificetions The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No.101 and revised by Attachment 2 to NPF 66, and the Environmental Protection Plan contained in A-ppendix B, both of which were attached to License No. NPF.37, dated February 14,1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Spechmations and the Environmental Protection Plan.

3.

This license amendment is effective as of the cate of its issuance.

FOR THE NUCJEAR REGULATORY COMMISSivN

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M, Davi ynch, Senior Project Manager Project Directorate ill-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: February 3, 1998 j

3,

ATTACHMENT TO LICENSE AMENDMENT NOS.101 AND 101 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 53-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

Remove Paaes insert Paaes Vill Vill 3/4 4-14 3!4 4-14 3/4 4-15 3/4 4-15 3/4 4-16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-17b 3/4 4-17b 3/4 417d 3/4 4-17d 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 3/4429 3/4 4-29 3/4 4-29a l

3/4 4-31 3/4 4-31 l

B 3/4 4-3 8 3/4 4-3 l

B 3/4 4-3b B 3/4 4-3b I

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M NG CONDITIONS F0 4 OPERATION AND SURVEILLANCR REOUIREMENTS SECTICH HE TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS...............

3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS.........................

3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY........................................

3/4 4-27 FICURE 3.4-1 Unit 1 (AFTER CYCLE 8) AND UNIT 2 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE RFACTOR COOLANT SPECIFIC ACTIVITY >l #Ci/ GRAM DOSE EQUIVALENT I-131....................................

3/4 4-29 FIGURE 3.4-2 Unit 1 THROUGH CYCLE 8 DOSE EQUIVALEh? I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >0.35 yci/ GRAM DOSE EQUIVALENT I-131....................................

3/4 4-29a TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM....................................

3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant 5ystem...................................

3/4 4-32 Pressurizer..............................................

3/4 4-33 Overpressure Protection Systems..........................

3/4 4-34 3/4.4.10 STRUCTURAL INTEGRITY.....................................

3/4 4-36 3/4.4.11 PEACTOR COOLANT SYSTEM VENTS.............................

3/4 4-37 BYRON - UNITS 1 & 2 VIII AMENDMENT NO.

101 i

t

REACTOR COOLANT SYSTEM l

SURVEILLANCE RE0VIREMENTS (Continued) 1)

All tubes that previously had detectable tube wall penetrations greater than 20 percent that have not been plugged or sleeved in the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, 2)

Tubes in those areas where experience has indicated potential

problems, 3)

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

If any selected tube does not I

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, 4)

For Westinghouse Model D4 steam generators, indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages, and 5)

For Westinghouse Model D4 steam generatprs, tubes which remain in l

service due to the application of the F criteria will be inspected, in the tubesheet region, during all future outages.

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 these samples include the tubes from those areas of the tube sheet array where tubes with imper Metions were previously found, and 2)

The inspections include those portions of the tubes where imperfec-tions were previously found.

d.

For Unit 1, through Cycle 8, implementation of the steam generator tube / tube support plate repair criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (0DSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length, For Westinghouse Model D4 and D5 steam generators, a random sample of at e.

least 20% of the total number of laser welded sleeves and at least 20%

of the total number of TIG welded sleeves installed shall be inspected for axial and circumferential indications at the end of each cycle.

In the event that an imperfection exceeding the repair limit is detected, an additional 20% of the unsampled sleeves shall be inspected, and if an imperfection exceeding the repair limit is detected in the second sample, all remaining sleeves shall be inspected. These inservice inspections will include the entire sleeve, the tube at the heat treated arca, and the tube to sleeve joints. The inservice inspection for the sleeves is required on all types of sleeves installed in the Byron and Braidwood Steam Generators to demonstrate acceptable structural integrity.

BYRON - UNITS 1 & 2 3/4 4-14 AMENDMENT N0.101

4 4

BEACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued)

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

Category Insoection Resulti C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

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

of the total tubes inspected are degraded tubes.

