ML20203B723
| ML20203B723 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 02/03/1998 |
| From: | Lynch D NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20203A279 | List: |
| References | |
| NUDOCS 9802240404 | |
| Download: ML20203B723 (20) | |
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'a UNITED STATES i
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30006 4001 1
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COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT No.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 92 License No. NPF-72 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated February 28,1997, as supplemented on November 13,1996, and March 20, June 24, August 19 and November 3,1997, complies with the 4
standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the i
Commission's regulations; i
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachmerit to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:
i 9802240404 900203 PDR ADOCK 05000456 P.
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ItGbdGaLfine.drsallena The Technical Specifications contained in Appendix A as revised through Amendment No. 92 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date c/ its issuance and shall be implemented in the first operating cycle after installation of the Babcock and Wilcox, Intemational replacement steam generators.
FOR THE NUCLEAR REGULATORY COMMISSION
/
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M. David Lynch, Seni ject Manager Project Directorate Ill-2 Division of Reactor Projects -lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 3, 1998 4
. pu.y p-4 UNITED STATES y
NUCLEAR RECULATORY COMMISSION WAsHINeToN, D.C. seseHe01
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COMMONWEALTH EDISON COMPANY DOCKET N.GLPTN 50457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 92 License No. NPF 77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated February 28,1997, as supplemented on November 13,1996, and March 20, June 24, August 1g and November 3,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; i
B.
The facility will operate in conformity with the application, the provisions of the i
Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will 5e conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordjince with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:
4
2-(2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 92 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date if its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION gA_
M. David Lynch, Senior Project Manager Project Directorate 111-2 Division of Reactor Projects - Ill/lV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications 4
Date of issuance: February 3,1998 a
ATTACHMENT TO LICENSE AMENDMENT NOS. 92 AND 92 FACILITY OPERATING LICENSE NOS. NPF 72 AND NPF-77 DOCKET NOS. STN $0 456 AND STN 50-457 Replace the fcFowing pages of the Appendix "A" Tenical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Pages insert Pages Vill Vill 3/4 4-14 3/4 4-14 3/4 4-15 3/4 4-15 3/4 4-16 3/4416 3/4417 3/4417 3/4 4-17c 3/4 4-17c 3/4 4-17d 3/4 4-17d 3/4 4-27 3/4 4-27 l
3/4 4-28 3/4428 3/4 4-29 3/4429 3/4 4-29a 3/4 4-31 3/4 4-31 B 3/4 4-3 8 3/4 4-3 8 3/4 4-3b B 3/4 4-3b B W4 &5 B W4 &5 G
'f LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION HER TABLE 3.4-2 REACTOR OOOLANT SYSTEM CHEMISTRY LIMITS...............
3/4 4-25 i
l TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIaExENTS........................................
3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY........................................
3/4 4-27 FIGURE 3.4-1 UNIT 1 (AFTER CYCLE 7s AND UNIT 2 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1 #CI/ GRAM DOSE EQUIVALENT I-131....................................
3/4 4-29 FIGURE 3.4-2 UNIT 1 THROUGH CYCLE 7 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >0.35 pCI/ GRAM DOSE EQUIVALENT I-131.........................
3/4 4-29a TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND l
ANALYSIS PROGRAM....................................
3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...................................
3/4 4-32 4
4 Pressurizer..............................................
3/4 4-33 Overpressure Protection Systems..........................
3/4 4-34 3/4.4.10 STRUCTURAL INTEGRITY.....................................
3/4 4-36 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................
3/4 4-37 4
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4 BRAIDWOOD - UNITS 1 & 2 VIII AMENDMENT NO. 92
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 1)
All tubes that previously had detectable tube wall pen'etrations greater than 20% that have not been plugged or sleeved in the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, 2)
Tubes in those areas where experience has indicated potential
- problems, 3)
A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube.
If any selected tube does not s
permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection, 4)
For Westinghouse Model D4 steam generators, indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages, and l
5)
For Westinghouse Model 04 steam generatprs, tubes which remain in l
service due to the application of the F criteria will be inspected, in the tubesheet region, during all future outages.
c.
The tubes selected as the second and third samples (if required by Table 4.4-R) 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 i
f 2)
The inspections include those portions of the tubes where imperfections were previously found.
d.
For Unit 1 Cycle 7, implementation of the steam generator tube / tube support plate repair criteria requires a 100-percent bobbin coil inspection for hot-leg and cold-leg tube support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length, e.
