ML20063M097
| ML20063M097 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 03/04/1994 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | Commonwealth Edison Co |
| Shared Package | |
| ML20063M099 | List: |
| References | |
| NPF-37-A-046, NPF-66-A-046, NPF-72-A-058, NPF-77-A-058 NUDOCS 9403090262 | |
| Download: ML20063M097 (26) | |
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,t UNITED STATES j.gf.')' !
WASHINGTON. D.C. 20566-0001 j
NUCLEAR REGULATORY COMMISSION s.,..,.j COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 46 License No. NPF-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated August 13, 1993, as supplemented by letters dated September 15, 1993, September 16, 1993, December 17, 1993, January 19, 1994, February 11, 1994, and February 24, 1994, 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 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:
9403090262 940304 PDR ADOCK 05000454 P
L 1 (2)
Technical Specifications The Technical Specifications contained in Appendix A as revised -
through Amendment No. 46 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 of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
- f. d ev f
ames E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 4, 1994 i
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UNITED STATES
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5 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666-0001 o
sg v.....f COMMONWEALTH EDISON COMPANY i
DOCKET NO. STN 50-455 BYRON STATION. UNIT NO 2 AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 46 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 August 13, 1993, as supplemented by letters dated September 15, 1993, September 16, 1993, December 17, 1993, January 19, 1994, February 11, 1994, and February 24, 1994, 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 l
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 l
conducted in compliance with the Commission's regulations;
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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 t
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 Specifi-cations as indicated in the attachment to this license amendment, and j
paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:
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Technical Specifications The Technical Specifications contained in Appendix A (NUREG-1113),
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as revised through Amendment No. 46 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 l
hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and i
the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION th9t$.
V James E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 4, 1994
A ATTACHMENT TO LICENSE AMENDMENT NOS. 46 AND 46 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. Overleaf pages have been provided for convenience and are marked with an asterisk.
Remove Paaes Insert Paaes 3/4 4-13 3/4 4-13 3/4 4-14
-3/4 4-14
- 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 4-18
- 3/4 4-18 3/4 4-19 3/4 4-19 B 3/4 4-3 B 3/4 4-3
REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.
1 APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one or more steam generators inoperable, restore the inoperable steam generator (s) to OPERABLE status priur to increasing T,y above 200*F.
SURVEILLANCE RE0VIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requinments of Specification 4.0.5.
4.4.5.1 Steam Generator Sample Selection and Inspection - 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.
4.4.5.2 Steam Generator Tube
- Samole Selection and Inspection - 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 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 except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the j
tubes inspected shall be from these critical areas; b.
The first sample ci tuces selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
- When referring to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.5.4.a.10.
BYRON - UNITS 1 & 2 3/4 4-13 AMENDMENT NO. 46
6 REACTOR COOLANT SYSTEM SURVEILLANCE RE0UIREMENTS (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,
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2)
Tubes in those areas where experience has indicated potential
- problems, 3)
At least 3% of the ' total number of sleeved tubes in all four steam generators or all of the sleeved tubes in the generator chosen' for the inspection-program, whichever is less.
These inspections will include both the tube and.the sleeve, and 4)
A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be 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
-kjected to a tube inspection, 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 imperfections were previously found.
The results of each sample inspection shall be classified into one of-the following three categories:
Cateaorv 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.
1 C-3 More than 10% of the total tubes inspected are i
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.
BYRON - UNITS 1 & 2 3/4 4-14 AMENDMENT NO. 46
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j REACTOR COOLANT SYSTEM 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 not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections, not including the pre-service inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that pre-viously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; b.
If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months.
The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a 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 accident requiring actuation of the Engineered Safety Features, or 4)
A Condition IV main steam line or feedwater line break.
i l
BYRON - UNITS 1 & 2 3/4 4-15
SURVEILLANCE RE0VIREMENTS (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 l
drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube or sleeve wall thickness, if l
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 a
tube or sleeve; l
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 l
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; I
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 I
repair limit imperfection depth is equal to 40% of the nominal l
wall thickness; I
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 l
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 l
U-bend to the top support of the cold leg.
For a tube that has been repaired by sleeving, the tube inspection shall include the sleeved portion of the tube, and i
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l BYRON - UNITS 1 & 2 3/4 4-16 AMENDMENT N0. 46
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (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 Repair refers to a process that reestablishes tube l
serviceability. Acceptable tube repairs will be performed by the following processes:
l a)
Laser welded sleeving as. described by Westinghouse report i
WCAP-13698, Rev. 1, or l
b)
Kinetic welded sleeving as described by Babcock & Wilcox
' Topical Report BAW-2045PA, Rev. 1.
