ML20117J058
| ML20117J058 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 08/19/1996 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20117J017 | List: |
| References | |
| NUDOCS 9609100138 | |
| Download: ML20117J058 (56) | |
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ATTACHMENT C MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 BYRON STATION UNITS 1 & 2 BRAIDWOOD STATION UNITS 1 & 2 REVISED t' AGES: REVISED PAGES:
3/4 4-13* 3/4 4-13*
3/4 4-14 3/4 4-14 3/4 4-14a*
3/4 4-15* 3/4 4-15*
3/4 4-16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-17a 3/4 4-17a 3/4 4-17b 3/4 4-17b 3/4 4-17c 3/4 4-17c 3/4 4-17d 3/4 4-17d B 3/4 4-3* B 3/4 4-3*
B 3/4 4-3a B 3/4 4-3a B 3/4 4-3b B 3/4 4-3b
- NOTE: THESE PAGES HAVE NO CHANGES BUT ARE INCLUDED FOR CONTINUITY.
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I 9609100138 960819 PDR ADOCK 05000454 P PDR
3 /4. 4. 5 STEAM GENERATORK W P NG CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY: M0' DES 1, 2, 3 and 4.
EllDN:
- With one or more steam generators inoperable, restore the inoperable steam generator (s) to OPERABLE status prior to increasing T , above 200*F.
SURVEILLANCE REDUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of Specificationaugmented the following 4.0.5. inservice inspection program and the requirements of 4.4.5.3 Steam Generator Samole selection and Insnection - Each steam generator least the minimum number of steam generators specified in 4.4.5.2 Steam Generator Tube
- 5= ale selection and Insnection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The l 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.
applying the expectations of 4.4.5.2.a through 4.4.5.2.c previous defects When or imperfections requiring reinspection. in the area repaired by the sleeve are not considered an area include at least 3% of the total number of tubes in all steam ge tubes except:selected for these inspections shall be selected on a random basis 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 shall include:to the preservice inspection) of each steam generator
- 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.
BYRON - UNITS 1 & 2 3/4 4-13' AMENDMENT N0. 58 l
REACTOR COOLANT SYSTDI
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SURVEILLANCE RI.alIREMENTS (Continued)
)
l 1) All tubes that previously had detectable tube wall penetrations i
greater than 20 percent that have not been plugged or sleeved in
!i , the affected area, and all tubes that previous'y had detectable e
sleeve wall penetrations that have not been plugged,
- 2) Tubes in those areas where experience has indicated potential problems, y
- 3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection,
- 4) For Unit 1, indications left in service as a result of application l' of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages, and
- 5) For Upit 1, tubes which remain in service due to the application of [
the F criteria wil7 be inspected, in the tubesheet region, during all future outages.
- c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
- 1) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and
- 2) The inspections include those portions of the tubes where imperfec-tions were previous 1 und.
- d. For Unit 1, through Cyc1 implementation of the steam generator tube / tube support plate r air criteria requires a 100 percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (0DSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length,
- e. A random sample of at least 20% of the total number of laser welded sleeves and at least 20% of the total number of TIG welded sleeves f
installed shall be inspected for axial and circumferential indications at the end of each cycle. In the event that an imperfection exceeding -
the repair limit is detected, an additional 20% of the unsampled sleeves shall be inspected, and if an imperfection exceeding the repair limit is detected in the second sample, all remaining sleeves shall be inspected. < '
These inservice inspections will include the entire sleeve, the tube at the heat treated area, and the tube to sleeve joints. The inservice inspection for the sleeves is required on all types of sleeves installed - f in the Byron and Braidwood Steam Generators to demonstrate acceptable structural integrity. I 1
BYRON - UNITS 1 & 2 3/4 4-14 AMENDMENTNO.p[
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REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 1The results of each sample inspection shall be classified into one of the following three categories:
Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
- C-2 One or more tubes, but not more than 1% of the total
! tubes inspected are defective, or between 5% and 10%
of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note: In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10% of wall thickness) further wall penetrations to be included in the above percentage calculations.
4.4.5.3 Inspection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
I 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 i observed degradation has not continued and no additional degradation has !
l 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 fran tube-to-tube sheet welds) in excess of the limits of g Specification 3.4.6.2c., or BYRON - UNITS 1 & 2 3/4 4-15 AMENDMENT NO. 83
i i . REACTOR COOLANT SYSTEN f SURVEILLANCE REQUIREMENTS (Continued)
- 2) A seismic occurrence greater than the Operating Basis Earthquake, l or 1
- 3) A Condition IV loss-of-coolant accident requiring actuation of the Engineered safety Features, or i
l 4) A Condition IV main steam line or feedwater line break.
4.4.5.4 Accentance Criteria
! a. As used in this specification:
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- 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 i
below 20% of the nominal tube or sleeve wall thickness, if detectable, may be considered as imperfections; l
Dearadation means a service-induced cracking,
wastage, wear or
] 2) general corrosion occurring on either inside or outside of a tube l
j or sleeve; l 3) Deoraded Tube means a tube or sleeve containing unrepaired
- imperfections greater than or equal to 20% of the nominal tube j
or sleeve wall thickness caused by degradation; 3
j 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 j the plugging or repair limit. A tube or sleeve containing an unrepaired defect is defective;
- 6) Pluaaina or Reoair 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 tr repair i
limit imperfection depth for the tubing and laser welded sleeves -
, is equal to 405 of the nominal wall thickness. The plugging limit i
imperfection depth for TIG welded sleeves is equal to 32% of the /
l For Unit 1, this definition does npt apply j nominal wall thickness.
to defects in the tubeshe that meet the criteria for an F tube; j For Unit 1, through Cyc1 this definition does not apply to tube support plate intersections for which the voltage-based plugging criteria are being applied. Refer to 4.4.5.4. for the repair limit applicable to these intersections; /3 j
l 7) Unserviceable describes the condition of a tube if it leaks or j
contains a defect large enough to affect its structural integrity i 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; 4
BYRON - UNITS 1 & 2 3/4 4-16 AMENOMENT NO. f3' 4
f.. REACTOR COOLANT SYSTEM i SURVEILLANCE REQUIREMENTS (Continued) i I 8) Tube insoection means an inspection of the steam generator tube !
- from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. For a tube that has been repaired by sleeving, the tube inspection shall include the sleeved portion of the tube, and 1
l g) Preservice Insoection means an inspection of the full length of
' each tube in each steam generator perfonned by addy current i
techniques prior to service to establish a baseline condition of the tubing. This inspection shall be perfonned prior to initial p0WER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
- 10) Tube Reoair refers to a process that reestablishes tube I serviceability. Acceptable tube repairs will be perfonned by the following processes:
l a) Laser welded sleeving as described in a Westinghouse Technical i Report currently approved by the NRC, subject to the !
limitations and restrictions as noted by the NRC staff, or j b) TIG welded sleeving as described in ABB Combustion Engineering, ' l Inc. Technical Reports: Licensing Report CEN-621-P, !
Revision 00, " Commonwealth Edison Byron and Braidwood Unit 1 ' -
and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, FINAL REPORT," April 1995, and Licensing Report CEN-627-P, Revision 00-P, " Verification of the Installation Process and ,
Operating Perfonnance of the ABB CEN0 Steam Generator Tube Sleeve for Use at Commonwealth Edison Byron and Braidwood - )
Units 1 and 2,* January 1996, subject to the limitations and l restrictions as noted by the NRC Staff. ;
gg Tube repair includes the removal of plugs that were previously installed as a corrective or preventative measure. A tube inspection per 4.4.5.4.a.8 is required prior to returning
' >previously plugged tubes to service. ;
'J3 4) For Unit I through Cycle he Tube Suncort Plate Pluaaino Limit l 1s used for the disposit on of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outer 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 attributed to outside I diameter stress corrosion cracking within the bounds of the bee'hn bel cold-le; teh ::;;:rt-plate with bobbin voltages less than or equal to the lower voltage repair limit (Note 1) will be i'
B u.sec k o s allowed to remain in service. Steam generator tubes, with degradation attributed to outside diameter stress corrosion l cracking within the bounds of the hot-leg-teh ::;p:rt plate--
with bobbin voltages less than or equal to 3.0 volts will be allowed to remain in service.
Loc.h ed- Tub e Mode l L+ers e c.6on s BYRON - UNITS 1 & 2 3/4 4-17 AMENDNENT NO.
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(4.4.5.4.a)
- 11) Locked-Tube Model Intersection means all steam generator hot- t leg tube-to-tube support plate intersections which have been :
analyzed to experience a tube support plate displacement less i
, than 0.1 inches during accident conditions, excluding the !
l following: l a) All tube-to-tube support plate intersections where IPC cannot be applied per Generic Letter 95-05; !
b) All Flow Distribution Baffle' intersections; l
c) All steam generator tube intersections adjacent to an intersection that contains a corrosion induced dent greater than 0.065 inches; and d) All tube-to-tube support plate intersections that will be displaced more than 0.1 inches during accident conditions due to failure of the steam generator internal structures. '
- 12) Free-Soan Model Intersection means all steam generator tube- !
to-tube support plate intersections to which the Locked-Tube l Model does not apply and which meet the criteria of Generic j Letter 95-05, excluding the following: i a) All tube-to-tube support plate intersections where IPC l cannot be applied per Generic Letter 95-05; and i b) All Flow Distribution Baffle intersections.
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- _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ __ ___ _ __ . =
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
L b) Steam generator tubes with degradation attributed to !
outside diameter stress corrosion cracking within the !
Ne LnC ModeI bouads of the e=1e 1:s t;ba : ;;:rt ;1 t:-with =
bobbin voltage greater than the lower voltage repair
~I~Mev sech,en s limit [ Note 1], will be repaired or plugged, except as noted in 4.4.5.4. below.
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c) Steam generator tu s with degradation attributed to '
L. Ocked- Tube- outside diameter stress corrosion cracking within the i bounds of the hg t 1;g td: ::;;;rt plate with a bobbin Mode.l hkrseche voltage greater than 3.0 volts will be repaired or
-' plugged. }
l d) Steam generator tubes, with indications of potential 1 degradation attributed to outside diameter stress pree 4P gege l corrosion cracking within the bounds of the : 1d le; 1 ytd: :e;; rt pl:t with a bobbin voltage greater than L h echens the lower voltage repair limit [ Note 1] but less than y hy or equal to the upper voltage repair limit [ Note 2], f may remain in service if a rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indication of outside diameter stress corrosion crackin bounds of the '- +g degradation
"" --- r4 plet within the with a -
bobbin voltage greater than the upper voltage repair /
i Flow bistA buMon limit [ Note 2) will be plugged or repaired.
