Text

Proposed Tech Specs,Implementing 3.0 Volt Bobbin Coil Probe, Voltage Based,Sg TSP Alternative Plugging Criteria Limit for Outside Diameter Stress Corrosion Cracking Indications at hot-leg TSP Intersections
	ML20086D999
Person / Time
Site:	Byron, Braidwood
Issue date:	07/07/1995
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20086D995	List: ML20086D991; ML20086D999;
References
	NUDOCS 9507110111
	Download: ML20086D999 (61)
	v • d • e

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ATTACH 1MNT C-l' )

l MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A j TECHNICAL SPECIFICATIONS OF.

FACILITY OPERATING LICENSES NPF-72, AND NPF-77 1

i BRAIDWOOD STATION UNITS 1 & 2 REVISED PAGES:

3/4 4-13*

3/4 4-14 3/4 4-14a* i 3/4 4-15* ,

3/4 4-16 !

. 3/4 4-17 ,

3/4 4-17a !

3/4 4-17b l 3/4 4-27 i

3/4 4-28 3/4 4-29 l 3/4 4-30 i 3/4 4 1 B 3/4-3* l B 3/4 4-3a

)

NOTE: THESE PAGES HAVE NO CHANGES BUT ARE INCLUDED FOR CONTINUITY.

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9507110111 950707 PDR- ADOCK 05000454 P PDR

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REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. l APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of i Specification 4.0.5.

]

4 . 4 . 5.1 Steam Generator Samole Selection and Insoection - Each steam generator l shall be determined OPERABLE during shutdown *y selecting and inspecting at least the minimum number of steam generators ., ecified in Table 4.4-1.

4.4.5.2 Steam Generator Tube

Samole Selection and Insoection - The steam generator tube minimum sample size, inspection result classification, and the l corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator. tubes shall be performed at the fre-quencies 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. When j applying the expectations of 4.4.5.2.a through 4.4.5.2.c, previous defects or 1 imperfections in the area repaired by.the sleeve are not considered an area requiring reinspection. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

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:

When referring to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.5.4.a.10.

BRAIDWOOD - UNITS 1 & 2 3/4 4-13 AMENDMENT NO. 46

EEACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued)

s. ;

1)' All tubes 'that previously had-detectable tube wall penetrations '

greater than 20% 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) At least 3% of the total number of sleeved tubes in all four steam generators or all of the sleeved tubes in the generator chosen for the. inspection program, whichever is less. These inspections will include both the tube and the sleeve, and .

4) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be perfonned on each selected tube. If any selected tube does not c permit the passage of the eddy current probe for a tube inspection, MWQ this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.

(c ) For Unit 1, tubes which rema'in in service due to the application of (

the F criteria will be inspected, in the tubesheet region, during all future outages. , k

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 1) areas of the tube sheet array where tubes with imperfections were previously found, and ' ,

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

M h Unit I Cycle 5, implementation'of the tube support plate int plugg iteria limit requires a 100% bobbin coil probe etion for !

all hot leg upport plate intersections and al leg u- intersections down lowest cold leg tu pport plate with outer diameter stress corrosion ng (0 ndications. An inspection

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using a rotating pancake coil OPERABILITY of tubes wit w-like bob be is required in order to sh'ow 1 signal amplitudes greater than 1.0 volt but han or equal to 2.7 vo . or tubes that will be administra plugged or repaired, no RPC inspection equired. The RPC s are to be evaluated to establish that the principa f ications can be characterized as ODSCC.

e. A random sample of at leas't 20 percent of .the total number of sleeves shall be inspected for axial and circumferential indications at the end of each cycle. In the event that an imperfection of 40 percent or greater depth is detected, an additional 20 percent of the unsaiupled sleeves shall be inspected, and if an imperfection of 40 percent or greater depth is detected in the second sample, all remaining sleeves shall be inspected.

These inservice inspections will include the entire sleeve and the tube at l

BRAIDWOOD - UNITS 1 & 2 3/4 4-14 AMENDMENT.NO. I

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INSERT A f (4.4.5.2.b)

! 5) For Unit 1, indications left in service as a result of i application of the tube support plate voltage-based l repair criteria shall be inspected by bobbin coil probe l during all future refueling outages.

l

INSERT B ;

i (4.4.5.2.d) 1

d. For Unit 1, 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 (ODSCC) indications. The determination of the.: lowest cold-leg tube _. support plate intersections having ODSCC indications shall be based 3 on the performance of'at least a 20 percent random- ;

sampling.of tubes inspected over.their full length. >

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REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) the heat treated area. The inservice inspection for the sleeves is required until the corrosion resistance for the laser welded or kinetically welded joints in tubes that bound the material parameters of the tubes installed in the steam generators has been demonstrated acceptable. If conformance with the acceptable criteria of Specification 4.4.5.4 for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service.

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

Cateaory Insoection Results 1

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 flore 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.

s BRAIDWOOD - UNITS 1 & 2 3/4 4-14 a AMENDMENT N0. 63

REACTOR COOLANT SYSTEM i

SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frecuencies - The above required inservice inspections of 4 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, not including the pre-service inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that pre-viously 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 I

/ maximum of once per 40 months; and ;

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c. Additional, unscheduled inservice inspections shall be performed on ;

each steam generator in accordance with the first sample inspection j specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions: l

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 :

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

3) A Condition IV loss-of-coolant accident requiring actuation of

the Engineered Safety Features, or

4) A Condition IV main steam line or feedwater line break.

)

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.BRAIDWOOD - UNITS 1 & 2 3/4 4-15

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REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) .

s.

4.4.5.4 Acceotance Criteria

a. As used in this specification: ,

1) Imoerfection 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)

Dearadation means a service-induced cracking,

wastage, wear or '

general corrosion occurring on either inside or outside of a l tube or sleeve; -

1

3) Deoraded Tube means a tube or sleeve containing unrepaired imperfections greater than or- equal to 205 of the nominal tube !

or sleeve wall thickness caused by degradation; l

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

5) Defect means an imperfection of such severity th'at it exceeds the plugging or repair limit. A tube or sleeve containing an unrepaired defect is defective;

6) Pluaaina or Renair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the affected area. The plugging or repair limit imperfection depth is equal to 405 of the nominal wall thickness. For Unit 1, this definition does not apply t defects in the tubesheet that meet the criteria for an F tu f:r Unit ! Cy:1: 5, thi: d:finiti:: d::: ::t :;;ly t th: r:g. n-h-

--)2yg Y Of the tub: :dj::t t; th: tri: ::p;;rt pht; int:ri;;; phsging M( crit:rh 1i:it, i.:., th: tut; ::pp;rt p ht int:r;;;:ti:n:.

[L geg p- !;::ift::ti:: 1.1.5.1.:.!! d:::rf5:: th: :;:!r li it for ::

withir th: td: : pp:rt pht: int:r:::th: :f th: t b:;

7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity 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; i

8) Tube inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the i 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 portion of the tube, and i

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BRAIDWOOD - UNITS 1 & 2 3/4 4-16 AMENDMENTNO.fr3

INSERT C (4.4.5.4.a.6)

For Unit 1, this definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to 4.4.5.4.a.11 for the repair limit applicable to these intersections;

t I

REACTOR C001. ANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) h.

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 tha tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques

' expected to be used during subsequent . inservice inspections.

10) Tube Renair refers to a process.that reestablishes tube serviceability. Acceptable tube repairs will be performed by

'ch . the following processes:

a) ' Laser welded sleeving as described in .a Westinghouse Technical Report currently approved by the NRC, subject to 4 the limitations and restrictions as noted by the NRC staff, 2_

or '

b) Kinetic welded sleeving as described in a Babcock & Wilcox Nuclear Technologie,s Technical Report currently aproved by r/

the NRC, subject to the limitations and restrictions as ' ' -

noted by' the NRC staff. I

/

Tube repair includes the removal of plugs that were previously 1 installed as a corrective or preventative measure. A tube j inspection per 4.4.5.4.a.8 is required prior to returning previously plugged tubes to service.

Tube Suonort Plate interim Pluaaina Criteria limit for Unit 1 Cycle 5 is used for the disposition of a steam generator tube r continued service that is experiencing ODSCC confinedpi' thin

.the hickness of the tube support plates. For applicat n of the t support plate interim plugging criteria limi , the d upon tube's standarddigositionin coil for continued probe signal amplitude service of willf)aw-be b, like indications. plant specific guidelines uje8 for all inspections sha be consistent with the ed y current guidelines Q\ x in Appendix.A of -13854 as appropria to accommodate the additional- informatio eeded to evalu e tube support plate j -- \ J signals with respect to e voltage arameters as specified in

! 3E Specification .4.4.5.2. Pen ng i orporation of the voltage l verification requirements in standard verifications, an j ASME standard calibrated ag3) st he laboratory standard will be

! utilized in Unit I steam nerator spections for consistent voltage normalization.

1. A tube can in in service with a w-like bobbin coil signal tude of less than~ or equal A1.0 volt, r regardi s of the depth of the tube wall phetration, prov Item 3 below is satisfied.

2. tube can remain in service with a flaw-like bobbi 2

signal amplitude greater than 1.0 volt but less tha(coil na i- -

equal to 2.7 volts provided an RPC inspection does not

[ . detect degradation and provided Item 3 below is satisfied.

u

! Q 'f' b JNtOAMENDMENTN0.[

BRAIDWOOD UNITS 1 & 2 3/4 4-17 -BNI-T-2 -AMEN 0 MENT-NO _ _ _ _ _ . _ . _ . _ . _ . _ _ _ _ _ _ _ . _

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REACTOR COOLANT SYSTEM -

SURVEILLANCE REOUIREMENTS (Continued) - -

h '

O.

3. 'The projected end of cycle distribution of crack '

l Indications is verified to result in total ~ primary to l econdary leakage less than 9.I'gpm. (includes operat al l an ccidentleakage). The basis for determinin pected . '

leak r s from the projected crack distribu is provided . CAP-14046, "Braidwood Unit chnical Support.

a for Cycle 5 S Generator Interim P ging Criteria" i dated May.1994.

i j 4. A tube with a flaw-like i coil signal amplitude of i

greater than 2.7 volt all lugged or repaired.

l Certain tubes iden ed in WCAP'14046, "Br

ood Unit 1

Technical Suppo or Cycle 5 Steam Generator rim Plugging

! Criteria," May 1994, shall be excluded from a ication of i the tub upport plate interim plugging criteria limit. t has j bee etermined that these tubes may collapse or deform j lowing a postulated LOCA + SSE.

