ML20029D664
| ML20029D664 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/02/1994 |
| From: | Assa R Office of Nuclear Reactor Regulation |
| To: | Farrar D COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 9405090208 | |
| Download: ML20029D664 (7) | |
Text
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May 2, 1994 Docket No. 50-456 Mr. D. L. Farrar Manager, Nuclear Regulatory Services Commonwealth Edison Company Executive Towers West III, Suite 500 I400 OPUS Place Downers Grove, Illinois 60515
Dear Mr. Farrar:
SUBJECT:
PUBLIC NOTICE OF APPLICATION'FOR AMENDMENT TO OPERATING LICENSE FOR BRAIDWOOD NUCLEAR PLANT, UNIT I.
The enclosed announcement has been forwarded to the Joliet News Herald and the Morris Daily Herald for publication. This announcement relates to your request for Emergency Technical Specification (TS) Amendment for Specification 3/4.4.5, " Steam Generators" dated April 25, 1994.
A separate notice will be published later in the Federal Reaister concerning the revision to the Steam Generator TS requirements.
Sincerely, original signed by G. Dick for Ramin A. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Enclosure:
Announcerrent cc w/ enclosure:
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May 2, 1994 Docket No. 50-456 Mr. D. L. Farrar j
Manager, Nuclear Regulatcry Services Commonwealth Edison Company Executive Towers West III, Suite 500 1400 OPUS Place Downers Grove, Illinois 60515
Dear Mr. Farrar:
SUBJECT:
PUBLIC NOTICE OF APPLICATION FOR AMENDMENT TO ORRATING LICENSE FOR BRAIDWOOD NUCLEAR PLANT, UNIT I.
The enclosed announcement has been forwarded to the Joliet News Herald and the Morris Daily Herald for publication. '. iis announcement relates to your request for Emergency Technical Spec'/1 cation (TS) Amendment for Specification 3/4.4.5, " Steam Generators" dated April 25, 1994.
A separate notice will be published later in the Federal Reaister concerning the revision to the Steam Generator TS requirements.
Sincerely, original signed by G. Dick for Ramin A. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Enclosure:
Announcement cc w/ enclosure:
See next page DISTRIBUTION:
Docket File NRC & Locals PDRs PDIII-2 Reading JRoe JDyer RAssa CHawes JZwolinski
)
DFC LA:PDlII-2 PM:PDIII-2 PD:PDIII-2 NAME CHAWESfj)\\O RAssa JDyer M DATE
(/ 9/94 6 / 2_/94 h/)/94
/ /94
/ /94
/ /94 YESh0;
/iES)NO kEh/NO COPY YES/N0 YES/NO YES/NO l
l PUBLIC NOTICE l
NRC STAFF CONSIDERING LICENSE AMENDMENT REQUESTS FOR BRAIDWOOD UNIT 1 The U.S. Nuclear Regulatory Commission (NRC) has received an application dated April 25, 1994, and supplements from Commonwealth Edison Company (CECO, l
the licensee) for emergency amendment to Facility Operating License No. NPF-72 for the Braidwood Nuclear Plant, Unit No.1, located in Will County, Illinois.
If approved, the amendment would implement a 1.0 volt steam generator (SG) tube interim plugging criteria (IPC) for the tube support plate elevation outer diameter stress corrosion cracking (0DSCC) for a limited period during Cycle 5.
l The NRC has determined that the licensee used its best efforts to make timely application for the proposed changes and that exigent circumstances do exist and were not the result of any fault of the licensee.
Ceco had no reason to anticipate the extensive number of tubes that would require plugging / repairing at the tube support plates due to ODSCC under the current Technical Specifications (TSs). After identifying the number of tubes requiring repair, Braidwood management considered various repair strategies that could be performed within the current Braidwood licensing basis. The licensee determined that these repair options were unheceptable because they would create operational limitations (unit derating) and may impact future tube repair efforts.
Subsequently, Braidwood management decided to apply an NRC approved IPC methodology to the inspection results. The licensee reviewed the number of tubes to be repaired using the existing TS criteria for tube repair and the number required by the IPC methodology and discussed its findings with the NRC.
Based on its analysis, the licensee has proposed to incorporate a 1.0 volt criterion into the Unit 1 SG inservice inspection and
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repair program for its current Unit I refueling outage. Currently, the licensee has plugged all tubes with indications greater than 2.7 volts and all confirmed indications greater than 1 volt.