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Affle. in4use 'oo s g gg e Certain intersections as identified in WCAP-14046, eWed b Section 4.7, will be excluded from application of the gFPyon oS%e voltage-based repair criteria as it is determined that these intersections may collapse or deform following a 7 Vo W d f yeg%8 M*h o. .
c u rk postulated LOCA + SSE event. N
{ f) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply ead of the limits id ad in 4.4.5.4.a.
C 4.4.5.4.a.)i. and 4.4.5.4.a.Wil for outsi iameter /
sta" corrosion cracking indications occurring in the /
steam generator tem-4egs. For outside diameter stress corrosion cracking indications occurring in the and 4.4.5.4.a.M.c app;y. l Thesteam generator mid-cycle et 1: ;, the limit repair lim are determined from the followin equations:
13 BYRON - UNITS 1 & 2 3/4 4-17a AMEN 0MENTNO./
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
Va= V" 1.0+NDE+Gr( U) /
CL
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Va=Vm -(V -Vm)( b)
Where:
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Vun - upper voltage repair limit Vtt,t -
Mer voltage repair limit V
aunt mid-cycle upper voltage repair limit based on tine into cycle /-
V,t,t -
mid-cycle lwer voltage repair limit /
based on V t and time into cycle At - length of Mme since last scheduled inspection during which Vunt and Vt ,t were implemented.
CL - cycle length (the time between two /
scheduled steam generator /
inspections)
V,t - structural limit voltage Gr -
average growth rate per cycle length NDE -
95-percent cumulative probability allowance for nondestructive examin-ation uncertainty (i.e., a value of /
20 percent has been approved by NRC) /
Implementation of these mid-cycle repair limitt should follow the same approach as in TS 4.4.5.4.a.}Ca,
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4.4.5.4.a.y.b,4.4.5.4.a.y.cand4.4.5.4.a.g.d.
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Note 1: The lower voltage repair limit is 1.0 volt for indications /
of outside diameter stress corrosion cracking occurring 44- /
in %e Ee-Span cold-leii t re;;:rt plate intersections.
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M* Note 2: The upper voltage repair limit for indications of outside '
diameter stress corrosion cracking occurring at celd-lag 4 s: :r;;:rt t pista intersections is calculated according to
/ i the methodology in Generic Letter 95-05 as supplemented.
yl) 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 determined to be 1.7 inches.
Js) F* Tube is g Unit I steam generator tube with degradation below the F distance and has no indications of degradation (i.e., no indication of cracking) within the F* distance.
Defects contained in an F tube are not dependant on flaw geometry.
BYRON - UNITS 1 & 2 3/4 4-17b AMEN 0MENTNO.J
SURVEILLANCE REOUIREMENTS (Continued)
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair in the affected area all tubes exceeding the plugging or repair limit) required by Table 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 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 shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the 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 l indication of an imperfection, and '
- 3) Identification of tubes plugged or repaired.
- c. Results of steam generator tube inspections which fall into I 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.
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- d. For implementation of the voltage based repair criteria o tube '
support plate intersections for Unit I through Cycle AF 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 distribution exceeds the leak limit (determined from ,
the licensing basis dose calculation for the postulated main f steamline 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 beyond the confines of the tube support plate.
- 4) If indications are identified at the tube support plate elevations that are attributable to primary water stress y corrosion cracking.
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BYRON - UNITS 1 & 2 3/4 4-17c l AMENDMENT NO. J7
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INSERT B (4 . 4. 5. 5. d)
- 5) If cracking is observed in the tube support plates.
- 6) If any tube which previously passed a 0.610 inch diameter bobbin coil eddy current probe currently fails to pass a 0.610 inch diameter bobbin coil eddy current probe.
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I SURVEILLANCE REOUIREMENTS (Continued)
J)I If the calculated conditional burst probability based on the ,- !
projected end-of-cycle (or if not practical, using the actual pasured end-of-cycle) voltage distribution exceeds 7 l 1
1 x 10' , notify the NRC and provide an assessment of the safety significance of the occurrence. f 1
f) internals inspection, if
- Following indications a steam generator.
detrimental to t he integrity of the load path necessary to support the 3.0 volt IPC are found, notify the ',-
NRC and provide an assessment of the safety significance of y the occurrence,
- e. The results of inspections of F* Tubes shall be reported to the i Commission prior to the resumption of plant operation. The report '
shall include:
- 1) Identification of F* Tubes, and
)
- 2) Location and size of the degradr. tion.
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k BYRON - UNITS 1 & 2 3/4 4-17d AMENDMENT NO. 77
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l . BASES
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l 3/4.4.5 STEAM GENERATORS I
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 l
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 the tubes in the event that there is evidence of mechanical
! damage or progressive degradation due to design, manufacturing errors, or
] inservice conditions that lead to corrosion. Inservice inspection of steam 4 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 j
coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant j
chemistry is not maintained within these limits, localized corrosion may likely
! result in stress corrosion cracking. The extent of cracking during plant l 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 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 i
reactor-to-secondary leakage of 150 gallons per day per steam generator can s
! readily be detected by radiation monitors of steam generator blowdown, l mainsteam lines, or the steam jet air ejectors. Leakage in excess of this i j limit will require plant shutdown and an unscheduled inspection, during which l the leaking tubes will be located and plugged or repaired by sleeving. The ,
technical bases for sleeving are described in the current Westinghouse or ABB l l
Combustion Engineering, Inc. Technical Reports.
l Wastage-type defects are unlikely with proper chemistry treatment of the i i
secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.
j Plugging or sleeving will be required for all tubes with imperfections i exceeding the plugging or repair limit of 40% of the tube, nominal wall thickness, excluding defects that meet the criteria for F tubes. A laser l welded sleeved tube must be plugged if a through wall penetration is detected j in the sleeve that is equal to or greater than 40% of the nominal sleeve thickness. TIG welded sleeved tubes must be plugged if a through wall penetration is detected in the sleeve that is equal to or greater than 32% of the nominal sleeve thickness. The plugging limit for the sleeve is derived
.l from Reg. Guide 1.121 analysis and utilizes a 20% allowance for eddy current uncertainty and additional degradation growth. Inservice inspection of sleeves i
l j is required to ensure RCS integrity. Sleeve inspection techniques are '
j described in the current Westinghouse or A88 Combustion Engineering, Inc.
Technical Reports. Steam Generator tube and sleeve inspections have l j demonstrated the capability to reliably detect degradation of the pressure j 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 i methods as better methods are developed and validated for commercial use.
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' AMENDMENT NO. 83
- 3YRON _ENITSJ 1.2 B 3/4 4-3 _ _ .
_ BASES m nmemm 3/4.4.5 STEAM GENERATORS (Continued) Nh M D&ro e chwn The voltage-based repair limits for Unit 1 in urveiITaNe' Requirement (SR) 4.4.5 implement the guidance in Generic Letter 95-05, " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" (Generic Letter 95-05) for Westinghouse-designed !
steam generators (SGs) with the exception of the specific voltag) limit. /
Generic Letter 95-05 discusses a 1.0 volt Alternate Plugging Criteria (APC) that can be applied to more than one cycle of' operation. Byron SR 4.4 A -
implements a 3.0 volt ht ic; Interim Plugging Criteria (IPC)fand a 1.0 volt /
- ld le;; IPC4for the Unit 1 SGs per WCAP-14273, " Technical Support for -
Oternative Plugging Criteria with Tube Expansion at Tube Support Plate /
Intersections for Braidwood-1 and Byron-1 Model D-4 Steam Generators" for a /
(specifiedoperatingcycle -the. he. - nH The voltage-based repai M f'S 4.4 M pl ca e only to el nkmechs [
Westinghouse-designed SGs with outside diameter stress corrosion cracking (00 SCC) 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 j extending outside the thickness of the support plate. Refer to Generic Letter 95-05 for additional description of the degradation morphology.
Application of the 3.0 volt h:t 1;; IPC requires verification of the integrity of load path necessary to support this IPC in accordance with the Byron /Braidwood Steam Generator Internals Inspection Plan.
Implementation of SR 4.4.5 requires a derivation of the voltage structural /
limit from the burst versus voltage empirical correlation and then the j 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 accountforthelgwer95/95-percenttoleranceboundfortubingmaterial 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. The u 7 repair limit for c:!d 1:; indications at the.ti gprt pl.ger m; V voltage,
, is j/
determined from the structural voltage limit by applying the followin g -
equation: ge _
'w psec/hons Rede. l V =V,t-V a ,-V.,
/.
where V , represents the allowance for flaw growth between inspections and V represents the allowance for potential sources of error in the measurement oY [
the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit is contained in Generic Letter 95-05.
BYRON - UNITS 1 & 2 B 3/4 4-3a AMENDMENT NO.
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REACTOR coolant SYSTEM y re n Mocdel _lodersecSon s BASES 3/4.4.5 STEAM GENERATORS (Continued)
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The mid-cycle equation in SR 4.4.5.4.a. .f should only be used during unplanned inspections in which eddy current data is acquired for indications /'.
___at the 4cid 1:; tub: ::;;;rt phte:. The voltage repair limit for indications /
at the Mt 1= tt : ;;;tJhiAag _during unplanned /
inspections. g_ TW MA l hhen s 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-cycla voltage distributions prior to returning the SGs to service. Note K1at if leakage 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-cyt.le voltage distribution should be provided per Generic Letter 95-05 section 6.b(c) criteria. /l-The maximum site allowable primary-to-secondary leakage limit for end-of-cycle main steamline break conditions incipdes the accident leakage from IPC
[
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in addition to the accident leakage from F on the faulted steam generator and the operational leakage limit of Specification 3.4.6.2.c. The operational leakage limit of Specification 3.4.6.2.c in 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 tubesheet grea 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 AS".F Boiler and Pressure Vessel Code,Section III, Subsection NB and Division I Appendices, for normal operating and faulted conditions.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-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 result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
/
BYRON - UNITS 1 & 2 B 3/4 4-3b AMENDMENTN0./7
!-* REACTOR CDOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION ,
3.4.5 Each steam generator shall be OPERABLE.
APPLICABILI1Y: MODES 1, 2, 3 and 4. .
ACTION:
With one or more steam generators inoparable, restora the inoperable steam generator (s) to OPERABLE status prior to increasing T,, above 200*F.
SURVEILLANCE REQUIREMENTS I
4.4.5.0 .Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.
