Text

Application for Amends to Licenses NPF-72 & NPF-77,revising Ts,By Renewing Current 1 Volt SG Tube Plugging Criteria Per GL 95-05
	ML20087H688
Person / Time
Site:	Braidwood
Issue date:	08/15/1995
From:	Saccomando D; COMMONWEALTH EDISON CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20087H692	List: ML20087H688; ML20087H710;
References
	GL-95-05, GL-95-5, NUDOCS 9508180185
	Download: ML20087H688 (28)
	v • d • e

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" rommonwealth litison Company 14(M Opus Place

. Downer > Grtne,il69515 August 15,1995 U.S. Nuclear Regulatory Commission Document Control Desk

- Washington, D.C. 20555 -

Attn: Document Control Desk

Subject:

Application for Amendment to Facility Operating Licenses:

Braidwood Nuclear Power Station Unit 1 and 2 NPF-72/77: NRC Docket Number 50-456/457

" Steam Generators"

References:

1.

D. Saccomando letter to Nuclear Regulatory Commission dated April 21,1995, transmitting Information on Interim Plugging Criteria Request for Braidwood Unit 1 Cycle 6 2.

D. Saccomando letter to the Nuclear Regulatory Commission dated Febraary 13,1995, transmitting a Proposed Technical Specification Amendment for Increasing Interim Plugging ~

Criteria for Braidwood and Byron Unit 1 In the Reference letter 2, Commonwealth Edison Company (Comed) submitted an application for a 3-volt Interim Plugging Criteria (IPC) for the Byron and Braidwood Unit 1 steam generators. We understand that the Nuclear Regulatory Commission (NRC) is ' currently reviewing this amendment request and at this i

time there still remain issues that need to be resolved prior to Staff approval.

Comed and the NRC have been actively engaged in the resolution of our request and currently have a meeting scheduled for August 17,1995, to discuss outstanding issues. Comed remains committed to applying the 3-volt IPC in the Fall of 1995; however, as stated in Reference 1, in the event that the 3-volt IPC request is denied for the fall refueling outage, Braidwood is submitting a Technical Specification amendment request to renew their current 1-volt steam generator tube plugging uiteria in accordance with Generic Letter 95-05, " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," August 3,1995 (Generic Letter 95-

~05).

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August 15,1995 3

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- Pursuant to 10 CFR 50.90, Commonwealth Edison Company.(Comed) proposes to 3

amend Appendix A, Technical Specifications of Facility Operating Licenses NPF-1

72 and NPF-77.'Please__ note that the proposed Technical Specification amendment 1

is applicable to Braidwood Unit 1 only, but because common Technical

.j Specification pages are used for both units, this request is being docketed for Braidwood Units 1 and 2.

In this amendment request, Braidwwd proposes to amend Braidwood Technical.

Specification (TS) 3/4.4.5, " Steam Generators," TS 3.4.8 " Specific' Activity", and the 1

associated Bases. The changes proposed to " Steam Generatora" renew the voltage-l based steam generator tube plugging limit of 1.0 volt, which was previously

.l approved for Braidwood _ Unit 1, Cycle 5. The change to " Specific Activity""

l continues to implement a reduced dose equivalent iodine concentration in the t

. reactor coolant. Other changes are being proposed which will make the Braidwood Technical Specification consistent with Generic Letter 95-05. Based on guidance

-l in Generic Letter 95-05, application of the 1-volt criteria will not be cycle specific.

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The amendment package consists of the following:

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i Attachment A:

Description and Safety Analysis of Proposed Changes to

.I Appendix A r

Attachment B:

Alternate Plugging Methodology l.t Attachment C:

Proposed Changes to the Technical Specification Pages l

for Braidwood Station Attachment D:

Evaluation of Significant Hazards Consideration -

l Attachment E:

Environmental Assessment Comed requests that this proposed license amendment request be approved to j

permit application of the 1.0-volt Alternate Plugging Criteria (APC) during AIR 05

-l should the 3.0-volt IPC request not be approved. Approval of the 1.0-volt APC or the 3.0-volt IPC request is required in order to declare the Braidwood Unit 1 steam generators operable prior to entering Mode 4, Hot Shutdown.

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NRC Document Control Desk August 15,1995 j

l It is worthy to note that the steam generator manways are scheduled for l'

reinstallation on Wednesday, October 25,1995. In order to minimize potential rework and scheduling impact, Comed respectfully requests that this amendment request be approved, if possible, on or before October 25,1995.

l To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects these statements are not based on my personal knowledge, but on information furnished by other Comed employees, contractor employees, and/or consultants. Such information has been reviewed in accordance'with company practice, and I believe it to be reliable.

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. i Please address any further comments or questions regarding this correspondence i

to this office.

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^h^py~g'yj^k'([hk'{']

'ncerely,

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MARY JO YACK i fooTARY PUBUC. STATE of RLINotSl,>

d l MY CoMMtStooN EXPmf td MS/97 <l Denise M. Sacco ando

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i Nuclear Licensing Administrator

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/b T M d S7fY Attachments 44^

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cc:

D. Lynch, Senior Project Manager-NRR R. Assa, Braidwood Project Manager-NRR S. Ray, Acting Senior Resident Inspector-Braidwood J. Martin, Regional Administrator-RIII Office of Nuclear Safety-IDNS l

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ATTACHMENT A DESCRIPTION _AND SAFETY ANALYSIS OF PROPOSED CHANGES TO APPENDIX A l

TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 DESCRIPTION OF THE PROPOSED CHANGE Commonwealth Edison Company (Comed) proposes to amend the following j

Braidwood Technical Specircations Specification 3/4.4.5 REACTOR COOIANT SYSTEM - STEAM GENERATORS Specification 3/4.4.8 REACTOR COOLANT SYSTEM - SPECIFIC ACTIVITY Bases for Specification 3.4.5 l

The changes proposed to TS 3.4.5 renew the voltage-based Alternate Plugging Criteria (APC) limit of 1.0 volt, which was previously approved for Braidwood Unit 1, Cycle 5.

