ML20078C421
| ML20078C421 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/24/1994 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078C425 | List: |
| References | |
| NPF-37-A-066, NPF-66-A-066 NUDOCS 9410310241 | |
| Download: ML20078C421 (13) | |
Text
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COMMONWEALTH EDIS0N COMPAN1 DOCKET NO. STN 50-454 BYRON STATION. UNIT NO. 1 1
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 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 1, 1994, as supplemented September 7, 1994 and September 17,1994 (two letters), with clarifying information submitted by letters dated September 22, 1994, September 23, 1994, September 30, 1994, October 17, 1994, and October 24, 1994, complie.; 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.
9410310241 941024 PDR ADOCK 05000454 P
. 2.
Accordingly, the license is amended by changes to the Techtical 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:
(2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 66 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 3 O CL.
Robert A. Capra, Director Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 24, 1994
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UNITED STATES j..
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c WASHINGTON, D.C. 2055MMX)1
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COMMONWEALTH EDIS0N COMPANY j
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DOCKET NO. STN 50-455 BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 66 License No. NPF-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for anenoment by Commonwealth Edison Company (the licensee) dated August 1, 1994, as supplemented September 7, 1994 and September 17,1994 (two letters), with clarifying information submitted by letters dated September 22, 1994, September 23, 1994, September 30, 1994, October 17, 1994, and October 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-66 is hereby amended to read as follows:
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. (2)
Technical Specifications The Technical Specifications contained in Appendix A (NUREG-lll3),
as revised through Amendment No. 66 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 inco*purated into this license. Attachment 2 contains a revision to Appendix A which is 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 TddafRobertA.Capfa, Director Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 24, 1994 i
ATTACHMENT TO LICENSE APJNDMENT NOS. 66 AND 66 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 00CKET 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.
Pages marked with an asterisk are provided for convenience.
Remove Paaes Insert Paaes 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 4-17a 3/4 4-17b B 3/4 4-3 B 3/4 4-3 B 3/4 4-3a
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued) 1 I
1)
All tubes that previously had detectable tube wall penetrations greater than 20% that have not been plugged or sleeved in the 1
affected area, and all tubes that previously had detectable sleeve wall penetrations that have not been plugged, 2)
Tubes in those areas where experience has indicated potential
- problems, 3)
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 subjected to a tube inspection.
i 5)
For Unit 1, tubes left in service as a result of application of t
the tube support plate plugging criteria shall be inspected by bobbin coil probe during all future outages.
The tubes selected as the second and third samples (if required by c.
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.
l d.
For Unit 1, Cycle 7 implementation of the tube support plate interim plugging criteria limit requires a 100% bobbin coil probe inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support plate with outer diameter stress corrosion cincking (ODSCC) indications.
The l
determination of the tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20%
random sampling of tubes inspected over their full length.
1 The results of each sample inspection shall be classified into one of the following three categories:
l Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.
BYRON - UNITS 1 & 2 3/4 4-14 AMENDMENT NO. 66
REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS (Continued)
C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10% of wall thickness) further wall penetrations to be included in the above percentage calculations.
4.4.5.3 Insoection Freauencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:
The first inservice inspection shall be performed after 6 Effective a.
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 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 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 Additional, unscheduled inservice inspections shall be performed on c.
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.
~
t BYRON - UNITS 1 & 2 3/4 4-15 AMENDMENT NO. 66
REACTOR COOLANT SYSTEM 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 l
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; 1
5)
Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube or sleeve containing an l
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 is equal to 40% of the nominal wall thickness; 4
For Unit 1 Cycle 7, 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; 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; I
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
't BYRON - UNITS 1 & 2 3/4 4-16 AMENDMENT N0. 66
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 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 limitations and restrictions as noted by the NRC staff, or b)
Kinetic welded sleeving as described in a Babcock & Wilcox Nuclear Technologies Technical Report currently approved by the NRC, subject to the limitations and 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.
