ML20202E631
| ML20202E631 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/25/1997 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Duquesne Light Co, Ohio Edison Co, Pennsylvania Power & Light Co |
| Shared Package | |
| ML20202E636 | List: |
| References | |
| DPR-66-A-208 NUDOCS 9712080097 | |
| Download: ML20202E631 (8) | |
Text
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WASHINGTON, D.C. 306eMKs01
....,6 DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY DOCKET NO. 50-334 BEAVER VALLEY POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 208 License No. DPR-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duquesne Light Company, et al. (the licensee) dated March 10, 1997, as supplemented July 28 and September 17, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will o)erate in conformity with the application, the provisions of tie Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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.
9712090097 971125 PDR ADOCK 05000334 P
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-66 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 208, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
In addition, the license i; amended by changes to paragraph 2.C.(10) to the Facility Operating License No. DPR-66 as follows:
(10)
Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No.208, are hereby incorporated into this license.
Duquesne Light Company shall operate the facility in accordance with the Additional Conditions.
3.
This license amendment is effective as of the date of its issuance, to be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION Jo n F. Stolz, Dir ~ or oject Directora 1-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attachments:
1.
Page 1 of Appendix C of License
- DPR-66 2.
Changes to the Technical Specifications Date of Issuance: November 25, 1997
- Page 1 of Appendix C is attached, for convenience, for the composite license to reflect this change.
. _ _ _ _ =. _ - - - -
l ATTACHMENT TO LICENSE AMENDMENT NO. 9nn FACIllTY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 1.
Revise Appendix C of the License as follows:
Remove Paae Insert Paae 1
1 2.
Replace the following pr 's of Appendix A Technical Specifications, with the enclosed pages as it.sicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 3/4 4-10b 3/4 4-10b B 3/4 4-2a B 3/4 4-2a B 3/4 4-2b B 3/4 4-2b B 3/4 4-2c B 3/4 4-2c
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v APPENDIX C
&QD1110NAL CONDITIONS 1
OPERATING LICENSE NO. DPR-66 Duquesne Light Company, Ohio Edison Company, and Pennsylvania Power Company shall comply with the following conditions on the schedules noted below:
Amendment Additional Condition Implementation Number Date 202 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.
Implementation implemented of this amendant shall include the relocation within 60 days of these technical specification requirements from April 14, to the appropriate documents, as described in 1997 the licensee's application dated September 9, 1996, and evaluated in the staff's safety evaluation attached to this amendment.
208 The licensee commits to perform the sost weld The amendment heat treatment of sleeve welds and tie shall be NRC-recommended inspections for repaired tubes implemented as described in the licensee's application within 60 days dated March 10, 1997, as supplemented July 28 from November 25, and September 17, 1997, and evaluated in the 1997.
staff's safety evaluation attached to this amendment.
1 Amendment No. 403 208
1 DPR-66 REACTOR COOLANT SYSTEM i
SURVEILLANCE REQUIREMENTS (Continued) u 6.
Pluacina 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 because it may become unserviceable prior to the next inspection.
The plugging or repair limit imperfection depths are specified in percentage of nominal wall thickness as follows:
's a.
Original tube wall 40%
This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied.
Refer to
- 4. 4. 5. 4. a.10 for the repair limit applicable to these interse'ctions.
b.
ABB Combustio Engineering TIG Welded sleeve wall 32%
c.
Westinghouse laser welded sleeve wall 25%
7.
Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an operating Basis Earthquake, a loss-of-coolant at,cident, or a steamline or feedwater line break as specified in 4.4.5.3.c, above.
8.
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot-leg side) completely around the U-bend to the top support to the cold-leg.
9.
Tube ReDair refers to sleeving which is used to maintain a
tube in-service or return a
tube to service.
This includes the removal of plugs that were installed as a corrective or preventive measure.
The following sleeve designs have been found acceptable:
a.
ABB Combustion Engineering TIG Welded Sleeves, CEN-629-P, Revision 02 and CEN-629-P Addendum 1.
b.
Westinghouse laser welded sleeves, WCAP-13483, Revision 1.
BEAVER VALLEY - UNIT 1 3/4 4-10b Amendment No. 208
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DPR-66 REACTOR COOLANT SYSTM BASES iii =
3/4.4.5 STEAM GENERATORS (Continued) operation would be limited by the limitation of steam generator tube leakage between the Primary Coolant System and the Secondary Coolant System (primary-to-secondary LEAKAGE = 150 gallons per day per steam generator).
