ML20101R386
| ML20101R386 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, Marble Hill  |
| Issue date: | 04/12/1996 |
| From: | Dick G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20101R391 | List: |
| References | |
| NPF-37-A-083, NPF-66-A-083, NPF-72-A-075, NPF-77-A-075 NUDOCS 9604160250 | |
| Download: ML20101R386 (24) | |
Text
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t UNITED STATES s
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2000H001
.....,o COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-454 BJRON STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 83 License No. NPF-37 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated May 17, 1995, as supplemented on January 17, March 8, March 18, April 4 and April 9, 1996, complies with the standards and reo.uirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 Comission; 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 Comission'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 accordance with 10 CFR Part 51 of the Comission'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 paragraphs 2.C.(2) and 2.C.(16) of Facility Operating License No.
NPF-37 are hereby amended to read as follows:
j i
- Page 7 is attached, for convenience, for the composite license to reflect this change.
9604160250 960412 PDR ADOCK 05000454 P
1 0 (2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 83 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.
4 (16) Steam Generator Sleevina Corrosion Testina The licensee shall conduct additional corrosion testing to establish the design life for the laser welded sleeved tubes in the presence of a crevice. The corrosion testing shall demonstrate the corrosion resistance for the laser welded joints i
in tubes that bound the material parameters in the steam generators. The corrosion testing results shall be reviewed and accepted by the Nuclear Regulatory Commission prior to the i
Beginning-of-Cycle 9.
If conformance with the requirements of the plant Technical Specifications for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
< Ely' t
George F. Dick, J., Sr. Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Attachments:
1.
License Page 7 2.
Changes to the Technical Specifications Date of Issuance:
April 12,1996
O i
j !
(16). Steam Generator Sleevina Corrosion Testina The licensee shall conduct additional corrosion testing to i
4 i
establish the design life for the laser welded sleeved I
tubes in the presence of a crevice. The corrosion testing shall demonstrate the corrosion resistance for the laser l
welded joints in tubes that bound the material parameters l
in the steam generators. The corrosion testing results shall be reviewed and accepted by the Nuclear Regulatory Commission prior to the Beginning-of-Cycle 9.
If conformance with the requirements of the plant Technical Specifications for tube structural integrity is not i
confirmed, the tubes containing the sleeves in question shall be removed from service.
D.
The facility requires exemptions from certain requirements of Appendices A, E and J to 10 CFR Part 50. These include (a) an 4
exemption from the requirements of Paragraph III.D.2(b)(ii) of Appendix J, the testing of containment air locks at times when containment integrity is not required (Section 6.2.6 of the l
SER), (b) an exemption from GDC-2 of Appendix A, the requirement that structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes (Section 3.10 of SSER #5), (c) an exemption from t
GDC-13 and GDC-17 of Appendix A, the requirement that instrumentation be provided to monitor variables and systems over their anticipated ranges, and the requirement that provisions be included to minimize the probability of losing electric power (Section 9.5.4.1 of SSER #5), (d) an exemption from GDC-19 of Appendix A, the requirement that the control room i
i have adequate radiation protection to permit access and occupancy under accident conditions (Section 6.5.1 of SSER #6),
and (e) an exemption from the requirement of Section IV.F of Appendix E that a full participation emergency planning exercise be conducted within one year before issuance of the first operating license for full power and prior to operation above 5%
of rated power (Section 13.3 of SSER #6). These exemptions are authorized by law and will not endanger life or property or the 4
l AMENDMENT NO. 83 J
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UNITED STATES pe 3
j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 2006H1001
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COMMONWEALTH EDISON COMPANY LOCKET NO. STN 50-455 BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 83 License No. NPF-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated May 17, 1995, as supplemented on January 17, March 8, March 18. April 4 and April 9, 1996, complies with the standards and requ!rements 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 (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 4
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 attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(5) of Facility Operating License No.
NPF-66 are hereby amended to read as follows:
"Page 3a is attached, for convenience, for the composite license to reflect this change.