This proposed license amendment request will modify Specification 3/4.4.8 to reduce Reactor Coolant System DOSE EQUIVALENT Iodine-131 (DE I-131) concentration for Braidwood Unit 1.

Technical Specification Bases Section 3/4.4.5, STEAM GENERATORS, will also be modified to reflect these changes.

DESCRIPTION OF THE CURRENT REQUIREMENT Specification 3/4.4.5 The Technical Specification Surveillance Requirements (TSSRs) associated with Specification 3.4.5 had previously required that any SG tube with an imperfection depth at or exceeding the plugging or repair limit of 40% of the nominal wall thickness be removed from service by plugging or repaired by sleeving in the affected area. Amendment 54, issued in September,1994, authorized the use of a 1 volt Interim Plugging Criteria 'IPC) for Braidwood Unit 1, Cycle 5.

Specification 3/4.4.8 The Technical Specification Action Requirement references a footnote which requires that for Unit 1 Cycle 5, RCS DE I-131 will be limited to 0.35 microCuries per gram (pCi/gm).

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BASES OF THE CURRENT REQUIREMENT Soecification 3/4.4.5 The TSSRs for inspection of the SG tubes ensure that the structural integrity of

' this portion of the RCS will be maintained. The program for inservice inspection of SG tubes is based on a modification of Regulatory Guide (RG) 1.83, " Inservice Inspection of PWR Steam Generator Tubes," Revision 1, July 1975. Inservice I

inspection of SG tubing is essential in order to maintain surveillance of the condition of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of SG tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

Specification 3/4.4.8 The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the Site Boundary will not exceed an appropriately small fraction of Title 10 Code of Federal Regulations Part 100 (10 CFR 100) dose guideline values following a SG tube rupture accident in conjunction with an assumed steady state reactor-to-secondary SG leakage rate of 1 gpm.

The current Braidwood Unit 1 Cycle 5 DE I-131 limit of 0.35 pCi/gm referenced in the footnote to TS 3.4.8.a is based en ensuring the resulting 2-hour doses at the Site Boundary will not exceed an appropriately small fraction of 10 CFR 100 dose guideline values with the predicted Main Steam Line Break (MSLB) leakage calculated as part of Braidwood Station's April 30,1994, D. Saccomando to W.

l Russell letter.

The leakage limits specified in the TS ensure that current offsite dose limits are maintained.

NEED FOR REVISION OF THE REQUIREMENT Braidwood Unit I has four Westinghouse Model D-4 SGs and Unit 2 has four Westinghouse Model D-5 SGs. The significant differences between the SG models i

are in the tube material and tube support materials and design. The D-4s have 0.75" thick carbon steel tube support plates with drilled hole tube supports. The D-5s have 1.125" thick stainless steel support plates with Quatrefoil tube supports. The D-4 SG tubes are mill annealed Inconel 600 which were hard rolled into the tubesheet during initial assembly. Subsequently, the D-4 tubes were shot peened in the tubesheet area and thermally stress relieved in the U-bend area.

l The D-5 tubes are heat treated Inconel 600 which were hydraulically expanded into the tube sheet during initial assembly. Over the past several refueling outages, the number of SG tubes plugged per outage has been increasing. Unit I k rala @rtmed :stemgens ilvoltpac.wpf ; S 1

9 has had more defective tubes than Unit 2 primarily due to the design differences between the D-4 and D-5 SGs as described above.

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In the most recent Braidwood Unit 1 Steam Generator Mid-Cycle Outage (AIM 05), conducted in the spring of 1995, a SG tube inservice inspection was performed in accordance with the current TSSR 4.4.5.0. The results of this inspection identified a total of 3935 bobbin coil indications at the tube support plate locations. Using a rotating pancake coil to confirm these indications and a 1 volt IPC,874 flaws were identified due to ODSCC at the TSPs in 815 SG tubes.

Tha 815 tubes were removed from service by plugging. This increased the overall phyging total for Braidwood Unit 1 to 1678 tubes or 9.2% of the tubes. Of the 16'id tubes plugged to date,1587 were plugged due to ODSCC at the tube support plate locations.

For the upcoming Braidwood Unit 1 Refueling Outage (A1R05), the prediction on the number of pluggable indications using the previous TSSR 4.4.5 acceptance criteria (40-percent through-wall) are approximately 2600 tubes. This would result in 4278 tubes (23.4-percent) being plugged or repaired.

With the approval to use the APC as proposed, the predicted number of tubes requiring removal from service by plugging or repair by sleeving would be reduced to approximately 975. This represents a " savings" of 1625 tubes that would have had to be plugged or repaired using the 40-percent through-wall criteria. This represents a savings of approximately $7.5M in plugging and sleeving repair costs alone. In addition, APC implementation saves a minimum of 20 days in critical path outage time and eliminates the associated replacement power costs. Also, permitting these tubes to remain in service maximizes RCS flow and heat transfer area availability and minimizes RCS loop asymmetries and loss of rated thermal power.

Calculations conducted for Braidwood have shown that the resulting 2-hour doses at the site boundaries will not currently exceed an appropriately small fraction of 10 CFR 100 dose guideline values in conjunction with the predicted MSLB leakage calculated in accordance with this submittal and a DE I-131 level of 1.0 pCi/gm.

The site allowable leakage calculated using a DE I-131 level of 1.0 pCi/gm is 9.4 gallons per minute (gpm). This leakage includes accident leakage and the allowed 0.1 gpm primary-to-secondary leakage of the 3 unfaulted SGs per TS 3.4.6.2.c.