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

Note:

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

4.4.5.3 Insoection Freouencies - 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 or initial operation following a steam generator replacement. Subsequent-inservice inspections shall be performed at intervals of not less than 12 nor more than 24 caler'ar months after the previous inspection.

If two consecutive inspections, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspe-tions demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; b.

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

The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and Additional, unscheduled inservice inspections shall be performed on each c.

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)

Reactor-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2c., or BYRON - UNITS 1 & 2 3/4 4-15 AMENDMENT N0. 101

REACTOR COOLANT SYSTEM

+

SURVEILLtNCE REQUIREMENTS (Continued) 2)

A seismic occurrence greater than the Operating Basis Earthquake, or 1

3)

A Condition IV loss-of-coolant eccident requiring actuation of the Engineered Safety Features, or 4)

A condition IV main steam line or feedwater line break.

4.4.5.4

&geeotance criteria a.

As used in this specification:

1)

Imoerfectien means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube or sleeve wall thickness, if detectable, may be considered as imperfections; 2)

Deoradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside o.e outside of a tube or sleeve; 3)

Deoraded Tube means a tube or sleeve containing unrepaired imperfections greater than or equal to 20% of the nominal tube l

or sleeve wall thickness caused by degradation; i

4) t Deoradation means the percentage of the tube or sleeve wall thickness affected or removed by degradation; 5)

Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tune or sleeve containing an unrepaired defect is defective; 6)

Pluccino or Reoair Limit means the imperfection depth at or beyond which the tube shall be removed from airvice by plugging or repaired by sleeving in the affected area. The plugging or repair limit imperfection depth for tubing is equal to 40% of the nominal wall thickness.

For Westinghouse Model D4 and D5 steam generators, the plugging or repair limit imperfection depth for laser welded sleeves is equal to 40s of the nominal sleeve wall thickness, and for TIG welded sleeves is equal to 32% of the nominal sleeve wall thickness. For Westinghouse Model D4 steam generators, this definition does not apply to defects in the tubesheet that meet the criteria for an F* tube; For Unit 1, through cycle 8, this definition does not apply to tube support plate intersections for which the voltage-based plugging criteria are being applied. Refer to 4.4.5.4.a.11 for the repair limit applicable to these intersections; 7)

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity l

in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above; BYRON - UNITS 1 & 2 3/4 4-16 AMENDMENT NO.101

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inued) 8)

Tube Insnection means an inspection of-the steam generator tube from the point of entry (hot leg side) completely around the U-bend -

to the top support of the cold leg.

For_a tube that_has been -

repaired by sleeving, the tube inspection _ shall: include the sleeved i

portion of the tube, and 4

g)

Preservice Insnection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniquis expected to be used during subsequent inservice inspections.

10)- Tube Renair refers to a process that reestablishes tube i

serviceability. Acceptable tube repairs for Westinghouse Model D4 j.

or D5 steam generators will be performed by the following processes:

I a)

Laser welded sleeving as described in a Westinghouse Technical Report currently approved by the NRC, subject to the limitations and restrictions as noted by the NRC staff, or b)

.TIG welded sleeving as described in ABB Combustion Engineering Inc. Technical Reports: Licensing Report CEN-621-P,-

Revision 00, " Commonwealth Edison Byron and Braidwood Unit 1 and 2 Steam Generators Tube Repair Using Leak Tight Sleaves, i

FINAL REPORT," April 1995, and Licensing Report CEN-627-P, Revision 00-P, " Verification of the Installation Process and-Operating Performance of the ABB CENO Steam Generator Tube i

Sleeve for Use at Commonwealth Edison Byron and Braidwood i

Units 1 and 2," d uary 1996, subject to the limitations and restrictions as noted by the NRC Staff.

Tube repair includes the removal of plugs that were previously installed as a corrective or preventative measure.