1 1 i (2)
Technical Snecificatirns The Technical Specifications contained in Appendix A (NUREG-lll3),
as revised through Amendment No. 83 and revised by Attachment 2 to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license. The license: shall spcrate the facility in accordance with the Technical Specifications and E;
the Environmental Protection Plan.
j (5)
Steam Generator Sleevina Corrosion Testina l
The licensee shall conduct additional corrosion testing to i
establish the design life for the laser welded sleeved tubes in the presence of a crevice. The corrosion testing shall 3
demonstrate the corrosion resistance for the laser welded joints
{
in tubes that bound the material parameters in the steam i
generators. The corrosion testing results shall be reviewed and i
accepted by the Nuclear Regulatory Commission prior to the Beginning-of-Cycle 8.
If conformance with the requirements of the plant Technical Specifications for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall i
be removed from service.
I 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMISSION i
j l
Geor
. Dick, J., Sr. Project Nanager l
Project Directorate III-2 Division of Reactor Projects - III/IV i
Office of Nuclear i;eactor Regulation 1
Attachments:
1.
License Page 3a 2.
Changes to the Technical Specifications Date of Issuance: April 12, 1996 i
=.
_ _ _ _ ~
l I
- 3a -
l (3)
Inttial Test Procras Any changes to the Initial Startup Test Program described in Chapter 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.
(4)
Reaulatory Guide 1.97. Revision 2 Como11ance The licensee shall submit by March 1, 1987, a preliminary i
report describing how the requirements of Regulatory Guide i
1.97, Revision 2 have been or will be met. The licensee shall submit by September 1,1987, the final report and a schedule for implementation (assuming the NRC approves the DCRDR by March 1, 1987).
(5)
Steam Generator Sleevina Corrosion Testina The licensee shall conduct additional corrosion testing to establish the design life for the laser welded sleeved tubes in the presence of a crevice. The corrosion testing shall demonstrate the corrosion resistance for the laser welded l
3 i
joints in tubes that bound the material parameters in the steam generators. The corrosion testing results shall be 4
reviewed and accepted by the Nuclear Regulatory Commission i
prior to the Beginning-of-Cycle 8.
If conformance with the requirements of the plant Technical Specifications for tube i
structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service.
I h.
2 1
1 i
AMENDMENT NO. 83 l
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ATTACHMENT TO LICENSE AMENDMENT NOS. 83 AND 83 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
j Remove Paaes Insert Paaes 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 8 3/4 4-3 B 3/4 4-3 i
{-
REACTOR COOLANT SVSTEM SURVEILLANCE REOUIREMENTS (Continued) i 1)
All tubes that previously had detectable tube wall penetrations greater than 20 percent that have not been plugged or sleeved in j
the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, 4
2)
Tubes in those areas where experience has indicated potential
- problems, 1
3)
A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be perforised on each selected tube.
If any selected tube does not permit the >assage of the eddy current probe for a tube inspection, this shall w recorded and an adja::ent tube shall be selected and i
subjected to a tube inspection, 4)
For Unit 1, indications left in service as a result of application l
i of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages, and l
l 5)
For Upit 1, tubes which remain in service due to the application of l
i the F criteria will be inspected, in the tubesheet region, during j
all future outages.
4 h
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:
i 1)
The tubes selected for these samples include the tubes from those i
areas of the tube sheet array where tubes with imperfections were i
previously found, and l
2)
The inspections include those portions of the tubes where imperfec-tions were previously found.
d.
For Unit 1, through Cycle 8, implementation of the steam generator tebe/ 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 (00 SCC) indications. The determination of the lowest cold-leg tube support plate intersections having 00 SCC indications shall be based on the performance of at least a i
20 percent random sampling of tubes inspected over their full length.
4 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 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 i
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 i
in the Byron and Braidwood Steam Generators to demonstrate acceptable structural integrity.
BYRON - UNITS 1 & 2 3/4 4-14 AMENDMENT NO. 83
~
REACTOR COOLANT SYSTEM SVRVEILLANCE RE0VIREMENTS (Continued)
The results of each sample inspection shall be classified into one of the fo' lowing three categories:
Category Inspection Results C-1 Less than 5% cf 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, l
4 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:
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-i tion.
If two consecutive inspections, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occLrred, 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 inspecticn frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a maximum of once per 40 months; and c.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
1)
Reactor-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2c., or BYRON - UNITS 1 & 2 3/4 4-15 AMENDMENT N0. 83
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 2)
A seismic occurrence greater than the Operating Basis Earthquake, or 7
3)
A Condition IV loss-of-coolant accident requiring actuation of the Engineered Safety Features, or j
4)
A Condition IV main steam line or feedwater line break.