However, in order to provide a defense in depth approach to application of this requested APC and to envelope any future increases in MSLB leakage due to tube degradation, Braidwood is lowering the RCS DE I-131 level to 0.35 pCi/gm for all future cycles until SG replacement. The site allowable leak rate calculated using 0.35 pCi/gm DE I-131 is 26.8 gpm. This leakage also includes accident leakage and the allowed 0.1 gpm primary-to-secondary leakage of the 3 unfaulted SGs per TS 3.4.6.2.c.

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This amendment request also proposes to modify the footnote to TS 3.4.8.a which limits Unit 1 Cycle 5 RCS DE l-laLto 0.35 pCi/gm. The footnote will be changed such that the 0.35 pCi/gm limit is not cycle specific. Analyses described in WCAP 14046 have shown that the resulting 2-hour dosee at the Braidwood site boundary will not exceed an appropriately small fraction of 10 CFR 100 dose guideline values with the predicted MSLB leakage calculated in accordance with Braidwood Station's April 30,1994, D. Saccomando to W. Russell letter.

DESCRIPTION OF THE REQUESTED REVISION 1

The changes proposed in the amendment are contained in seven inserts to the surveillance requirements for the Braidwood Technical Specifications and Bases.

The inserts are applicable to Unit 1 but not Unit 2, as indicated. The inserts reflect the option to allow tubes to remain in service using a voltage-based APC for ODSCC indications in the tube support plate region. Using APC also results in changes to the sample selection, inspection criteria, and reporting requirements.

The term, " Tube Support Plate Alternate Plugging Criteria Limit", is revised and includes the equation to be used to determine repair criteria for other than a full operating cycle. Clari6 cations are made to existing definitions to reference APC, as appropriate.

Braidwood's probability of tube burst limit is decreased from 2.5x10-2 to 1.0x10'8 consistent with Generic Letter 95-05, " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" (Generic Letter 95-05).

The footnote to Braidwood TS 3.4.8.a which limits RCS DE I-131 to 0.35 pCi/gm for Unit 1 Cycle 5 is being changed such that this limit is no longer cycle specific.

TSSR 4.4.5.2.b, Substeps 5) and 6) are being renumbered for consistency with Byron Station. Other changes are being made to Braidwood TS to make them more consistent with the requirements of Generic Letter 95-05.

Specification 4.4.5.2. Steam Generator Tube Samole Selection and Inanection Changes to this section of the surveillance requirements will require that all tubes remaining in service due to the application of APC shall be included as part of the tubes to be inspected in addition to the sample selection made in accordance with existing criteria. Also, the surveillance requirements will specify how APC will be implemented.

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-1 Insert "A" adds a section to Specification 4.4.5.2.b, requiring all tubes in which the tube support plate APC limit is applied be inspected in each scheduled refueling :

- outage. This Insert is being locatalprior to the. existing Substep 5) for.

consistency with Byron Station. Insert "A" reads as follows "For Unit 1, indications left in service as a result 'of application of the tube ~

support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages."

Insert "B" revises section 4.4.5.2.d describing the inspections associated with the implementation of APC. Insert "B" reads as follows:

"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 3

plate intersections having ODSCC indications shall be based on the

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performance of at least a 20-percent random sampling of tubes inspected

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over their full length."

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>i Soecification 4.4.5.4. Accentance Criteria Insert "C" adds to the definition of" Plugging or Repair Limit", Specification 4.4.5.4.a.6, to identify that this definition does not apply for Unit 1 in the region of the tube subject to the TSP APC limit, i.e. the TSP intersections, and that l

Specification 4.4.5.4.a.11 describes the repair limit for use within the TSP intersection of the tube. Insert "C" reads as follows:

"For Unit 1, this definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied.

j Refer to 4.4.5.4.a.11 for the repair limit applicable to these intersections."

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Insert "D" revises Specification 4.4.5.4.a.11 defining the TSP APC limit. Insert "D" reads as follows:

"For Unit 1, the Tube Supoort Plate Pluaaina 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.

i At tube support plate intersections, the plugging (repair) limit is based on' maintaining steam gen'erator tube serviceability as described below:

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Steam generator tubes,~ whose degradation is attributed to outside -

. a.

diameter stress corrosion cracking within the bounds of the tube support plate with.behhin voltages less than or equal to the lower L

. voltage repair limit [ Note 1], will be allowed to remain in service.

b.

Steam generator tubes, whose degradation is attributed to outside i diameter stress corrosion cracking within'the bounds of the 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.a.11.c below.

Steam generator tubes, with indications of potential ~ degradation c.

attributed to outside diameter stress corrosion cracking within the :

bounds of the tube support plate with a bobbin voltage greater than j

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 degradation. Steam generator tubes, with indication of outside diameter stress corrosion

. cracking degradation with a bobbin voltage greater than the' upper voltage repair limit [ Note 2] will be plugged or repaired.

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 determined that these intersections may collapse or deform -

following a postulated LOCA + SSE event.

If an unscheduled mid-cycle inspection is performed, the following e.

mid-cycle repair limits apply instead of the limits identified in 4.4.5.4.a.11.a, 4.4.5.4.a.11.b, and 4.4.5.4.a.11.c. The mid-cycle repair -

limits are determined from the following equations:

V" Va=

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Va= Vm -(V,g-Vm)(

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Where:

Vun upper voltage repair limit

=

Vm lower voltage repair limit

=

V mid-cycle upper voltage repair limit based

=

xun on time into cycle V

mid-cycle lower voltage repair limit based

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a on V and time into cycle uun n o. w.i a.a e,....

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' length of time since last scheduled ~

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. inspection during which Vvat and V ac L

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cycle length (the time between two

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scheduled ~ steam generator inspections)

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' structural limit voltage t

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Gr.

average growth rate per ' cycle length

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NDEL 95-percent cumulative probability l

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uncertainty (i.e., a value of 20Jpercent has -

been approved by the NRC)-

2 Implementation of these mid-cycle repair limits should follow the..

i same approach as in TS 4.4.5.4.a.11.a, 4.4.5.4.a.11.b,' and -

4.4.5.4.a.11.c.

i Note 1:

The lower voltage repair limit is 1.0 volt for 3/4-inch diameter tubing.

i Note 2:

The upper voltage repair limit is calculated according to the~

'l methodology in Generic Letter 95-05, as supplemented.

f Specification 4.4.5.5 Renorts

- Insert "E" adds reporting requirement 4.4.5.5.d to identify the reports, including ;

j content ' nd time period, to be submitted to the Commission associated with the j

a implementation of APC. Insert "E" reads'as follows:

_j 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:

1.