A tube inspection per-4.4.5.4.a.8 is required prior to returning previously plugged tubes to evice.

i 11)

For Unit 1 through Cycle 8, the Tube Suonort Plate Pluaaina Liuit 1s used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing pred"minantly axially oriented outer diameter stress corrosion crackicg confined within the thickness of the tube support plates. At tube support plate 1-intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:

a)

Steam generator tubes, with degradation attributed to outside diameter stress co' rosion cracking within the bounds of the 1

cold-leg tube support plate with bobbin voltages less than or equal to the lower voltage repair limit [ Note 1] will be allowed to remain in service. Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the hot-leg tube support plate with bobbin voltages less than or equal to 3.0 volts will be allowed to remain in service.

BYRON - UNITS 1 & 2 3/4 4-17 AMENDMENT N0.101

REACTOR COOLANT SYSTEM

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SURVEILLANCE REOUIREMENTS (Continued) i V

M

1. 0 +NDE+Gr ( CL-A c )

m)(C U)

Va= Vm (V

-V m

Where:

V upper voltage repair limit i

V,t,t t

lower voltage repair limit V

mid-cycle upper voltage repair ot limit based on time into cycle V

mid-cycle lower voltage repair at limit based on V and time ot into cycle length of time since last At

=

scheduled inspection during which V,t and V,t were t

implemented.

4 CL cycle length (the time between

=

two scheduled steam generator inspections)

V,t structural limit voltage Gr average growth rate per cycle length NDE 9"-percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC) t Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.a.ll.a, 4.4.5.4.a.11.b, 4.4.5.4.a.ll.c and 4.4.5.4.a.11.d.

Note 1:

The lower voltage repair limit is 1.0 volt for indications-of outside diameter stress corrosion cracking occurring at cold-leg tube support plate intersections.

Note 2:

The upper voltage repair limit for indications of outside diameter stress corrosion cracking occurring at cold-leg tube support plate intersections is calculated according to the methodology in Generic Letter 9E-05 as supple:nented.

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12)

F* Distance is the distance into the tubesheet of a Westinghouse Model D4 steam generator from the secondary face of the tubesheet or the top of the last hardroll, whichever is further into the tubesheet, that has been determined to be 1.7 inches.

13)

F* Tube is a Westinghouse Model D4 steam generator tube with l

degradation below the F* distance and has no indications of degradation (i.e., no indication of cracking) within the F*

distance. Defects contained in an F* tube are not dependant on flaw geometry.

BYRON - UNITS 1.& 2 3/4 4-17b AMENDMENT NO. 101

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R'ECT05COOLANTSYSTEM

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SURVEILLANCE REQUIREMENTS (Continued) 5)

If the calculated conditionel burst piobability based on the

. projected end-of-cycle (or if cot practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10 2,

notify the NRC and provide an assessment of the safety significance of the occurrence.

6)

Following a steam generator internals inspection, if indications detrimental to the integrity of the load path necessary to support the 3.0 volt IPC are found, notify the NRC and provide an assessment of the safety significance.of the occurrence.

e.

The results of inspections of Westinghouse Nodel D4 steam generators' F* Tubes shall be reported to the Commission prior to the r1esu.r.ption of plant operation.

The report shall include:

1)

Identification :of f* Tubes, and 2) location and size of the degradation.

1 BYRON - UNITS 1 & 2 3/4 4-17 d AMENDMENT N0.101

REACTOR COOIANT SYSTEM 3/4.4.8.. SPECIFIC ACTIVITY r

LIMITING CONDITION FOR OPERATION y

3 A.8 The specific activity of the reactor coolant shall be limited to m.

Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131**,

and b.

Less than or equal to 100/E microcuries per gram of gross radioactivity.

L71'LICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2 and 3**

a.

With the specific activity of the reactor coolant granteer than 1 cleroCurie per gram DOSE EQUIVALENT I-131** for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the lituit line shown on less than 500, Figure 3.4-1, be in at least HOT STANDBY with Tag F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and b.