I 4.4.5.4 Accentance Criteria a.
As used in this specification:
1)
Imoerfection means an exception to the dimension:, finish or i
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 j
detectable, may be considered as imperfections; 2)
Dearadation means a service-induced cracking,
wastage, wear or l
general corrosion occurring on either inside or outside of a tube or sleeve; 3)
Dearaded Tube means a tube or sleeve containing unrepaired imperfections greater than or equal to 20% of the nominal tube i
or sleeve wall thickness caused by degradation; 4)
% 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 j
the plugging or repair limit. A tube or sleeve containing an i
unrepaired defect is defective; l
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 or repair limit imperfection depth for the tubing and laser welded sleeves is i
equal to 40% of the nominal wall thickness. The plugging limit imperfection depth for TIG welded sleeves is equal to 32% of the nominal wall thickness. For Unit 1, this definition does npt apply to defects in the tubesheet that meet the criteria for an F tube; I
For Unit 1, through Cycle 8, this definition does not apply to tube support plate intersections for which the voltage-based plugging criteria are being applied.
Refer to 4.4.5.4.a.11 for the repair limit applicable to these intersections; l
7)
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant i
accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above; BYRON - UNITS 1 & 2 3/4 4-16 AMEN 0 MENT NO. 83 we
l REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) i 8)
LLe Insnection means an inspection of the steam generator tube h
4 from the point of entry (hot leg side) completely around the U-bend i
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 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, i
- 10) Tube Renair refers to a process that reestablishes tube I
serviceability. Acceptable tube repairs 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 j:
limitations and restrictions as noted by the NRC staff, or b)
TIG welded sleeving as described in AB8 Combustion Engineering i
Inc. Technical Reports: Licensing Report CEN-621-P, i
Revision 00, " Commonwealth Edison Byron and Braidwood Unit 1 and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, 4
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 1 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 previously plugged tubes to service.
l 11)
For Unit I through Cycle 8, the Tube Suncort Plate Pluaaina limit is used for the disposition of an alloy 600 steam generator tube i
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 j
diameter stress corrosior, cracking within the bounds of the cold-leg tube support plate with bobbin voltages less than or equal to the lower voltage repair limit [ Note 1] will be allowed to remain in service.
Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the hot-leg tube support plate with bobbin voltages less than or equal to 3.0 volts will be 4
allowed to remain in service.
l BYRON - UNITS 1 & 2 3/4 4-17 AMENOMENT NO. 83
I REACTOR C0OLANT SYSTEM j
BASES i
3/4.4.5 STEAN GENERATORS 4
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-i tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection l
of steam generator tubing is essential in order to maintain surveillance of the 4
conditions of the tubes in the event that there is evidence of mechanical i
damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary 4
coolant will be maintained within those chemistry limits found to result in i
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 i
operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-i 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 i
an adequate margin of safety to withstand the loads imposed during normal l
operation and by postulated accidents. Operating plants have demonstrated that l
reactor-to-secondary leakage of 150 gallons per day per steam generator can l
readily be detected by radiation monitors of steam generator blowdown, l
n;Mnsteam lines, or the steam jet air ejectors.
Leakage in excess of this i
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 l
Combustion Engineering, Ir.c. Technical Reports.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolar.t. 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 i
exceeding the plugging or repair limit of 40% of the tube, nominal wall i
thickness, excluding defects that meet the criteria for F tubes. A laser i
welded sleeved tube must be plugged if a through wall penetration is detected 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 j
the nominal sleeve thickness. The plugging limit for the sleeve is derived i'
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 ensuro RCS integrity. Sleeve inspection techniques are I
described in the current Westinghouse or A88 Combustion Engineering, Inc.