Tube repair includes the removal of plugs that were previously
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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.
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.
j 4.4.5.5 Reports j.
a.
Within 15 days following the completion of each inservice inspection l
of steam generator tubes, the number of tubes plugged or repaired in l
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 perc. 't of waEthickness penetration for each indication of an taperfectibn, 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 investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
i BYRON - UNITS 1 & 2 3/4 4-17 AMENDMENT N0. 46
e TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS T0 BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection Yes No. of Steam Generators per Unit Four First Inservice Inspection Two l
Second & Subsequent Inservice Inspections One TABLE NOTATION 1.
The inservice inspection may be limited to one steam generator _on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators'in the plant) if the results of the.first or previous inspections indicate that all. steam generators are performing in a like manner.
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be-more severe than-those in other steam generators.
Under such circumstances the sample sequence shall be modified to inspect,the most severe conditions.
Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections..The fourth and subsequent inspections shall follow the instructions described above.
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BYRON - UNITS 1 & 2 3/4 4-18 l
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TABLE 4.4-2 asy STEAM GENERATOR TUBE INSPECTION E
i 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION e5 d
Sample Size Result Action Required Result Action Required Result Action Required A mininum of C-1 None N.A.
N.A.
N.A.
N.A.
S Tubes per l
N C-2 Plug or repair C-1 None N.A.
N.A.
I S. G.
defective tubes and i
C-2 Plug or repair C-1 None l
inspect additional defective tubes 2S tubes in this and inspect C-2 Plug or repair S. G.
additional 4S defective tubes tubes in this S. G.
C-3 Perform act. ion for C-3 result of first sample C-3 Perform action for N.A.
N.A.
w
'E C-3 result of first p-sample m
C-3 Inspect all tubes in All other None N.A.
N.A.
this S. G., plug or S. G.s are repair defective C-1 tubes and inspect 2S tubes in each Some S. G.s Perform action for N.A.
N.A.
other S. G.
C-2 but no C-2 result of additional second sample Notification to NRC S. G. are C-3
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nI 50.72 (b) 21 of 10
{nspect all tubes N.A N.A.
CFR Part 50
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plug or repair
>M defective tubes.
5 Notification to M
NRC pursuant to 5
5 50.72(b)(2) of 5
N, Where N is the number of steam gene ators i. the unit, and n is the number of steam o
generators inspected during an inspection
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-based on a modifgram for inservice inspection of steam generator tubes is tained. The pro ication of Regulatory Guide 1.83 Revision 1.
Inservice inspectionofsteamgeneratortubingisessentialinordertomaintainsurveil-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 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 between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage - 500 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 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plant i
shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged or repaired by sleeving.
The technical bases for sleeving i
are described in Westinghouse report WCAP-13698 Rev. I and Babcock & Wilcox Topical Report BAW-2045PA Rev.1.
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.
If a sleeved tube is found to contain a through wall penetration in the sleeve of equal to or greater than 40% of the nominal wall thickness, the
)
tube must be plugged.
The 40% plugging limit for the sleeve is derived from j
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 Westinghouse Report WCAP-13698 Rev. I and Babcock & Wilcox Topical Report BAW-2045PA Rev. 1.
Steam Generator tube and sleeve inspections have demonstrated the capability to reliably detect degradation that has penetrated 20% of the pressure retaining portions of the tube or sleeve wall thickness.
Commonwealth Edison will validate the adequacy of any system that is used for periodic inservice inspection of the sleeves and, as deemed appropriate, will upgrade testing methods as better methods are developed and validated for commercial use.
Whenever the resuits of any steam generator tubing inservice inspection fall into Category t'-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 4-3 AMENDMENT NO. 46
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UNITED STATES Qq~{[g
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NUCLEAR REGULATORY COMMISSION l
WASHINGTON, D.C. 20566 0001 s.....f COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 58 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 August 13, 1993, as supplemented by letters dated l
September 15, 1993, September 16, 1993, December 17, 1993, January 19, 1994, February 11, 1994, and February 24, 1994, 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; l
B.
The facility will operate in conformity with the application, the l
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 l
requirements have been satisfied.
l 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:
a l
l (2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 58 and the Environmental Protection Plan l
contained in Appendix B, both of which are attached hereto, are I
hereby incorporated into this license. The licensee shall operate l
the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION fa*u f V 9ames E. Dyer, Director Project Directorate III Division of Reactor Projects - III/IV/V l
Office of Nuclear Reactor Regulation 1
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 4, 1994 l
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UNITED STATES
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,E WASHINGTON, D.C. 20666-0001 v.
COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 58 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 August 13, 1993, as supplemented by letters dated September 15, 1993, September 16, 1993, December 17, 1993, January 19, 1994, February 11, 1994, and February 24, 1994, 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 l
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; l
l D.
The issuance of this amendment will not be inimical to the common j
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 Specifi-cations as indicated in the attachment to this license amendment, and l
paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:
1 l
l (2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 58 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 P1an.
3.
This license amendment is effective as of the date if its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION HGCf.hf James E. Dyer, Director Project Directorate III-2 l
Division of Reactor Projects - III/IV/V l
Office of Nuclear Reactor Regulation i
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 4, 1994 l
I ATTACHMENT TO LICENSE AMENDMENT NOS. 58 AND SR FACILITY OPERATING LICENSE N05. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.
The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Overleaf pages have been provided for convenience and are marked with an asterisk.
Remove Paaes Insert Paaes 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4-14
- 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 4-18 3/4 4-18 3/4 4-19 3/4 4-19 l
B 3/4 4-3 8 3/4 4-3 I
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REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS 2
1 LIMITING CONDITION FOR OPERATION a
j 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3 and 4.
LCTION:
With one or more steam generators inoperable, restore the inoperable steam generator (s) to OPERABLE status prior to increasing T,y above 200*F.
i SURVEILLANCE RE0VIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of l
the following augmented inservice inspection program and the requirements of Specification 4.0.5.
1
~
4.4.5.1 Steam Generator Sample Selection and Inspection - 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.
4.4.5.2 Steam Generator Tube
- Samole Selection and Insoection - 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 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 tha expectations of 4.4.5.2.a through 4.4.5.2.c, previous defects or imperfections in the area repaired by the sleeve are not considered an area requiring reinspection. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
l a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas; b.
The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
- When referring to a steam generator tube, the sleeve shall be considered a part of the tube if the tube has been repaired per Specification 4.4.5.4.a.10.
BRAIDWOOD - UNITS 1 & 2 3/4 4-13 AMENDMENT NO. 58
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continuea) 1)
All tubes that previously had detectable tube wall penetrations greater than 20% that have not been plugged or sleeved in the affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, 2)
Tubes in those areas where experience has indicated potential
- problems, l
3)
At least 3% of the total number of sleeved tubes in all four steam generators or all of the sleeved tubes in the generator chosen for the inspection program, whichever is less, These inspections will include both the tube and the sleeve, and 4)
A tube inspection (pursuant to Specification 4.4.5.4a.8) shall be l
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.
c.
The tubes selected as the second and third samples (if required by Table i
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 l
2)
The inspections include those portions of the tubes where imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Cateqory 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 i
tubes are defective.
Note:
In all inspections, previously degraded tubes or sleeves l
must exhibit significant (greater than 10% of wall thickness) further wall penetrations to be included in the above percentage calculations.
l I
BRAIDWOOD - UNITS 1 & 2 3/4 4-14 AMENDMENT NO. 58
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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 not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections, not including the pre-service inspection, result in all inspection.results falling into the C-1 category or if two consecutive inspections demonstrate that pre-viously observed degradation has.not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; b.
If the results of the inservice inspection-of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be' increased to at least once per 20_ months.
The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.; the interval may then be extended to a 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 accident requiring actuation of the Engineered Safety Features, or 4)
A Condition IV main steam line or feedwater line break.
l BRAIDWOOD - UNITS 1 & 2 3/4 4-15
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4 SURVEILLANCE RE0VIREMENTS (Continued) 4.4.5.4 Acceptance Criteria l
a.
As used in this specification:
1)
Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that required by fabrication drawings or specifications.
Eddy-current testing indications below 20% of the nominal tube or sleeve wall thickness, if i
j 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 1
imperfections greater than or equal to 20% of the nominal tube j
or sleeve wall thickness caused by degradation; i
i 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 j
unrepaired defect is defective; 6)
Pluaaina or Repair Limit means the imperfection depth at or i
beyond which the tube shall be removed from service by plugging j
or repaired by sleeving in the affected area. The plugging or repair limit imperfection depth is equal to 40% of the nominal wall thickness; 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-j coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above; j
l 8)
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the i
U-bend to the top support of the cold leg.
For a tube that has been repaired by sleeving, the tube inspection shall include the j
sleeved portion of the tube, and 1
5 l
I 4
I BRAIDWOOD - UNITS 1 & 2 3/4'4-16 AMENDMENT NO. 58
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j SURVE1LLANCE RE0VIREMENTS (Continued) 9)
Preservice Insoection means an inspection of the full length of each tube in each steam generator performed by eddy current i
techniques prior to service to establish a baseline condition l
of the tubing.