With the specific activity cf the reactor coolant greater than 100/E microcuries per gram, be in at least HOT STANDBY with T less than avn 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

't

• With T,y greater than or equal to 500*F.

• • For Unit 1 through Cycle 8, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 microcuries per gram.

BYRON - UNITS 1 & 2 3/4 4-27 AMENDMENT NO. 101

JtEACTOR COOIANT SYSTEM N.

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LIMITING CONDITION FOR OPERATION MON fContinued1 MODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than s :-

1 microcurie per gram DOSE EQUIVALENT I-131* or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.

SURVEILLANCE REOUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of l

Table 4.4-4.

I I

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• For Unit 1 through Cycle 8, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 microCuries per gram.

BYRON - UNITS 1 & 2 3/4 4-28 AMENDMENT No. 101

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FOR L?JIT 1(AFTER CCCLE fi UNIT 2 DOSE EQUNALENT l 131 REACTOR COOLANT SPEC 5J '

ilVITY LIMIT VERSUS PERC8ENT OF RATED THERMAL POWER WITH THE REACTOR r %

ECIFIC ACTNITY >1 pCl/ GRAM DOSE EQlm ALLNt i l

BYRON UNITS 1 & 2 3!4429 AMENDMENT NO.101

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DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC i

ACTIVITY >0.35 pCl/ GRAM DOSE EQUIVALENT l-131 l

l BYRON - UNITS 1 & 2 3/4 4-29a AMENDMENT NO.101 i

TABLE 4.4-4 (Continued)

TABLE NOTATIONS i

Until the specific activity of the Reactor Coolant System is restored within its limits.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radioiodines. The total specific activity shall be the sum of the degasred beta-gamma activity and the total of all identified gaseous activities in the sample within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level.

The latest available data may be used for pure beta-emitting radionuclides.

A radiochemical analysis for E shall consist of the quantitative measurement of the specific activity for each radionuclida, except for radionuclides with half-lives less than 10 minutes anu all radio-iodines, which is identified in the reactor coolant. The specific activitiet for thsse individual radionuclides shall be used in the determination of E far the reactor coolant sample. Determination of the contributors to E shall be based upon these energy peaks identifiable with a 95% confidence level.

• • " For Unit I through Cycle 8, reactor coolant DOSE EQUIVALENT I-131 l

will be limited to 0.35 microcuries per gram.

I l

BYRON - UNITS 1 & 2 3/4 4-31 AMENDMENT N0.101

- _ _. ~. _ _ _ _ _ _ _ _ _ _ _ _.. _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. _ _ _ _. _ _ _ _ _.

mEacram conLaMT araTEM am ena j

2/4.4.5 STEAM GENERA' TORS 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 i

based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice i

l inspection of steam generator tubing is essential in order to maintain eurveillance of the conditions of the tubes in the event that there is evidence

]

of mechanical damage or progressive degradation due to design, manufacturing t

errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides_a muans of characterising the nature j

and cause of any tube degradation so that corrective measures can be taken.

I The plant is espected to be operated in a manner such that t.he secondarv coolant will be maintained within those chemistry limits found to result in i

negligible corrosion of the steam generator tubes.

If the escondary coolant chemistry is not maintained within these limits, localised corrosion may likely result in stress corrosion cracking. Tn.e extent of cracking during plant j

operation would be limited by the limitation of steam generator tube leakage l

between the Reactor Coolant System and the Secondary Coolant System (reactor-l to-secondary leakage = 150 gallons per day per steam generator). Cracks having a reactor-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. operating plants have demonstrated that

[

reactor-to-secondary leakage of 150 gallons per stay per steam generator can j

readily be detected by radiation monitors of steam generator blowdown, j

mainsteam lines, or the steam jet air ejectors. 1,eakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired by sleeving. The l

technical bases for sleeving are described in the current Westinghouse or ABB i

combustion Engineering, Inc. Technical Reports which are applicable for

{

Westinghouse Model D4 and D5 steam generators only.