12) F* DistarLqe is the dist'nce a into the tubesheet from the j secondary face of the tubesheet or the top of the last hardroll, whichever is further into the tubesheet, that has been j determined to be 1.7 inches.

l 13) F* Tube'is a Unit I steam generator tube with degradation below j the F' distance and has no indications of degradation (i.e., no I i indication of cracking) within the F" distance. Defects contained in an F* tube are not dependant on flaw geometry.

l b. The steam generator shall bb determined OPERABLE after completing the j corresponding actions (plug or repair in the affected area all tubes l exceeding the plugging or repair limit) required by Table 4.4-2.

4.4.5.5 Reoorts .

j a. Within 15 days following the completion of each inservice inspection i of steam generator tubes, the number of tubes plugged or repaired in

[ each steam generator shall be reported to the Commission in a Special :

Report pursuant to Specification 6.9.2;

b. The complete results of the steam generator tube inservice inspection l shall be submitted to the Commission in a Special Report pursuant to i Specification 6.9.2 within 12 months following the completion of the

( inspection. This Special Report shall include: l

1) Number and extent of tubes inspected, L 2) Location and percent of wall-thickness penetration for each l indication of.an imperfection, and

)

<. 3) Identification of tubes plugged or repaired.

J l BRAIDWOOD - UNITS 1 & 2 3/4 4-17a AMENDMENTNO.f[

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INSERT D i

! (4.4.5.4.a.11) l 11. For Unit 1, the Tube Suncort Plate Pluacina Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is l experiencing predominantly axially oriented outside l diameter stress corrosion cracking confined within the l thickness of the tube support plates. At tube support l

plate 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 corrosion cracking within the bounds of the 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, I

with degradation attributed to outside diameter stress corrosion cracking within the bounds of the hot-leg tube cupport plate with bobbin voltages less than or equal to 3.0 volts will be allowed to remain in service.

b. Steam generator tubes with degradation attributed to outside diameter stress corrosion cracking within the bounds of the cold-leg tube support plate with a bobbin voltage greater than the lower voltage repair limit (Note 1], will be repaired or plugged, except as noted in 4.4.5.4.11.d below.

c. Steam generator tubes with degradation attributed to outside diameter stress corrosion cracking within the bounds of the hot-leg tube support plate with a bobbin voltage greater than 3.0 volts will be repaired or plugged.

d. Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the cold-leg tube support plate with a bobbin voltage greater than the lower voltage repair limit [ Note 1] but less than or equal to the upper voltage repair limit

[ Note 2], may remain in service if a rotating pancake coil inspection does not detect 1

c 1

. INSERT D 4 (continued) degradation. Steam generator tubes, with indication of outside diameter stress corrosion cracking degradation within the bounds of the cold-leg tube support plate I with a bobbin voltage greater than the upper voltage repair limit (Note 2] will be plugged

, or repaired.

e. Certain intersections as identified in WCAP-14046, Section 4.7, will be excluded from

' application of'the voltage-based repair

criteria as it is determined that these

! intersections may collapse or deform

! following a postulated LOCA + SSE event.

f. If an uuscheduled mid-cycle inspection is

! performed, the following mid-cycle repair

! limits apply instead of the limits identified in 4.4.5.4.11.a, 4.4.5.4.11.b and 4.4.5.4.11.d for outside diameter stress corrosion cracking indications occurring in the steam generator cold-legs. For outside diameter stress corrosion cracking

! indications occurring in the steam generator i hot-legs, the limits in 4.4.5.4.11.a and l 4.4.5.4.11.c apply. The mid-cycle repair "

i

limits are determined from the following equations:

l

~

Vst y n 1.0+NDE+Gr( U)

CL l

Va= Vm -(Vu -Vm) (C U)

Where:

u Vme = upper voltage repair r

limit Vua = lower voltage _ repair limit Vma = mid-cycle upper voltage repair limit based on time into cycle i

2 l I

l l

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L l

INSERT D (continued)

Vmat = mid-cycle lower voltage repair limit based on V,,t and time into cycle At = length of time since last scheduled inspection during which Van and Vat were implemented.

CL = cycle length (the time between two scheduled steam generator t

inspections) l Vo s

= structural limit voltage Gr = average growth rate per '.

cycle length NDE = 95-percent cumulative

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

Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.11.a, 4.4.5.4.11.b, 4.4.5.4.11.c and 4.4.5.4.11.d. l Note 1: The lower voltage repair limit is 1.0 volt for l indications of outside diameter stress corrosion :

cracking occurring at cold-leg tube support plate I intersections.

1 Note 2: The upper voltage repair limit for indications of l i

outside diameter stress corrosion cracking occurring at cold-leg tube support plate intersections is calculated according to the methodology in the May 30, 1995 Frank J. Miraglia memorandum to Edward L. Jordan requesting CRGR review of Generic Letter 95-XX, " Voltage Based

! Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."

3

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REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) -

g.

c. Results of steam generator tube inspections which fall into' Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investi-gations conducted to. determine cause of the tube degradation and corrective measures taken to prevent recurrence.

. or Unit 1 Cycle 5, the results of. inspection for all tubes in '

t e support plate interim plugging criteria limit h en Qhd applie 6.9.2 followin 1 be reported to the Commission pursua letion of the steam gener Specification tube inservice Lb inspection and prior cle 5 operati . he report shall include:

TvSE 1. Listing of the applica

2. Location ~ . cable intersections per and extent of degr ,on (voltage), and

. Projected' Steam Line Break (MSLB) Leakage.

e. , The results of inspections of F* Tubes shall be reported to the Commission prior to the resumption of plant operation. The report shall include:

(

1) Identification of F* Tubes, and k,

2) location and size of the degradation.

. (

BRAIDWOOD - UNITS 1 & 2 3/4 4-17 b AMENDMENTNO.fr3

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INSERT E l

L (4.4.5.5.d) l l d. For implementation of the voltage based repair criteria to tube support plate intersections for Unit 1, notify the staff prior to returning the steam generators to service should any of the following conditions arise:

l 1. If estimated leakage based on the projected l end-of-cycle (or if not practical, using the l- actual measured end-of-cycle) voltage distribution exceeds the leak limit (determined from the licensing basis dose l

calculation for the postulated main steam line break) for the next operating cycle ~.

2. If circumferential crack-like indications are detected at the tube support plate intersections.

(- 3. If indications are identified that extend l beyond the confines of the tube support plate.

4. If indications are identified at the tube support plate elevations that are i attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based.on the projected end-of-cycle (or if not practical, using the actual 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, i

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I REACTOR COOLANT SYSTEM ,

3/4.4.8 SPECIFIC ACTIVITY ,

4 LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant'shall be limited to:

a. Le han or equal to'1 microcurie per gram DOSE EQUIVALENT I-131f,Ch an -

b. Less than or equal to 100/5 microcuries per gram of gross radioactivity.

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

A_GILOR:

MODES 1, 2 and 3*: l

a. With the specific activity of the reactor coolant greater than .

1 I microcurie per gram DOSE EQUIVALENT I-131*for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months l during one continuous time interval or exceeding the limit line j shown on Figure 3.4-1, be in at least HOT STANDBY with T,y, less e than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and ,

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

With T,y, greater than or equal' to 500*F.

- For Unit 1 Cy:1: 5, react'or coolant DOSE EQUIVALENT I-131 will be limited C ) l to 0.35 microcuries per gram. ,

. ly BRAIDWOOD - UNITS 1 & 2 _

3/4 4-27 [ AMENNENT N0. Ji

i' REACTOR COOLANT SYSTEM -

LIMITING CONDITION FOR OPERATION l

ACTION (Continued) ,

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

With the s)ecific activity of the reactor coolant greater than 1 microcurie ,

Oi per gram D)SE EQUIVALENT I-1317or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of l !

Item 4.a) of Table 4.4-4 until the specific activity o'f the reactor coolant is restored to within its limits. C d

SURVEILLANCE I.Juu1REMENTS

?

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 !

Table 4.4-4.

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BRAIDWOOD - UNITS 1 & 2 3/4 4-28 AMENDMENTNO.)0

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! DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY

! * . LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR

% COOLANT SPECIFIC ACTIVITY > 1pci/ GRAM DOSE EQUIVALENT I-131' l l 1 Ec< k h \; ) ec Gee Oc N $9ec M c b .s % bl

(,/Sr cdv-e E{u\c.T-C \ l l BRAIDWOOD - UIHTS 1 & 2 3/4 4-29 ht J$hM ** -

P L__ _ .

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,y,, , i , , , , , , ; OPERATION i , , !

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5 ii l i i e i ! I i i i i i i 150

' ' ' ' ' ' ' ' ' ' ' ' ' ' '\ ' ' ' ' ' ' ' ' '

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2 100 :3lgj ,i

. ACCEPTABLE OPERATION FOR UNIT 2 .

i :\ i i i . i , ! i

j i i \ F UNACCEPTABLE OPERATION FOR UNIT 1 i ei ! i i i e i i , i iN i i 11 iiiiiiii i E f; i i i i \i i I i iiii i i i iie i i ' i \i i: i , , i i d ,

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ERCEli 7 UTED *E MAL COWER

_ _ _ _ _ . . _ . _ . _ _ _ . _ . . _ . . . - _ _ _ - _ _ . _ . _ . . . _ , . _ ._.__ _ . _ ._ _._.._ - ..-._ _ _..._ _ ._._. _ _. _ _ .._ _mm -

Z .,e t=

i TABLE 4.4-4 i

)

REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE -

AND ANALYSIS PROGRAM l

o

TYPE OF MEASUREMENT AND ANALYSIS SAMPLE AND. ANALYSIS FREQUENCY MDOES IN WHICH SAMPLE -

k

1. Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months afb ANALYSIS REQUIRED Determination ** 1, 2, 3, 4 H

} 2. Isotopic Analysis for DOSE EQUIVA-i e' Once per 14 days '

I " LENT I-131 Concentration 1 i 3. .

i Radiochemical for E Determination *** Once per 6 months

i 1

4. Iisotopic Analysis for Iodine l Including I-131, I-133, and I-135 a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , ~

! whenever the specific If, 2f, 30, 4#, 5#

1 activity exceeds 1

1

8 pCf/g. M DOSE

EQUIVALENT I-131 *

l t or'100 4 pCf/ gram g

of gross radioactivity, and .

l

b) One sample between 2 1, 2, ~ 3 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following i a THERMAL POWER change exceeding 15% .

i

of the RATED THERMAL i POWER within.a 1-hour i period. '

e i s, i

i 7

t./

j d i

v i l 4,

9

l .

j .

{ ,

'~

j TABLE 4.4-4 (Continued) ,

^

j TABLE NOTATIONS.

i

Until the specific activity of the Reactor Coolant System is restored 4

within its limits.