For Westinghouse Model D4 and D5 steam generators, 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 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 in the Byron and Braidwood Steam Generators to demor. strate acceptable structural integrity.
BRAIDWOOD - UNITS 1 & 2 3/4 4-14 AMENDMENT NO. 92
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Insoection Freauenci n - 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 riticality or initial operation following a steam generator
< slacement. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor enre than 24 calendar months after the previous inspection.
If two consecutive inspections, not including the )reservice inspection, result in all inspection results falling into tie C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradatic.n has occurrad, the inspection interval may be extended to a maximum of once per 40 months; b.
If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and c.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1)
Reactor-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2c., or 2)
A seismic occurrence greater than the Operating Basis Earthquake, or 3)
A Condition IV loss-of-coolant accider.t 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 AMENDMENT NO.92 l
REACTOR CCCI.hMT SYSTEM SURVIILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance criteria a.
As used in this specification 1)
Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawlags or specifications.
Bddy-current testing indications below 20% of the nominal tube or sleeve well thickness, if detectable, may be considered as Luperfections; 2)
Dearadation means a service-induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeves 3)
Dearaded 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) t Dearadation means the percentage of the tube or sleeve wall thickness affected or removed by degradation; 5)
Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube or sleeve containing an unrepaired defect is defective; 6)
Pluaaina or Renair Limit means the imperfection depth at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the affected area.
The plugging or repair limit imperfoetion depth for tubing is equal to 40% of the nominal wall thickness.
For Westinghouse Model D4 and D5 steam generators, the plugging or repair limit imperfection depth for laser wulded sleeves is equal to 40% of the nominal sleeve wall thicknssa, and for TIG welded sleeves is equal to 32% of the nominal sleeve wall thickness.
For Westinghouse Model D4 steam generators, this definition does not apply to defects in the tubesheet that meet the criteria for an F* tube.
For Unit 1 Cycle 7, this definition does not apply to the tube support plate intersections for which the voltage-based repair criteria are tcing applied, fafer to 4.4.5.4.m.13 for the repair limit applicable to the so intersections;
")
Unserviceable describes the ca ctition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operatang Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified la 4.4.5.3c., above; 8)
Tube Inson.ction means an inspection of the steam generator tube from the *:aint of entry (hot leg side) completely around the U-band 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 BRAIDWOOD - UNITS 1 & 2 3/4 4-16 AMENDMENT NC. 92
REACTOR COOLANT SYSTEM SURVElllANCE RE0VIREMENTS (Continued) 9)
Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
- 10) Tube Reoair refers to a process that reestablishes tube serviceability. Acceptable tube repairs for Westinghouse Model D4 or D5 steam generators will be performed by the following processes:
a)
Laser welded sleeving as described in a Westinghouse Technical Report currently approved by the NRC, subject to the limitations and rastrictions as noted by the NRC staff, or b)
TIG welded sleeving as described in ABB Combustion Engineering Inc. Technical Reports:
Licensing Report CEN-621-P, Pevision 00, " Commonwealth Edison Byron and Braidwood Unit I 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 Performance of the ABB CEN0 Steam Generator Tube Sleeve for Use at Commonwealth Edison Byron and Braidwood Units I and 2," January 1996, subject to the limitations and l
restrictions as noted by the NRC Staff.
Tube repair includes the removal of plugs that were previously installed as a corrective or preventative measure.
A tube inspection per 4.4.5.4.a.8 is required prior to returning I
previously plugged tubes to service, 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 pl:te displacement less than 0.1 inches during accident conditions, excluding the following:
a)
All tube-to-tube support plate intersections where IPC can r.ot 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.
BRAIDWOOD - UNITS 1 & 2 3/4 4-17 AMENDMENT N0. 92
REAC' TOR COOLANT SYSTEM BURVEfLLhMCE REOUIREMENTs fContinuadi Note 2:
The upper voltage repair limit for indications of outside diameter stress corrosion cracking occurring in the Free-span Model intersections is calculated according to the methodology in Generic Letter 95-05 as supplemented.
14)
F Distance is the distance into the tubesheet of a Westinghouse Model D4 steam generator from the secondary face of the tubesheet or the top of the last hardroll, whichever is further into the tubesheet, that has been determined to be 1.7 inches.
15)
F Tube is a Westinghouge Model D4 steam generator tube with l
degradation below the F distance and has no indications of degradation (i.e., no indication of cracking) within the F*
distance. Defects contained in an F* tube are not dependant on flaw geometry.
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 Within 15 days following the completion of each inservice inspection a.