I Technical Reports. Steam Generator tube and sleeve inspections have demonstrated the capability to reliably detect degradation of the pressure l
retaining portions of the tube or sleeve wall thickness. Commonwealth Edison 4
will val'date the adequacy of any system that is used for periodic inservice i
inspection of the sleeves and, as deemed appropriate, will upgrade testing methods as better methods are developed and validated for commercial use.
i, l
BYR6N - UNITS 1 & 2 B 3/4 4-3 AMEN 0 MENT NO. 83
ssM*h p
-d UNITED STATES i
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 206eM001
/
COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO. 1 AMfNDMENT TO FACILITY OPERATING LICENSE Amendment No. 75 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 May 17, 1995, as supplemented on January 17, March 8, March 18, April 4 and April 9,1996, 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; 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 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 accordance with 10 CFR Part 51 ef the Comr.;ission'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 paragraphs 2.C.(2) and 2.C.(6) of Facility Operating License No.
NPF-72 are hereby amended to read as follows:
- Page 4 is attached, for convenience, for the composite license to reflect this change.
) 1 l
(2)
Technical Snecifications The Technical Specifications contained in Appendix A as revised j
through Amendment No. 75 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.
(6)
Steam Generator Sleevina Corrosion Testina The licensee shall conduct additional corrosion testing to establish the design life for the laser welded sleeved tubes in the presence of a crevice. The corrosion testing shall demonstrate the corrosion resistance for the laser welded joints in tubes that bound the material parameters in the steam generators. The corrosion testing results shall be reviewed and accepted by the Nuclear Regulatory Commission prior to the Beginning-of-Cycle 7.
If conformance with the requirements of the plant Technical Specifications for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service.
i 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~
L Ramin R. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Attachments:
1.
License Page 4 2.
Changes to the Technical Specifications Date of Issuance:
April 12. 1996 I
t
- _. -. _. _ =. -
- = - -. -.
4 (4)
Initial Startun Test Procram Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR i
50.59 shall be repert:d in accordance with 50.59(b) within one month of such change.
(5) Reaulatory Guide 1.97. Revision 2 Como11ance The licensee shall submit the final report and a schedule for implementation within six months of NRC approval of the DCRDR.
l (6)
Steam Generator Sleevina Corrosion Testina 4
The licensee shall conduct additional corrosion testing to establish the design life for the laser welded sleeved tubes in l
the presence of a crevice. The corrosion testing shall demonstrate the corrosion resistance for the laser welded joints in tubes that bound the material parameters in the steam generators. The corrosion testing results shall be reviewed and I
accepted by the Nuclear Regulatory Commission prior to the Beginning-of-Cycle 7.
If conformance with the requirements of the plant Technical Specifications for tube structural integrity l
1s not confirmed, the tubes containing the sleeves in question l
shall be removed from service.
D.
The facility requires an exemption from the requirements of Appendix J to 10 CFR Part 50, Paragraph III.D.2(b)(ii), the testing of containment air locks at times when containment integrity is not i
required (SER Section 6.2.6).
This exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security.
This exemption is hereby granted. The special circumstances regarding this exemption are identified in the referenced section of the safety evaluation report and the supplements thereto. This exemption is granted pursuant to 10 CFR 50.12. With this exemption, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
An exemption was previously granted pursuant to 10 CFR 70.24.
The exemption was granted with NRC materials license No. SNM-1938, issued October 8, 1985, and relieved the licensee from the requirement of having a criticality alarm system. Therefore, the licensee is l
exempted from the criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.
E.
The licensee shall implement and maintain in effect all provisions of i
the approved fire protection program as described in the Final Safety Analysis Report, as supplemented and amended, and as approved in the SER dated November 1983 and its supplements, subject to the following provision:
l The licensee may make changes to the approved fire protection program without prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
AMENDMENT NO. 75 4
t@ Cifug ge 4
UNITED STATES s
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20606-4001
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COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 75 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 May 17, 1995, as supplemented on January 17, March 8, March 18, April 4 and April 9, 1996, 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 1; 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 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 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 attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(5) of Facility Operating License No.
NPF-77 are hereby amended to read as follows:
- Page 4 is attached, for convenience, for the composite license to reflect this change.
o (2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised l
through Amendment No. 75 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License j
No. NPF-72, dated July 2, 1987, are hereby incorporated into this i
license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection i
Plan.
j (5)
Steam Generator Sleevina Corrosion Testina j
4 The licensee shall conduct edditional corrosion testing to 4
establish the design life for the laser welded sleeved tubes in the presence of a crevice. The corrosion testing shall demonstrate the corrosion resistance for the laser welded joints in tubes that bound the material parameters in the steam generators. The corrosion testing results shall be reviewed and accepted by the Nuclear Regulatory Commission prior to the 1
l Beginning-of-Cycle 7.
