ML20029E120
| ML20029E120 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/06/1994 |
| From: | Saccomando D COMMONWEALTH EDISON CO. |
| To: | Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9405160285 | |
| Download: ML20029E120 (5) | |
Text
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Ccmmonwulth Edison O /
1400 Opus Placo Oy Donners Grove, Illinois 60515
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May 6, 1994 Mr. William Russell, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Attn:
Document Control Desk
Subject:
Braidwood Station Unit 1 Additional Information Regarding Emergency Technical Specification Amendment for Technical Specification 3/4.4.5 NRC Docket No. 50-456
Reference:
1)
Commonwealth Edison Company (CECO)
Teleconference with Nuclear Regulatory Commission (NRC) dated May 6, 1994 2)
D.
Saccomando letter to W.
Russell dated May 2,
1994, transmitting Additional Information Regarding Emergency Technical Specification Amendment for Steam Generator IPC 3)
D.
Saccomando letter to W. Russell dated April 30, 1994, transmitting Supplemental Information Regarding Emergency Technical Specification Amendment for Steam Generator IPC 4)
D.
Saccomando letter to W.
Russell dated April 25, 1994, transmitting Request for Emergency Technical Specification Amendment for Specification 3/4.4.5 5)
R. Martin (:NRC) letter to D.
Rehn (Catawba Nuclear Station) dated December 16, 1993, which transmitted Amendment Nos.111 and 105 for Catawba Units 1 and 2, Docket Nos. 50-413 and 50-414
Dear Mr. Russell,
Reference 4 and the subsequent supplement transmitted CECO's request to process an Emergency Technical Specification Amendment to Specification 3/4.4.5 for Braidwood Unit 1.
The proposed amendment modifies the Technical Specification to incorporate a 1.0 volt steam generator tube interim plugging criteria (IPC) for Cycle 5.
Per the Reference 1 teleconference with the NRC, CECO agreed to submit / clarify the following information:
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'{l PDR ADOCK 05000456 P
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e W.. Russell May 6, 1994 1)
Reference 2 provided clarification to questions that the Staff asked during a May 2, 1992 teleconference. The response provided was in part based upon preliminary information, and has since been revised.
Therefore, this attachment provides information to supersede that submitted in Reference 2.
2)
The coverletter which was transmitted in Reference 3, acknowledged that the proposed revision to the Technical Specification pages which were transmitted with Reference 4 would need to reflect the conditions addressed in the April-30, 1994 supplement.
Braidwood proposed verbiage for a footnote for Technical Specification 4.4.5.0.
Subsequent.to that supplement, and with the Staff's concurrence, Braidwood would like to propose the following verbiage to be used in lieu of that submitted on April 30, 1994.
Proposed footnote for-Specification 3.4.8:
"For Unit 1 Cycle 5, the steam generators will be considered OPERABLE for the first 100 calendar days of operation with Tna greater than 500 F.
During that time, reactor coolant DOSE EQUIVALENT I-131 will be limited to 0.35 microCuries per gram."
3)
CECO acknowledges that the proposed revision to the Technical Specification pages which were transmitted with Reference 4 will need to be revised to specify.the eddy current inspection guidelines used during the Unit 1 inspection.
The plant specific-guidelines used are consistent with those centained in Appendix A of WCAP-13854, which was referenced in the Amendment Numbers 111 and 105 for Catawba Units 1 and 2, respectively (See' Reference 5).
Proposed footnote for Technical Specification 4.4.5.4.a.11 "The plant specific guidelines used for all inspections shall be consistent with the eddy current guidelines in Appendix A of WCAP-13854."
4)
Regarding Braidwood's independent oversight-of the steam generator inspections during this' refueling ~ outage, all safety related work associated with the steam generators was conducted in accordance with the Quality Assurance Program requirements identified in 10CFR50 Appendix B.
This was verified by the OnSite Quality Verification (SQV) Group by sampling of various activities throughout both the current refueling outage (A1RO4) as well as the recent Unit'l Forced Outage in October, 1993 for a leaking 1C steam generator k : nl 3 :brwd : lr<t al3 t 2
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W, Russell May 6, 1994 tube.
Field Monitoring Reviews were conducted by SQV on steam generator activities including eddy current testing, tube plugging, and plug verification. - One Level III Finding was identified through these reviews for usage of an incorrect procedure revision which was immediately corrected by Westinghouse.
CECO believes that these items should fully clarify the issues discussed earlier today. Please address any comments or questions regarding this matter to this office.
Sincerely, e
NN') [
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Denise M.
Saccomando Nuclear Licensing Administrator cc:
R. Assa, Braidwood Project Manager-NRR S.
Dupont, Senior Resident Inspector-Braidwood J. Martin, Regional Administrator-Region III Office of Nuclear Facility Safety-IDNS l
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ATTACHMENT The following information supersedes that submitted in Reference 2.
l CECO would like to clarify that the finalized version of WCAP-14046 will contain the following information:
i Additions to Section 8.2 of WCAP-14046 RPC sample inspection of more than 100 TSP intersections with dents (at Braidwood-1 these are typically mechanically induced " dings") or artifact / residual signal that could potentially mask a 1.0 volt bobbin signal.
Any RPC flaw indications in this sample will be plugged or repaired.
1 The NRC will be informed, prior to plant restart from the refueling outage, of any unexpected inspection findings relative to the assumed characteristics of the flaws at the TSP intersections.
This includes any detectable circumferential indications or detectable indications extending outside the thickness of the TSP.
100% bobbin coil, full length inspection of all active tubes with a 0.610 inch diameter bobbin probe for all straight length tubing.
l RPC inspection of all bobbin indications greater than the 1.0 volt repair limit (actual implemented was all bobbin indications).
RPC inspections were performed with a'0.620 inch diameter, 3 coil motorized RPC probe.
Addition to Section 7.2.1 of WCAP-14046 An RPC sampling plan was performed to inspect TSP intersections with dent signals greater than 5 volts and artifact / residual signals that could potentially mask bobbin indications of about 1.0 volt.
Denting in Braidwood-l is minor and most of the dents represent mechanical dings-rather than corrosion induced denting.
The RPC sampling plan was performed on all identified hot leg dents > 5.0 volts in S/Gs A and B.
It included 21 dents (18 in S/G A, 3
in S/G B) at TSP intersections.
There are only 6 dents in S/G C (one additional dent was in a tube plugged for other causes) and 2 in S/G D left in service above 5 volts that-were not RPC inspected.
The RPC sample included 40 mix residuals in S/G A and 41 in S/G B.
The mix residuals inspected had greater than a one volt signals and were
.l manually selected to represent the larger residual signals.
In addition to this RPC sampling plan, 85 intersections with no bobbin indications were RPC inspected.
No RPC flaw kinlastrwd:1perai3:4 m.
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ATTACHMENT (CONTINUED) indications were found in the RPC sampling-plan.
In both this RPC sample and the RPC inspection of bobbin flaw J
indications, no circumferential indications or indications I
extending outside of the TSP thickness were detected.
Limiting the RPC sampling to only S/G A and B left only 8 dented TSP intersections uninspected in S/Gs C and D.
Reviews of data indicate that all 8 of those dent indications were present in previous outages.
The uninspected dent indications lead to a negligible risk of leakage or rupture due to the small number of dents, the fact that no flaw indications were found at the inspected dent locations and the fact that a conservative POD of 0.6, independent of voltage, is applied for the SLB leak rate and
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the tube burst probability estimates.
Similarly, uninspected mix residuals in S/Gs C and D would have negligible concern for leakage or burst considerations.
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