Axial cracks having a primary-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 primary-to-secondary LEAKAGE of 150 gallons per day per steam generator can readily be detected.
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 s3eeving are described in the approved vendor reports listed in Surveillance Requirement 4.4.5.4.a.9.
Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant.
However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging or repair will be required of all tubes with imperfections exceeding the plugging or repair limit.
Degraded steam generator tubes may be repaired by the installation of sleeves which span the degradet tube section.
A steam generator tube with a sleeve installed meets the structural requirements of tubes which are not degraded, therefore, the sleeve is considered a part of the tube.
The surveillance requirements identify those sleeving methodologies approved for use.
If an installed sleeve is found to have through wall penetration greater than or equal to the plugging limit, the tube must be plugged.
The plugging limit for the sleeve is derived from R.G.
1.121 analysis which utilizes a 20 percent allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional degradation growth.
Steam generator tube inspections of operating plants have denonstrated the capability to reliably detect degradation that has penetrated 20 percent of the original tube wall thickness.
The voltage-based repair limits of these surveillance requirements (SR) implement the guidance in Generic Letter (GL) 95-05 and are applicable only to Westinghouse-designed steam generators (SGs) with outside diameter stress corrosion cracking (ODSCC) located at the tube-to-tube support plate intersections.
The voltage-based repair limits are not applicable to other forms of SG tube degradation nor are they applicable to oDSCC that occurs at other locations within the SG.
Additionally, the repair criteria apply only to indications where the degradation mechanism is dominantly axial ODSCC with BEAVER VALLEY - UNIT 1 B 3/4 4-2a Amendment No. 208
l DPR-66 REACTOR COOLANT SYSTEM i
BASES 1
3/4.4.5 STEAM GENERATORS (Continued) no NDE detectable cracks extending outside the thickness of the support plate.
Refer to GL 95-05 for additional description of the degradation morphology.
Implementation of these SRs requires a derivation of the voltage structural limit from the burst versus voltage empirical correlation and then the subsequent derivation of the voltage repair limit from the structural limit (which is then implemented by this surveillance).
The voltage structural limit is the voltage from the burst pressure / bobbin voltage correlation, at the 95-percent prediction interval curve reduced to account for the lower 95/95-percent tolerance bound for tubing material properties at 650*F (i.e.,
the 95-percent LTL curve).
The voltage structural limit must be adjusted downward to account for potential degradation growth during an operating interval and to account for NDE uncertainty.
The upper voltage repair limit; VURL, is determined from the structural voltage limit by applying the following equation:
Vunt = VsL ~ Vor - VNDE where Vor represents the allowance for degradation growth between inspectiona and VnDr represents the allowance for potential sources of error in the measurement of the bobbin coil voltage.
Further discussion of the assumptions necessary to determine the voltage repair limit are discussed in GL 95-05.
Safety analyses were performed pursuant to Generic Letter 95-05 to l
determine the maximum MSLB-induced primary-to-secondary leak rate that could occur without offsite doses exceeding a small fraction of 10 CFR 100 (concurrent iodine spike), 10 CFR 100 (pre-accident iodine spike), and without control room doses exceeding GDC-19.
The current value of this allowable leak rate and a summary of the analyses are provided in Section 14.2.5 of the UFSAR.
The mid-cycle equation in SR 4.4.5.4.10.d should only be used during unplanned inspections in which eddy current data is acquired for indications at the tube support plates.
SR 4.4.5.5 implements several reporting requirements recommended by GL 95-05 for situations which the NRC wants to be notified prior to returning the SGs to service.
For the purposes of this reporting requirement, leakage and conditional burst probability can be calculated based on the as-found voltage distribution rather than the projected end-of-cycle (EOC) voltage distribution (refer to GL 95-05 BEAVER VALLEY - UNIT 1 B 3/4 4-2b Amendment No. 208
DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) for more information) when it is not practical to complete these calculations using the projected EOC voltage distributions prior to returning the SGs to service.
Note that if leakage and conditional burst probability were calculated using the measured EOC voltage distribution for the purposes of addressing the GL section 6.a.1 and 6.a.3 reporting criteria, then the results of the proje::ted EOC voltage distribution should be provided per the GL section 6.b (c) criteria.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 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.
BEAVER VALLEY - UNIT 1 B 3/4 4-2c Amendment No. 208