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

l **A gross radioactivity analysis shall consist of the quantitative measurement j of the total specific activity of the reactor coolant except for radionuclides

! with half-lives less than 10 minutes and all radiofodines. 'The total ,

specific activity shall be the sum of the degassed beta gamma activity and l

]

the total of all identified gaseous activities in the sample within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months '

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 l radio-nuclides. .

) ***Aradiochemicalana'lysisforIshailconsistofthequantitativemeasurement of the specific activity for each radionuclide, except for radionuclides

with half-lives less than 10 minutes and all radio-iodines, which is identified i in the reactor coolant. The specific activities for these individual

! radionuclides shall be used in the determination of E for the reactor i coolant sample. Determination of the' contributors to I shall be based upon j these energy peaks identifiable with a 95% confidence level.

1 1

UD 6 W a \, rec ccr cuAc 4 @ N N C '1 % b- E j t s3ca w o n %cenes Sec gea . ,

b

=

I 1

l l

L i T.

1 T

4

1. 3RAIDWOOD - UNITS 1 & 2 3/4 4-31 k.u. h ^kms .

1 i REACTOR COOLANT SYSTEM BASES l

i 3/4.4.5 STEAM GENERATORS

The Surveillance Requirements for inspection of the steam generator tubes l ensure that the structural integrity of this portion of the RCS will be main-

) tained. The program for inservice inspection of steam generator tubes is I based on a modification of Regulatory Guide 1.83, Revision 1. Inservice

inspection of steam generator tubing is essential in order to maintain surveil-

] lance of the conditions of the tubes in the event that there is evidence of j mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection l of steam generator tubing also provides a means of characterizing the nature j and cause of any tube degradation so that corrective measures can be taken.

! The plant is expected to be operated in a manner such that the secondary j coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant a chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during i plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System

! (reactor-to-secondary leakage - 150 gallons per day per steam generator).

Cracks having a reactor-to-secondary leakage less than this limit during j operatioh will have an adequate margin of safety to withstand the loads imposed i during normal operation and by postulated accidents. Operating plants have

! demonstrated that reactor-to-secondary leakage of 150 gallons per day per steam

! generator can readily be detected by radiation monitors of steam generator

{ blowdown, mainsteam lines, or the steam jet air ejectors. Leakage in excess of

this limit will require plant shutdown and an unscheduled inspection, during l which the leaking tubes will be located and plugged or repaired by sleeving.

t The technical bases for sleeving are described in the current Westinghouse or

{ Babcock & Wilcox Nuclear Technologies Technical Reports.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it i will be found during scheduled inservice steam generator tube examinations.

j Plugging or sleeving will be required for all tubes with imperfections

( exceeding the plugging or repair limit of 40% of the tube nominal wall 4 thickness, excluding defects that meet the criteria for F* tubes. If a sleeved l l tube is found to contain a through wall penetration in the sleeve of equal to

or greater than 40% of the nominal wall thickness, the tube must be plugged.

The 40% plugging limit for the sleeve is derived from Reg. Guide 1.121 analysis

{ and utilizes a 20% allowance for eddy current uncertainty and additional i degradation growth. Inservice inspection of sleeves is required to ensure RCS I integrity. Sleeve inspection techniques are described in the current

, Westinghouse ur Babcock & Wilcox Nuclear Technologies Technical Reports. Steam

Generator tube and sleeve inspections have demonstrated the capability to i reliably detect degradation that has penetrated 20% of the pressure retaining

portions of the tube or sleeve wall thickness. Commonwealth Edison will

! validate the adequacy of any system that is used for periodic inservice 1 inspection of the sleeves and, as deemed appropriate, will upgrade testing methods as better methods are developed and validated for commercial use.

~

l - :l l

4 .

I BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 AMENDMENT NO. 63

^

i j REACTOR COOLANT SYSTEM

~

{ S

{ %wG

v j 7 / .4.5 ~ STEAM GENERATORS (continued) f L - c n.4+ i r.op' . e ...u._ . .. 4-__4._ ....__ u___... .. ___ ______4-.

4 5 -

.,6 4 ". T.iK 4 . K. ' I'.6:'~.'r'TE'. '!"':"L-"' "4 T."': , G' Z ~ ~J" _' " !I" _a i -

7...'......

~ ":".&.:" D. G.. . ' e:"l'.,..E. . .' .~ :. a. "..'. :". . V. T, '. ". W. . . ' ' ;. &... ' .'. .~ ~. ~ ." .:

4. a., :.7. . ', 7. . ". " ,'. ':.', "

includes the accident leakage fro C in addition to the accident leakage from F on the faulted steam generator 8 the operational leakage limit of Specification 3.4.6.2.c. The. operational leakage limit of Specification 3.4.6.2.c in each of the three, remaining intact steam generators shall include the operational leakage from F For Unit 1, plugging or repair is not required for tubes with degradation

(

l '

within the tubesheet area which fall under the alternate tube plugging criteria )

defined as F . The F" Criteria is based on " Babcock & Wilcox- Nuclear j

l Technologies (BWNT) Topical Report BAW-10196 P."

l i

F* tubes meet the structural integrity requirements with appropriate margins for safety as specified in Regulatory. Guide 1.121 and the ASME Boiler i and Pressure Vessel Code',Section III,. Subsection NB'and Division I Appendices, 3 for normal operating and faulted conditions.

l Whenever the results of any steam generator tubing inservice inspection j fall into' Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resumption of plant operation. Such i

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

,i Specifications, if necessary.

l

\ .

6h M\ A (\% - U d b * '

l -Q - E..A 83 C ck W.,&c. Lue Aec k scow 1

2 1

I

\

j

BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 a AMENbMENTNO.

i------.-.-.-...---.---. - - - _ , . . . - - - - . . . - . . - . . - - - --

l l'

i l

i

! INSERT G f (Bases 3/4.4.5)

The voltage-based repair limits for Unit 1 in Surveillance Requirement (SR) 4.4.5 implement the guidance in the NRC's May 30 Memorandum on voltage based repair criteria for Westinghouse steam generators (SG) with the exception of the specific voltage limit. The May 30 t Memorandum discusses a 1.0 volt Alternate Plugging Criteria (APC). Braidwood SR 4.4.5 implements a 3.0 volt APC for ;

Unit 1 SGs per WCAP-14273, " Technical Support for Alternative Plugging Criteria with Tube Expansion at Tube Support Plate Intersections for Braidwood-1 and Byron-1 l Model D-4 Steam Generators."

l The voltage based repair limits of SR 4.4.5 are l applicable only to Westinghouse-designed SGs with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-TSP intersections. The voltage-based repair limits

.are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG. Additionally, the repair criteria apply only <

to indications where the degradation mechanism is dominantly axial ODSCC with no significant cracks extending outside the thickness of the support plate. Refer to the NRC's May 30 Memorandum on voltage based repair criteria for Westinghouse SG for additional description of the degradation morphology.

Implementation of SR 4.4.5 requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of ,

the voltage repair limit from the structural limit (which is then implemented by this surveillance).

The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 6500F (i.e., the 95-percent LTL curve). The voltage structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty.

1

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

l INSERT G

(continued)

L l The upper voltage repair' limit for cold-leg indications at the tube support plate; Vat , is determined from the structural voltage limit by applying following equation:

Van =Vst-Vor-Vg, l

l where Vca represents the allowance for flaw growth between inspections and Vm, represents the allowance for potential ,

sources of error in the measurement of the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit is contained in the NRC's May 30 Memorandum on voltage based repair criteria for Westinghouse SGs.

The mid-cycle equation in SR 4.4.5.4.ll.f should only be used during unplanned inspections 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 plate remains 3.0 volts during unplanned inspections.

SR 4.4.5.5 implements several reporting requirements recommended by the NRC's May 30 Memorandum for situations which the NRC wants to be notified prior 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 rather than the projected end-of-cycle voltage distribution (refer to the May 30 Memorandum for more information) when it is not practical to complete these calculations using the ;

projected end-of-cycle voltage distributions prior to '

returning the SGs to service. Note that if leakage and conditional burst probability were calculated using the l measured end-of-cycle voltage distribution for the purposes )

of addressing the May 30 Memorandum generic letter 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 the May 30 Memorandum generic letter section i 6.b(c) criteria.

l i

l 2

f

- -. . . . - . ._- .. .. . . . .-- - . . = . _ . . - . . . . . - - .- .-

ATTACHMENT C-2 l MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX'A ,

t- TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES.

NPF-37 AND NPF-66 l l l

l BYRON STATION UNITS 1 & 2

REVISED PAGES l 3/4 4-13*

3/4 4-14 '

3/4 4-15*

3/4 4-16 3/4 4-17 3/4 4-17a l 3/4 4-17b ,

3/4 4-27 ,

3/4 4-28 3/4 4-29 3/4 4-30 i 3/4 4-31 B 3/4 4-3*

B 3/4 4-3a !

NOTE: THESE PAGES HAVE NO CHANGES BUT ARE INCLUDED FOR CONTINUITY.

i I

i

l

)

l INSERT A Not Used INSERT B Not Used INSERT C Not Used

~ ~ - ' ' - - - - - - - -

I

~~ REACTOR'C00LANT'SYSTfM' 3/4.4.5 STEAM GENERATORS t

LIMITING CONDITION FOR OPERATION 1 4 l l

I 3.4.5 Each steam generator shall be OPERABLE.

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

l .

4: ACIl0N: .

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

j SURVEILLANCE REDUIREMENTS I

i 1

j 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by perfomance of

] the following augmented inservice inspection program and the requirements of Specification 4.0.5.

j 4.4.5.1 Steam Generator Samole Selection and Insoection - Each steam generator l j shall be determined OPERABLE during shutdown by selecting and inspecting at l least the minimum number of steam generators specified in Table 4.4-1.

)

i l 4.4.5.2 Steam Generator Tube

Samole Selection and Insoection - The steam l

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 fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be 4

verified acceptable per the acceptance criteria of Specification 4.4.5.4. When applying the expectations of 4.4.5.2.a through 4.4.5.2.c, previous defects or

imperfections in the area repaired by the sleeve are not considered an area

! requiring reinspection. The tubes selected for each inservice inspection shall

! include at least 3% of the total number of tubes in all steam generators; the

! tubes selected for these inspections,shall be selected on a random basis

except

i

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; I l 4

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

i shall include:

t

' *When referring to a steam generator tube, the sleeve shall be considered a j part of the tube if the tube has been repaired per Specification 4.4.5.4.a.10.