4 . 4 . 5.1 Steam Generator Samole Selection and Insoection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.
i 4 . 4 . 5. 2 Steam Generator Tube
- Samole Selection and Insnection - The steam ,
generater 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 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 i 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: j e
- When referring to a steam gener>. tor tube, the sleeve shall be considered a i part of the tube if the tube has been repaired per Specification 4.4.5.4.a.10.
i a
BRAIDWOOD - UNITS 1 12 3/4 4-13 AMENDMENT NO. 46 ,
, -SURVEILLANCE REQUIREMENTS (Continued) j
- 1) All tubes that previously had detectable tube wall penetrations greater than 20% that have not been plugged or sleeved in the t
affected area, and all tubes that previously had detectable sleeve j wall-penetrations that have not been plugged, j
2LTubes' in those areas where experiencelas_ indicated potential i problems, .
i 2
- 3) A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not
)
i i
permit the passage of.the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and 4 .
subjected to a tube inspection, j 4)
- . For Unit 1, indications left in service as a result of application of //
4 the tube support plate voltage-based repair criteria shall be i
inspected by bobbin coil probe during all future refueling outages, and I
f i
- 5) F>r Upit 1, tubes which remain in service due to the application of the F criteria will be inspected, in the tubesheet region, during f
- all future outages.
- c. The tubes selected as the second and third samples (if required by Table 2
- 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided
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
- 2) The inspections include those portions of the tubes where imperfectio were previously found.
- d. For Unit 1 Cycle +, 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 i stres's corrosion cracking lowest cold-leg tube suppor(DDSCC) indications.
t plate intersections TheODSCC having determination of the indications
- shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length. ;
! e. A random sample of at least 20% of the total number of laser welded i sleeves and at least 20% of the total number of TIG welded sleeves installed shall be inspected for axial and circumferential indications at the end of each cycle. In the event that an imperfection exceeding the [;:
1 repair limit is detected, an additional 20% of the unsampled sleeves shall '
l be inspected, and if an imperfection exceeding the repair limit is -
i detected in the second sample, all remaining sleeves shall be inspected, i These inservice inspections will include the entire sleeve, the tube at ;
s the heat treated area, and the tube to sleeve joints. The inservice 3
inspection for the sleeves is required on all types of sleeves installed ;
in the Byron and Braidwood Steam Generators to demonstrate acceptable
. structural integrity. j l 4'
i BRAIDWOOD - UNITS 1 & 2 3/4 4-14
- AMENDMENT NO.
i,
,-. -. , , - . , . , . . - - --w .-..,..e , -
. , _ , , ., a < .-~
~
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued)
The reWts of each sample ins?ection shall be classified into one of the following three categories:
Cateoory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or mo.re 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.
BRAIDWOOD - UNITS 1 & 2 3/4 4-14a AMENDMENT NO. 75
l
^
l REACTOR COOLANT SYSTEM I -
(
SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - 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 L not less than 12 nor more than 24 calendar months after the previoys 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 t 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
- 2) A seismic occurrence greater than the Operating Basis Earthquake, i 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.
BRAIDWOOD - UNITS 1 & 2 3/4 4-15
- L SURVEllt ANCE RE0VIREMENTS (Continued) '
4.4.5.4 Acceptance Criteria
- a. As used in this specification:
- 1) Imoerfection means an exception tT"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; 3
- 2) Deoradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve;
- 3) Deoraded Tube means a tube or sleeve containing unrepaired imperfections greater than or equal to 20% of the nominal tube 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 er sleeve containing an unrepaired defect is defective;
- 6) Pluacino or Repair 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 for the tubing and -laser welded sleeves is equal to 40% of the nominal wall thickness. The plugging limii. imperfection depth for TIG welded sleeves is equal to 32% of the nominal wall thickness. For Unit 1, this definition does not apply to defects in the tubesheet that meet the criteria for an F tube. For Unit 1 Cycle-Er, this definition does not apply to the tube support plate intersections for which the voltage-based repair criteria are being applied. Refer to 4.4.5.4.a. for the repair limit applicable to these intersections; ,3 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;
- 8) Tube inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg. For a tube that has been repaired by sleeving, the tube inspection shall include the sleeved portion of the tube, and l
l BRAIDWOOD - UNITS 1 & 2 3/4 4-16 AMENDMENT N0.
REACTOR C00LANT SYSTEM
,. SURVEILLANCE REQUIREMENTS (Continued)
- 9) Preservice Insnection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial
-POWER OPERATION using the equig .6 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 the
- following processes
i a) Laser welded sleeving as described in a Westinghouse Technical i
Report currently approved by the NRC, subject to the l limitations and restrictions as noted by the NRC staff, or i
L b) TIG welded sleeving as described in ABB Combustion Engineering Inc. Technical Reports: Licensing Report CEN-621-P, j
- Revision 00, " Commonwealth Edison Byron and Braidwood Unit 1 i
and 2 Steam Generators Tube Repair Using Leak Tight Sleeves,
{
FINAL REPORT," April 1995, and Licensing Report CEN-627-P, ( ,/
4 Revision 00-P, " Verification of the Installation Process and Operating Performance of the ABB CEN0 Steam Generator Tube Sleeve for Use at Commonwealth Edison Byron and Braidwood Units 1 and 2," January 1996, subject to the limitations and ,
- l. restrictions as noted by the NRC Staff.
{nfo / A Tube repair includes the removal of plugs that were previously installed as a corrective or preventative measure. A tube inspection per 4.4.5.4.a.8 is required prior to returning prev,iously plugged tubes to service.
E^*
s -
4+> ""''*'"'"""'"'""'
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 thi'ckness 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 attributed to outside diameter stress corrosion cracking within the bounds of the f'"g/** y#'j,( ) 4:oM-leg tub: : ppert plate with bobbin voltages less than or bb et6ms i
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 Act kg t;b :r;;;rt pht:
withbobbinvoltageslessthanorejualto3.Lyogit will be allowed to remain in service. 'Lo Meh hb4 in Miech ms b) Steam generator tubes with degradation uted to outside diameter stress corrosion cracking within the bounds of the
> -cold hg tub: ::;;;rt phu 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. .d below.
L3 1
/
BRAIDWOOD - UNITS 1 & 2 3/4 4-17 AMENDMENTNO.J6
INSERT A (4.4.5.4.a)
- 11) Locked-Tube Model Intersection means all steam generator hot-leg tube-to-tube support plate intersections which have been analyzed to experience a tube support plate displacement less than 0.1 inches during accident conditions, excluding the following:
a) All tube-to-tube support plate intersections where IPC cannot be applied per Generic Letter 95-05; b) All Flow Distribution Baffle intersections; c) All steam generator tube intersections adjacent to an intersection that contains a corrosion induced dent greater than 0.065 inches; and d) All tube-to-tube support plate intersections that will be displaced more than 0.1 inches during accident conditions due to failure of the steam generator internal structures.
- 12) Free-Span Model Intersection means all steam generator tube-to-tube support plate intersections to which the Locked-Tube Model does not apply and which meet the criteria of Generic Letter 95-05, excluding the following:
a) All tube-to-tube support plate intersections where IPC cannot be applied per Generic Letter 95-05; and b) All Flcv Distribution Baffle intersections.
t.-
/
SURVEfttANCE REOUIREMENTS (Continued) b ad ~v c)
\
h),J T,Ja,redieaf Steam generator tubes with degradation attributed to !
outside diameter stress corrosion cracking within the '
bounds of the het-leg tub: : ;;;rt pl:te with a bobbin voltage greater tham 3.0 volts will be repaired or l plugged.
~
d) Steam generator tubes, with indications of potential degradation attributed to outside diameter stress j
corrosica cracking within the bounds of the cold-leg-tube.
Fm 4 - .apport-plat + with a bobbin voltase areater than the M, h l '[,,/,rg/a,, lower voltage repair limit [ Note l'! 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 s
degradation within the bounds of the cold-leg teb
' support-plate with a bobbin voltage greater than the 4 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 flM D,TNuM'I these intersections may collapse or deform following a i i
postulated LOCA + SSE event. 4 BMu inWataw Mg C) f If an unscheduled id-cycle inspection is performe i the I D# fre#f g/,,. limits identifle.d n 4.4.5.4.a.4+?a,p yfollowing mid-cycle !
o# / """d .4.5.4.a. b and volhye-h w d' 4.4.5.4.a.R.d Tor outside diameter stress corrosion cracking indications occurring in the ste m generator l
og,raucna ecold-legs. For outside diameter stress corrosion cracking indications occurring in the stea gener-ator dot-legs, the limits in 4.4.5.4.a.th and 4.4.5.4.a. W c apply. The mid-cycle repair limits are detennined from the following equations:
y* J $
Va=
1.0+RDE+Gr( )
Va= Vm -(Vm -Vm) (G t)
BRAIDWOOD - UNITS 1 & 2 3/4 4-17a AMENDMENT NO.
!- l
. REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) i i
Where: 1 I V o upper voltage repair limit V,,t t
- lower voltage repair limit V
ot mid-cycle upper voltage repair limit based on time into cycle V et -
mid.-cycle lower voltage repair limit based on V t and time into cycle At - length of Ume since last scheduled l inspection during which V ,t and Vt ,t were implemented.
CL -
cycle length (the time between two scheduled steam generator inspections)
V,t - structural limit voltage Gr -
average growth rate per cycle length /
NDE -
95-percent cumulative probability d 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.a.fi.a .b, 4.4.5.4.a.-It.c and 4.4.5.4.a.H.d. N4.4.5.4.a. l
. t- t i N 3(Lte 1: The lower voltage repair limit is 1.0 volt for indications of
/ie th' fru Spo, outs'ide diameter stress corrosion cracking occurring at-cold leg l gg itse--suppert phte intersections.
I
- Note 2: The upper voltage repair limit for indications of outside diameter l l
stress corrosion cracking occurring at--c+1d--leg tube :;uppert plate intersections is calculated according to^the methodology in Generic Letter 95-05 as supplemented. )
hh tP) 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 determined to be 1.7 inches.
+3) F* Tube is a Unit I steam generator tube with degradation below the F distance and has no indications of degradation (i.e., no indication of cracking) within the F* distance.
Defects contained in an F* tube are not dependant on flaw geometry,
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair in the affected area all tubes exceeding the plugging or repair limit) required by Table 4.4-2.
BRAIDWOOD - UNITS 1 & 2 3/4 4-17b AMENDMENT N0.
! .e
. REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.5 Reports
- a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number o.f 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 shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the 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. For implementation of the voltage based repair criteria to tube support plate intersections for Unit 1 Cycle-fr, 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 distribution exceeds the leak limit (determined from the licensing basis dose 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 beyond the confines of the tube support plate.
- 4) If indications are idcatified at the tube support plate elevations that are attribatable to primary water stress corrosion cracking.
I1Jer/
I 0) y l BRAIDWOOD - UNITS 1 & 2 3/4 4-17c AMENDMENT N0.
INSERT B (4.4.5.5.d)
- 5) If cracking is observed in the tube support plates.
- 6) If any tube which previously passed a 0.610 inch diameter bobbin coil eddy current probe currently fails to pass a 0.610 inch diameter bobbin coil eddy current probe.