The licensee has evaluated the requested amendments against the standards in 10 CFR 50.92 and the NRC staff has made a proposed (preliminary) determination that the requested amendments involve no significant hazards considerations. Under NRC regulations, this means that operation of the facility in accordance with the proposed amendments would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The licensee's analyses are summarized below:
Incorocrate 1.0 Volt Criterion 1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed license amendment request to implement SG IPC for Braidwood, Unit 1, Cycle 5 meets the requ'irements of Regulat'ory Guide (RG) 1.121 by demonstrating that tube leakage is acceptably low and tube burst is a i
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l highly improbable event during normal operation or a main steam line break (MSLB) event over the limited time period o.f this license amendment.
l Under accident conditions, conservatively assuming main steam line break (MSLB), significant margins exist for free span burst considerations for
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voltage growth in excess of 95% cumulative probability.
For the largest confirmed indications left in service, the projected voltage at 95% growth is 2.6 volts at the end-of-cycle (EOC) 5, compared to the 4.54 volts structural limit for a free span burst pressure of 1.43 times the steam line break pressure differential.
Even at 99% cumulative probability, the observed voltage growth is bounded by 2.7 volts and the structural limit is satisfied for the 1.0 volt rotating pancake coil (RPC) confirmed indications left ir, service.
In addition, the following analyses were done to provide assurance that there are additional margins provided by the following considerations:
A demonstration of limited tube support plate (TSP) displacement was done which reduces the likelihood of a tube burst.
Limited TSP l
displacement would also reduce leakage compared to free span indications.
The Electric Power Research Institute (EPRI) Performance Demonstration Program analyzed the performance of some 20 eddy current data analysts evaluating data from a unit with 3/4" inside diameter and 0.049" wall thickness tubes. This data demonstrated the voltage dependence of the probability of detection (POD) and argues for a P0D of greater than 0.6 for ODSCC indications larger than 1.0 volt.
Ceco's risk evaluation of the operability of Braidwood, Unit 1, Cycle 5 compared core damage frequency, with containment bypass, with and 1
without the IPC applied at Braidwood Unit 1.
The total Braidwood core damage frequency is estimated to be 2.74E-5 per reactor year with a total contribution from containment bypass sequences of 2.9E-8 per reactor year in the current individual plant evaluation (IPE).
Operation with the requested IPC resulted in an ir.significant increase in the MSLB with containment bypass sequence frequency.
The analyses presented above apply to a full cycle of operation.
Because plant operations approved by the proposed amendment would be for a significantly shorter period, the probability of an accident is much i
less than that calculated for the full cycle.
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To support the restart and limited operation of Braidwood Unit 1, reactor coolant system (RC5) ' dose equivalent (DE) Iodine 131 (I-131) will be limited to 0.35 microcuries per gram (uc/gm).
Therefore, since implementation of tha 1.0 volt IPC for a limited time period for Braidwood, Unit 1, Cycle 5 does not adversely affect steam generator tube integrity and results in acceptable dose consequences for a t
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worst case postulated accident, the proposed license amendment request does not result in any increase in the probability or consequences of an accident l
previously evaluated within the Braidwood Updated Final Safety Analysis Report.
2.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed steam generator tube interim plugging criteria does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism which could result in an accident outside the tube support plate elevations; no ODSCC is occurring outside de thickness of the tube support plates. Neither a single or multiple tube rupture event would be expected in a steam generator in which the IPC has been applied.
Ceco will implement a maximum ieakage rate limit of 150 gallons per day i
(gpd) through any one SG to help preclude the potential for excessive leakage during all plant conditions.
The RG 1.121 criterion for establishing operational leakage rate limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture during faulted plant conditions.
The 150 gpd limit will provide for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is associated with the longest permissible free span crack length.
Since tube burst due to ODSCC is precluded during normal operation due to the proximity of the TSP to the tube and the potential exists for the crevice to become uncovered during MSLB conditions, the leakage from the maximum permissible crack must preclude tube burst at MSLB conditions.
Thus, the 150 gpd limit provides for plant shutdown prior to reaching critical crack lengths for MSLB conditions.
3.
The proposed change does not involve a significant reduction in a margin of safety.
Upon implementation of RG 1.121 criteria even under the worst case postulated accident conditions, the occurrence,of ODSCC at the TSP elevations is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions.
The distribution of crack indications at the TSP elevations are confirmed to result in acceptable primary-to-secondary leakage during all plant conditions for the limited time period of this license amendment and that radiological consequences are not adversely impacted.