4 BYRON - UNITS 1 & 2 3/4 4-13 AMEN 0 MENT NO. 58 I

l , SURVEILLANCEREQUIREMENTS(Continued)

1) All tubes that previously had detectable tube wall penetrations s.

greater than 20% that have not been plugged or sleeved in the ,

- affected area, and all tubes that previously had detectable sleeve i- wall penetrations that have not been plugged, 4

j 2) Tubes in those areas where experience has indicated potential problems,

! 3) At least 3% of the total number' of sleeved tubes in all four steam i generators or all of the sleeved tubes in the generator chosen for i the inspection program, whichever is less. These inspections will l include both the tube and the sleeve, and '

i 4) A tube inspection (pursuant to Sp'ecification 4.4.5.4a.8) shall .be j performed on each selected tube. If any selected tube does not t permit the passage of the eddy current probe for a tube inspection,

! this shall be recorded and an adjacent tube shall be selected and j subjected pection.

L co u A 5) For Unit 1 in service as a result of application of the i fwne 4 g# suoe support plate *pluhin; criteria shall be inspected by bobbin W , '- coil probe during all futu *m -

NGs

6) For Upit 1, tubes which resa n s ce due to the application of i the F criteria will be inspected, in the tubesh all future outages.

j -

! 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 l i tube inspection provided. ;

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

2) The inspections include those parfia__ bes where imperfec-tions were previously found. g3cc gm,# w i

d. For Unit 1, Cycl: 7 implementation of the tube succort plate 4ntec4m-g l tg -a* gl ;;i criteria Mm+t-requires a 10$*55551n coil pr:b; inspection for

! e+ hohgatube support pla ntersections :;d :11 ::ld 1 ; int:r;;;

l t'::: dow~n to the lowest col tube support plate with outeQiaggiar j cA oJA-\%~ stress corrosion cracking (O C) indications. The determination of theS j p ube t support plate intersections having ODSCC indications shall be based

on the performance of at least a gilrandomsamplingoftubesinspe l g e o ctM - % over their full length. gem Nh we4q

e. A random sample of at least 20 percent of the total number of sleevesu

shall be inspected for axial and circumferential indications at the end i of each cycle. In the event that an imperfection of 40 percent or l

! greater depth is detected, an additional' 20 percent of the unsampled ;

l sleeves shall be inspected, and if an imperfection of 40 percent or l greater depth is detected in the second sample, all remaining sleeves l

i shall be inspected. These inservice inspections will include the entire

sleeve and the tube at the heat treated area. The inservice inspection i for the sleeves is required until the corrosion resistance for the laser l welded or kinetically welded joints in tubes that bound the material

/

i BYRON - UNITS 1 & 2 3/4 4-14 AMENDMENT NO./2 1

_.__.____._________.___...___.~_....._.;.____..-___._._,-

! REACTOR COOLANT SYSTEM

. SURVEILLANCE RE0VIREMENTS (Continued) l parameters of the tubes installed in the steam generators has been

. demonstrated acceptable. If conformance with the acceptable criteria of

Specification 4.4.5.4 for tube structural integrity is not confirmed, 2

the tubes containing the sleeves in question shall be removed from j service.

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

following three categories

Category Insoection Results t

'C-1 Less than 5% of tlm 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 significani. (greater than 10% of wall thickness) further wall penetrations to be included in the above percentage calculations.

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

b. If the results of 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

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) 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 NO. 72 l

i REACTOR COOLANT SYSTEM l -

l SURVEILLANCE REOUIREMENTS (Continued)

I

! 2) A seismic occurrence greater than the Operating Basis Earthquake, 4

! or ,

3) . A Condition IV loss-of-coolant accident requiring actuation of the Engineored Safety Features, or

~

4) A Condition IV main steam line or feedwater line break.

4.4.5.4 Acceptance Criteria

a. As used in this specification: ,

1) Imperfection 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 or outside of a tube or slee've;

3) Deoraded Tube means a tube or sleeve containing unrepaired imperfections greater than or equal to 20% of the nominal tube or sleeve wall thickness caused by degradation;

4) % 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 tube or sleeve containing an

. unrepaired defect is defective; ,

6) Pluaoina or Reoair Limit means the. imperfection depth at or beyond l which the tube shall be removed from service by plugging or l repaired by sleeving in the affected area. The plugging or repair limit imperfection depth is equal to 40% of the nominal wall i

thickness. . For Unit 1, this definition does not appl o in the tubesheet that meet the criteria for an F* tu 1 defectsr ("

9.w 1 For Unit 1 Cy:': 7, this definition does not appl uBe support I plate intersections for which the voltage-based criteria are being applied. Refer to 4.4.5.4.a.ll 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 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;

8) Tube Insoection means an inspecticn 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 portion of the tube, and BYRON - UNITS 1 & 2. 3/4 4-16 AMENDMENT NO.

[ REACTOR COOLANT SYSTEM l

SURVEILLANCE REQUIREMENTS (Continued) i j 9) Prese'rvice Insnection means an inspection of the full length of -

F each tube in each steam generator perfomed by addy current

! techniques prior to service to establish a baseline condition of l the tubing. This inspection shall be performed prior to initial.

POWER OPERATION using the equipment.and techniques expected to, i be used during subsequent inservice inspections.

t i 10) Tube Reoair refers to a process that reestablishes tube

- serviceability. Acceptable tube repairs will be performed by j the following processes: -

i a) Laser welded. sleeving as described in a Westinghouse j Technical Report currently approved by the NRC, subject to

the limitations and restrictions as noted by the NRC staff, l

or i b) Kinetic welded sleeving as described in a Babcock & Wilcox

} Nuclear Technologies Technical Report currently approved by

the NRC, subject to the limitations and restrictions as

] noted by the NRC staff.

i i Tube repair includes the removal.of plugs that were previously j 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 service.

i i 1 For Unit 1 Cycle 7, the Tube Sunnart Plate Interim Pluaaina iteria Limit is.used for the disposition of a steam generato tu for continued service that is experiencing outer diane er

! ,stres corrosion cracking confined within the thickness the

l. tube su rt plates. At tube support plate intersecti s, the L._ repair lim is based on maintaining steam generator ube l1

!(

h A4(cn. serviceabili as described below: i

\

$ a) Degradation a ributed to outside diamet r stress corrosion i g igt M) cracking within a bounds of the tu support plate with bobbin voltage les than or equal 1.0 volt will be allowed to remain in rvice.

l -

b) Degradation attributed to side diameter stress corrosion j cracking within the bou o he tube support plate with l bobbin voltage greate han 1. olt will be repaired or i plugged except as ed in 4.4.5. . 11)c) below.

i i c) Indications o otential degradation a t buted to outside j diameter st ss corrosion cracking withi e bounds of the

{ tube su rt plate with a bobbin voltage gre ter than 1.0 l volt less than or equal to 2.7 volts may remain in t ser eifarotatingpancakecoi1~inspectiondoe%erot l ect degradation. Indications of outside diamet corrosion cracking degradation with bobbin voltage gre(tress ater than 2.7 volts will be plugged or repaired.

j if -

BYRON - UNITS 1 & 2 3/4 4-17 AMENDMENTNO.Ji6

~ ~~ '~

i REACTOR COOLANT SYSTEM

- SURVEILLANCE REQUIREMENTS (Continued) '

i- h I d) Certain intersections as identified in WCAP-14046, Section

4.7, will be excluded from application of the voltage-based ,

repair criteria as it is detemined that these intersecti may collapse or deform following a postulated LOCA+SSE ent.

I e) , as a result of leakage due.to a mechanism oth than i at the tube support plate intersection, some other cause, unscheduled mid-cycle insp~ection i performed, the i followi epair criteria apply'instead o .4.5.4.11)c). If j bobbin volt is within expected limi , the indication can i remain in serv . The expected bo n voltage limits'are j determined from following eq ion: '

AU +V' i g

V< " (V -V 1+(0.2 U) l where:

1- V '- measured voltage

V voltage at BOC l A = time period of operation to unschedule tage

! = cycle length (full operating cycle length re

! operating cycle is the time between two i scheduled steam generator inspections) l Vu - 4.5 volts l

12) F* Distance is the distance into the tubesheet from the secondary face of the tubesheet or the top of the last

~

( '

hardroll, whichever is' further into the tubesheet, that has

. been detemined to be 1.7 inches. ,

l 13) F* Tube is a Unit I steam generator tube with degradation below )

indication of cracking) within the F* distance. Defects the F' distance and has no indications o contained in an F* tube are not dependant on flaw geometry.

h b. The steam generator shall be determined OPERABLE after completing j the corresponding actions (plug or repair in the affected area all l tubes exceeding the plugging or repair limit) required by Table i 4.4-2.

4.4.5.5 Reports

! a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in j' each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; i b. The complete results of the steam generator tube inservice

inspection shall be submitted to the Commission in a Special Report j-

)- /

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

t.

l i

l l INSERT D V

(4.4.5.4.a.11)

11. For Unit 1, the 1X11xa Succort Plate Pluccina Limit is used for the disposition of an alloy 600 steam generator tube for continued ~ service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the ,

thickness of the-tube support plates. At tube support ,

. plate intersections, the plugging (repair) limit is based on maintaining steam generator tube

! serviceability as described below:

a. Steam generator tubes, with degradation I attributed to outside diameter stress corrosion cracking within the bounds of the 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.

b. Steam generator tubes with degradation attributed to outside diameter stress corrosion cracking within the bounds of the cold-leg tube support plate with a bobbin voltage greater than the lower voltage repair limit [ Note 1], will be repaired or plugged, except as noted in 4.4.5.4.ll.d below. ,

c. Steam generator tubes with degradation attributed to outside diameter stress corrosion cracking within the bounds of the hot-leg tube support plate with a bobbin voltage greater than 3.0 volts will be repaired or plugged.

d. Steam generator tubes, with indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the cold-leg tube support plate with a bobbin voltage greater than the lower voltage repair limit [ Note 1] but lass than or equal to the upper voltage repair limit

[ Note 2], may remain in service if a rotating pancake coil inspection does not detect 1

F i INSERT D (continued)

I degradation. Steam generator tubes, with 4 indication of outside diameter stress ;

i- corrosion cracking degradation within the :

bounds of the cold-leg tubo support plate with a bobbin voltage greater than the upper i voltage repair limit-[ Note 2] will be plugged i

or repaired. [

a ?

i ,

e. Certain intersections as identified in WCAP- ;

j 14046, Section 4.7, will be excluded from l f application of the voltage-based repair +

1.

criteria as it is determined that these intersections may collapse or deform

! following a postulated LOCA + SSE event. ,

f !

f- f. If an unscheduled mid-cycle inspection is

performed, the following mid-cycle repair

! limits apply instead of the limits identified in 4.4.5.4.11.a, 4.4.5.4.11.b and j 4.4.5.4.11.d for outside diameter stress I corrosion cracking indications occurring in the steam generator cold-legs. For outside diameter stress corrosion cracking indications occurring in the steam generator !'

p hot-legs, the limits in 4.4.5.4.11.a and r 4.4.5.4.11.c apply. The mid-cycle repair L limits are determined from the following l equations:

V" Va= _

1.0+NDB+Gr( )

CL L

I Va= Vm - ( Vun~ Vm)I CL- g At) l l

I Where:

Vmm = upper voltage repair

.- limit l

l- Vua = lower voltage repair l- limit l Vmma = mid-cycle upper' voltage E repair limit based on time into cycle

.2 l

l-i.

l., 4..-...-,,,,, . ..-.,_ , m , ,...._...m..,. _ , . . _ - . . , . _ , , , _ , _ _ . , - , , _ . . . . - . , , , . , . ,

0 L

L INSERT D (continued) ,

Vmat = mid-cycle lower. voltage ,

o repair limit based on :

L V ,mt and time into cycle !