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 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 hoplementation of the voltage based repair criteria to tube support plate intersections for Unit 1 through cycle 7, notify the staff prior to returning the steam generators to service should any of the following conditions arises BRAIDWOOD - UNITS 1 & 2 3/4 4-17c AMENDMENT NO. 92
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 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 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 corrosion cracking.
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.
7)
If the calculated conditional burst probability based on the projected end-of-cycle (or if not. practical, using the actu measured end-of-cycle) voltage distribution exceeds 1 x 10',a1 notify the NRC and provide an assessment of the safety significance of the occurrence.
8)
Following a steam generator internals inspection, if indications detrimental to the integrity of the load path necessary to support the 3.0 volt IPC are found, notify the NRC and provide an assessment of the safety significance of the occurrence.
e.
The results of inspections of Westinghouse Model D4 steam generators l
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 2)
Location and size of the degradation.
BRAIDWOOD - UNITS 1 & 2 3/4 4-17d AMENDMENT N0. 92
REACTOR COO 1 ANT SYSTEM 3/4.4.8 EPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to a.
Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131**, and Lessthanorequalto100/5microcuriespergramofgross b.
radioactivity.
APPLICABILITY: Moots 1, 2, 3, 4, and 5.
ACTION:
MODES 1, 2 and 3**
l a.
With the specific activity of the reactor coolant greater than 1 microcurin per gram DOSE EQUIVALENT I"131** for more Wan 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on 1***
than 500, Figure 3.4-1, be in at least HOT STANDBY with Tag F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and b.
Withthespecificactivityofthereactorcoolantgreaterthan100/5 microcuries per gram, he in at least HOT STANDBY with T less than en 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, s
- With T., greater than or equal to 500*F.
- For Unit 1 through Cycle 7, reactor coolant DOSE EQUIVALENT I-131 will be l
limited to 0.35 microcuries per gram.
BRAIDWOOD - UNITS 1 & 2 3/4 4-27 AMENDMENT No. 92 4
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REAC'!OR COOIANT SYSTEM I
LIMITING CONDITION FOR OPERATION ACTION icontinuedt j
MODES 1, 2, 3, 4, and 5:
I with the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131* or greater than 100/E microcuries per gram, perform the sampling and analysis requirwaents of Item 4.a) of Table 4.4-4 until the specific activity of the reactor coolan+. is restored to within its limits.
i i
i BURVEf t.1.ANCE REOUIREMENTS 4
a 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of 4
Table 4,4-4.
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- For Unit I through cycle 7, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 microcuries per gram.
BRAIDWOOD - UNITS 1 & 2 3/4 4-23 AMENDMENT NO. 92
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BRAIDWOOD - UNITR 1 & 2 3/4429 AMENDMENT NO. 92 L
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BRAIDWOOD - LINITS 1 & 2 3/4 4 29a AMENDMENT NO. 92 n..m_,,,
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TABLE 4.4-4 (Continued)
TABLE NOTATIONS Until the specific activity of the Reactor Coolant System is restored within its limits.
Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radiciodines. The total specific activity shall be the sum of the dogassed beta-gamma activity and the total of all identified gaseous activitime in the sample within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the sample is taken and extrap91sted back to when the sample was taken. Determination of the contr.ibutors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest ave.ilable data may be used for pure beta-emitting radionuclides.
A radiochemical analysis for E shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radio-iodines, which is identified in the reactor coolant. The specific activities for these individual radionuclides shall be used in the determination of E f9r the reactor coolant sample. Determination of the contributors to E shall be besed upon these energy peaks identifiable with a 95% confidence level.
For Unit 1 through cycle 7, reactor coolant DOSE EQUIVALENT I-131 will bu limited to 0.35 microcuries per gram.
f BRAIDWOOD - UNITS 1 & 2 3/4 4-31 AMENDKENT NO.92
j mEacTom cooLauf arsTEM i
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4 i
l 3/4.4.5 STEAM GEMEEAMRs l
The surveillance Requirements for inspection of the steam generator tubes l
ensure that the structural integrity of this portion of the RCS will be F
maintained. The program for inservice inspection of steam generator tubes is
[
based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain
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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 i
of steam generator tubing also provides a means of characterising the nature and cause of any tube degradation so that-corrective measures can be taken.