If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle)' voltage dist-ibution exceeds the leak limit (determined from the licensing basis dose calculation for the postulated main steam line break) for the next operating cycle.

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2.

If circumferential crack-like indications are detected at the tube i

support plate intersections.

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Ifindications are identified that extend beyond the confines of the tube support plate.

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Ifindications are identified at the tube support plate. elevations that are attributable to primary water stress corrosion cracking.

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. 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 10-8, notify the NRC and.

L provide ~an assessment of the safety significance of the occurrence."

l Specification 3.4.8. Specific Activity The reference to Cycle 5 in the footnote to TS 3.4.8.a will be deleted. This footnote now reads:

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"**For Unit 1, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 L

microCuries per gram."

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A footnote with wording identical to that described above has also been added to the applicable TS ACTION statements and the notations for Table 4.4-4.

1 Figure 3.4-1 of TS 3.4.8 is being revised to include a new transient Iodine limit curve for Unit 1 based on the new Unit 1 DE I-131 level of 0.35 pCi/gm. A '

footnote is added to identify Unit I curve applicability when RCS Specific Activity -

is greater than 0.35 pCi/gm DE I-131.

1 Bases 3/4.4.5. Staam Generators Insert "F" adds a discussion to the Bases section of Technical Specifications to refer to the dispositioning of tubes in accordance with APC. Insert "F" reads as follows:

"The voltage-based repair limits for Unit 1 of Surveillance Requirement (SR).

j 4.4.5 implement the guidance in Generic Letter 95-05.

The voltage based repair limits of SR 4.4.5 are applicable only to Westinghouse-designed SGs with outside diameter stress corrosion cracking L

(ODSCC) located at the tube-to-tube support plate 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 Generic Letter 95-05 for additional description of the degradation morphology.

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l Implementation of SR 4.4.5 requires a derivation of the voltage structural l

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).

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The voltage structural limit is the voltage from the burst pressure 1 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 d'ownward to account for potential flaw growth during an operating interval and to account for NDE uncertainty. The upper voltage repair limit, Vm, is determined from the structural voltage limit by applying the following equation:

Va=V -Va,-Vuum st where Vax represents the allowance for flaw growth between inspections and V g represents the allowance for potential sources of error in the 3p measurement of bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit is contained in Generic Letter 95-05.

The mid-cycle equation in SR 4.4.5.4.a.11.e should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.

SR 4.4.5.5 implements several reporting requirements recommended by Generic Letter 95-05 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 Generic Letter 95-05 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 ifleakage and conditional burst probability were calculated using the measured end-of-cycle voltage distribution for the purposes of addressing Generic Letter 95-05 sections 6.a.1 and 6.a.3 reporting criteria, then the results of the projected end-of-cycle voltage distribution should be provided per Generic Letter 95-05 section 6.b (c) criteria.

The specific changes to these Technical Specifications and associated bases are included in Attachment C.

BASES FOR THE REVISED REQUIREMENT The technical bases for the changes proposed in this amendment request are contained in the following documents:

Generic Letter 95-05, " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking".

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The August 18,1994 epproval cf ths Braidwood request f:r a 1.0 v:lt

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Interim Plugging Criteria for 3/4" diameter SG tubing.

WCAP 14046,."Braidwood Unit 1 Technical Support for Cycle 5 Steam -

Generator Interim-Pluggm' g"Driteria," Revision 3, March 1995,-

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i Electric Power Research Institute (EPRI) Report NP-7480-L, " Steam i

Generator Tubing Outside Diameter Stress. Corrosion Cracking at Tube i

s Support Plates - Database for Alternate Repair Criteria,3/4 Inch Tubing."

j Volume 2, October 1993, j

Westinghouse Document SG-95-01-003, " Byron Unit 1 End-of-Cycle 6 Interim Plugging Criteria Report." January 17,1995.

j WCAP 14277, "SLB Leak Rate and Tube Burst Probability Analysis i

Methods for ODSCC at TSP Intersections." January 1995 The completion of a satisfactory review assuring the structural integrity of Braidwood SG tubing during the next cycle operation.

To support this request for amendment, Braidwood will remove tubes, as appropriate, from Unit 1 SGs for laboratory examination, leak, and burst testing.'

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Guidance from Generic Letter 95-05 on tube selection will be followed-1 Analysis required by the APC methodology will be completed to demonstrate leak '

and burst capabilities using AIR 05 inspection results and Cycle 5 growth rates.

1 The bases of the APC approach includes,in part:

.1 Determination of a beginning of cycle (BOC) voltage distribution for Cycle 6 with application of a POD of 0.6 in accordance with Generic Letter 95-05.

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-1 Prediction of an end of cycle (EOC) voltage distribution by applying Cycle 5 growth rates to the BOC distribution through Monte Carlo simulations.

Application of a log-logistic probability ofleakage (POL) function.

Application of the EPRI leak rate versus voltage correlation (conditional leak rate model).

Calculation of the EOC leak rate and comparison with the site allowable leak rate for off-site dose consideration.

Calculation of the EOC tube burst probability and comparison with the allowable burst probability per Generic Letter 95-05.

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Braidwood will be del: ting the ref:rence to Cycle 5 in the footnote of TS 3.4.8.a, adding footnotes limiting Unit 1 DE I-131 for the applicable TS ACTION statements and Table 4.4-4, and revising Figure 3.4-1 to incorporate the Unit 1 Iodine limit.