- 11) For Unit 1 Cycle 7, the Tube Support Plate Interim Pluaaina Criteria Limit is used for the disposition of a steam generator tube for continued service that is experiencing outer diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:
a)
Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltage ioss than or equal to 1.0 volt will be allowed to remain in service, b)
Degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltage greater than 1.0 volt will be repaired or plugged except as noted in 4.4.5.4.a.ll)c) below.
c)
Indications of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 1.0 volt but less than or equal to 2.7 volts may remain in service if a rotating pancake coil inspection does not i
detect degradation.
Indications of outside diameter stress corrosion cracking degradation with bobbin voltage greater than 2.7 volts will be plugged or repaired.
BYRON - UNITS 1 & 2 3/4 4-17 AMENDMENT N0. 66
~
- e SURVEILLANCE RE0VIREMENTS (Continued) d)
Certain intersections as identified in WCAP-14046, Section 4.7, will be excluded from application of the voltage-based repair criteria as it is determined that these intersec-tions may collapse or deform following a postulated LOCA+SSE event.
e)
If, as a result of leakage due to a mechanism other than ODSCC at the tube support plate intersection, or some other cause, an unscheduled mid-cycle inspection is performed, the following repair criteria apply instead of 4.4.5.4.ll)c).
If bobbin voltage is within expected limits, the indication can remain in service. The expected bobbin voltage limits are determined from the following equation:
f(V,z-Va)+V a
1+ (0.2) ( A c) where:
V measured voltage i
V voltage at B0C Alx time period of operation to unscheduled outage CL cycle length (full operating cycle length where operating cycle is the time between two scheduled steam generator inspections)
V 4.5 volts
=
u b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair in the affected area all tubes exceeding the plugging or repair limit) required by Table 4.4-2.
4.4.5.5 Reports Within 15 days following the completion of each inservice inspection a.
of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
1)
Number and extent of tubes inspected, 2)
Location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes plugged or repaired.
BYRON - UNITS 1 & 2 3/4 4-17 a AMENDMENT NO. 66
SVRVEILLANCE RE0VIREMENTS (Continued) c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
d.
For Unit 1 Cycle 7, implementation of the voltage-based repair criteria to tube support plate intersections, reports to the Staff shall be made as follows:
1)
Notify the Staff prior to returning the steam generators to service should any of the following conditions arise:
a)
If estimated leakage based on the actual measured end-of-cycle voltage distribution would have exceeded the leak limit (for postulated main steam line break utilizing licensing basis assumptions) during the previous operation cycle.
b)
If circumferential crack-like indications are detected at the tube support plate intersections, c)
If indications are identified that extend beyond the confines of the tube support plate.
d)
If the 1 X 10', calculated conditional burst probability exceeds
, notify the NRC and provide an assessment of the safety significance of the occurrence.
2)
The final results of the inspection and the tube integrity evaluation shall be reported to the Staff pursuant to 1
Specification 6.9.2 within 90 days following restart.
1 I
l BYRON - UNITS 1 & 2 3/4 4-17b AMENDMENT NO. 66
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-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 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 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 Babcock & Wilcox Nuclear Technologies Technical Reports.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.
However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging or sleeving will be required for all tubes with imperfections exceeding the plugging or repair limit of 40% of the tube nominal wall thickness.
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 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 Babcok & Wilcox Nuclear Technologies Technical Reports.
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.
BYRON - UNITS 1 & 2 B 3/4 4-3 AMENDMENT NO. 66
HEACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) for Unit 1 Cycle 7, tubes experiencing outer diameter stress corrosion cracking within the thickness of the tube support plates will be dispositioned in accordance with Specification 4.4.5.4.a.ll.
The operating period may be adjusted to less than the full operating cycle to meet the maximum site allowable primary-to-secondary leakage limit for End of Cycle Main Steam Line Break conditions. The leakage limit,12.8 gpm, includes the accident leakage from a faulted steam generator and the operational leakage of the three remaining intact steam generators equal to the Specification 3.4.6.2.c leakage limit.
j Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resumption of plant operation.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
BYRON - UNITS 1 & 2 B 3/4 4-3 a AMENDMENT NO. 66