ML20137G667
| ML20137G667 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 03/28/1997 |
| From: | Lynch M NRC (Affiliation Not Assigned) |
| To: | Johnson I COMMONWEALTH EDISON CO. |
| References | |
| TAC-M96904, NUDOCS 9704010461 | |
| Download: ML20137G667 (5) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20065-0001
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March 28, 1997 3
Ms. Irene Johnson, Acting Manager Nuclear Regulatory Services 3
Comonwealth Edison Company Executive Towers West III 1400 Opus Place, Suite 500 Downers Grove, IL 60515
SUBJECT:
EXTENSION OF THE BYRON, UNIT 1, OPERATING CYCLE BETWEEN STEAM GENERATOR TUBE EDDY CURRENT INSPECTIONS (TAC NO. M96904)
Dear Ms. Johnson.
By letter dated December 20, 1996, as supplemented on January 7,1997, the Comonwealth Edison Company (Comed) requested a change in its previous 1
comitment to limit operation of Byron Station, Unit 1 (Byron 1). The prior comitment made by Comed in June 1996, would limit the present operating cycle (i.e., Cycle 8) to 1.1 effective full power years (EFPY).
This is equivalent i
to 448.5 days above s hot-leg temperature (T,) of 500 degrees Fahrenheit, no which represents the operating interval between the eddy current inspections in October 1994 and April 1996. The proposed change in commitment made in your letter dated December 20, 1996, would extend the current operational cycle for Byron 1 to 540 days. At that time, Comed would enter an outage to replace the unit's steam generators.
Comed's prior comitment to limit the present operating cycle to 448.5 days i
was based on its assessment of indications of circumferentially-oriented steam generator tube degradation identified at tubesheet hardroll expansions during eddy current examinations completed in the three most recent outages (i.e., October 1994, October 1995 and April 1996). Comed concluded that because a number of circumferential indications identified in the 1996 inspection were evident at that time but not detected in the previot:s 1995 examination, a 448.5 day cycle length was justified on the basis that the degradation that occurred during this operating period had not adversely affected the structural or leakage integrity of these tubes. This conclusion was supported by in-situ pressure test results in the 1996 inspection outage 4
that demonstrated a sample of tubes containing the most significant circumferential indications that were detected in a "look back" review of the 1994 outage data, had adequate margins for structural and leakage integrity.
The NRC staff had previously agreed with Comed's conclusion iegarding a proposed 448.5 day cycle length for Byron l's Cycle 8, as stated in "Sumary i
of Meeting Discussing Steam Generator Tube U-Bend Flaws and Length of
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Byron 1 Operating Cycle - June 20, 1996," dated July 3,1996. However, the staff also stated in this meeting sumary that should Comed later request a 01004s NRC HLE CENTER COPY i
9704010461 970328 PDR ADOCK 05000454 P
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cycle length past 1.1 EFPY, it should provide additional justification for this request. Subsequently, Comed submitted in your letter dated October 18, 1996, a technical justification for extending the present Byron 1 operating cycle from 448.5 days to 600 days. The basis for the request was a proposed i
voltage-based assessment for determining the conditional tube burst t
probability and primary-to-secondary tube leakage under accident conditions i
attributable to circumferential cracking. The methodology employed in your
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evaluation was similar to that specified in NRC Generic Letter (GL) 95-05.
Based on the results of your analysis, Comed concluded that Byron I would i
meet the safety margins specified in GL 95-05 for structural integrity for the duration of 600 days for the present fuel cycle. Additionally, Comed i
stated in its letter dated October 18, 1996, that the a:,timate of leakage
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4 attributable to circumferential cracking under postulated accident conditions, i
plus that attributable to outer diameter stress corrosion cracking at the l
intersections of the steam generator tubes with the tube support plates, was
-less than the current licensing basis for the maximum site allowable primary-l to-secondary leakage of 36.5 gallons per minute.
j The methodology presented in your letter dated October 18, 1996, was similar j
to that previously proposed by Comed in your letter dated August 2,1996, in which you requested that the Braidwood, Unit 1, mid-cycle steam generator tube 4
inspection scheduled for October 1996, be deleted.
After extensive review, the NRC stated in a letter dated October 3,1996, that there was insufficient information in the August 2,1996, proposal to allow it to make a decision at l
that time regarding the Braidwood, Unit 1, mid-cycle inspection.
_ Comed later revised its request for the preser,t Byron, Unit 1, operating cycle from 600 days to 540 days in a letter dated December 20, 1996..Since the
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basic methodology for addressing circumferential cracking at-the tubesheet j
hardroll-expansions did not change appreciably from the original submittal on this subject on August 2, 1996, the staff continued its review of this i
proposed methodology. While this proposed methodology appears to address the major aspects of other approved methods of voltage-based steam generator tube repair criteria (i.e., GL 95-05) and the results support an extended cycle length greater than 448.5 days, the staff was unable to complete a thorough and in-depth review of all aspects of the analytical cycle length assessment.
As a result, the staff considered the merits of your quantitative technical assessment as well as other measures taken during recent inspections to demonstrate adequate tube structural and leakage integrity in making a decision on the acceptability of an extended operating cycle for Byron 1.
Although the NRC staff recognizes that a thorough review of your technical assessment was not feasible in the time frame for evaluating the appropriate cycle length for Byron 1, additional qualitative considerations included in Comed's letter dated January 7,1997, support an extended operational cycle.
As indicated in this latest letter, the two previous inspections of the
j I. Johnson.