If conformance with the requirements cf the plant Technical Specifications for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service.
3.
This license amendment is effective as of the date if its issuance.
FOR THE NUCLEAR REGULATORY COPMISSION l
Ramin R. Assa, Project Manager Project Directorate III-2 do Division of Reactor Projects - III/IV e
Office of Nuclear Reactor Regulation
(
Attachments:
1.
License Page 4 2.
Changes to the Technical Specifications Date of Issuance:
April 12, 1996 t
r
(4)
Initial Startun Test Procram Any changes to the Initial Test Program described in Section 14 l
of the FSAR made in accordance with the provisions of 10 CFR i
50.59 shall be reported in accordance with 50.59(b) within one month of such change.
i j
(5) Steam Generator Sleevina Corrosion Testina The licensee shall conduct additional corrosion testing to i
establish the design life for the laser welded sleeved tubes in I
the presence of a crevice. The corrosion testilig shall l
demonstrate the corrosion resistance for the laser welded joints l
~
in tubes that bound the material parameters in the steam generators.
The corrosion testing results shall be reviewed and accepted by the Nuclear Regulatory Commission prior to the l
Beginning-of-Cycle 7.
If conformance with the requirements of the plant Technical Specifications for tube structural integrity is not confirmed, the tubes containing the sleeves in question shall be removed from service.
D.
The facility requires an exemption from the requirements of Appendix J to 10 CFR Part 50, Paragraph III.D.2(b)(ii), the testing of containment air locks at times when containment integrity is not required (SER Section 6.2.6).
This exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security.
The staff's environmental assessment was published on May 19, 1988 (53 FR 17995).
This exemption was grtnted in the low power license and is continued for the full power license. The special circumstances regarding this exemption are identified in the referenced section of the Safety Evaluation Report and the supplements thereto.
This exemption is granted pursuant to 10 CFR 50.12. With this exemption, the facility j
will operate, to the extent authorized herein, in conformity with the 1
application, as amended, the provisions of the Act, and the rules and
)
l regulations of the Commission.
An exemption was previously granted pursuant to 10 CFR 70.24. The exemation was granted with NRC materials license No. SNM-1938, issued Octou r 8, 1985, and relieved the licensee from the requirement of having a criticality alans system. Therefore, the licensee is j
exempted from the' criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.
4 i
i AMEN 0 MENT NO. 75
1 ATTACHMENT TO LICENSE AMENDMENT NOS. 75 AND 75' FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457
{
i Replace the following pages of the Appendix "A" Technical Specifications with i
the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Paaes Insert Paaes 3/4 4-14 3/4 4-14 3/4 4-14a 3/4 4-14a i
3/4 4-16 3/4 4-16 3/4 4-17 3/4 4-17 8 3/4 4-3 B 3/4 4-3 l
1 l
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
\\
1)
All tubes that previously had detectable tube wall penetrations greater than 20% that have not been plugged or sleeved in the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, l.
2)
Tubes in those areas where experience has indicated potential l
- problems, d
3)
A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be perfomed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall
>e recorded and an adjacent tube shall be selected and subjected to a tube inspection, t
4)
For Unit 1, indications left in service as a result of application of l
i the tube support plate voltage-based repair criteria shall be i
inspected by bobbin coil probe during all future refueling outages, l
and l
5)
For Unit 1, tubes which remain in service due to the application of l
the F* criteria will be inspected, in the tubesheet region, during 1
]
all future outages.
t The tubes selected as the second and third samples (if required by Table l
c.
i 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
I i
1)
The tubes selected for these samples include the tubes from those j
areas of the tube sheet array where tubes with imperfections were previously found, and 2)
The inspections include those portions of the tubes where j
imperfections were previously found.
d.
For Unit 1 Cycle 6, 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 lowest cold-leg tube suppor(00 SCC) indications. The determination of the l
t plate intersections having 00 SCC indications i
shall be based on the performance of at least a 20 percent random sampling
]
of tubes inspected over their full length.