1 s 1 4
[ -
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued)
- 7) \; - h If the calculated conditional burst probability based on the f /
J 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.
- 8) % Following a steam generator. internals inspection, if. I indications detrimental to the integrity of the load path necessary to support the 3.0 volt IPC are found, notify the NRC and provide an assessment of the safety significance of the occurrence.
l
- 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 l 2) Location and size of the degradation.
l l
i 1
l BRAIDWOOD - UNITS 1 & 2 3/4 4-17d AMENDMENT NO. 69 l
i' REACTOR COOLANT SYSTEM l . BASES i
j 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes j ensure that the structural integrity of this portion of the RCS will be main-tained?"The program for inservice inspection of'1 Team generator tubes is i based on a modification of Regulatory Guide 1.83, Revision 1. Inservice i
! 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 mechanical damage or progressive degradation due to design, manufacturing
- errors, or inservice conditions that lead to corrosion. Inservice inspection i
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 chemistry is not maintained 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 i
between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 150 gallons per day per steam generator). Cracks having i
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 i
reactor-to-secondary leakage of 150 gal.lons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown, i
mainsteam lines, or the steam jet air ejectors. Leakage in excess of this
' limit will require plant shutdown and an unscheduled inspection, during which j
the leaking tubes will be located and plugged or repaired by sleeving. The technical bases for sleeving are described in the current Westinghouse or ABB
- Combustion Engineering, Inc. 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, exclu' ding defects that meet the criteria for F tubes. A laser welded sleeved tube must be plugged if a through wall penetration is detected in the slene that is equal to or greater than 40% of the nominal sleeve thickness. TIG welded sleeved tubes must be plugged if a through wall penetration is detected in the sleeve that is equal to or greater than 32% of the nominal sleeve thickness. The plugging limit for the sleeve is derived from Reg. Guide 1.121 analysis and 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 ABB Combustion Engineering, Inc.
Technical Reports. Steam Generator tube and sleeve inspections have l demonstrated the capability to reliably detect degradation of the. pressure l retaining portions of the tube or slee"e wall thickness. Commonwealth Edison will validate the adequacy of any system that is used for periodic intervice inspection of the sleeves and, as deemed appropriate, will upgrade testin methods as better methods are developed and validated for commercial use.g BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 AMENDMENT NO. 75
- n. .
. i
< ~ w r-- -r 3/4.4.5 STEAM GENERATORS (continued) ( .
The voltage-based repair limits for Unit 1 in Syrveillance Requirement (SR) 4.4.5 implement the guidance in Generic Letter 95-05, " Voltage-Based Repair Criteria for_ Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" (Generic Letter 95-05) for Westinghouse-designed steam generators (SGs) with the exception of the specific voltage limit. Generic Letter 95-05 discusses a 1.0 volt Alternate Plugging Criteria (APC) that can be applied to more thar one cycle of operation. Braidwood SR 4.4.5 implements a 3.0 volt h:t 1 ; !nterim Plugging Criteria (IPC)'and a 1.0 volt : ld 10; IPC+for the Unit 1 SG! per_WCAP-14273, " Technical Support for Alternative Plugging Criteria with Yube Expansion at Tube Support Plate Intersections for Braidwood-1 and %yron-1 Model D-4 Steam Generators" for a
-, specified_ operating,cych ~
M de hee-Scan n ku} h tyse b 1 )
e vWa'g'e-based repafrTin;its of SR 4.4.5 are applicable only to Westinghouse-designed SGs with outside diameter stress corrosion cracking (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 00 SCC 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.
Application of the 3.0 volt h:t 1 ; IPC requires verification of the l' l integrity of the load path necessary to support this IPC in accordance with the Byron /Braidwood Steam Generator Internals Inspection Plan.
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 / j 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. The upper voltage repair limit for ::ld'l:; indications at the t;h ;;;;;-t l V , is determined from the structural voltage limit by applying _the l pl:tc; fElowingequation: g g hj l
V -V,t-V,r-V, M *"* N '
where V , represents the allowance for flaw growth between inspections and V represents the allowance for potential sources of error in the measurement oT the bobbin coil voltage. Further discussion of the assumptions necessary to determine the voltage repair limit is contained in Generic Letter 95-05.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3a AMENDMENT NO. 69
- e. .
REACTOR COOLANT SYSTEM BASES
- 3 3/4.4.5 STEAM GENERATORS (continued)
[
U in SR 4.4.5.4.a.-H.f should only be used during I unplanned inspections in which eddy current data is acquired for indications at the cal flag tube suppert pletes. The voltage repair limit for indications atthehet-legtebesuppertplet_e5 remains atJJLvol_ts t during unplanned inspections. /_ A, 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 if leakage 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. i The maximum site allowable primary-to-secondary leakage limit for end-of-1 cycle main steamline break conditions includes the accident leakage from IPC in addition to the accident leakage from F* on the faulted steam generator and the operational leakage limit of Specification 3.4.6.2.c. The operational leakage limit of Specification 3.4.6.2.c in 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 i
within the tubesheet 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." i F* tubes meet the structural integrity requirements with app moriate i margins for safety as specified in Regulatory Guide 1.121 and the ASME Boiler and Pressure Vessel Code,Section III, Subsection NB and Division I Appendices, for normal operating and faulted conditions.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission 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 result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
BRAIDWOOD - UNITS 1 & 2 B 3/4 ',-3b AMENDMENT NO. 69
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 Commonwealth Edison (Comed) has evaluated this proposed amendment and determined that it involves no significant hazards 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 does 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 Comed proposes to amend Braidwood and Byron Technical Specification (TS) 3/4.4.5, " Steam Generators" and the Bases for TS 3/4.4.5.
The changes proposed to TS 3/4.4.5 will renew the 3.0 volt bobbin coil probe, Steam Generator (SG) Tube Support Plate (TSP) Interim Plugging Criteria (IPC) limit for Outside Diameter Stress Corrosion Cracking (ODSCC) indications at hot-leg TSP intersections as described in WCAP-14273, " Technical Suoport for Alternative Plugging Criteria with Tube Expansion at Tube Support Plate Intersections for Braidwood-l and Byron-1 Model D4 Steam Generators," as supplemented, (Locked-Tube Model Intersections) and approved by the Nuclear Regulatory Commission (NRC) in Amendments 69 and 77 for Braidwood and Byron, respectively.
This proposal will also renew a 1.0 volt IPC to be applied to ODSCC indications at the cold-leg TSP intersections, and specific hot-leg intersections, (Free-Span Model Intersections) in accordance with Generic Letter 95-05, " Voltage-Based Repair D-1
Criteria for Westinghouse Steam Generator Tubes Affected By Outside Diameter Stress Corrosion Cracking," August 3, 1995 (GL 95-05). Administrative changes will also be made to TS 3/4.4.5 and to the Bases for TS 3/4.4.5 to further clarify the proper ;
application of the SG tube plugging and repair criteria described in this amendment request. This renewal will be applicable for Braidwood Unit 1 Cycle 7 and for Byron Unit 1 Cycle 9.
t 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.
This amendment request proposes to renew the SG tube i plugging / repair criteria previously approved by the NRC in l Amendments 69 and 77 to Braidwood and Byron Technical ~
Specifications, respectively.
The previously evaluated applicable accidents are steam generator tube burst and Main Steam Line Break (MSLB). The postulated MSLB outside of containment but upstream of the Main Steam Isolation Valve (MSIV) represents the most limiting radiological condition relative to the IPC. The pctential impact on public health and safety as a result of renewing the SG tube interim plugging criteria contained in the current Braidwood and Byron Technical Specifications is very low as discussed below. Tube burst due to predominantly axially oriented ODSCC at the TSP intersections 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, i.e., MSLB, the tubes and TSP may ,
move relative to each other. This can expose the crack length '
portion to free-span conditions. Testing has shown that the burst pressure correlates to the crack length that is exposed to the free-span, regardless of the length that is still contained within the TSP bounds.
Therefore, a more appropriate methodology has been established for addressing leakage and burst considerations. This methodology is based on limiting potential TSP displacements (Locked-Tube Model Intersections) during postulated MSLB events, thus reducing the free-span exposed crack length to minimal levels. The tube expansion process employed in conjunction with this tube plugging criteria is designed to provide postulated TSP displacements that result in negligible tube burst probabilities due to the minimal free-span exposed crack lengths. The tube expansions were performed during the first outage that the 3.0 volt IPC was applied (Braidwood refuel outage A1R05 - Fall 1995, and Byron midcycle outage BlP02 - Fall 1995) . These expansions will be inspected in accordance with an eddy current inspection D-2
)
probe that is sensitive to axial and circumferential indications.
This program will ensure the integrity of the expansions for the additional cycle of operation. It has been demonstrated that axial indications in the expansion region will not result in a reduction of the load carrying capability of the expanded tubes.
Thermal hydraulic modeling was used to determine TSP loading during MSLB conditions. A safety factor was 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.
Additional models were used to verify the thermal hydraulic results.
Assessment of the tube burst probability for the Locked-Tube Model Intersections was based on a conservative assumption that all hot-leg TSP intersections (32,046) contained through wall cracks equal to the postulated TSP displacement and that the crack lengths were located within the boundaries of the TSP.
Alternatively, it was assumed that all hot-leg TSP intersections contained through wall cracks with lengths 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 through wall indications, given a uniform MSLB TSP displacement of 0.31 inches, was calculated to be lx104 This is a factor of 1000 less than the GL 95-05 burst probability limit of lx10-2, Therefore, the functional design criteria for tube expansion was 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". This design goal results in a total tube burst probability of lx104 for all 32,046 postulated through wall indications.
Additional tubes were expanded to provide redundancy for the required expansions.
The structural limit for the Locked-Tube Model Intersection SG tube repair criteria was 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. Based on the current voltage distributions and growth rates, Monte Carlo projections were performed for Braidwood Unit 1 and Byron Unit 1 for the additional cycle of operation that this proposed amendment is requesting. The End of Cycle (EOC) voltage projections for Braidwood Unit 1 Cycle 7 predict that the maximum voltage to be seen will be less than 10.5 volta. The number of indications predicted greater than ten volts at the end of Cycle 7 for Braidwood Unit 1 is 0.3. The EOC voltage projections for Byron Unit 1 Cycle 9 predict that the maximum voltage to be seen will be less than 13.5 volts. The number of indications predicted greater than ten volts at the end of Cycle 9 for Byron Unit 1 is D-3
1 4.59.
l Using a tensile rupture probability for a ten volt indication of i 3x104, the probability of tensile rupture from the predicted 0.3 I indications at Braidwood is 1-(1-3x104) " = 9.0x10". The probability of tensile rupture from the predicted 4.59 indications at Byron is 1-(1-3x104)4d' = 1.38x104 Both of these probabilities result in a negligible contribution to the total l burst probability when compared to the 1x10-2 GL 95-05 limit.