Calculations conducted for Braidwood have shown that the resulting 2-hour doses at the site boundaries will not currently exceed an appropriately small fraction of 10 CFR 100 dose guideline values in conjunction with the predicted MSLB leakage calculated in accordance with this submittal and a DE I-131 level of 1.0 pCi/gm.

The site allowable leakage calculated using a DE I-131 level of 1.0 pCi/gm is 9.4 gpm. This leakage includes accident leakage and the allowed 0.1 gpm primary-to-secondary leakage of the 3 unfaulted SGs per TS 3.4.6.2.c. However, in order to provide a defense in depth approach to application of this requested APC and to envelope any future increases in MSLB leakage due to tube degradation, Braidwood is lowering the RCS DE I-131 levels to 0.35 pCi/gm for all future cycles until SG replacement. The site allowable leak rate calculated using 0.35 pCi/gm DE I-131 is 26.8 gpm. This leakage also includes accident leakage and the allowed 0.1 gpm primary-to-secondary leakage of the 3 unfaulted SGs per TS 3.4.6.2.c.

IMPACT OF THE PROPOSED CHANGE The 1 volt IPC was approved for use at Braidwood Unit 1 in Amendment 54, dated August 18,1994. This amendment proposal is for renewal of those previously approved limits, and to make format changes which will comply with Generic Letter 95-05. With the implementation of this proposed license amendment request, the Braidwood Unit 1 SGs will continue to satisfy the requirements of Regulatory Guide 1.121. There will be no significant reduction in the margin of safety to protect the health and safety of the public. Based on current projections, approximately 2600 tubes with ODSCC would require repair under previous repair criteria during AIR 05. Implementation of a 1.0 volt APC at Braidwood Unit I will save approximately 1625 tubes from repair. This represents a savings of approximately $7.5M in plugging and sleeving repair costs alone. In addition, APC implementation saves a minimum of 20 days in critical path outage time and eliminates the associated replacement power costs. RCS loop asymmetries and the loss of rated thermal power due to excessive plugging and sleeving are minimized through APC application and RCS flow and available heat transfer area are maximized.

Lowering the Unit 1 RCS DE 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-hour dose rates at the Braidwood 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.

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SCHEDULE REQUIREMENTS Comed requests that this proposed license amendment request be approved to permit APC application during A1R05. Approval of this proposed license amendment request is requirelin order to declare the Braidwood Unit 1 SGs operable prior to entering Mode 4, Hot Shutdown. Based on the current outage schedule, Braidwood Unit 1 is predicted to be ready to enter Mode 4 on Tuesday, November 7,1995. It is worthy to note that the steam generator manways are scheduled for reinstallation on Wednesday, October 25,1995. In order to minimize potential rework and scheduling impact, Comed respectfully requests that this amendment be approved on or before October 25,1995.

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A'ITACHMENT B BRAIDWOOD UNIT 1 STEAM GENERATOR ALTERNATE PLC6CING CRITERIA METHODOLOGY INTRODUCTION Braidwood Unit I contains four Westinghouse Model D-4 steam generators (SG).

Each generator has 4578 mill annealed Inconel 600 U-tubes that are 3/4" diameter with 0.043" nominal wall thickness. Following the Mid-Cycle (AI'405) eddy current tube inspection, 3935 bobbin coil indications were identified at tube support plate locations. Of these,874 indications in 815 tubes exceeded the 1.0 volt interim plugging criteria (IPC). Subsequently, all 815 tubes were removed from service by plugging, thus increasing the overall plugging total to 1678 tubes or 9.2% of the tubes. Of the 1678 tubes plugged to date,1587 were plugged due to outside diameter stress corrosion cracking (ODSCC) at the tube support plate locations. The number of repairs are projected to be significant in future outages should the current Technical Specification plugging limit of 40% through-wall be applied. Braidwood Station is therefore requesting Technical Specification changes to renew a voltage-based Alternate Plugging Criteria (APC) for ODSCC at tube support plate intersections for Unit 1. The requested repair limits and inspection requirements are based on Generic Letter 95-05, " Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking". Additional considerations from the Braidwood SER for IPC, the Braidwood technical support document for IPC (WCAP-14046), the EPRI technical support documents for ODSCC at support plates, and Generic Letter 95-05 have been incorporated into the Braidwood Unit 1 APC methodology.

Some of the features of the Braidwood APC methodology are:

A 1.0 volt APC limit.

Calculation of the site specific maximum allowable primary-to-secondary leakage during a Main Steam Line Break (MSLB) event based on a small fraction of 10 CFR 100 limits at the site boundary.

Calculation of the tube structural limit is identical to the method applied during the Braidwood Unit 1 IPC implementation and is based on maintaining a margin of safety of 1.43 against tube failure under postulated accident conditions and maintaining a margin of safety of 3 against burst during normal operation.

Enhanced eddy current inspection guidelines have been implemented to increase detectability and minimize voltage variability.

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i The method: logy for calcul: ting primary-to-secondary leakage from ths steam generator tubes during a postulated MSLB is described in WCAP 14277. This method has been approved by the NRC.

Beginning of cycle (BOCJ and end of cycle (EOC) voltage distributions are determined with a Monte Carlo simulation to account for voltage growth during the cycle and a probability of detection (POD) correction of 0.6 to evaluate tube burst and leakage during a MSLB event.

Removal of a sample of steam generator tubes to evaluate the mode of degradation and provide additional data points for leak rate and burst correlations.