Byron 1 steam generators utilized more sensitive eddy current prober (i.e.,
Plus Point probes) and data analysis software with features to facilitate the detection of circumferential indications.
The staff believes that these 4
improved inspection measures have enhanced your ability to detect indications of circumferential cracking at an earlier stage than in inspections completed prior to 1995. Consequently, circumferential flaws that could become potentially significant defects at the end of Cycle 8 were more likely to be detected and removed from service in the most recent outage than in previous inspections. Also, Comed pulled and conducted a metallurgical examination of ten tubes with circumferential flaws from the Byron I steam generators in the October 1995 outage.
This led to increased analyst awareness in the most recent inspection as a result of the destructive metallurgical analyses of the m tubes. The staff also notes that Comed completed in-situ pressure testing of a sample of tubes containing the most significant circumferential indications identified during the April 1996 outage. All of the expansion-transition circumferential indications had structural and leakage integrity margins in excess of the guidelines in NRC Regulatory Guide 1.121.
The staff has considered both the qualitative considerations and the quantitative cycle length assessment of Byron 1 and concluded that the Comed proposal to extend the maximum permissible operating time from 448.5 days to 540 days above a T,, of 500'F, is acceptable.
In addition to considering these factors which support an extended operational cycle for Byron 1, the NRC staff notes that other defense-in-depth measures will also facilitate the safe operation of the Byron I steam generators through the remainder of Cycle 8.
Specifically, primary-to-secondary leakage monitoring equipment and procedures at Byron 1, will enable you to detect and react to conditions potentially indicative of more advanced steam generator tube degradation.
The Byron 1 Technical Specifications limit primary-to-secondary leakage to a relatively low level of 150 gallons per day through any one steam generator. Also, Comed has indicated its intent to lower the maximum permissible dose equivalent (DE) concentration of iodine-131 to 0.20 microcuries per gram of coolant to provide additional margin to the maximum site allowable primary-to-secondary leakage limit.
Specifically, this proposed measure decreases the potential radiological dose consequences of a main steam line break accident with concurrent primary-to-secondary leakage.
This last measure to lower the primary coolant activity was submitted to the NRC as a license amendment request dated January 31, 1997.
If approved, this reduction in the DE concentration of iodine-131 would remain in effect until the end of the present Byron 1 operating cycle.
In summary, the staff finds that your proposal to operate Byron 1 for 540 days in Cycle 8 at a T,1 above 500*F before conducting eddy current inspections of 3
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I. Johnson Byron Station Commonwealth Edison Company Unit Nos. I and 2 cc:
Michael I. Miller, Esquire Chairman, Ogle County Board Sidley and Austin Post Office Box 357 One First National Plaza Oregon, Illinois 61061 Chicago, 1111nois 60603 Mrs. Phillip B. Johnson Regional Administrator, Region III 1907 Stratford Lane U.S. Nuclear Regulatory Commission Rockford, Illinois 61107 801 Warrenville Road Lisle, Illinois 60532-4351 Attorney General 4
500 South Second Street Illinois Department of Springfield, Illinois 62701 Nuclear Safety Office of Nuclear Facility Safety EIS Review Coordinator 1035 Outer Park Drive U.S. Environmental Protection Agency Springfield, Illinois 62704 77 W. Jackson Blvd.
Chicago, Illinois 60604-3590 Document Control Desk-Licensing Commonwealth Edison Company Commonwealth Edison Company 1400 Opus Place, Suite 400 Byron Station Manager Downers Grove, Illinois 60515 4450 North German Church Road Byron, Illinois 61010 Mr. William P. Poirier, Director i
Westinghouse Electric Corporation Kenneth Graesser, Site Vice President Energy Systems Business Unit Byron Station Post Office Box 355, Bay 236 West Commonwealth Edison Station Pittsburgh, Pennsylvania 15230 4450 N. German Church Road Byron, Illinois 61010 Joseph Gallo Gallo & Rcss 1250 Eye St., N.W.
Suite 302 Washington, DC 20005 Howard A. Learner Environmental' law and Policy Center of the Midwest 203 North LaSalle Street Suite 1390 j
Chicago, Illinois 60601 I
U.S. Nuclear Regulatory Commission Byron Resident Inspectors Office 4448 North German Church Road Byron, Illinois 61010-9750 1
Ms. Lorraine Creek i
Rt. 1, Box 182 Manteno, Illinois 60950
L I. Johnson the tubesheet hardroll expansions, is acceptable.
The staff understands that this will allow Byron, Unit 1, to operate without a scheduled outage until about mid-December 1997.
If you have any questions on this matter, please contact M. David Lynch at (301) 415-3023.
Sincerely, ORIGINAL SIGNED BY:
M. David Lynch, Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV i
Docket No. STN 50-454 cc w/ encl:
See next page Distribution: Docket' File PUBLIC PD!li-2 R/F J. Roe, JWR R. Capra E. Adensam, EGA1 M. D. Lynch G. Dick R. Assa C. Moore S. Bailey OGC, 015B18 ACRS, T2E26 B. McCabe, 017G21 P. Rush, 07D4 R. Lanksbury, Rill S. Coffin, 07D4 E. Sullivan, 07D4 B. Sheron, 07D26 K. Karwoski, 07D4 J. Strosnider, 07D4 DOCUMENT NAME: BYRON \\BY96904. FIN To receive a copy of this document, f r;(icate in the box:
"Cu e Copy without enclosures "E" = Copy with enclosures "N" a No copy 0FFICE SM
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