Cellular corrosion is a more limiting mode of degradation at the TSPs with respect to affecting the tube structural limit.
Tensile tests that measure the force required to sever a tube with cellular corrosion and uncorroded cross sectional areas are l used to establish the lower bound structural limit. Based upon
! these tests, a lower bound 95% confidence level structural l 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 I 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 I volts. Accounting for voltage growth and Non-Destructive Examination (NDE) uncertainty, the full IPC upper limit exceeds ten volts. However, for added conservatism a single voltage repair limit of 3.0 volts for the Locked-Tube Model Intersection indications is specified in the current plugging / repair criteria.
All indications at the Locked-Tube Model Intersections with bobbin coil probe voltages greater than 3.0 volts will be plugged or repaired.
The free-span tube burst probability must be calculated for the indications at the Free-Span Model Intersections. The total burst probability must be within the requirements of GL 95-05.
The free-span structural voltage limit is calculated using
! correlations from the database described in GL 95-05, with the inclusion of the recent Byron, Braidwood, and South Texas tube i pull results. The structural limit for the Free-Span Model l
Intersections is 4.745 volts. The lower voltage repair limit for the indications at the Free-Span Model Intersections continues to be 1.0 volt. The upper voltage repair limit for the indications at the Free-Span Model Intersections will be calculated in accordance with GL 95-05.
Since IPC will not be applied to indications at the Flow Distribution Baffle (FDB), no leakage or burst analyses are required for these indications.
Per GL 95-05, MSLB leak rate and tube burst probability analyses are required to be performed prior to returning the unit to power. The results of these analyses are to be included in a report to the NRC within 90 days of restart. If allowable limits D-4 i
on leak rates and burst probability are exceeded, the results are to be reported to the NRC and a safety assessment of the significance of the results is to be performed prior to returning the SGs to service.
A site specific calculation has determined the site allowable leakage limit for Braidwood and Byron. These limits use the recommended Dose Equivalent 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 leakage rate calculation methodology 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 the 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 the leak rate, as described in GL 95-05. The database used for the leak and burst correlations is consistent with that described in GL 95-05 with the inclusion of the Byron Unit 1, Braidwood Unit 1, and South Texas 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 NDE uncertainty.
The Electric Power Research Institute (EPRI) leak rate I correlation has been used. This 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 for the Locked-Tube Model Intersections, the tube is constrained by the TSP and tube burst is precluded.
However, the flanks of the crack may 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 (6.0 gallons per minute). The bounding leak rate value was then applied to a leak rate calculation methodology that accounts for the MSLB leak rate contribution from IRB indications to the total leak rate calculated as described above. Results indicate that the IP3 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 GL 95-05. When this is done, the dose at the site boundary D-5
resulting from the predicted leakage will be a small fraction (less than 10%) of the 10 CFR 100 limits. l l
Modification of the Braidwood and Byron TS to clarify application I of the proposed tube plugging / repair criteria is purely administrative and will not have any effect on the probability or ,
consequences of an accident previously evaluated. i Operating experience over the last cycle with this plugging ;
criteria applied has not revealed any unpredicted or unusual l
effects. >
l For these reasons, renewal of the current Braidwood and Byron j tube plugging criteria does not adversely affec; SG tube integrity and results in acceptable dose consequences. By effectively eliminating tube burst at the Locked-Tube Model TSP intersections, the likelihood of a tube rupture is substantially reduced and the probability of occurrence of an accident previously evaluated is reduced. !
1 This conclusion is not affected by foreign or domestic plant SG i experiences (NRC Information Notice 96-09 and its supplement). !
As the following evaluation shows, these experiences are not relevant to Braidwood or Byron.
A foreign unit detected eddy current signal distortions in one area of the top TSP during a 1995 inspection. The steam generators had been chemically cleaned in 1992. Visral l inspection showed that a small section of the top TSP had broken j free and was resting next to the steam generator tube bundle 4 wrapper. The support plate showed indications of metal loss. l The chemical cleaning process used by the foreign unit was developed by the utility and differs significantly from the modified EPRI/SGOG process performed at Byron Unit 1 in 1994.
The foreign chemical cleaning process, coupled with the specific application of the process, resulted in TSP corrosion of up to 250 mils compared to a maximum of 2.16 mils (11 mils maximum allowed) measured at Byron. During the Byron eddy current inspection performed after the chemical cleaning, no distortion of the tube support plate signals was reported. Therefore, these differences in cleaning processes imply that this foreign experience is irrelevant to the effects of the chemical cleaning process on the TSPs at Byron. Chemical cleaning of the SGs has l not occurred at Braidwood. I i
A number of units have experienced TSP cracking associated with )
severe tube denting due to TSP corrosion at the tube-to-TSP crevice. WCAP 14273, Section 12.4, shows that a diametral ,
reduction of a SG tube of 0.065 inches is required to develop '
stress levels above yield in the TSP ligaments at dented intersections. The bobbin voltage range associated with a one D-6
i
. i 5
-mil radial dent is twenty to twenty-five volts. i i
Although Braidwood Unit 1 and Byron Unit 1 have not seen corrosion induced denting, a 0.610 inch diameter bobbin coil probe will be used as a go/no-go gauge to assess dents at the !
Locked-Tube Model Intersections, if they occur in the future. If ,
a tube has a dent at a Locked-Tube Model TSP intersection that ,
fails to pass the go/no-go test probe, IPC will not be applied to that intersection. In addition, if the dent is determined to be corrosion induced, the Free-Span Model repair criteria will be applied to the intersections adjacent to the dented intersection.
IPC repair limits will not be applied to tubes with dents greater than 5.0 volts since dent signals of this magnitude could mask a 1.0 volt ODSCC signal. Tube intersections with corrosion induced l dents greater than 5.0 volts and the intersections adjacent to ;
l such an intersection were not selected for tube expansion to j preclude adverse effects of the failure of such a tube on limiting TSP displacement. If corrosion induced denting, either greater than 5.0 volts or such that the tube is unable to pass a !
0.610 inch diameter bobbin coil probe, are detected at an intersection adjacent to an expanded intersection, the dented I intersection will be inspected by an EPRI developed technique to
! determine if the TSP is cracked. If a crack-like indication is identified in a TSP, a plus point inspection will be conducted per the EPRI TSP program. If the plus point inspection verifies the existence of a crack-like indication, the effect of that indication on TSP displacement will be evaluated. If this l evaluation shows that TSP displacement would be greater than 0.1 inches during a MSLB event, the effected area will either be mechanically corrected or the Free-Span Model criteria will be applied to the affected area. Based on the information presented above, the SG tube denting experience at other plants is not relevant to Braidwood or Byron.
A foreign utility's SGs have experienced cracking at the top TSP.
The cause of the cracking appears to be the configuration of the single anti-rotation device, connected between the SG shell and wrapper, and the wrapper internals. The single anti-rotation device carries the full load associated with the wrapper to shell motion. This rotational load is believed to be transferred to the TSP via the wrapper internals. The Byron /Braidwood Unit 1 SG design (D-4) uses three anti-rotation devices to spread the ;
rotational load. The D-4 wrapper internals are configured such !
that this load is not directly transmitted to the TSP. ]
l No top TSP cracking has been detected at Braidwood Unit 1 or j Byron Unit 1 and very few (<1%) of the ODSCC indications in the i SG tubes at Braidwood and Byron, to date, have been at the top l TSP elevation. Nevertheless, an analysis was performed to assess the impact of cracking of the top TSP. The results show an
! increase in the deflection of the top TSP for a very limited
- number of tubes to greater than the 0.10" limit used in the 3.0 D-7
.* l I
volt IPC analysis. The deflections of the lower support plates l l
also increased, but remain within the 0.10" limit. Thus, a large '
majority of the Locked-Tube Model indications continue to be '
bounded by the existing analysis even with a cracked top TSP. .
The Locked-Tube Model repair criteria will not be applied to any !
SG tube ODSCC indication where the TSP has been shown to be '
displaced by more than 0.1 inches during accident conditions. i In response to these experiences at foreign and domestic utilities, Comed developed an inspection plan for the SG internals to identify if indications detrimental to the load path components existed. This inspection plan was carried out at Braidwood during refueling outage A1R05 (Fall 1995) and at Byron during the midcycle outage B1P02 (Fall 1995) and refuel outage B1R07 (Spring 1996). These inspections revealed no degradation of the SG load path components necessary to support implementation of the 3.0 volt IPC. Inspections will be performed during the upcoming refuel outages at Braidwood Unit 1 and Byron Unit 1 to further ensure the integrity of the SG load path components necessary to support implementation of the 3.0 volt IPC.
A domestic utility reported several distorted TSP signals over the past three refueling outages' SG tube inspections. It was determined that these signals were associated with the TSP geometry in an area where an access cover is welded to the TSP.
l These signal distortions are not cttributed to TSP cracking or degradation. Since the distorted signals were due to TSP geometry which did not indicate or result in a defect of the TSP, there is no increase in the probability or consequences of an I accident previously evaluated due to Braidwood Unit 1 and Byron Unit 1 steam generator TSP geometries which may result in distorted eddy current signals.
One foreign unit observed a dislocation of the tube bundle wrapper when they were unable to pass sludge lancing equipment through a hand hole in the wrapper. The dislocation appears to be a result of improper attachment of the wrapper to the support structure. SG sludge lance operations have been successfully performed at Braidwood Unit 1 and Byron Unit 1 which indicates that no problem with the wrapper attachment exists. The foreign unit's wrapper support design is significantly different than that used on Braidwood Unit 1 and Byron Unit 1. Therefore, a similar wrapper dislocation will not occur and the foreign experience is not applicable to Braidwood or Byron. An inspection was conducted during the last Braidwood Unit 1 and Byron Unit I refueling outages which verified this conclusion.
Comed will continue to apply a maximum primary-to-secondary leakage limit of 150 gallons per day (gpd) through any one SG at Braidwood and Byron to help preclude the potential for excessive leakage during all plant conditions. The RG 1.121 criterion for i D-8 l
1 i
establishing operational leakage limits that require plant shutdown are based on detecting a free-span crack prior to it !
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 l associated with the longest permissible free-span crack length. '
Therefore, the proposed amendment does not result in any (
significant increase in the probability or consequences of an ;
accident previously evaluated within the Braidwood and Byron ,
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. !
This amendment request proposes to renew the SG tube !
plugging / repair criteria previously approved by the NRC in l Amendments 69 and 77 to Braidwood and Byron Technical '
Specifications, respectively.