2 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 installed shall be ins the end of each cycle.pected for axial and circumferential indications at In the event that an imperfection exceeding the repair limit is detected, an additional 20% of the unsampled sleeves shall i
l be inspected, and if an imperfection exceeding the repair limit is detected in the second sample, all remaining sleeves shall be inspected.
l 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 demonstrate acceptable structural integrity, i
i BRAIDWOOD - UNITS 1 & 2 3/4 4-14 AMEN 0 MENT NO. 75
1 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
I i
The results of each sample inspection shall be classified into one of the following three categories:
Cateaory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes l
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 i
degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected 4
2 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 4
in the above percentage calculations.
1 i
v i
4 BRAIDWOOD - UNITS 1 & 2 3/4 4-14a AMENDMENT N0. 75
~
SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance Criteria a.
As used in this specification:
1)
Imoerfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube or sleeve wall thickness, if detectable, may be considered as imperfections; 2)
Dearadation means a service-induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve; 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)
% 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 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 limit 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 6, 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.11 for the repair limit applicable to these intersections; 7)
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural 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 BRAIDWOOD - UNITS 1 & 2 3/4 4-16 AMENDMENT N0. 75
s 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 equipment and techniques expected to be used during subsequent inservice inspections.
- 10) Tube Renair refers to a process that reestablishes tube i
serviceability. Acceptable tube repairs will be performed by the i
following processes:
a)
Laser welded sleeving as described in a Westinghouse Technical Report currently approved by the NRC, subject to the j
limitations and restrictions as noted by the NRC staff, or b)
TIG welded sleeving as described in A88 Combustion Engineering Inc. Technical Reports:
Licensing Report CEN-621-P, Revision 00, " Commonwealth Edison Byron and Braidwood Unit 1 l
and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, l
FINAL REPORT," April 1995, and Licensing Report CEN-627-P, Revision 00-P, " Verification of the Installation Process and i
Operating Performance of the ABB CENO Steam Generator Tube Sleeve for Use at Commonwealth Edison Byron and Braidwood Units 1 and 2," January 1996, subject to the limitations and j
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 i
inspection per 4.4.5.4.a.8 is required prior to returning j
previously plugged tubes to service.
11)
For Unit 1 Cycle 6, the Tube Sunoort Plate Pluaaina Limit is used i
for the disposition of an alloy 600 steam generator tube for j
continued service that is experiencing predominantly axially i
oriented outside diameter stress corrosion cracking confined within i
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.
l a.
Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the cold-leg tube support plate with bobbin voltages less than or equal to the lower voltage repair limit [ Note 1] will be allowed to remain in service. Steam generator tubes, with degradation attributed to outside diameter stress corrosion cracking within the bounds of the hot-leg tube support plate with bobbin voltages less than or equal to 3.0 volts will be allowed to remain in service.
b.
Steam generator tubes with degradation attributed to outside diameter stress corrosion cracking within the bounds of the cold-leg tube support plate with a bobbin voltage greater than the lower voltage repair limit (Note 1), will be repaired or plugged, except as noted in 4.4.5.4.a.11.d below.
I BRAIDWOOD - UNITS 1 & 2 3/4 4-17 AMENDMENT NO. 75
~ _ _ _ _ _ _ _
BASES b
3/4.4.5 STEAM GENERATORS i
j The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-i tained. 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 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 i
errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature j
and cause of any tube degradation so that corrective measures can be taken.
i 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 4
j 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 between the Reactor Coolant System and the Secondary Coolant System (reactor-l to-secondary leakage = 150 gallons per day per steam generator).
Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 150 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown, mainsteam lines, or the steam jet air ejectors.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during 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 i
j Combustion Engineering, Inc. Technical Reports.
i l
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it l
will be found during scheduled inservice steam generator tube examinations.
Plugging or sleeving will be required for all tubes with imperfections i
i exceeding the plugging or rapair limit of 40% of the tube nominal wall thickness, excluding defects that meet the criteria for F* tubes.
A laser i
i 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 4
penetration is detected in the sleeve that is equal to or greater than 32% of l
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.
l Technical Reports.
Steam Generator tube and sleeve inspections have demonstrated the capability to reliably detect degradation of the pressure l
retaining portions of the tube or sleeve wall thickness.
Commonwealth Edison will validate the adequacy of any system that is used for periodic inservice inspection of the sleeves and, as deemed appropriate, will upgrade testing methods as better methods are developed and validated for commercial use.
8 BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 AMENDMENT N0. 75