At = length of time since last '

scheduled inspection during which Vum, and V tu, ;

were implemented. .

CL = cycle length (the time' between two scheduled steam generator inspections) l Vi = structural limit voltage Gr. = average growth rate per i cycle length -

NDE =. 95-percent cumulative l probability allowance for !

nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by NRC) :

Implementation of these mid-cycle repair limits should follow the same approach as in TS 4.4.5.4.11.a, 4.4.5.4.11.b, 4.4.5.4.11.c and 4.4.5.4.11.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 the May 30, 1995 Frank J. Miraglia memorandum to Edward L. Jordan requesting CRGR review of Generic Letter 95-XX, " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."

3

_ . . . _ _ . , _ . _ . .~. . _ . . _ _ . _ . _ _ _ , , _ _ . _ . . .

REACTOR COOLANT SYSTEM :x SURVEILLANCE REQUIREMENTS (Continued) ,

pursuant to Specification 6.9.2 within 12 months foilowing the 3 completion of the inspection. This Special Report shall include: ,

1) Number and extent of tubes inspected,

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

3) Identification of tubes plugged or repaired.

c. Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant' to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

d. 7qr Unit 1 Cycle 7, implementation of the voltage-based repair crlteria to tube support plate intersections, reports to the Staff shall'he made as follows:

1) Not e Staff prior-to returning the steam generator to service ld any of the following conditions aris .

a) If estimated leakage based on the actual asured end-of-cycle voltag stribution would have ceeded the leak c f.Cd g limit (for post ted main steam li break utilizing L ^g' licensing basis as tions) dur the previous operation cycle.

b) If circumferential crac indications are detected at

, the tube support plat nters ions. ,

c) If indications identified.that tend beyond the confines of tube support plate.

d) If the f culated conditional burst probabhi{y exceeds 1X safe)ty significance of the occurrence.0' , notify the NRC and pro 2)

/

e final results of the inspection and the tube integrity evaluation shall be reported to the Staff pursuant to Specification 6.9.2 witMn 90 days following restart.

e. The results of inspections of F* Tubes shall be reported to the Comission prior to the resumption of plant operation. The report ( ,

shall include:

Identification of F*~ Tubes, and 1)

(

2) Location and size of the degradation.

t BYRON - UNITS 1 & 2. 3/4 4-17b AMENDMENT NO. J2 L___________-___-. .__ -- --.- -.. - _ _ - . - . . . . - . . - - - . - _ . . - - .

. - - - - . _ . -. .. - - . - . . - - = _ _ - . _

(-

L INSERT E (4.4.5.5.d) i

d. For implementation of the_ voltage based repair l criteria to tube support plate intersections for ~

Unit 1, notify the staff prior to returning the steam generators to service should any of the .

following conditions arise: j 1.. If estimated leakage based on the projected end-of-cycle (or if not practical, using the ;

actual measured end-of-cycle) voltage

! distribution exceeds the leak limit !

(determined from the licensing basis dose I

calculation for the postulated main steam line break) for the next operating cycle. ;

2. If circumferential crack-like indications are i detected at the tube support plate l intersections.

3. If indications ~are identified that-extend )

beyond the confines of the tube support 1 plate.

4. If indications are identified at the tube support plate elevations that are i attributable to primary water stress corrosion cracking.

5. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual 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.

I i

I

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

J REACTOR COOLANT SYSTEM s-3/4.4.8 SPECIFIC ACTIVITY l LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to: ,

a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-13I , b and

b. Less than or equal to 1004 microcuries per gram of gross radioactivity.

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

ACTION:

MODES 1, 2 and 3*: -

With the specific activity of the reactor coolant areater than

$/ CS

a. '

1 microcurie per gram DOSE EQUIVALENT I-131*for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months l during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T'V9 less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and

b. With the specific activity of the reactor coolant greater than 100 6 (

microcuries per gram, be in at least HOT STANDBY with Tavg less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

/

With I greater than or equal to 500*F.

E5 \ re <%- CcdoG I' bN g o 3 ci,c Cu Ae . }e e Q + l BYRON - UNITS 1 & 2 3/4 4-27 AMENDMENT NO. I I

l l

,,1

REACTOR COOLANT SYSTEM 4-LIMITING CONDITION FOR OPERATION ACTION (Continued)

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

With the specific activitv of the reactor coolant greater than 1 microcurie K.~ per gram DOSE-EQUIVALENT I-13Por greater than 100/E microcuries per gram, l ;

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. ,

Q J SURVEILLANCE REQUIREMENTS '

'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 1 Table 4.4-4.

8 P

e s T- l. b \, recoc- cAa ht"X (%LMan 7-R\ u..W L khd

~

s e 3 s c.s.u h . 3c- n mcc.. .

)

. /

BYRON - UNITS 1 & 2 3/4 4-28 AMENDMENT NO. I

iis i l . i , e i

T ,

, i e i I

( l i

L i i/

g \ l iA L' ,i ,

1

= 1 j' t i 1-- \ \1 / l I g \ 1 1

/ i

_ < 4 - i

.J 3 .\ \ , i

.8 g N 'L -

i Meg \ed 5: s i /

}' .

j g 200 x* N / ' '

-C 4 '

\

UNACCE ABLE '

TilW O s \ OPER. ION j

!f: N ( / ;

, o o i /

1 W s -

T ,

C i i

> b i

( \s /

l 8 s f , ,

z 150 x s f ,

j s T / i o 'y f \T l O/ /

A s ( *

>= f g g i E / \ g i 4 / s \ 6 6 a,

100

/

/ s \ ,

s % i6 i i

, e s T I i m 1 / s X i i 7 / sh i i

/ -n T i i

{ '

'z AC9A!PTABLE \

- $ OPERATION s , i ,

< 50 ' ' ' ' '

\

} } /

s l 1 J g /

\ l I j i w \ '

i j g / x i '

s

)/

, +

v>

j o

g /i 4 6 6N i

O 20 30 40 50 60 70 80 90 1

PERCENT OF RATED THERMAL POWER 1

i Figure 3.4-1 Dose Equivalent 1-131 Reactor Coolant Specific Activity Limit Versus Percent of Rated Thermal Power with the i Reactor Coolant S

! - Equivalent 1-131* pecific Activity >1pCi/ Gram Dose ,,

f',* _

We (kw (3 hc ~

O,hS ec1.hawaf' 9 o.%gC,/Gec,mhc bg cbBh l BYRON ' UNITS 1 & 2 3/4 4-29 ht h htCT kc, a , _ -- - . _ . _. _ _ _ . _ _ _ . _ _ . . . . . . - . . _ _ _ . . _ -- _ _. ,._. . _ ,. - .---- _ . - .. . - -..-.. ,.,. ..__.

INSERT F '

s (Figure 3.4-1) i i \ l l l l I i i ! I i , t- : , i i l i ; I l 1 i . ,

i . i3,i ! i i,iiiiii- i i i . , i i i i i i . . , , i i e i i\ . ! ii: i i , i

i ! . i i * , i i i ' . - i i i i ii\ i 4 ;,ii , i i -

, + i . i i

' I ' ' ' ' '

250 i iii ^i .,'''''' i i i i ,

i .

I ' '

O i! l e IL i i I 6 i ) l I e i i . i i g i i i . i iN i i i , i , i i .;i ,

i , , . i , , , ,

x 1 ; i i i i i \i i i i i i ,

i i !- i i i i i i , -

O . I i i i i\.i,i i i i ie i . i i i ; I ! - .

j 3 i I i l i i i\ i l t 4 l i l l I

, i ! l l I i i i I \iiiiiiiiil i . i . i I , , . i ,

j i ! I i l l : ! i\ l i l I ' t ' i i i 4

i 4 i i i i i i i 'l 4 i i i i ii i e i i '

i i e i i i e i i i i i a i i i , , j iUNACCEPTABLEl' , ;e g 200 , , , , , , , , , ,;,y,,,3 3 , , , 3 i j OPERATION ; , ,

i j i i i i i i i i 6 i i i iX l t i 1 . i i , ! i e i i i i i i t i e i o i i i i i i i iiiii! \it i i i i i ' I I i i i ! i i i i i 4 ' . i I ' ' I i I l l i i i l i i e i i 'N i ! i i ! l i I i l u i 6 i ! i i i i i i i , i \i i ,

l I i i e i i i i i i i i

, s t i i i i i i i i i i I i .

11I i . i ! i ' i i i i i ! I i i i j i i i i iie i i i i i i i !\1 : I e i i ! I i i i i i i 1 t i i i i i i . t i e i i i \l I i i i i i i e iii ,

W i ; i i i i i i i i e i i i e i\l l i e i iI . 1 I i i i ' !

- ' I ' ' ' ' ' ' ' ' ' ' ' ' I' ' ' ' ' ' ' ! ' ' ' ' '

150 E i ! I i i i t i ! ! ! I i i i i i i t i i i l i i . i i 3 i i i i i ! I i i i i i i l \ l l i i li i i t i i i O i ii!l i i i i i i i -

i ii i \ i e i ii i i i i i i 8 '

V i l i i i i i ! : , I ( ! l l l l \il l l l l t i i i l i * '

> i i ieii! i i t i i i i ie i i iN . i i i i i e i i i i i

$ \i i t i i i e i i i 4 i e i ii i ! I \l i i i i i i ! , i ! I y l X,iiiii,t i ! ! i i t I t i i Xie i i i e i i i y*

i\i i ! I i ! i i i l l 6 i i l i i i\i ! I i i i i i i i +

i \i! i j i i i i i i ! i i t i i i i i \1 i i i ! I i i i ! i i

<\ii, i\ i i i . i i , i 2 100 , , ,g i , ACCEPTA8LE OPEPATION FOR UNil' 2 -

6 A i i . i i , i, ,

1 l i iXi UNACCEPTABLE OPERATION FOR UNIT 1 i t\ ! i i ! i i i i

i i i iN i i\ i i i i i i i i i

3 ! i ! i \i i i i i t !ii i i i i i i e i i\i i i i . ! i j i i i i\! i i i i i i ! I i i ! i i i i\ l i i I ! I i g i i i a i i\! i i i i i i i i i i I iii T i i i a i i !