The plant is espected to be operated in a manner such that the secondary-l coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant i
4 l
chemistry is not maintained within these limits, localised 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 generater). Cracks having a reactor-to-secondary leakage less than this limit during operation will have l
an adequate margin of safety to withstand the loads imposed during normal l
operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 150 gallons per day per steam generator can j
readily be detected by radiation monitors of steam generator blowdown,
[
mainsteam lines, or the steam jet air ejectors.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired by sleeving. The technical bases for sleeving are described in the current Westinghouse or ABB Combustion Engineering, Inc. Technical Reports whb:h are applicable to l_
Westinghouse Model D4 and D5 steam generators only.
t Wastage-type defects are unlikely with proper chemistry treatment of the l
secondary coolant. However. even if a defect should develop in service, it j
will be found during scheduled inservice steam generator tube examinations.
Plugging or sleeving will be required for all tubes with imperfections 4
exceeding the plugging or repair limit of 40% of the tube nominal wall j
tpickness, excluding defects that meet the criteria for Westinghouse Model D4
{
F - tubes. Acceptable plugging criteria for Westinghouse Model D4 and D5 i
sleeved tubes are: 1) a laser welded sleeved tube must be plugged if a through wall penetratiol is detected in the sleeve that is equal to or. greater that 40%
]
of the nominal sleeve thickness, and 2) TIG welded sleeved tubes must be l
1 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 j
the sleeve is derived from Reg. Guide 1.121 analysis and utilises a 20%
i allowance for eddy current uncertainty and additional degradation growth.
(
Inservice inspection of sleeves is required to ensure RCS integrity. Sleeve j
inspection techniques are described in the current Westinghouse or ABB
(
Combustion Engineering, Inc. Technical Reports.
Steam Generator tube and I
sleeve inspections have demonstrated the capability to reliably detect i
degradation of the pressure retaining portions of the tube or sleeve wall thickness. Commonwealth Edison will validate the adequacy of any system that is used for periodic inservice inspection of the sleeves and, as deemed-appropriate, will upgrade testing methods as better methods are developed and j
validated for cosumercial use.
BRAIDWOOD - UNITS 1 & 2 3 3/4 4-3 AMENDMENT NO. 92 3
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM G EERATORS (Continued)
The mid-cycle equation in SR 5.4.a.13.f should only be used during unplanned inspections in which ed'
-arrent data is acquired for indications at the Free-Span Model Intersections.
The voltage repair limit for indications at the Locked-Tube Model Intersections remains at 3.0 volts during unplanned inspections.
SR 4.4.5.5 implements several re>orting requirements recommended by Generic Letter 95-05 for situations witch the NRC wants to be notified prior to returning the SCs to service.
For the purposes of this reporting requirement, leakage and conditional burst )robability can be calculated based on the as-found voltage distribution ratier than the projected enc'-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.
The maximum site allowable primary-to-secondary leakage limit for end-of-i cycle main steamline break conditions ipcludes 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 rpmaining intact steam generators shall include the operational leakage from F For Westinghouse Model D4 steam generators, plugging or repair is not requiredfortubeswithdegradationwithinthetupesheetareawhichfallunder 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 l
2 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 hppendices, for normal operating and faulted conditions.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.9.2 prior to resumption of plant operation.
Such cases will be considered by the Commission or, 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 4-3 b AMENDMENT NO. 92
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)
BASES OPERATIONAL LEAKAGE (Continued) i The Surveillance Requirements for RCS pressure isolation valves provide i
added assurance of valve integrity thereby reducing the probability of gross I
valve failure and consequent intersystem LOCA. Leakage from the RCS pressure Isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of-the allowed limit.
1 4
3/4.4.7 CHEMISTRY l
The limitations on Reactor Coolant System chemistry ensure that corrosion i
of the ReacHF. Coolant System is minimized and reduces the potential for i
Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, I
' chloride, and fluoride limits are time and temperature dependent. Corrosion i.
studies show that operation may be continued with_ contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for i
i the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits l
provides time for taking corrective actions to restore the contaminant concen-j trations to within the Steady-State Limits.
1 The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
I 3/4.4.8 SPECIFIC ACTIVIII i
i The limitations on the specific activity of the reactor coolant ensure i
that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an
- appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-l state reactor-to-secondary steam gs aerator leakage rate of 1-gm.
For Unit 1 through Cycle 7, the limitations on the specific activity of tie reactor i
coolant ensure that the resulting 2-hour off-site doses will not exceed an j
appropriately small fraction of the 10 CFR Part 100 dose guideline values following a Main Steam Line Break accident in conjunction with an assumed steady-state primary-to-secondary steam generator leakage rate of 150 gpd from each enfaulted steam generator and maximem site allowable primary-to-secondary l
1eakage from the faulted steam generator. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Braidwood Station, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-5 AMENDMENT NO. 92 m---
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