This. inspection'shall be performed prior to initial POWER OPERATION using the equipment and techniques j
expected to be used during subsequent inservice inspections.
4 j
- 10) Tube Repair refers to a process that reestablishes tube serviceability.
Acceptable tube repairs will be performed by the following processes:
1 a)
Laser welded sleeving as described by Westinghouse report WCAP-13698, Rev. 1,-or i
b)
Xinetic welded sleeving as described by Babcock & Wilcox i
Topical Report BAW-2045PA, Rev.1.
i Tube repair includes the removal of plugs that were previously-j inst alled as a corrective or preventative measure. A tube 1
inspection per 4.4.5.4.a.8 is required prior to returning previously plugged tubes to service.
b.
The steam generator shall be determined OPERABLE after completing j
the corresponding actions (plug or repair in the affected area all j
tubes exceeding the plugging or repair limit) required by 1
Table 4.4-2.
(
]
4.4.5.5 R_e. port s a.
Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in j
each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; l
i b.
The complete results of-the steam generator tube inservice inspection I
shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months'following the completion of the I
inspection.
This Special Report shall include:
1 1)
Number and extent of tubes inspected, 2) location and percent of wall-thickness penetration for each indication of an imperfection, and i
J 3)
Identification of tubes plugged or repaired.
j c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation.
This report shall provide a description of investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
i BRAIDWOOD - UNITS 1 & 2 3/4 4-17 AMENDMENT NO. 58 1
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n TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection Yes No. of Steam Generators per Unit Four First Inservice Inspection Two Second & Subsequent Inservice Inspections l
One TABLE NOTATION 1.
The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner.
Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators.
Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
Each of the.other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections.
The fourth and subsequent inspections shall follow the instructions described above.
l l
BRAIDWOOD - UNITS 1 & 2 3/4 4-18 l
TABLE 4.4-2 h
STEAM GENERATOR TUBE INSPtCTION U
88 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required b
A mininum of C-1 None N.A.
N.A.
N.A.
N.A.
d S Tubes per 1
C-2 Plug or repair C-1 None N.A.
N.A.
l S. G.
defective tubes and b
inspect additional C-2 Plug or repair C-1 None l
defective tubes 2S tubes in this and inspect C-2 Plug or repair l
S. G.
additional 4S defective tubes tubes in this S. G.
C-3 Perform act. ion for C-3 result of first sample R
C-3 Perform action for N.A.
N.A.
C-3 result of first sample G
C-3 Inspect all tubes in Allother None N.A.
N.A.
this S. G., plug or S. G.s are repair defective C-1 tubes and inspect 2S tubes in each Some S. G.s Perform action for N.A.
N.A.
other S. G.
C-2 but no C-2 result of additional second sample Notification to NRC S. G. are C-3 pursuant t Additional Inspect all tubes N.A N.A.
g I
S. G. is C-3 in each S. G. and 0
g plug or repair l
W defective tubes.
9 Notification to NRC pursuant to
,5 150.72(b)(2) of i
10 CFR Part 50 ai co N 1lhere N is the number of steam generators in the unit, and n is the number of steam S-3y
~
n" generators inspected during an inspection
l REACTOR COOLANT SYSTEM BASES i
i i
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained.
The program for inservice inspection of steam generator tubes is 1
based on a modification of Regulatory Guide 1.83 Revision 1.
Inservice 1
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 corro:, ion.
inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant 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-to-secondary leakage = 500 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 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plant f'
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 Westinghouse report WCAP-13698 Rev. I and Babcock & Wilcox Topical Report BAW-2045PA Rev. 1.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However even if a defect should develop in service, it willbefoundduringscheduledinservicesteamgeneratortubeexaminations.
Plugging or sleeving will be required for all tubes with imperfections exceeding the plugging or repair limit of 40% of the tube nominal wall 1
i thickness.
If a sleeved tube is found to contain a through wall penetration in the sleeve of equal to or greater than 40% of the nominal wall thickness, the 4
tube must be plugged.
The 40% plugging limit for the sleeve is derived from Reg. Guide 1.121 analysis and utilizes a 20% allowance for eddy current uncertainty and additional degradation growth.
Inservice inspection of sleeves is reguired to ensure RCS integrity.
Sleeve inspection techniques are described in Westinghouse Report WCAP-13698 Rev. I and Babcock & Wilcox Topical Report BAW-2045PA Rev. 1.
Steam Generator tube and sleeve inspections have demonstrated the capability to reliably detect degradation that has penetrated 4
20% of the pressure retaining portions of the tube or sleeve wall thickness.
Commonwealth Edison will validate the adequacy of any system that i.s 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.
l 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.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 AMENDMENT NO. 58
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