INSPECTION CRITERIA Inservice inspection of the steam generator tubing is an essential element in maintaining the reliability and structural integrity of the tubing. Inservice inspection also provides a means to characterize the nature and cause of any tube degradation so that corrective measures can be implemented. In the past, Braidwood Station has performed inspections on 100% of the tubes each refueling outage since plant startup with inspection and analysis guidelines consistent with '

EPRI Steam Generator Inspection Guidelines and Regulatory Guide 1.83. A common Byron /Braidwood eddy current guideline document was created and implemented. Prior to the Braidwood Unit 1 Cycle 4 refueling inspection, the Byron /Braidwood guidelines were updated and used in support of the Braidwood Unit 1 IPC effort. The update incorporated inspection requirements from draft NUREG-1477 and is consistent with those contained in Appendix A of WCAP-13854, referenced in Amendment Numbers 111 and 105 for Catawba Units 1 and 2, respectively. The guidelines have been updated further in accordance with Generic Letter 95-05. The enhanced inspection guidelines are intended to increase flaw detectability and reduce voltage variability such that consistent and accurate voltages and growth rates are obtained.

Technical Specification Surveillance Requirement (TSSR) 4.4.5.2 requires bobbin coil inspection to be performed on 100% of the hot leg tubes down to the lowest cold leg TSP elevation having ODSCC. A minimum of a 20% random sample is also to be inspected over the full length of the tube. Rotating pancake coil (RPC) inspections are to be performed on the following indications:

All TSP indications greater than 1.0 volts.

All TSP intersections that contain dents greater than 5.0 volts and a 20%

sample of dents between 2.5 volts and 5.0 volts. 'If Primary Water Stress Corrosion Cracking (PWSCC) or Circumferential Cracking is detected,100%

of the dents between 2.5 volts and 5.0 volts will be inspected.

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All intersecti:ns with 1 rga mixed re:iduab th:t could cau e a 1.0 volt signal to be missed or misread.

All intersections with interfering signals from copper deposits. Braidwood does not have significant copper deposits in the SGs. Guidance on conducting RPC inspections for interference signals due to copper has been included in the station inspection guidelines.

Any flaw-like indication confirmed by RPC at intersections with dent signals >2.5 volts, large mixed residuals, or copper deposits will result in the tube being repaired by sleeving or plugged. In addition, APC will not be applied to any crack-like indication in a wedge area or the Flow Distribution Baffle.

i In addition, the followmg data acquisition and analysis requirements will be met:

i The bobbin coil will be calibrated against a reference standard in the laboratory by direct testing or through use of a transfer standard.

The voltage response of new bobbin coil probes for the 40% to 100%

American Society of Mechanical Engineers (ASME) through-wall holes will not differ from the nominal voltage by more than i 10%.

Probe wear will be controlled by either an inline measurement device or tnrough the use of a periodic wear measurement. When utilizing the periodic wear measurement approach,if a probe is found to be out of specification (15%), all tubes inspected since the last successful calibration will be reinspected with a new calibrated probe.

Data analysts will be trained in the use of the Comed Byron and Braidwood Stations Units 1 and 2 Eddy Current Analysis Guidelines and qualified through site specific testing. Data analyst performance will be consistent with the assumptions for analysts measurement variability utilized in the tube integrity evaluations.

Quantitative noise criteria (resulting from electrical noise, tube noise, calibration standard noise) will be included in the data analysis procedures.

Data failing to meet these criteria will be rejected, and the tube will be reinspected.

Data analysts will review the mixed residuals on the standard itself and take action as necessary to minimize the residuals.

A 0.610 inch diameter bobbin coil probe will be utilized for the inspection.

If a 0.610 inch diameter probe will not pass through a portion of a tube, APC will not be applied to the portion of the tube that is inspected by a smaller probe.

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Data enalysts will be trained on the potential for primary water stress corrosion cracking to occur at TSP intersections. The data analysts will be sensitized to identify indications attributed to primary water stress corrosion cracking.

Whenever possible, a 3-coil RPC probe will be used to determine degradation orientation. Notification to the Staffis required prior to returning the SGs to service of any unexpected RPC results relative to the assumed characteristics of flaws at tube support plates. This includes reporting any detectable circumferential indications, PWSCC, or detectable indications extending beyond the thickness of the tube support plate. Indications with bobbin coil voltages greater than 1.0 volt but less than or equal to the upper voltage repair limit may remain inservice if degradation is not found with RPC. Indications of ODSCC degradation greater than the upper voltage repair limit will be plugged or repaired.

Safety-related work associated with steam generator repair at Braidwood Station is conducted in accordance with the Quality Assurance Program requirements identified in Appendix B to 10 CFR Part 50. Braidwood Station's Site Quality Verification (SQV) group typically provides an independent oversight of steam generator inspection and repair activities during refueling outages. Field Monitoring Reviews are conducted on steam generator activities, including eddy current testing and tube repair activities.

INSPECTION RESULTS The eddy current data from the 815 tubes that were plugged due to ODSCC from the previous outage (Cycle 5, Mid-Cycle outage in the Spring of 1995) were reanalyzed using the APC guidelines described in INSPECTION CRITERIA above to determine a voltage distribution. The indications found during the Cycle 5 Mid-Cycle outage were traced back to the Cycle 4 refueling outage and reanalyzed to obtain a growth rate for the first part of Cycle 5. The average growth rate of all of the indications for the first part of Cycle 5 was 0.28 volts over the entire cycle, or 0.40 volts /EFPY, with the largest single growth rate being 4.12 volts over the cycle (5.77 volts /EFPY). Likewise, the largest single voltage amplitude detected was 5.13 volts. These results are summarized in Table B-1 below:

Parameter Voltage for Voltage per Cycle 5 EFPY Average Growth Rate 0.28 0.40 Maximum Growth Rate 4.12 5.77 Maximum Voltage 5.13 n/a Table B-1 Braidwood Unit 1 Cycle 5 Plugged Tube Data krnla:byrbwd stesgens:1voltpac.wpf 19

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7 nThe Braidwood Unit 1 Cyclo 5 growth rates and ' amplitudes f:r plugged tubes are' '

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_ - consistent with industry data described in the EPRI PWR Steam Generator Tube n..