I Renewal of the proposed steam generator tube plugging criteria l 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 i beyond the thickness of the tube support plates and IPC is not allowed to be applied to indications that extend beyond the thickness of the tube support plate. Neither a single nor multiple tube rupture event would be expected in a SG 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 one tube bursting, two tubes bursting, etc. For the Locked-Tube Model TSP Intersections, the maximum total probability of burst, by design, is estimated to be lx104 with all tube expansions ,
functional. The burst probability for the Free-Span Model TSP intersections will be dependent on the number and size of indications at these applicable intersections. The total burst probability will be within the limit specified in GL 95-05.
Accounting for the unlikely event of a failure of the expanded tubes, 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 i intersections contain through wall indications. This level of burst probability is considered to be negligible when compared to the GL 95-05 limit of lx104 D-9
L l In addressing the combined effects of a Loss Of Coolant Accident l
(LOCA) during a Safe Shutdown Earthquake (SSE) on the SG as l required by General Design Criteria (GDC) 2, it has been l
determined that tube collapse may occur in the steam generators at some plants. The tube support plates may become deformed as a ;
result of lateral loads at the wedge supports located at the I periphery of the plate due to the combined effects of the LOCA l rarefaction wave and SSE loadings. The resulting pressure '
l differential on the deformed tubes may cause some of the tubes to l collapse. There are two issues associated with SG tube collapse. '
First, the collapse of SG tubing reduces the Reactor Coolant System (RCS) flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA which, in turn, may potentially increase the Peak Clad Temperature (PCT). Second, there is a potential that partial through wall cracks in the SG tubes could progress to through wall cracks during tube deformation or collapse. The tubes subject to collapse have been identified via a plant specific analysis and are excluded from application of any voltage-based criteria. This analysis is included in revision 3 to WCAP-14046 which was submitted to the NRC June 19, 1995.
l l
Modification of the Braidwood and Byron Technical Specifications to clarify application of the proposed tube plugging / repair ;
criteria is purely administrative and will not create the possibility of a new or different kind of accident from any accident previously evaluated.
Operating experience over the last cycle with this plugging criteria applied has not revealed any unpredicted or unusual effects.
SG tube integrity will continue to be maintained following renewal of the 3.0 volt IPC voltage repair limit through inservice inspection, tube repair and primary-to-secondary leakage monitoring. By effectively eliminating tube burst at the Locked-Tube Model TSP Intersections, the potential for multiple tube ruptures is essentially eliminated.
Comed has evaluated industry experiences with TSP degradation, eddy current signal distortions, and component misalignment.
Eddy current signal distortions due to TSP geometry are not ,
indicative of TSP degradation and do not result in any kind of I new or different accident.
The component misalignment experienced by one unit is not applicable to Braidwood Unit 1 or Byron Unit 1 and, thus, will not result in any kind of new or different accident. Specific limitations, as discussed in response to Question 1, will be applied to indications at the Locked-Tube Model Intersections which contain dents. These limitations ensure that the integrity l of the SG tubes is maintained consistent with the current 1
D-10
I .
i
~
analyses should tube denting or TSP cracking occur.
Therefore, renewal of the current tube plugging / repair criteria at Braidwood Unit 1 and Byron Unit 1 will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The proposed change.does not involve a significant reduction in a margin of safety.
Th; use of the voltage-based, bobbin coil, tube support plate plugging criteria with tube expansion at Braidwood Unit 1 and Byron Unit 1 is demonstrated to maintain SG 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.
Reducing the probability or the consequences of steam generator tube rupture is accomplished by determining an eddy current inspection voltage value which represents a limit for leaving an axial, crack-like indication at an in service SG tube TSP intersection. Tubes with ODSCC voltage indications beyond this limiting value must be removed from service by plugging or repaired by sleeving. Implementation of a 3.0 volt IPC voltage repair limit for the Locked-Tube Model Intersections has been evaluated and shown not to present a credible potential for a
-steam generator tube rupture event during normal or faulted plant conditions, even with worst case assumptions. The total tube burst probability will include a contribution from the indications at the Locked-Tube Model Intersections and from I indications at the Free-Span Model Intersections. The projected EOC voltage distribution of crack-like indications at the TSP elevations will be confirmed to result in acceptable primary-to-secondary leakage during all plant conditions such that radiological consequences are not adversely impacted.
Addressing RG 1.83 considerations, implementation of the increased Locked-Tube Model 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 TSP elevations.
For the leak and burst assessments, the population of indications !
in the EOC voltage distribution is dependent on the POD function.
The purpose of the POD function is to account-for new indications that may develop over the cycle, and to account for indications I not identified by the data analyst. In implementing this !
proposed IPC renewal, Comed will continue to use the conservative GL 95-05 POD value of 0.6 for all voltage amplituda ranges.
i D-11 I l
l !
I l
e i Modification of the Braidwood and Byron Technical Specifications to clarify application of the proposed tube plugging / repair criteria is purely administrative and will not reduce any safety margins. j Operating experience over the last cycle with this plugging !
criteria applied has not revealed any unpredicted or unusual effects. >
Implementation of the TSP elevation repair limits will decrease the number of tubes which must be repaired. Installation of steam generator tube plugs or sleeves reduces the RCS flow margin. Thus, implementation of the IPC will maintain the margin ,
of flow that would otherwise be reduced in the event of increased tube plugging or sleeving.
As discussed previously, Comed has evaluated industry experiences with TSP degradation, eddy current signal distortions, and component misalignment. Eddy current signal distortions at tube support plates will be evaluated to attempt to determin.e the cause of the distortion. A signal distortion alone will not result in reduction in the margin of safety. The foreign unit that experienced the component misalignment was of a significantly different design than the Braidwood Unit 1 and l Byron Unit 1 steam generators. Analysis of the design differences shows that component misalignment of that type is not applicable to Braidwood Unit 1 or Byron Unit l'and, thus, will -
not result in a reduction in the margin of safety. An inspection l was conducted during the last Braidwood Unit 1 and Byron Unit 1 refueling outages which verified this conclusion.
Specific limitatichs, as discussed previously, will be applied to indications at the Locked-Tube Model Intersections which contain dents. These limitations conservatively treat-indications as free-span to ensure that the integrity of the SG tubes is i
maintained consistent with current analyses should tube denting or TSP cracking occur. Application of the 3.0 volt Locked-Tube Model Intersection IPC and the 1.0 volt Free-Span Model Intersection IPC at Braidwood Unit 1 and Byron Unit 1, with the limitations specified, will not result in a reduction in a margin l of safety.
l j Thus, the imp]ementation of this amendment does not result in a significant reduction in a margin of safety.
Therefore, based on the above evaluation, Comed has concluded that these changes involve no significant hazards considerations.
i i
i D-12 i
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 Commonwealth Edison (Comed) has evaluated this proposed amendment and determined that it involves no significant hazards 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 does 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 Comed proposes to amend Braidwood and Byron Technical Specification (TS) 3/4.4.5, " Steam Generators" and the Bases for TS 3/4.4.5.
The changes proposed to TS 3/4.4.5 will renew the 3.0 volt bobbin coil probe, Steam Generator (SG) Tube Support Plate (TSP) Interim Plugging Criteria (IPC) limit for Outside Diameter Stress Corrosion Cracking (ODSCC) indications at hot-leg TSP intersections as described in WCAP-14273, " Technical Support for Alternative Plugging Criteria with Tube Expansion at Tube Support Plate Intersections for Braidwood-l and Byron-1 Model D4 Steam Generators," as supplemented, (Locked-Tube Model Intersections) and approved by the Nuclear Regulatory Commission (NRC) in Amendments 69 and 77 for Braidwood and Byron, respectively.
This proposal will also renew a 1.0 volt IPC to be applied to ODSCC indications at the cold-leg TSP intersections, and specific hot-leg intersections, (Free-Span Model Intersections) in accordance with Generic Letter 95-05, " Voltage-Based Repair D-1 i
I
t Criteria for Westinghouse Steam Generator Tubes Affected By ;
Outside Diameter Stress Corrosion Cracking," August 3, 1995 (GL t 95-05). Administrative changes will also be made to TS 3/4.4.5 and to the Bases for TS 3/4.4.5 to further clarify the proper '
l application of the SG tube plugging and repair criteria described in this amendment request. This renewal will be applicable for Braidwood Unit 1 Cycle 7 and for Byron Unit 1 Cycle 9.
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.
This amendment request proposes to renew the SG tube plugging / repair criteria previously approved by the NRC in l Amendments 69 and 77 to Braidwood and Byron Technical Specifications, respectively.
The previously evaluated applicable accidents are steam generator tube burst and Main Steam Line Break (MSLB). The postulated MSLB outside of containment but upstream of the Main Steam Isolation Valve (MSIV) represents the most limiting radiological condition relative to the IPC. The potential impact on public health and safety as a result of renewing the SG tube interim plugging criteria contained in the current Braidwood and Byron Technical l Specifications is very low as discussed below. Tube burst due to ;
predominantly axially oriented ODSCC at the TSP intersections 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. i i
During accident conditions, i.e., MSLB, the tubes and TSP may move relative to each other. This can expose the crack length portion to free-span conditions. Testing has shown that the burst pressure correlates to the crack length that is exposed to the free-span, regardless of the length that is still contained within the TSP bounds.
Therefore, a more appropriate methodology has been established for addressing leakage and burst considerations. This methodology is based on limiting potential TSP displacements (Locked-Tube Model Intersections) during postulated MSLB events, thus reducing the free-span exposed crack length to minimal levels. The tube expansion process employed in conjunction with this tube plugging criteria is designed to provide postulated TSP displacements that result in negligible tube burst probabilities i due to the minimal free-span exposed crack lengths. The tube !
expansions were performed during the first outage that the 3.0 i volt IPC was applied (Braidwood refuel outage A1R05 - Fall 1995, and Byron midcycle outage BlP02 - Fall 1995) . These expansions will be inspected in accordance with an eddy current inspection D-2 l
probe that is sensitive to axial and circumferentjal indications.
This program will ensure the integrity of the expansions for the ,
' additional cycle of operation. It has been demonstrated that axial indications in the expansicn region will not result in a l reduction of the load carrying capability of the expanded tubes. l Thermal hydraulic modeling was used to determine TSP loading ;
during MSLB conditions. A safety factor was conservatively l 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. t Additional models were used to verify the thermal hydraulic ;
results. ,
Assessment of the tube burst probability for the Locked-Tube .
Model Intersections was based on a conservative assumption that p all hot-leg TSP intersections (32,046) contained through wall l cracks equal to the postulated TSP displacement and that the l l crack lengths were located within the boundaries of the TSP.
Alternatively, it was assumed-that all hot-leg TSP intersections contained through wall cracks with lengths equal to the thickness i 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 through wall indications, given a uniform MSLB TSP displacement of 0.31 inches, was calculated to be lx104 This is a factor of 1000 i less than the GL 95-05 burst probability limit of lx10-2, ;
Therefore, the functional design criteria for tube expansion was 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". This design goal results in a total tube burst probability of lx104 for all 32,046 postulated through wall indications.
