ML20073G329
| ML20073G329 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 09/29/1994 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073G331 | List: |
| References | |
| NPF-37-A-064, NPF-66-A-064, NPF-72-A-055, NPF-77-A-054 NUDOCS 9410040109 | |
| Download: ML20073G329 (19) | |
Text
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A uto y*,
UNITED STATES j
NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 206S0001 49.....,o COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.64~
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 July 6, 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 Optrating License No. NPF-37 is hereby amended to read as follows:
9410040109 940929 PDR ADOCM 05000454 P
. (2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 64 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 M
Ro rt. 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:
September 29, 1994 l
l
Aug g
UNITED STATES j
j NUCLEAR REGULATORY COMMISSION WASHINGTON D.C. 20666-0001 a
g COMMONWEALTH EDIS0N COMPANY DOCKET N0. STN 50-455 BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 64 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 July 6, 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 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:
. (2)
Technical Soecifications l
The Technical Specifications contained in Appendix A (NUREG-1113),
as revised through Amendment No. 64 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 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
-D Rob rt 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: September 29, 1994
ATTACHMENT TO LICENSE AMENDMENT NOS. 64 AND 64 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 i
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.
The page indicated by an asterisk is provided for convenience only.
I Remove Paaes Insert Paaes 3/4 4-17 3/4 4-17 8 3/4 4-3 B 3/4 4-3
- B 3/4 4-4
- B 3/4 4-4 1
i
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 9)
Preservice Insoettien means an inspection of the full length of l
each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of t'he 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 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.
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 Reoorts a.
Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted to tb C; mission in a Special Report pursuant to Specification 6.9.2 witnin 12 months following the completion of the inspection.
This Special Report shall include:
1)
Number and extent of tubes inspected, 2) location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes plugged or repaired.
c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of' pl?nt operation. This report shall provide a descr'iption of investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
,BYRDW - L' NITS 1 & 2 3/4 4-17 AMENDMENT NO. 64
BASES
]/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-inspectionofsteamgeneratortubingisessentialinordertomaintainsurveil-1 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 i
of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary l
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 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 havin a reactor-to-secondary leakage less than this limit during operationwi$1haveanadequatemarginofsafetytowithstandtheloadsimposed 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 l
Technologies Technical Reports.
I Wastage-type defects are unlikely with proper chemist-i treatment of the i
secondary coolant.
However even if a defect should develop in service, it i
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 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 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.
Commeruealth Edison will validate the adequacy of any system that is used for periodic inservice inspection of the sleeves and, as deemed apropriate, will upgrade testing methods as better methods are developed and validated for commercial use.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission
)ur-suant to Specification 6.9.2
)rior to resumption of plant operation.
Suci cases will be considered by tie 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.
4 BYRON - UNITS 1 & 2 B 3/4 4-3 AMENDMENT NO. 64
~
REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any ma~gnitudo is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm.
This threshold value is sufficiently low to ensure early detection of additional leakage.
The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break.
The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents.
The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating valve in the supply line fully open at a nominal RCS pressure of 2235 psig.
This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in the safety analyses.
The 1 gpm leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure.
It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of. time, verification of valve integrity is required.
Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, those valves should be tested periodically to ensure low probability of gross failure.
BYRON - UNITS 1 & 2 8 3/4 4-4
9 Ktivg
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t UNITED STATES j
NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 2066tM001
\\*...+/
_(OMMONWEALTH EDISON COMPANY
)
DGCKET NO. STN 50-456
)
BRAIDWOOD STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 License Nc. NPF-72 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated July 6, 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 realth and safety of the public, and (ii) that such activitier
.i' be conducted in compliance with the Commission's regulation:-
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 I
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-72 is hereby amended to read as follows:
=
. (2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 55 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
~
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:
September 29, 1994 k
ATTACHMENT TO LICENSE AMENDMENT NO. 55 FACILITY OPERATING LICENSE NO. NPF-72 DOCKET NO. STN 50-456 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.
Remove Paces Insert P_gggi 3/4 4-17 3/4 4-17 8 3/4 4-3 8 3/4 4-3
REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS (Continued) 9)
Preservice Insoection 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 techniaues expected to be used during subsequent inservice inspections.
- 10) Tube Renair refers to a process that reestablishes tube serviceability. Acceptable tube repairs will be performed by the following processes:
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 aproved 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) Tube SuoDort Plate Interim Pluaaino Criteria Limit for Unit 1 Cycle 5 is used for the disposition of a sterm generator tube for continued service that is experiencing ODSCC confined within the thickness of the tube support plates.
For application of l
the tube support plate interim plugging criteria limit, the i
tube's disposition for continued service will be based upon j
standard bobbin coil probe signal amplitude of flaw-like indications. The plant specific guidelines used for all inspections shall be consistent with the eddy current guidelines in Appendix A of WCAP-13854 as appropriate to accommodate the additional information needed to evaluate tube support plate signals with respect to the voltage parameters as specified in Specification 4.4.5.2.
Pending incorporation of the voltage verification requirements in ASME standard verifications, an ASME shndard calibrated against the laboratory standard will be utilized in Unit I steam generator inspections for consistent voltage normalization.
1.