Repair Limits - Technical Support Document for ODSCC at Tube Support Plates, TR-100407 Revision 1. This data indicates that Braidwood Unit 1 ODSCC tube condition is supportive of an AMsIn' ce the data is comparable to other plants with an approved IPC. Comparison of typical Braidwood Unit 1 inspection results 1

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._ ith the res'ults from other plants that have axial ODSCC_ degradation confirmed w

through tube pulls, indicates that the Braidwood Unit 1 degradation is consistent -

with known ODSCC.

' A sample of degraded. tube intersections and a sample of non-degraded tube -

intersections from Braidwood Unit I were removed during the last refueling outage in April,1994. The tube pull results confirmed the crack morphology for degradation at the tube support plates and provided additional data points for leak and burst correlations. Further tubes will be removed per the direction of Generic Letter 95-05.

STEAM GENERATOR TUBE INTEGRITY The purpose of the Technical Speci6 cation repair limit is to ensure that tubes accepted for continued service will retain' adequate structural and leakage.

integrity during normal, transient, and postulated accident conditions, consistent with General Design Criteria 14,15,31, and 32 of 10 CFR Part 50, Appendix A.

Structural integrity is defined as maintaining adequate margins against ross F

failure, rupture, and collapse of the steam generator tubing. Regulatory Guide 1.121, " Basis for Plugging Degraded PWR Steam Generator Tubes" requires a i

structural margin of safety of 1.43 against tube failure under postulated accident ;

conditions and a margin of safety of 3 against burst during normal operation.

The proposed repair limit for Braidwood Unit 1 involves the implementation of the

-voltage based eddy current signal amplitude of 1.0 volt for ODSCC occurring :

within the thickness of the tube support plates. This method was previously approved for Braidwood Unit 1, Cycle 5. This is a change from the traditional -

depth-based criteria of 40% which is currently required by Braidwood Technical Speci6 cations. The proposed APC for Braidwood Unit 1 meets the requirements of Regulatory Guide 1.121 by demonstrating that tube leakage is acceptably low (resulting in offsite doses that are a small fraction of 10 CFR 100 limits) and tube

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burst is a highly improbable event during normal operation as well as during a postulated MSLB event.

The Braidwood Unit 1 methods for determining the structural and leakage tube integrity for normal operating and MSLB conditionn is described in WCAP-14277.

This method has been approved by the NRC.

~ Application cf APC begins with acquisition of eddy current data using the APC guidelines described in INSPECTION CRITERIA above and generating a voltage distribution. Growth rates are determined by comparing voltage amplitudes of the current and previous outage eddy current data. The BOC distribution is generated by applying a POD of 0.6 as described in Generic Letter 95-05. The predicted EOC distribution is determined by increasing the BOC voltages by allowances for non-destructive examination (NDE) uncertainties and voltage k:nla:byrtud:steegens:1voltpac.wpf 20

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growth through e Monte Carlo cimulation. These BOC cnd EOC distributions are used to assess the probability of burst, probability ofleakage and overall leak rates for normal operation and MSLB conditions. A degree of conservatism is inherent in these distributions due to the use of a constant POD of 0.6. Data collected by the EPRI PerformancTDamonstration Program indicates that the POD is larger than 0.6 for indications greater than 1.0 volt and increases with voltage amplitude.

For normal operating conditions, tube burst is precluded due to the constraining effects of the support plate in the area of the ODSCC affected portion of the tube.

However, during a MSLB event, support plate displacement may expose a portion of a crack to free span conditions and increase the probability of burst. Accident burst assessments are conservatively determined assuming the cracks are located entirely in the free span. The cumulative probability of burst for all indications left inservice, considering a POD of 0.6, must demonstrate a probability ofless than 1 X 10-2/ reactor year, in accordance with NUREG-0844, over the entire cycle of operation. The EPRI database in the report for 3/4" tubing serves as the basis for the log-linear relationship between burst pressure and bobbin voltage. Using the lower 95% confidence level of the burst pressure and bobbin voltage correlation, the bobbin voltage corresponding to the free span structural limit is 4.75 volts for a burst pressure of 3657 psi. This burst pressure corresponds to the MSLB differential pressure with a 1.43 safety margin consistent with Regulatory Guide 1.121.

The steam generator MSLB accident tube leakap assessment consists of a probability ofleakage model and a model predi.: ting leak rate as a function of voltage, assuming a leak occurs. The probability ofleakage (POL) model is based on a single function form, which is the 1.og-logistic. The use of the log-logistic function form for IPCs was determined to be acceptable by the Staffin the May 7, 1994 SER for Braidwood Unit 1. Any non-conservatism which may be associated with the use of the log-logistic function as compared to other function forms is small in comparison to the conservatisms inherent in the existing methodology for estimating the radiological consequences ofleakage induced by a postulated MSLB.

The leak rate versus voltage correlation is based on a linear regression fit of the logarithms of the corresponding leak rate and voltage data (known as the conditional leak rate model). The linear regression fit for the leak rate versus voltage correlation was deemed to be valid by the NRC in guidance from Generic Letter 95-05, when a p-value test result ofless than 5% can be demonstrated.

WCAP-14046, Section 6.6 determined that the conditional leak rate model p-value test result was significantly less than 5%, thus demonstrating a valid correlation.

The overall MSLB leak rate is obtained by applying the POL correlation and the conditional leak rate correlation to the POD adjusted EOC voltage distribution.

An upper bound 95% confidence limit is factored into the final leak rate value.

The final overall leak rate must be less than the maximum site allowable leak rate.

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The n;rmal operational tube leakage is implicitly assured by the allowable limits on the operational leak rate as specified in the Braidwood Technical Specifications.