I Wastage-type defects are unlikely 91th proper chemistry treatment of the i

secondary coolant.

However, even if a defect should develop in service, it l

will be found during scheduled inservice steam generator tube examit.ations.

[

Plugging or sleeving will be r6 quired for all tubes with imperfections j

exceeding the plugging or repair limit of 40% of the tube nominal wall thickness, excluding defects that meet the criteria for Westinghouse Model D4 l

F tubes. Acceptable plugging criteria for Westinghouse Model D4 and D5 sleeved tubes are: 1) a laser welded sleeved tube _must be_ plugged if a through p

wall penetration is detected in the sleeve that is equal to or greater than 40%

of the nominal sleeve thickness, and 2) TIG welded sleeved tubes must be l

plugged if a through wall penetration is detected in the sleeve that is equal to or greater than 32% of the nominal sleeve thickness.

The plugging limit for the. sleeve is derived from Reg. Guide 1.121 analysis and utilises a 20%

allowance for eddy current uncertainty and additional degradation growth.

Inservice inspection of sleeves is required to ensure RCS integrity. Sleeve inspection techniques are described in the current Westinghouse or ABB Combustion Engineering, Inc. Technical Reports.

Steam Generator tube and sleeve inspections have demonstrated the capability to reliably detect degradation of the pressure retaining portions of the tube or sleeve wall j

thickness. Commonwealth Edison will talidate the adequacy of any system that is used for periodic inservice inspect ion of the sleeves and, as deemed appropriate, will upgrade testing mett ods as better methods are developed and validated for coemercial use.

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I BrRoM - UNITS 1 & 2 B /4 4-3 AMENDMENT NO.

101

o e

REACTOR COQ). ANT SYSTEM BAsrs 3/4.4.5 STEAM GENERATORS (Continued)

The mid-cycle equation in SR 4.4.5.4.a.11.f should only be used during unplanned inopections in which eddy current data is acquired for indications at the cold-leg tube support plates. The voltage repair limit for indications at the hot-leg tube support plates remains at 3.0 volts during unplanned inspections.

SR 4.4.5.5 implements several reporting requirements recommended by Generic Letter 95-05 for situations which the NRC wants to be notified pricf to returning the SGs to service.

For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution ratner than the projected end-of-cycle voltage distribution (refer to Generic Letter 95-05 for more information) when it is J

not practical to complete these calculations using the projected end-of-cyclo

)

voltage distributions prior to returning the SGs to service.

Note that if leakage and conditional burst prob sility were calculated using the measured end-of-cycle voltage distribution for the purposes of addressing Generic Letter 95-05 sections 6.a.1 and 6.a.3 reporting criteria, then the results of the projected end-of-cycle voltage distribution should be provided per Generic Letter 95-05 section 6.b(c) criteria.

The maximum site allowable primary-to-secondary leakage limit for end-of-cycle main steamline break conditions ipeludes the accident leakage from IPC in addi^ ion to the accident leakage from F on the faulted steam generator and the operational leakage limit of Specification 3.4.6.2.c.

The operational leakage limit of specification 3.4.6.2.c in sach of the three rpmaining intact steam generators shall include the operational leakage from F.

For Westinghouse Model D4 steam generators, plugging or repair is not l

required for tubes with degradation within the tubesheet area which f all under the alternate tube plugging criteria defined as F*.

The F* criteria is based on " Babcock and Wilcox Nuclear Technologies (DWNT) Topical Report BAW-10196P."

F tubes meet the structural integrity requirements with appropriate margins for safety as specified in Regulatory Guide 1.121 and the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB and Division I Appendices, for normal operating and faulted conditions.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to specification 6.9.2 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 oddy-current inspection, and revision of the Technical Specifications, if necessary.

BYRON - UNITS 1&2 B 3/4 4-3b AMENDMENT NO.101 i

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