ML20072G967
| ML20072G967 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 08/18/1994 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072G970 | List: |
| References | |
| NUDOCS 9408250166 | |
| Download: ML20072G967 (6) | |
Text
-
'ff g
UNITED STATES i
)yJ. I j NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D.C. 205S0001 s% /
4....
COMMONWEALTH EDISON COMPANY OOCKET N0, STN 50-456 BRAIDWOOD STATION. UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 54 License No.' NPF-72 I
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated June 20, 1994, as supplemented August 18, 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 acco' dance with 10 CFR r
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:
9408250166 940810 DR ADOCK 05000456
.PDR.
i t
. (2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 54 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 stlo. lW Robert A. Capra, Director Project Directorate III-2 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 18, 1994
l 1
s i
ATTACHMENT TO LICENSE AMENDMENT No. 54 FACILITY OPERATING LICENSE N0, NPF-72 1
Q0CKET N0. 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 Paaes 3/4 4-17 3/4 4-17 3/4 4-17a 3/4 4-17a 3/4 4-27 3/4 4-27
~
REACTOR COOLANT ~ SYSTEM
' SURVElllANCE REQUIREMENTS (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 techniques expected to be used during subsequent inservice inspections.
10)
Tube Reoair refers to a process that reestablishes tube serviceability. Acceptable tube repairs will be performed by the following processes:
a)
Laser welded sleeving as described by Westinghouse report WCAP-13698, Rev. 1, or 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 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 Succort Plate Interim Pluaaina Criteria Limit for Unit 1 Cycle 5 is used for the disposition of a steam generator tube for continued service that is experiencing ODSCC confined within the thickness of the tube support plates.
For application of the tube support plate interim plugging criteria limit, the tube's disposition-for continued service will be based upon 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 standard calibrated against the laboratory standard will be utilized in Unit I steam generator inspections for consistent voltage normalization.
l 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 V' LIT 1 - AMENDMENT NO. 54
REACTOR COOLANT SYSTEM SURVElllANCE REQUIREMENTS (Continued) 3.
The projected end of cycle distribution of crack indications is verified to result in total primary to secondary leakage less than 9.1 gpm (includes operational and accident leakage).
The basis for determining expected leak rates from the projected crack distribution is provided in WCAP-14046, "Braidwood Unit 1 Technical Support for Cycle 5 Steam Generator Interim Plugging Criteria" dated May 1994.
4.
A tube with a flaw-like bobbin coil signal amplitude of greater than 2.7 volts shall be plugged or repaired.
Certain tubes identified in WCAP-14046, "Braidwood Unit 1 Technical Support for Cycle 5 Steam Generator Interim Plugging Criteria," dated May 1994, shall be excluded from application of the tube support plate interim plugging criteria limit.
It_has been determined that these tubes may collapse or deform following a postulated LOCA + SSE.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair in the affected area all tubes exceeding the plugging or repair limit) required by Table 4.4-2.
4.4.5.5 Reports Within 15 days following the completion of each inservice inspection a.
of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection.
This Special Report shall include:
1)
Number and extent of tubes inspected, 2) location and percent of wall-thickness penetration for each indication of an imperfection, and 3)
Identification of tubes plugged or repaired, c.
Results of steam generator tube inspections which fall into Category C-3 shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days and prior to resumption of j
plant operation. This report shall provide a description of investi-gations conducted to determine cause of the tube degradation and j
corrective measures taken to prevent recurrence.
BRAIDWOOD - UNITS 1 & 2 3/4 4-17a UNIT 1 - AMENDMENT NO. 54
MACTOR COOLANT SYSTEM 3/4.4,8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
a.
Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and**
b.
Less than or equal to 100/5 microCuries per gram of gross radioactivity.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1, 2 and 3*:
With the specific activity of the reactor coolant greater than a.
1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T,,,less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and b.
With the specific activity of the reactor coolant greater than 100/l microcuries per gram, be in at least HOT STANDBY with T,,, less than 500"F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- With T,y, greater than or equal to 500'F.
- For Unit 1 Cycle 5, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 microcuries per gram.
BRAIDWOOD - UNITS 1 & 2 3/4 4-27 UNIT 1 - AMENDMENT NO. 54