The primary-to-secondary leak rate limit through any one steam generator was reduced by Amendment 50 to Braidwood Technical Specifications from 500 gallons per day (gpd) to 150 gpd with an overall limit of 600 gpd through all 4 steam generators. The 150 gpd limit in support of APC is based on the guidance provided in Generic Letter 95-05 and the EPRI Technical Report for ODSCC (TR-100407 Rev 1). These limits are based on the ability to detect primary-to-secondary leakage at normal operating conditions 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. The longest permissible crack is the length that provides a factor of safety of 1.43 against burst at faulted conditions. Alternate crack morphologies can correspond to the structural voltage limit of 4.75 volts, so a unique crack length is not defined by the burst pressure versus voltage correlation. Consequently, burst pressure versus through-wall crack length correlations are used to def'me the longest permissible crack for evaluating leakage limits. This evaluation is discussed in Section 8.3 of WCAP-14046. During normal operating conditions, tube burst is precluded due to the proximity of the support plate to the area of degradation.

Therefore, with the reduced limit and the leak rate monitoring program, reasonable assurance that a significant leak could be detected and the appropriate operator actions would occur in a timely manner are provided.

TUBES NOT APPLICABLE TO APC For a combined seismic event (SSE) and loss of coolant accident (LOCA) condition, designated LOCA+SSE, the potential exists for yielding of the tube support plate in the vicinity of the tube support plate-to-wrapper wedge locations. This deformation may lead to opening of pre-existing tight through-wall cracks or propagation of pre-existing non-through-wall cracks that would result in primary-to-secondary leakage. Therefore, tubes located in these susceptible areas are excluded from APC consideration. The tubes susceptible to collapse during a LOCA+SSE event are identical for Byron Unit 1 and Braidwood Unit 1 since both units contain Model D-4 steam generators with similar configurations. The tube lists and tube maps contained in Section 4.7 of WCAP-14046 are applicable to both Byron Unit I and Braidwood Unit 1.

OPERATIONAL MEASURES Braidwood Station's April 25,1994, and Byron Station's August 1,1994, request for a 1.0 volt IPC contained a description of enhanced operational and procedural measures that Braidwood and Byron Station had taken to ensure e. defense-in-depth approach against SG tube failures and detection of flaws that would exceed steam line break leakage limits. The measures remain in place at both Braidwood and Byron, and are summarized below.

Actions have been taken to mitigate the corrosive environment in the TSP

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crevices and to increase the likelihood that future growth rates and crack morphologies will be within expected bounds.

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Thialdrt cnd alarm setpoints on the main steam' lina cnd iteam jet air y

ejector radiation monitors have been lowered to ensure early positive j

' indication of primary to secondary leakage.

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Chamistry procedures have%n revised to facilitate."quiek counts" of:

chemistry samples to 'give rapid confirmation of SG leakage.

SG chemistry sampling frequencies have been increased to hourly when j

c primary-to-secondary leakage is detected, and then' reduced to not'less l

frequently than once per day once leakage stabilizes.

s In order to quick!y determine'if SG leakage is increasmg during a tube leak.

j event, Braidwood Operating Abnormal Procedure (BwOA SEC-8) and Byron Operating Abnormal Procedure (BOA SEC-8), have been revised to reqmre that radiation monitors be' checked at an increased frequency when SG-i leakage is detected.

j Tube rupture, tube leakage, and main stcam line break' scenarios are conducted frequently in the simulator. These scenarios include varying radiation monitor responses as appropriate.

Byron and Braidwood Emergency Procedures require continuous monitoring for SG tube leakage. BwOA SEC-8, and BOA SEC-8 require continued ;

monitoring ofleakage during a shutdown to ensure detection ofincreasing leakage.

' Control Room daily surveillances have been revised to require that hourly trend readings of steam jet air ejector radiation monitor activity levels be 1

reviewed on a daily basis.

For both Braidwood and Byron, TS 3.4.6.2.c has been' changed to limit primary-to-secondary leakage to 600 gallons per day total reactor-to--

secondary leakage through all SGs not isolated from the Reactor Coolant System, and 150 gallons per day through any one SG.

Generic Letter 95-05 Review

- Comed will implement all the requirements contained in Generic Letter 95-05.

l Below is a list which summarizes some of the key requirements contained in this Generic Letter.

Exclusion ofIntersections i

APC will not be applied to the following intersections:

LOCA + SSE tubes (Wedge area).

Dents greater than 5.0 volts.

Dents 2.5 volts to 5.0 volts with crack-like indications.

Large mixed residuals that could~cause a 1.0 volt indication in a tube i

to be missed or misread.

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! Intersections with interfering copper signals.'

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- Flow distribution baHles.

PWSCC or Circumferential crack-like indications at TSP.

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n Repair Criteria

The following indications / tubes will be repairedi All indications greater than the upper voltage limit.

All indications between the lower voltage limit and the upper voltage limit that'are confirmed by RPC.

. Tubes with known leakage.

RPC confirmed flaws indicative of ODSCC/PWSCC at locations that are excluded fmm APC as described above.

APC Vola-e Limit Determination A lower voltage limit of 1.0 volt.

An upper voltage limit is determined by reducing the structural.

voltage limit by voltage growth and NDE uncertainty.

Determined prior to each outage.

Use the larger of the site specific growth rate or 30%/EFPY.

NDE uncertainty of 20% of the BOC voltage.

Voltaae Growth Distribution Growth rates determined by indications identi6ed at two successive inspections, except that indications that grow from NDD to a relatively large voltage will be included (eg. 2.0 volts).

Current cycle growth rates:will be used if the current inspection or the current and previous inspections used IPC guidelines.

The most limiting growth rates will be used from' the last two inspection cycles.

Negative growth rates will be included as zero growth..

Re-evaluation of previous cycle data will be compensated for changes made in data acquisition guidelines.

Effects of chemical cleaning will be evaluated, if performed.

Tube Pulla A minimiun of two tubes and 4 TSP intersections have been removed.

A minimum of one additional tube (minimum 2 TSP intersections) will be removed following 34 effective full power months or 3 refueling cycles, whichever is shorter.

Alternatively, Comed will participate in a NRC approved industry.

ML20087H688

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