A tube can remain in service with a flaw-like bobbin coil signal amplitude of less than or equal to 1.0 volt, regardless of the depth of the tube wall penetration, provided Item 3 below is satisfied.
2.
A tube can remain in service with a flaw-like bobbin coil signal amplitude greater than 1.0 volt but less than or equal to 2.7 volts provided an RPC inspection does not detect degradation and provided Item 3 below is satisfied.
BRAIDWOOD - UNITS 1 & 2 3/4 4-17 UNIT 1 - AMENDMENT NO. 55
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-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, j
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 = 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. 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 Babcock & 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.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 UNIT I - AMENDMENT NO. 55
us2u9 y
UNITED STATES NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20066-0001
%...../
COMMONWEALTH EDIS0N COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENS1 Amendment No. 54 License No. NPF-77 1.
The Nuclear Regulatory Comission (the Commission). has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated July 6, 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 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 1
paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:
l
. 1 (2)
Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 54 and the Environmental Protection Plan contLined in Appendix B, both of which were attached to License Ho. NPF-72, dated July 2, 1987, are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date if its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION J
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:
September 29, 1994 I
l 1
ATTACHMENT TO LICENSE AMENDMENT NO. 54 FACILITY OPERATING LICENSE NO. NPF-77 DOCKET NO. 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.
The page indicated by an asterisk is provided for convenience only.
Remove Paaes Insert Paaes 3/4 4-17 3/4 4-17 3 3/4 4-3 B 3/4 4-3
- B 3/4 4-4
- B 3/4 4-4 I
4 i
E SCTOR COOLANT SYSTEM SURVEfLLANCE REOUIREMENTS (Con?inued) 9)
Preservice insoection 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 inttial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.
.10) Tube Reonir refers to a process that reestablishes tu %
serviceability. Acceptable tube repairs will be performed by the following processes:
I a)
Laser welded sleeving as described in a Westinghause r
Technical Report currently approved by the NRC, sabject to the limitations and restrictions as noted by the hRC staff, or b)
Kinetic welded sleeving as described in a Babcock & Wilcox Nuclear Technologies Technical Report currently aproved 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.
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 Recorts a.
Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted to th? Wission in a Special Report pursuant to Specification 6.9.2 witnic.: months following the completion of the inspection.
This Special Report shall include:
1)
Number and extent of tubes inspected, 2)
Location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes plugged or repaired.
c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation. This report shall provide a description of investi-gations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.,
BRAIDWOOD'- UNITS 1 & 2 3/4 4-17 UNIT 2 - AMENDMENT NO. 54
REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural inte rit of this portion of the RCS will be main-tained. The program for inserv ce nspection 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-i lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection i
of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant i
chemistry is not maintained within these limits, localized corrosion ma likely result in stress corrosion cracking. The extent of cracking du ng plant operation would be limited by the limitation of steam generator tube leaka e between the Reactor Coolant System and the Secondary Coolant System reac or-to-secondary leakage - 500 gallons per day per steam enerator racks having a reactor-to-secondary leakage less than this li it during).
o eration will have an ade uate mar in of safety to withstand the loads imposed d ring normal operation an by post lated 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, durin located and plugged or repaired by sleeving. g which the leaking tubes will be 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 willbefoundduringscheduledinservicesteamgeneratortubeexaminations.
Plugg ng or sleevin will be required for all tubes with imperfections excee ing the plugg ng 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 gged. The 40%
ging limit for the sleeve is derived from Reg. Guide 1.
analysis and u zes a 20% allowance for eddy current uncertainty and additional degradation growth.
Inservice inspection of sleeves is required to ensure RCS integ ty. Sleeve inspection techniques are described in the current Westin ouse or Babcock & Wilcox Nuclear Technologies Technical Reports. Steam Generator tube and sleeve inspections have demonstrated the capabilit to reliably detect degradation that has enetrated 20% of the ressure retain n portions of the tube or sleeve wall th ckn ns.
Conssonwealt Edison will val date 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 results of any steam generator tubing inservice inspection fall into Cate C-3, these results will be reported to the Commission ur-suant to Spect tion 6.9.2 prior to resumption of plant operation. Suc cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, d revision of the Technicaltests, laboratory examinations additional eddy-current inspection, an Specifications, if necessary.
BRAIDWOOD - UNITS 1 & 2 B 3/4 4-3 UNIT 2 - AMENDMENT NO. 54
=
REACTOR COOLANT SYSTEM 8ASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.
Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 gpm.
This threshold value is sufficiently low to ensure early detection of additional leakage.
The total steam generator tube leakage limit of 1 gpa for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose i
guideline values in the event of either a steam generator tube rupture or steam line break.
The 1 gpm limit is consistent with the assumptions used in t
the analysis of these accidents.
The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.
The 10 gpa IDENTIFIED LEAKAGE limitation provides allowance for a limited 1
amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow f
supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating i
valve in the supply line fully open at a nominal RCS precsure of 2235 psig.
This limitation ensures that in the event of a LOCA, the Safety Injection flow will not be less than assumed in th9 safety analyses.
The 1 gpa leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series check valve failure.
It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required.
Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment, those valves should be tested periodically to ensure low probability of gross failure.
BRAIDWOOO - UNITS 1 & 2 B 3/4 4-4
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