Loss of Coolant Accident (LOCA) coincident with a Safe Shutdown Earthquake (SSE) on the SG (as required by GDC 2), may cause a tube collapse in the SGs at some plants. A number of tubes have been identified, in the
" wedge" locations of the SG TSPs, to represent a potential for tube collapse during a LOCA + SSE event.
These tubes have been excluded from application of the voltage based SG TSP IPC.
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I Addressing RG 1.83, " Inservice Inspection of PWR Steam Generator Tubes,"
Revision 1, July 1975, considerations, implementation of the bobbin coil probe voltage based IPC of 1.0 volt is supplem'ented by: enhanced eddy current inspection guidelines to provide consistency in voltage normalization, a 100%
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eddy current inspection sample size at the tube support plate elevatians, and RPC inspection requirements for the larger indications left in service to characterize the principal degradation as ODSCC.
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i Implementation of the SG TSP IPC will decrease the number of tubes which must be repaired.
Therefore, the margin of flow that would otherwise be reduced in the event of increased tube plugging would be maintained.
If the proposed determinations that the requested license amendments involve r.o significant hazards considerations become final, the Nhc will issue the amendments without firrt offering an opportunity for a public hearing.
An opportunity for a hearing will be published in the Federal Reaister et a later date and any hearing request will not delay the effective date of the amendments.
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If the NRC staff decides in its final determinations that the amendments do involve a significant hazards consideration, a notice of opportunity for a
~i prior hearing will be published in the Federal Reaister and, if a hearing is granted, it will be held before the amendment is issued.
Comments on the proposed determinations of no significant hazards i
considerations may be telephoned to James E. Dyer, Director, Project Directorate III-2, by collect call to 1-(301)-504-1995 i
to the Rules and Directives Review Branch, Division of Freedom of Informacionor suo and Publication Services, Office of Administration, U.S. Nuclear Ryulatory Commission, Washington, DC 20555. All comments received by close of business on May 5, 1994, will be considered in reaching a final determination.
Copies j
of the applications may be examined at the NRC's Local Public Document Room, j
located at Wilmington Township Public Library, 201 S. Kankakee Street, i
Wilmington, Illinois 60481, and at the Commission's Public Document Room, the
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Gelman Building, 21?0 L Street, NW, Washington, DC 20555.
Mr. D. L. Farrar Braidwood Station Commonwealth Edison Company Units 1 and 2 CC:
Mr. William P. Poirier Chairman Westinghouse Electric Corporation Will County Board of Supervisors Energy Systems Business Unit Will County Board Courthouse Post Office Box 355, Bay 236 West Joliet, Illinois 60434 Pittsburgh, Pennsylvannia 15230 Ms. Lorraine Creek Joseph Gallo, Esquire Rt. 1, Box 182 Hopkins and Sutter Manteno, Illinois 60950 888 16th Street, N.W., Suite 700 Washington, D.C.
20006 Attorney General l
l 500 South 2nd Street Regional Administrator Springfield, Illinois 62701 U. S. NRC, Region III l
801 Warrenville Road Michael Miller, Esquire l
Lisle, Illinois 60532-4351 Sidley'and Austin One First National Plaza l
Ms. Bridget Little Rorem Chicago, Illinois 60690 Appleseed Coordinator 117 North Linden Street George L. Edgar l
Essex, Illinois 60935 Newman & Holtzinger, P.C.
1615 L Street, N.W.
Mr. Edward R. Crass Washington, D.C.
20036 Nuclear Safeguards and Licensing Division Illinois Dept. of Nuclear Safety Sargent & Lundy Engineers Office of Nuclear Facility Safety 55 East Monroe Street 1035 Outer Park Drive Chicago, Illinois 60603 Springfield, Illinois 62704 U. S. Nuclear Regulatory Commission Commonwealth Edison Company Resident Inspectors Office Braidwood Station flanager Rural Route #1, Box 79 Rt. 1, Box 04 Braceville, Illinois 60407 Braceville, Ili Sois 60407 Mr. Ron Stephens EIS Review Coordinatuc Illinois Emergency Services U.S. Environmental Protection Agency and Disaster Agency 77 W. Jackson Blvd.
l 110 East Adams Street Chicago, Illinois 60604-3590 i
Springfield, Illinois 62706 Howard A. Learner Environmental Law and Policy Center of the Midwest 203 North LaSalle Street Suite 1390 Chicago, Illinois 60601 1