\ l i i i i i ii' o

5 !

i i e i !\ i

! i i i i i i i e i i .

I i I i Ni UNIT 2 iM -- -

" j i i i i i i :\ } i ! i i i ,

I i i i l I l 6 i i I q p .l i i i i ii\ii! i I i i i i i i i e i i i I i i i i '

h *j l i i l '

skiiiiil ! ' fI i I I ! ! I i i i i i

.: i i i i , i i \! i i ! i ; i;ii. i i i i i i i i i -

! . i i i i i\ii,I e i i i i t i i i i i i ! I i i i i ACCEPTABLE i . :N i e i i ! i i ! l I i ! ' i ' i ' ' '

CPEGAilCN i i \l 1 ! i i i i i i ' i i i I I I i '

' ' UNIT 1 . '/i" i I iN

, j , i i i i i i l I i l i i i i t I i i it i . , iie i i i i i i i i ! I l 6 i i . I i 1 i i i i i

. . , i ! e i i i I i i e i ! e i l i i ' ' i l I I i l i ! i * '

ll t i > i iiiiiI e i , i i i i l. ) i i i i i l I l ! i a a i i '

0q 20 30 A0 50 60 70 80 30 CERCENT y :ATED mERVAL COWER

TABLE 4.4-4 m

5

~

REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE E AND ANALYSIS PROGRAM c TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN milch SAMPLE .

AND ANALYSIS AND ANALYSIS REQUIRED FREQUENCY

1. Gross Radioactivity At'least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I, 2, 3, 4

, Determination **

1 m 2. Isotopic Analysis for DOSE EQUIVA- Once per 14 days 1 LENT I-131 Concentration ,

3. Radiochemical for i Determination *** Once per 6 months

1

4. Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 1#, 2#, 3#, 4#, 5#

Including I-131, I-133, and I-135 whenever the specific

, activity exceeds 1 .

pCl/ gram DOSE.

to EQUIVALENT I-13 [

D or 100/E pC1/ gram

+ of gross radioactivity, g and b) One sample between 2 1, 2, 3.

! and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15%

of the RATED THERMAL POWER within a 1-hour y period.

i &

l'e '

a b

r o

et-

}*

o

J .

j

^

TABLE 4.4-4 (Continued)

TABLE NOTATIONS -

1 #Until the specific activity of'the Reactor Coolant System is restored j within its limits.

j

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

} **A gross radioactivity analysis shall consist of the quantitative measurement l

] of the total specific activity of.the reactor coolant except for radionuclides 1

with half-lives less than 10 minutes and all radiciodines. The total i j specific activity shall be the sum of the degassed beta gamma activity arid )

j the total of all identified gaseous activities in.the sample within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months j 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 wi.th a 95% confidence i level. The latest available data may be used for pure beta-emitting j radio-nuclides.

i ***A radiochemical analysis for I shall consist.of the quantitative measurement

of the specific. activity for each radionuclide, except for radionuclides j with half-lives less than 10 minutes and all radio-iodines, which is identified 1 i in the reactor coolant. The specific activities for these individual i radionuclides shall be used in the determination of I for the reactor i

l coolant sample. Determination of the contributors to I shall be based upon l these energy peaks identifiable with a 95% confidence level.

i

%%d E bh \ , reaw - cu\m' D2 bMLEU MD b\ kot.

j b.w %h % 0.% Ntrc Cdr. Se\ e A cam .

i j e i

l l~

i i

i f

i BYRON - UNITS 1 & 2 3/4 4-31 D*06 "

i

REACTOR COOLANT SYSTEM s

BASES

~

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection I of steam generator tubing is essential in order to maintain surveillance of the conditions of tbc iuoes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

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

negligible corrosion of the steam generator tubes. If the secondary coolant l chemistry is not mair.:ained within these 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 Peactor Coolant System and the Secondary Coolant System (reactor-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 day per steam generator can readily be detected by radiation monitors of steam generator blowdown, mainsteam lines, or the steam jet air ejecters. Leakage 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 technical bases for sleeving are described in the current Westinghouse or Babcock & Wilcox Nuclear Technologies Technical Reports.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect 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 plugging or repair limit of 40% of the tube nominal wall

> thickness, excluding defects that meet the criteria for F' tubes. If a sleeved l tube is found to contain a through wall penetration in the sleeve of equal to or greater than 40% of the nominal wall thickness, the tube must be plugged.

The 40% plugging limit for the sleeve is derived from Reg. Guide 1.121 analysis and utilizes 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 Babcok & Wilcox Nuclear Technologies Technical Reports. Steam Generator tube and sleeve inspections have demonstrated the capability to reliably detect degradation that has penetrated 20% of the pressure retaining portions of the tube or sleeve wall thickness. Commonwealth Edison will validate the adequacy of any system that is used for periodic inservice

- inspection of the sleeves and, as deemed appropriate, will upgrade testing methods as better methods are developed and validated for commercial use.

BYRON - UNITS 1 & 2 B 3/4 4-3 AMENDMENT NO. 72

REACTOR COOLANT SYSTEM BASES ',

3 /4. 4. 5 STEAM GENERATORS (Continued)'

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AEl'3(4 i r Unit 1 Cycl: 7, t h:: ::ptriin:ir.; ::t:r di---t:r :tr::: ::rr:: ion-Or : king within th: thicia::: Of th: t h; ;;;p;rt pl:t:: will L: di:p::itica:d in ::: rd:::: with S;;;ift:: tic; t.t.5.t.i.11. Th: :---' tin; p ried ::y 5:

dje:t:d t: 1::: th:: th: fell :;;r: tin :y:1:'t: :: maximum site allowable primary-to-secondary leakage Iimit for End ycle Main Steam Line Brep.onditionss' The 1 :h ;: 'i=it,12.2 ;;1/ includes the accident leakage -

frothfPC in addition to the accident leakage from F on the faulted steam generltor and the operational leakage limit of Specification 3.4.6.2.c. The operational leakage limit of Specification 3.4.6.2.c in each of the three remaining intact steam generators shall include the operational leakage from F.

For Unit 1, plugging or repair is not required for tubes with degradation within the tybesheet area which fall under the alternate tube plugging criteria defined as F . The F* Criteria is based on " Babcock & Wilcox Nuclear Technologies (BWNT) Topical Report BAW-10196 P."

F* tubes meet the structural integrity requirements with appropriate margins for safety as specified in Regulatory Guide 1.121 and the ASME Soiler 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 pur- I suant 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

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result in a requirement for analysis, laboratory examinations, tests,,

additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

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1 BYRON - UNITS 1 & 2 8 3/4 4-3a AMEADMENT NO.,J2 1 :

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l INSERT G

! (Bases 3/4.4.5) l The voltage-based repair limits for Unit 1 in

! Surveillance Requirement (SR) 4.4.5 implement the guidance l in the NRC's May 30 Memorandum on voltage based repair l criteria for Westinghouse steam generators (SG) with the exception of the specific voltage limit. The May 30 l

Memorandum discusses a 1.0 volt Alternate Plugging Criteria

! (APC). Braidwood SR 4.4.5 implements a 3.0 volt APC for Unit 1 SGs per WCAP-14273, " Technical Support for Alternative Plugging Criteria with Tube Expansion at Tube Support Plate Intersections for Braidwood-l and Byron-1 Model D-4 Steam Generators."

The voltage based repair limits of SR 4.4.5 are applicable only to Westinghouse-designed SGs with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-TSP intersections. The voltage-based repair limite are not applicable to other forms of SG tube degradation nor are they applicable to ODSCC that occurs at other locations within the SG. Additionally, the repair criteria apply only i to indications where the degradation mechanism is dominantly axial ODSCC with no significant cracks extending outside the ,

thickness of the support plate. Refer to the NRC's May 30 i Memorandum on voltage based repair criteria for Westinghouse l SG for additional description of the degradation morphology.

Implementation of SR 4.4.5 requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).

The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 650 F (i.e., the 95-percent LTL curve). The voltage structural limit must be adjusted downward to account for potential flaw growth during an operating interval and to account for NDE uncertainty.

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INSERT G (continued)

The upper voltage repair limit for cold-leg indications at the tube support plate; Van, is determined from the

! structural voltage limit by applying following equation:

Vuat=Vst-Va,-Ves where Van represents the allowance for flaw growth between inspections and Vos represents the allowance for potential sources of error in the measurement of the bobbin coil l voltage. Further discussion of the assumptions necessary to

! determine the voltage repair limit is contained in the NRC's May 30 Memorandum on voltage based repsir criteria for Westinghouse SGs.

The mid-cycle equation in SR 4.4.5.4.11.f should only be used during unplanned inspections in which eddy current i data is acquired for indications at the cold-leg tube support plates. The voltage repair limit for indications at the hot-leg tube support plate remains 3.0 volts during unplanned inspections.

SR 4.4.5.5 implements several reporting requirements recommended by the NRC's May 30 Memorandum for situations which the NRC wants to be notified prior to returning the i SGs to service. For the purposes of this reporting l requirement, leakage and conditional burst probability can l be calculated based on the as-found voltage distribution rather than the projected end-of-cycle voltage distribution (refer to the May 30 Memorandum for more information) when ,

it is not practical to complete these calculations using the projected end-of-cycle voltage distributions prior to returning the SGs to service. Note that if leakage and conditional burst probability were calculated using the measured end-of-cycle voltage distribution for the purposes of addressing the May 30 Memorandum generic letter 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 the May 30 Memorandum generic letter section ;

6.b(c) criteria.

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l ATTACHMENT D EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 l

Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards L considerations. According to Title 10 Code of Federal Regulations Section 50 Subsection 92 Paragraph c (10_CFR 50.92 (c)), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

A. INTRODUCTION Commonwealth Edison (Comed) proposes to amend Byron and Braidwood Technical Specification (TS) 3.4.5, " Steam Generators," the bases for TS 3.4.5, and TS 3.4.8, " Specific Activity."

The changes proposed to TS 3.4.5 will implement an increased voltage, bobbin coil probe, Steam Generator (SG)

Tube Support Plate (TSP) Alternate Plugging Criteria (APC) limit for Outside Diameter Stress Corrosion Cracking ;

(ODSCC) indications at the hot-leg TSP intersections.