Additional tubes were expanded to provide redundancy for the required expansions.
The structural limit for the Locked-Tube Model Intersection SG tube repair criteria was 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. Based on the current voltage ;
distributions and growth rates, Monte Carlo projections were performed for Braidwood Unit 1 and Byron Unit 1 for the l additional cycle of operation that this proposed amendment is j requesting. The End of Cycle (EOC) voltage projections for l l Braidwood Unit 1 Cycle 7 predict that the maximum voltage to be l j seen wil1 be less than 10.5 volts. The number of indications !
- predicted greater than ten volts at the end of Cycle 7 for Braidwood Unit 1 is 0.3. The EOC voltage projections for Byron Unit 1 Cycle 9 predict that the maximum voltage to be seen will
- be less than 13.5 volts. The number of indications predicted
! greater than ten volts at the end of Cycle 9 for Byron Unit 1 is i
- D-3 4
_ ._ _ - . _ _ _ _ _ _ _ - _ ._ ____.___._ _ _ . _ _ _ . = _
. + l 4.59. I Using a tensile rupture probability for a ten volt indication oi ,
3x104, the probability of tensile rupture from the predicted 0.3 I indications at Braidwood is 1-(1-3x104)'3 = 9.0x10". The '
probability of tensile rupture from the predicted 4.59 indications at Byron is 1-(1-3x104)* " = 1.38x104 Both of these ,
probabilities result in a negligible contribution to the total !
burst probability when compared to the 1x10-2 GL 95-05 limit. :
i Cellular corrosion is a more limiting mode of degradation at the j TSPs with respect to affecting the tube structural limit. l Tensile tests that measure the force required to sever a tube with cellular corrosion and uncorroded cross sectional areas are !
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. l This limit meets the Regulatory Guide (RG) 1.121, " Basis for l Plugging Steam Generator Tubes," structural requirements based I upon the normal operating pressure differential with a safety l factor of 3.0 applied. Due to the limited database supporting j this value, the structural limit was conservatively reduced to 20 -
volts. Accounting for voltage growth and Non-Destructive ,
Examination (NDE) uncertainty, the full IPC upper limit exceeds ten volts. However, for added conservatism a single voltage repair limit of 3.0 volts for the Locked-Tube Model Intersection indications is specified in the current plugging / repair criteria. I All indications at the Locked-Tube Model Intersections with -
bobbin coil probe voltages greater than 3.0 volts will be plugged or repaired.
The free-span tube burst probability must be calculated for the indications at the Free-Span Model Intersections. The total burst probability must be within the requirements of GL 95-05.
The free-span structural voltage limit is calculated using correlations from the database described in GL 95-05, with the inclusion of the recent Byron, Braidwood, and South Texas tube pull results. The structural limit for the Free-Span Model Intersections is 4.745 volts. The lower voltage repair limit for the indications at the Free-Span Model Intersections continues to be 1.0 volt. The upper voltage repair limit for the indications at the Free-Span Model Intersections will be calculated in accordance with GL 95-05.
Since IPC will not be applied to indications at the Flow Distribution Baffle (FDB), no leakage or burst analyses are required for these indications.
Per GL 95-05, MSLB leak rate and tube burst probability analyses are required to be performed prior to returning the unit to power. The results of these analyses are to be included in a report to the NRC within 90 days of restart. If allowable limits D-4 )
l
s on leak rates and burst probability are exceeded, the results are to be reported to the NRC and a safety assessment of the significance of the results is to be performed prior to returning the SGs to service.
A site specific calculation has determined the site allowable leakage limit for Braidwood and Byron. These limits use the recommended Dose Equivalent 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 leakage rate calculation methodology 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 the 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 the leak rate, as described in GL 95-05. The database used for the leak and burst correlations is consistent with that described in GL 95-05 with the inclusion of the Byron Unit 1, Braidwood Unit 1, and South Texas 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 NDE uncertainty.
The Electric Power Research Institute (EPRI) leak rate correlation has been used. This 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 for the Locked-Tube Model Intersections, the tube is constrained by the TSP and tube burst is precluded.
However, the flanks cf the crack may open up to contact the l Inside Diameter (ID) of the TSP hole and result in a primary-to- i secondary leak rate potentially exceeding that obtained from the l 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 (6.0 gallons per minute). The bounding leak rate value was then applied to a leak rate calculation methodology that accounts for the MSLB leak rate i contribution from IRB indications to the total 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 GL 95-05. When this is done, the dose at the site boundary D-5
. _ _ __ . _ _ _ _ - _ _ _ . . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ m . ._ _ _ _ _ _~
W resulting from the predicted leakage will be a small fraction [
(less than 10%) of the 10 CFR 100 limits. l Modification of the Braidwood and Byron TS to clarify application I of the proposed tube plugging / repair criteria is purely l administrative and will not have any effect on the probability or !
l consequences of an accident previously evaluated. ,
l t
i Operating experience over the last cycle with this plugging !
criteria applied has not revealed any unpredicted or unusual l
effects.
l For these reasons, renewal of the current Braidwood and Byron :
tube plugging criteria does not adversely affect SG tube j integrity and results in acceptable dose consequences. By '
effectively eliminating tube burst at the Locked-Tube Model TSP intersections, the likelihood of a tube rupture is substantially [
reduced and the probability of occurrence of an accident i previously evaluated is reduced.
This conclusion is not affected by foreign or domestic plant SG experiences (NRC Information Notice 96-09 and its supplement). ;
~
l As the following evaluation shows, these experiences are not relevant to Braidwood or Byron.
I l
A foreign unit detected eddy current signal distortions in one :
area.of the top TSP during a 1995 inspection. The steam j generators had been chemically cleaned in 1992. Visual ;
inspection showed that a small section of the top TSP had broken .
free and was resting next to the steam generator tube bundle !
l wrapper. The support plate showed indications of metal loss. !
l t The chemical cleaning process used by the foreign unit was developed by the utility and differs significantly from the modified EPRI/SGOG process performed at Byron Unit 1 in 1994.
The foreign chemical cleaning process, coupled with the specific i l application of the process, resulted in TSP corrosion of up to 250 mils compared to a maximum of 2.16 mils (11 mils maximum ,
allowed) measured at Byron. During the Byron eddy current inspection performed after the chemical cleaning, no distortion
- j of the tube support plate signals was reported. Therefore, these differences in cleaning processes imply that this foreign experience is irrelevant to the effects of the chemical cleaning j process on the TSPs at Byron. Chemical cleaning of the SGs has !
not occurred at Braidwood.
A number of units have experienced TSP cracking associated with severe tube denting due to TSP corrosion at the tube-to-TSP crevice. WCAP 14273, Section 12.4, shows that a diametral ,
- reduction of a SG tube of 0.065 inches is required to develop ;
i stress levels above yield in the TSP ligaments at dented 1 intersections. The bobbin voltage range associated with a one
?
4 D-6 A
I
mil radial dent is twenty to twenty-five volts.
l Although Braidwood Unit 1 and Byron Unit 1 have not seen !
l corrosion induced denting, a 0.610 inch diameter bobbin coil
- picbe will be used as a go/no-go gauge to assess dents at the i i
.ked-Tube Model Intersections, if they occur in the future. If ;
l o tube has a dent at a Locked-Tube Model TSP intersection that fails to pass the go/no-go test probe, IPC will not be applied to that intersection. In addition, if the dent is determined to be corrosion induced, the Free-Span Model repair criteria will be
, applied to the intersections adjacent to the dented intersection.
l IPC repair limits will not be applied to tuoes with dents greater i
than 5.0 volts since dent sigaals of this magnitude could mask a .
l.0 volt ODSCC signal. Tube intersections with corrosion induced i dents greater than 5.0 volts and the intersections adjacent to i such an intersection were not selected for tube expansion to '
preclude adverse effects of the failure of such a tube on ,
limiting TSP displacement. If corrosion induced denting, either !
greater than 5.0 volts or such that the tube is unable to pass a O.610 inch diameter bobbin coil probe, are detected at an intersection adjacent to an expanded intersection, the dented I intersection will be inspected by an EPRI developed technique to l
- determine if the TSP is cracked. If a crack-like indication is 1 l identified in a TSP, a plus point inspection will be conducted per the EPRI TSP program. If the plus point inspection verifies )
- the existence of a crack-like indication, the effect of that l indication on TSP displacement will be evaluated. If this evaluation shows that TSP displacement would be greater than 0.1 inches during a MSLB event, the effected area will either be mechanically corrected or the Free-Span Model criteria will be applied to the affected area. Based on the information presented above, the SG tube denting experience at other plants is not relevant to Braidwood or Byron.
l A foreign utility's SGs have experienced cracking at the top TSP.
! The cause of the cracking appears to be the configuration of the j
single anti-rotation device, connected between the SG shell and i wrapper, and the wrapper internals. The single anti-rotation device carries the full load associated with the wrapper to shell motion. This rotational load is believed to be transferred to l the TSP via the wrapper internals. The Byron /Braidwood Unit 1 SG i design (D-4) uses three anti-rotation devices to spread the l rotational load. The D-4 wrapper internals are configured such that this load is not directly transmitted to the TSP.
No top TSP cracking has been detected at Braidwood Unit 1 or j Byron Unit 1 and very few (<1%) of the ODSCC indications in the SG tubes at Braidwood and Byron, to date, have been at the top TSP elevation. Nevertheless, an analysis was performed to assess the impact of cracking of the top TSP. The results show an increase in the deflection of the top TSP for a very limited number of tubes to greater than the 0.10" limit used in the 3.0 D-7 l
4 volt IPC analysis. The deflections of the lower support plates also increased, but remain within the 0.10" limit. Thus, a large majority of the Locked-Tube Model indications continue to be bounded by the existing analysis even with a cracked top TSP.
The Locked-Tube Model repair criteria will not be applied to any SG tube ODSCC indication where the TSP has been shown to be displaced by more than 0.1 inches during accident conditions.
In response to these experiences at foreign and domestic utilities, Comed developed an inspection plan for the SG internals to identify if indications detrimental to the load path components existed. This inspection plan was carried out at Braidwood during refueling outage A1R05 (Fall 1995) and at Byron during the midcycle outage B1P02 (Fall 1995) and refuel outage BIR07 (Spring 1996) . These inspections revealed no degradation of the SG load path components necessary to support implementation of the 3.0 volt IPC. Inspections will be performed during the upcoming refuel outages at Braidwood Unit 1 and Byron Unit 1 to further ensure the integrity of the SG load path components necessary to support implementation of the 3.0 volt IPC.