The changes proposed to TS 3.4.8 involve reducing Reactor Coolant System (RCS) dose equivalent Iodine-131 (I-131) for Unit 1 at both Byron and Braidwood. .

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For Braidwood and Byron, additional changes are proposed to make TS more consistent with the Model TS contained in the May 30, 1995 Frank J. Miraglia memorandum to Edward L.

Jordan requesting CRGR review of Generic Letter 95-XX,

" Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" (May 30 Memorandum).

For Byron Unit 1 and Braidwood Unit 1, Comed is requesting implementation of an APC for ODSCC indications at hot-leg TSP intersections. This APC will increase the current plugging criteria voltage for ODSCC occurring at hot-leg TSP intersections up to a maximum of 3.0 volts. Selected SG tubes will be expanded above and below the hot-leg TSP to limit TSP movement during a Main Steam Line Break (MSLB) to reduce tube burst probabilities to negligible levels.

For Byron, the equation for determining voltage acceptance criteria for an unplanned outage is revised for conformance to the May 30 Memorandum. For Braidwood, the equation for mid-cycle unplanned outage voltage acceptance criteria is added to the specification for conformance with the May 30 Memorandum.

Braidwood's probability of tube burst limit is decreased from 2. 5x10-2 to 1. 0x10-2 consistent with the May 30 Memorandum.

Byron and Braidwood will be adding footnotes to TS 3.4.8 and revising Figure 3.4-1 to reduce the Unit 1 dose equivalent I-131 limit from 1.0 microcurie per gram (pci/gm) to 0.35 pci/gm.

Finally, bases changes are being made to Braidwood and Byron TS in order to accurately reflect the changes made to the individual specifications.

B. NO SIGNIFICANT HAZARDS ANALYSIS

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Tube burst is precluded during normal operating plant conditions since the tube support plates are adjacent to the degraded regions of the tube in the tube to tube support plate crevices. During accident conditions, ie., MSLB, relative tube to TSP movement may occur, which can expose a crack length portion to freespan conditions. Testing has shown that the burst pressure correlates to the crack length that is exposed to the freespan, regardless of the length that is still contained within the TSP bounds. Therefore, a 2

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more appropriate methodology has been established for addressing leakage and burst considerations that is based on limiting potential TSP displacements during postulated MSLB events, thus reducing the freespan exposed crack length to minimal levels. The tube expansion process to be employed in conjunction with this TS change is designed to provide postulated TSP displacements that result in negligible tube burst probabilities due to the minimal freespan exposed crack lengths.

l A thermal hydraulic model was developed to determine TSP

loading during MSLB conditions. A safety factor of 2 was l conservatively applied to these loads to envelope the collective uncertainties in the analyses. Various operating conditions were evaluated and the most limiting operating condition was used in the analyses. An additional independent model was used to verify the thermal hydraulic results.

The tube burst probability assessnent used a conservative assumption that all hot-leg TSP intersections (32,046) contained throughwall cracks equal to the postulated displacement and that the crack lengths were located within the TSP edge. Alternatively, it was assumed that all hot-leg TSP intersections contained throughwall cracks with length equal to the thickness of the TSP. The postulated TSP motion was conservatively assumed to be uniform and equal to the maximum displacement calculated. The total burst probability for all 32,046 throughwall indications given a uniform MSLB TSP displacement of 0.31" is calculated to be lx10-5 This is a factor of 1000 less than the May 30 Memorandum burst probability limit of lx10-2 Therefore, the functional design criteria for tube expansion is to limit the TSP motion to 0.31" or less. However, the design goal for tube expansion limits the TSP MSLB motion to less than 0.1", which results in a total tube burst probability of 1x10-2 for all 32,046 postulated throughwall indications.

Additional tubes will be expanded to provide redundancy should the required expansions fail. This redundancy ensures that the maximum TSP displacement is limited to 0.31" to meet the functional design criteria.

l The structural limit for the hot-leg SG tube repair criteria with tube expansion is based on axial tensile loading requirements to preclude axial tensile severing of the tube.

Axially oriented ODSCC does not significantly impact the axial tensile loading of the tube, therefore, the more limiting degradation mode with respect to affecting the tube structural limit at TSPs is cellular corrosion. Tensile tests that measure the force required to sever a tube with cellular corrosion and uncorroded cross sectional areas are 3

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used to establish the lower bound structural limit. Based upon these tests, a lower bound 95% confidence level structural voltage limit of 37 volts was established for cellular corrosion. This limit meets the Regulatory Guide (RG) 1.121, " Basis for Plugging Steam Generator Tubes,"

structural requirements based upon the normal operating pressure differential with a safety factor of 3.0 applied.

Due to the limited database supporting this value, the structural limit was conservatively reduced to 20 volts.

Accounting for voltage growth and Non Destructive l Examination (NDE) uncertainty, the full APC upper limit j exceeds 10 volts.

However, for added conservatism a single voltage repair limit for hot-leg indications is specified in this request.

All hot-leg indications with bobbin coil probe voltages 1 greater than the hot-leg voltage repair limit will be I plugged or repaired.

The freespan tube burst probability must be calculated for j the cold-leg TSP indications to be within the requirements )

of the May 30 Memorandum. The freespan structural voltage limit is calculated using correlations from the database described in the May 30 Memorandum, with the inclusion of the recent Byron and Braidwood tube pull results. This structural limit is 4.75 volts. The lower voltage repair !

limit for cold-leg indications continues to be 1.0 volt.

The upper voltage repair limit for cold-leg indications will be calculated in accordance with the May 30 Memorandum.

Per the May 30 Memorandum, MSLB leak rate and tube burst probability analyses are required prior to returning to power and are to be included in a report to the Nuclear Regulatory Commission (NRC) within 90 days of restart. If allowable limits on leak rates and burst probability are exceeded, the results ar; to be reported to the NRC and a safety assessment of the significance of the results is to be performed pricr to returning the steam generators to i service.

A postulated MSLB outside of containment but upstream of the Main Steam Isolation Valve (MSIV) represents the most

, limiting radiological condition relative to the APC. The ODSCC voltage distribution at the TSP intersections are projected to the end of the cycle and MSLB leakage is j

calculated. A site specific calculation has determined the

allowable MSLB leakage limit for the Byron 1 and Braidwood 1 sites. These limits use the recommended Iodine-131

transient spiking values consistent with NUREG-0800, ,

} " Standard Review Plan" and ensure site boundary doses are ]

within a small fraction of the 10 CFR 100 requirements. The -

projected MSLB leakage rate calculation methodology )

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described in WCAP-14046, "Braidwood Unit 1 Technical Support for Cycle 5 Steam Generator Interim Plugging Criteria," and WCAP 14277, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," will be used to calculate End Of Cycle (EOC) leakage. This method includes a Probability Of Detection (POD) value of 0.6 for all voltage amplitude ranges and uses the accepted leak rate versus bobbin voltage correlation methodology (full Monte Carlo) for calculating leak rate, as described in the May 30 Memorandum. The database used for the leak and burst correlations is consistent with that described in the May 30 Memorandum with the inclusion of the Byron 1 and Braidwood 1 tube pull results. The EOC voltage distribution is developed from the POD adjusted Beginning Of Cycle (BOC) voltage distributions and uses Monte Carlo techniques to account for variances in growth and uncertainty.

The Electric Po.. - Research Institute (EPRI) leak rate correlation is based on free span indications that have burst pressures above the MSLB pressure differential. There ,

is a low but finite probability that indications may burst at a pressure less than MSLB pressure. With limited TSP motion due to tube expansion, the tube is constrained by the TSP and tube burst is precluded. However, the flanks of the crack open up to contact the Inside Diameter (ID) of the TSP ;

hole and result in a primary to secondary leak rate potentially exceeding that obtained from the EPRI correlation. This phenomenon is known as an Indication Restricted from Burst (IRB) condition.

Comed has performed laboratory testing to determine the bounding leak rate obtainable in an IRB condition. The bounding leak rate value was then applied in a leak rate calculation methodology that accounts for the MSLB leak rate contribution from IRB indications to the total MSLB leak rate calculated as described above. Results indicate that -

the IRB contribution to the total leak rate value is negligible, however, Comed will conservatively add a leakage contribution due to IRBs in addition to the leakage calculated in accordance with the May 30 Memorandum. When this is done, the dose at the site boundary resulting from the predicted leakage is shown to be significantly less-than ;

10% of 10 CFR 100 limits. '

Modification of the Byron and Braidwood Specifications for conformance with the May 30 Memorandum requirements is primarily administrative and does not impact any accidents l previously evaluated. For Braidwood, the decrease in the allowed burst probability from 2.5x10-2 to 1. 0x10-2 is conservative. Byron Station has previously incorporated this requirement.

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Lowering the Unit 1 RCS dose equivalent I-131 limit from 1.0 pci/gm to 0.35 pci/gm is conservative, provides a defense in depth approach to implementation of this APC and ensures that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose rates at the Braidwood and Byron site boundaries will not exceed an appropriately small fraction of 10 CFR 100 dose guideline values with the predicted MSLB leakage calculated in accordance with this submittal until SGs are replaced at both sites.

Increasing the APC voltaga repair limit to a maximum of 3.0 volts for the hot-leg suppcrt plate intersections does not adversely affect steam generator tube integrity and results in acceptable dose consequences. By effectively eliminating tube burst at hot-leg TSP intersections, the likelihood of a tube rupture is substantially reduced and the probability of occurrence of an accident previously evaluated is reduced.

Therefore, the proposed amendment does not result in any i significant increase in the probability or consequences of an accident previously evaluated within the Byron Unit 1 and Braidwood Unit 1 Updated Final Safety Analysis Report (UFSAR).

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Implementation of the proposed steam generator tube plugging criteria with tube expansion does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism which could result in an accident outside of the region of the tube support plate elevations as ODSCC does not extend beyond the thickness of the tube support plates. Neither a single or multiple tube rupture event would be expected in a steam generator in which the plugging criteria has been applied.

The tube burst assessment involves a Monte Carlo simulation of the site specific voltage distribution to generate a total burst probability that includes the summation of the probabilities of 1 tube bursting, 2 tubes bursting, etc.

For the hot-leg TSP intersections, the total probability of burst, by design, is estimated to be 1x104 with all tube expansions functional. Accounting for the unlikely event of expansion failures, a sufficient number of redundant expansions exist to ensure that the burst probability remains below lx104 This includes the conservative assumption that all 32,046 hot-leg TSP intersections contain throughwall indications. This level of burst probability is considered to be negligible when compared to the May 30 Memorandum limit of 1x104 6

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I In addressing the combined effects of Loss Of Coolant Accident (LOCA) + Safe Shutdown Earthquake (SSE) on the SG

as required by General Design Criteria (GDC) 2, it has been j determined that tube collapse may occur in the steam generators at some plants. The tube support plates may 1

become deformed as a result of lateral loads at the wedge

. supports located at the periphery of the plate due to the

! combined effects of the LOCA rarefaction wave and SSE i loadings. The resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.

j There are two issues associated with SG tube collapse.

4 First, the collapse of SG tubing reduces the RCS flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA i which, in turn, may potentially increase Peak Clad i

Temperature (PCT). Second, there is a potential that

- partial throughwall cracks in tubes could progress to throughwall cracks during tube deformation or collapse. The tubes subject to collapse have been identified via a plant i specific analysis and excluded from application of the i l voltage-based criteria. This analysis is included in )

revision 3 to WCAP-14046 which was submitted to the NRC June 19, 1995.

Comed will continue to apply a maximum primary to secondary ;

leakage limit of 150 gallons per day (gpd) (0.1 gallons per minute (gpm)) through any one SG at Byron and Braidwood to help preclude the potential for excessive leakage during all plant conditions. The RG 1.121 criterion for establishing operational leakage limits that require plant shutdown are based on detecting a free span crack prior to resulting in primary-to-secondary operational leakage which could potentially develop into a tube rupture during faulted plant conditions. The 150 gpd limit provides for leakage detection and plant shutdown in the event of an unexpected single crack leak associated with the longest permissible free span crack length.

Tube burst is precluded during normal operation due to the proximity of the TSP to the tube and during a postulated MSLB event with tube expansion. The 150 gpd limit provides a conservative limit for plant shutdown prior to reaching critical crack lengths should significant crack extension unexpectedly occur outside the thickness of the TSP.

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Lowering the Unit 1 RCS dose equivalent I-131 limit from 1.0 pci/gm to 0.35 pci/gm is conservative, provides a defense in depth approach to implementation of this APC and ensures that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose rates at the Braidwood and Byron site boundaries will not exceed an appropriately small fraction of 10 CFR 100 dose guideline values with the predicted MSLB leakage calculated in accordance with this submittal until SGs are replaced at both sites.

Modification of the Byron and Braidwood Specifications for conformance with the May 30 Memorandum requirements is primarily administrative and will not alter the plant design basis. For Braidwood, the decrease in the allowed burst probability from 2. 5x10-2 to 1. 0x10-2 is conservative. Byron Station has previously incorporated this requirement.

With implementation of an increased APC voltage repair limit (up to a maximum of 3.0 volts) using tube expansion for the hot-leg support plate intersections, steam generator tube integrity continues to be maintained through inservice inspection, tube repair and primary to secondary leakage monitoring. By effectively eliminating tube burst at hot-leg TSP intersections, the potential for multiple tube ruptures is essentially eliminated. Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.

3. The proposed change does not involve a significant reduction in a margin of safety.

The use of the voltage-based bobbin coil tube support plate elevation plugging criteria with tube expansion at Byron Unit 1 and Braidwood Unit 1 is demonstrated to maintain steam generator tube integrity commensurate with the criteria of RG 1.121.

RG 1.121 describes a method acceptable to the NRC staff for meeting GDC 14, 15, 31, and 32 by reducing the probability or the consequences of steam generator tube rupture. This is accomplished by determining an eddy current inspection I voltage value which represents a limit for leaving a SG tube in service. Tubes with flaw-like voltage indications beyond this limiting value must be removed from service by plugging or repaired by sleeving. Upon implementation of an increased APC voltage repair limit (up to a maximum of 3.0 volts) for the hot-leg, even under the worst case conditions, the occurrence of ODSCC at the tube support plate elevations is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions. The EOC distribution of crack indications at the tube support plate elevations will be confirmed to 8

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result in acceptable primary to secondary leakage during all plant conditions such that radiological consequences are not adversely impacted. I l

Addressing RG 1.83 considerations, implementation of the ;

increased hot-leg tube support plate intersection bobbin coil voltage based repair criteria is supplemented by:

enhanced eddy current inspection guidelines to provide consistency in voltage normalization and a 100% eddy current inspection sample size at the affected tube support plate elevations.

For the leak and burst assessments. the population of indications in the voltage distribution is dependant on the POD function. The purpose of the POD function is to account for indications that may not be identified by the data analyst. In implementing this proposed APC, Comed will use the very conservative May 30 Memorandum POD value of 0.6 for all voltage amplitude ranges.

Lowering the Unit 1 RCS dose equivalent I-131 limit from 1.0 ,

pci/gm to 0.35 pci/gm is conservative, provides a defense in ;

depth approach to implementation of this APC and ensures l that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose rates at the Braidwood and Byron site boundaries will not exceed an appropriately small fraction of 10 CFR 100 dose guideline values with the predicted MSLB leakage calculated in accordance with this ;

submittal until SGs are replaced at both sites.

Modification of the Byron and.Braidwood Specifications for conformance with the May 30 Memorandum requirements is primarily administrative and will not reduce any safety margins. For Braidwood, the decrease in the allowed burst probability from 2.5x10-2 to 1. 0x10-2 is conservative. Byron Station has previously incorporated this requirement.

Implementation of the tube support plate elevation repair limits will decrease the number of tubes which must be repaired. The-installation of steam generator tube plugs or sleeves reduces the RCS flow margin. Thus, implementation of the alternate plugging criteria will maintain the margin of flow that would otherwise be reduced in the event of )

increased tube plugging. 1 l

Thus, the implementation of this amendment does not result in a significant reduction in a margin of safety.

Therefore, based on the above evaluation, Commonwealth Edison has concluded that these changes involve no significant hazards considerations.

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l ATTACHMENT E i ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 Commonwealth Edison Company (Comed) has evaluated this proposed license amendment request against the criteria for identification of licensing and regulatory actions requiring

! environmental assessment in accordance with Title 10, Code .

i of Federal Regulations, Part 51, Section 21 (10 CFR 51.21).

Comed has determined that this proposed license amendment request meets the criteria for a categorical exclusion set forth in 10- CFR 51.22 (c) (9) . This determination is based upon the following:

1. .The proposed licensing action involves the issuance of an amendment to a license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect to installation or use of ;

a facility component located within the restricted area, as defined in 10 CFR 20, or which changes an inspection or a surveillance requirement. This proposed license amendment request changes Byron and Braidwood Technical Specification (TS) 3.4.5,

" Steam Generators," the bases for TS 3.4.5, and TS 3.4.8, " Specific Activity."

The changes proposed to TS 3.4.5 will implement a 3.0 volts bobbin coil probe, voltage based, Steam i Generator (SG) Tube Support Plate (TSP) Alternate i Plugging Criteria (APC) limit for Outside Diameter I Stress Corrosion Cracking (ODSCC) indications at j the hot-leg TSP intersections.

The changes proposed to TS 3.4.8 involve reducing Reactor Coolant System (RCS) dose equivalent Iodine-131 (I-131) for both Byron and Braidwood.

For Byron and Braidwood, additional changes are proposed to make TS more consistent with the Model TS contained in the May 30, 1995 Frank J. Miraglia memorandum to Edward L. Jordan requesting CRGR review of Generic Letter 95-XX, " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" (May 30 Memorandum).

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2. This proposed license amendment request involves no significant hazards considerations; 3.

There is no significant change in the types or significant increase in the amounts of anyeffluent that

4. There is no significant increase in individual or cumulative occupational radiation exposure.

neither an Therefore, pursuant to 10 CFR 51.22(b), l dment environmental impact statement nor an environmentaass request.

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2. This proposed license amendment request involves i no significant hazards considerations; t

3. There is no significant change in.the types'or significant increase in the. amounts of any effluent that may be released offsite; and

4. There is no significant increase in individual or -

cumulative occupational radiation exposure.

Therefore,' pursuant to 10 CFR 51.22(b), neither an i environmental impact statement nor an environmental-assessment is necessary for this proposed license amendment request.

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ATTACHMENT F INDICATION RESTRICTED FROM BURST LEAK RATE CALCULATION METHODOLOGY Commonwealth Edison and the Electric Power Research l Institute (EPRI) have conducted a program to evaluate the impact of outside diameter stress corrosion cracking (ODSCC) tube indications in the support plate crevice area, which are unable to burst, regardless of crack geometry, because of tube support plate constraint. These indications are i known as indications restricted from burst (IRBs). The I objective of the program was to define a bounding leak rate for such indications and to use that bounding leak rate in the appropriate leak rate calculation under main steam line break (MSLB) conditions for an end-of-cycle (EOC) distribution of indications in the support plate. Based upon the results of this test program, a bounding leak rate for 7/8" and 3/4" tubing has been determined to be 5.0 gpm (June 20, 1995, letter from D. Saccomando to NRC to the Office of Nuclear Reactor Regulation " Additional Information Pertaining to the Application for Amendment to Facility Operating Licenses: Byron Nuclear Power Station, Units 1 and 2 NPF 37/66: NRC Docket Nos. 50-454/455, Braidwood Nuclear j Power Station, Units 1 and 2 NPF-72/77, NRC Docket Nos. 50- 1 456/457, " Steam Generators" and the final test program report to be submitted by July 14, 1995). j The leak rate calculation to be used for the Byron and Braidwood 3 volt Alternate Plugging Criteria (APC) uses the freespan leakage correlations developed by the Electric Power Research Institute and defined in the May 30, 1995, i memorandum from F. J. Miraglia to E. L. Jordan, Chairman, l Committee to Review Generic Requirements (CRGR), " Request !

For CRGR Review Of Generic Letter 95-XX, ' Voltage-Based Repail Criteria For Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking' " 1 (May 30 Memorandum). To assure that IRBs are conservatively included in this calculation, a leakage term is substituted for the freespan leakage for those indications which may leak at a higher rate under MSLB conditions. The EOC voltage distribution is predicted using a beginning-of-cycle voltage distribution, a Probability of Detection (POD) of 0.6, and by applying a growth rate, as defined in the May 30 Memorandum. The EOC leak rate is then predicted by calculating the freespan leakage for the EOC distribution indications that are not IRBs. The IRB leakage is determined by the probability of burst correlation in the 1

May 30 Memorandum, and, for those indications predicted to burst over the entire range of voltages, a 5.0 gpm leak rate !

is added.

Calculations completed by Comed indicate that the impact of IRBs on the EOC leak rate calculation is not significant.

However, Comed has chosen to use the freespan leak rate plus IRB leak rate methodology to more directly include IRBs in the calculation, i

i 2

1

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

ML20086D999

Text

Dearadation means a service-induced cracking,

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