A domestic utility reported several distorted TSP signals over the past three refueling outages' SG tube inspections. It was determined that these signals were associated with the TSP geometry in an area where an access cover is welded to the TSP.
These signal distortions are not attributed to TSP cracking or degradation. Since the distorted signals were due to TSP i geometry which did not indicate or result in a defect of the TSP, !
there is no increase in the probability or consequences of an accident previously evaluated due to Braidwood Unit 1 and Byron Unit 1 steam generator TSP geometries which may result in l distorted eddy current signals.
One foreign unit observed a dislocation of the tube bundle wrapper when they were unable to pass sludge lancing equipment through a hand hole in the wrapper. The dislocation appears to be a result of improper attachment of the wrapper to the support i structure. SG sludge lance operations have been successfully 1 performed at Braidwood Unit 1 and Byron Unit 1 which indicates that no problem with the wrapper attachment exists. The foreign unit's wrapper support design is significantly different than that used on Braidwood Unit 1 and Byron Unit 1. Therefore, a similar wrapper dislocation will not occur and the foreign i experience is not applicable to Braidwood or Byron. An inspection was conducted during the last Braidwood Unit 1 and Byron Unit 1 refueling outages which verified this conclusion.
Comed will continue to apply a maximum primary-to-secondary leakage limit of 150 gallons per day (gpd) through any one SG at Braidwood and Byron to help preclude the potential for excessive leakage during all plant conditions. The RG 1.121 criterion for D-8
O !
}
establishing operational leakage limits that require plant !
shutdown are based on detecting a free-span crack prior to it !
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 I associated with the longest permissible free-span crack length. !
Therefore, the proposed amendment does not result in any ;
significant increase in the probability or consequences of an accident previously evaluated within the Braidwood and Byron {
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 j
evaluated.
This amendment request proposes to renew the SG tube !
plugging / repair criteria previously approved by the NRC in ;
Amendments 69 and 77 to Braidwood and Byron Technical ;
Specifications, respectively. i Renewal of the proposed steam generator tube plugging criteria !
with tube expansion does not introduce any significant changes to i the plant design basis. Use of the criteria does not provide a mechanism which could result in an accident outside of the region ,
i of the tube support plate elevations as ODSCC does not extend ,
beyond the thickness of the tube support plates and IPC is not !
allowed to be applied to indications that extend beyond the ;
thickness of the tube support plate. Neither a single nor '
multiple tube rupture event would be expected in a SG in which the plugging criteria has been applied.
The tube burst assessment involves a Monte Carlo simulation of I I
the site specific voltage distribution to generate a total burst probability that includes the summation of the probabilities of one tube bursting, two tubes bursting, etc. For the Locked-Tube Model TSP Intersections, the maximum total probability of burst, by design, is estimated to be 1x10" with all tube expansions functional. The burst probability for the Free-Span Model TSP intersections will be dependent on the number and size of indications at these applicable intersections. The total burst probability will be within the limit specified in GL 95-05.
Accounting for the unlikely event of a failure of the expanded tubes, 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 through wall indications. This level of burst probability is considered to be negligible when compared to the GL 95-05 limit of 1x10-2, D-9 I
. i
- i In addressing the combined effects of a Loss Of Coolant Accident (LOCA) during a Safe Shutdown Earthquake (SSE) on the SG as required by General Design Criteria (GDC) 2, it has been- ;
determined that tube collapse may occur in the steam generators '
at some plants. The tube support plates may 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 i rarefaction wave and SSE loadings. The resulting pressure j differential on the deformed tubes may cause some of the tubes to :
collapse. There are two issues associated with SG tube collapse. '
First, the collapse of SG tubing reduces the Reactor Coolant System (RCS) flow area through the tubes. The reduction in flow l area increases the resistance to flow of steam from the core during a LOCA which, in turn, may potentially increase the Peak Clad Temperature (PCT). Second, there is a potential that partial through wall cracks in the SG tubes could progress to through wall cracks during tube deformation or collapse. The tubes subject to collapse have been identified via a plant specific analysis and are excluded from application of any l voltage-based criteria. This analysis is included in revision 3 to WCAP-14046 which was submitted to the NRC June 19, 1995.
Modification of the Braidwood and Byron Technical Specifications to clarify application of the proposed tube plugging / repair criteria is purely administrative and will not create the I possibility of a new or different kind of accident from any accident previously evaluated.
1 Operating experience over the last cycle with this plugging l criteria applied has not revealed any unpredicted or unusual i effects. I SG tube integrity will continue to be maintained following renewal of the 3.0 volt IPC voltage repair limit through inservice inspection, tube repair and primary-to-secondary leakage monitoring. By effectively eliminating tube burst at the Locked-Tube Model TSP Intersections, the potential for multiple tube ruptures is essentially eliminated.
Comed has evaluated industry experiences with TSP degradation, eddy current signal distortions, and component misalignment.
Eddy current signal distortions due to TSP geometry are not indicative of TSP degradation and do not result in any kind of new or different accident.
The component misalignment experienced by one unit is not applicable to Braidwood Unit 1 or Byron Unit 1 and, thus, will not result in any kind of new or different accident. Specific limitations, as discussed in response to Question 1, will be ;
applied to indications at the Locked-Tube Model Intersections l which contain dents. These limitations ensure that the integrity l of the SG tubes is maintained consistent with the current D-10
o analyses should tube denting or TSP cracking occur. i Therefore, renewal of the current tube plugging / repair criteria at Braidwood Unit 1 and Byron Unit 1 will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The proposed change does not involve a significant reduction in a margin of safety. j l
The use of the voltage-based, bobbin coil, tube support plate l plugging criteria with tube expansion at Braidwood Unit 1 and Byron Unit 1 is demonstrated to maintain SG 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, i and 32 by reducing the probability or the consequences of steam l generator tube rupture.
Reducing the probability or the consequences of steam generator tube rupture is accomplished by determining an eddy current inspection voltage value which represents a limit for leaving an axial, crack-like indication at an in service SG tube TSP intersection. Tubes with ODSCC voltage indications beyond this limiting value must be removed from service by plugging or repaired by sleeving. Implementation of a 3.0 volt IPC voltage repair limit for the Locked-Tube Model Intersections has been evaluated and shown not to present a credible potential for a steam generator tube rupture event during normal or faulted plant conditions, even with worst case assumptions. The total tube burst probability will include a contribution from the indications at the Locked-Tube Model Intersections and from indications at the Free-Span Model Intersections. The projected EOC voltage distribution of crack-like indications at the TSP elevations will be confirmed to result in acceptable primary-to-secondary leakage during all plant conditions such that radiological consequences are not adversely impacted.
Addressing RG 1.83 considerations, implementation of the increased Locked-Tube Model 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 TSP elevations.
For the leak and burst assessments, the population of indications in the EOC voltage distribution is dependent on the POD function.
The purpose of the POD function is to account for new indications that may develop over the cycle, and to account for indications not identified by the data analyst. In implementing this proposed IPC renewal, Comed will continue to use the conservative GL 95-05 POD value of 0.6 for all voltage amplitude ranges.
D-11
!n
- Modification of the Braidwood and Byron Technical Specifications to clarify application of the proposed tube plugging / repair I criteria is purely administrative and will not reduce any safety margins. :
Operating experience over the last cycle L, ' 1 is plugging !
criteria applied has not revealed any unpreo. ec or unusual l effects.
l l Implementation of the TSP elevation repair limits will decrease {
l the number of tubes which must be repaired. Installation of !
steam generator tube plugs or sleeves reduces the RCS flow margin. Thus, implementation of the IPC.will maintain the margin l of flow that would otherwise be reduced in the event of increased l tube plugging or sleeving.
As discussed previously, Comed has evaluated industry experiences !
l with TSP degradation, eddy current signal distortions, and component misalignment. Eddy current signal distortions at tube support plates will be evaluated to attempt to determine the l cause of the distortion. A signal distortion alone will not l result in reduction in_the margin of safety. The foreign unit L that experienced the component misalignment was of a l significantly different design than the Braidwood Unit 1 and i Byron Unit i steam generators. Analysis of the design differences shows that component misalignment of that type is not
- applicable to Braidwood Unit 1 or Byron Unit 1 and, thus, will not result in a reduction in the margin of safety. An inspection was conducted during the last Braidwood Unit 1 and Byron Unit 1 l refueling outages which verified this conclusion.
i Specific limitations, as discussed previously, will be applied to
! indications at the Locked-Tube Model Intersections which contain l dents. These limitations conservatively treat indications as
! free-span to ensure that the integrity of the SG tubes is L maintained consistent with current analyses should tube denting l or TSP cracking occur. Application of the 3.0 volt Locked-Tube l Model Intersection IPC and the 1.0 volt Free-Span Model ,
Intersection IPC at Braidwood Unit 1 and Byron Unit 1, with the l limitations specified, will not result in a reduction in a margin L of safety.
l Thus, the implementation of this amendment does not result in a j significant reduction in a margin of safety. '
Therefore, based on the above evaluation, Comed has concluded that these changes involve no significant hazards considerations.
l 2
l D-12 l
l _. _ _
[ l i
, e l
! o l
ATTACHMENT E l
ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 1
l l
Commonwealth Edison Company (Comed) has evaluated this proposed l license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with Title 10, Code of Federal l Regulations, Part 51, Section 21 (10 CFR 51.21). Comed has i determined that this proposed license amendment request meets the l criteria for a categorical exclusion set forth in !
10 CFR 51. 22 (c) ( 9 ) . This determination is based upon the j following: j
- 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 cr 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) 3/4.4.5, " Steam Generators" and the Bases for 3/4.4.5. 1 i
l The changes proposed to TS 3/4.4.5 will renew the 3.0 volt bobbin coil probe, Steam Generator (SG) Tube Support Plate (TSP) Interim Plugging Criteria (IPC) limit for Outside Diameter Stress Corrosion Cracking (ODSCC) indications at hot-leg (Locked-Tube Model) TSP intersections as approved by the Nuclear Regulatory Commission (NRC) in Amendments 69 and 77 for Braidwood and Byron, respectively.
This renewal will also continue to require a 1.0 volt IPC be applied to ODSCC indications at the cold-leg and specific hot-leg TSP intersections (Free-Span Model),
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 (GL 95-05). Administrative E-1 l
! a
. 3 o !
changes will be made to TS 3/4.4.5 and to the Bases for !
3/4.4.5 to clarify the proper application of the SG i tube plugging and repair criteria described in this l amendment request. This renewal will be applicable for Braidwood Unit 1 Cycle 7, and for Byron Unit icle 9. ,
- 2. This proposed license amendment request involves no '
significant hazards considerations as demonstrated in i Attachment D;
- 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 environmental impact statement nor an environmental assessment is necessary for ,
i this proposed license amendment request. !
l I
E-2