ML17331B271
| ML17331B271 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/25/1994 |
| From: | John Hickman Office of Nuclear Reactor Regulation |
| To: | Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| References | |
| TAC-M85971, NUDOCS 9403040208 | |
| Download: ML17331B271 (16) | |
Text
Docket No. 50-315 February 25, 1994 Mr. E.
E. Fitzpatrick, Vice President Indiana Michigan Power Company c/o American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43215
Dear Mr. Fitzpatrick:
SUBJECT:
DONALD C.
COOK NUCLEAR PLANT, UNIT NO.
1 PUBLIC NOTICE OF APPLICATIONS FOR AMENDMENTS (TAC NO. M85971)
The enclosed announcement has been forwarded to the South Haven Tribune and the Herald-Palladium for publication.
This announcement relates to 'your two applications dated February 15, 1994, for amendments to Facility Operating Licence No. DPR-58, to modify the Technical Specifications to incorporate F*
steam generator (SG) tube plugging criterion and to incorporate 2.0 volt interim SG tube support plate plugging criterion for fuel cycle 14.
A separate notice will be published later in the Commission's biweekly Federal
~Re ister notice concerning the above applications.
Sincerely, Original signed by:
John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation
Enclosure:
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Docket No. 50-315 Mr. E.
E. Fitzpatrick, Vice President Indiana Michigan Power Company c/o American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43215
Dear Hr. Fitzpatrick:
SUBJECT:
DONALD C.
COOK NUCLEAR PLANT, UNIT NO.
1 PUBLIC NOTICE OF APPLICATIONS FOR AMENDMENTS (TAC NO. H85971 /
The enclosed announcement has been forwarded to the South Haven Tribune and the Herald-Palladium for publication.
This announcement relates to your two applications dated February 15, 1994, for amendments to Facility Operating Licence No.
DPR-58, to modify the Technical Specifications to incorporate F*
steam generator (SG) tube plugging criteriop~and to incorporate 2.0 volt interim SG tube support plate plugging criterion for fuel cycle 14.
A separate notice will be published laPer in the Commission's biweekly Federal
~Re ister notice concerning the above applications.
Sincerely,
Enclosure:
Announcement cc w/enclosure:
See next page DISTRIBUTION Docket File NRC
- 8. Loca PDRs PD3-1 Rd JRoe JZwol i ki LMars JHickman CJamerson
OPA OC/LFDCB LMiller, RIII
- RStrasma, RIII o
John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation I
OFFICE NAME LA:PD31 CJamerso PH' man:
BC:EMCB JStrosn er OGC D: PD31 LBMarsh DATE W/~ 94 2-
>~/94 94 OFFICIAL RECORD COPY FILENAME: G: iWPDOCSiDCCOOKiCOPUBLC. NOT 1
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bottom of the transition between the roll expansion and the unexpanded tube.
The safety factors used in the verification of the strength of the degraded tube are consistent with the safety factors in the ASHE Boiler and Pressure Vessel Code used in SG design.
The F* distance has been verified by testing to be greater than the length of roll expansion required to preclude both'ube pullout and significant leakage during normal and postulated accident conditions.
Implementation of the tubesheet plugging criterion'ill decrease the number of tubes which must be taken outi of service with tube 'plugs or repaired with sleeves.
Bo8th plugs and sleeves reduce the RCS flow margin;
- thus, implementation of the F* criterion will maintain the margin of flow that would otherwise be reduced in the event of "increased plugging or sleeving.
If the proposed determinations that the requested license amendments involve no significant hazards considerations become final, the NRC will issue the amendments without first offering an opportunity for a public hearing.
An opportunity for a hearing will be published in the Federal Re<eister at a later date and any hearing request will nnt delay the effective date of the amendments.
If the NRC staff decides in its final determinations that the amendments do involve a significant hazards consideration, a notice of opportunity for a prior hearing will be published in the Federal Re ister and, if a hearing is granted, it will be held before the amendment is issued.
Comments on the proposed determinations of no significant hazards considerations may be telephoned to Ledyard (Tad) Harsh, Director, Project Directorate III-l, by collect call to l-(301)-504-1991.
or submitted in writing to the Rules and Directives Review Branch, Division of Freedom of Information and Publication Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
All comments received by close of business on Harch 7,
- 1994, will be considered in reaching a final determination.
Copies of the applications may be examined at the NRC's Local Public Document Room ~<located i at the Haud Preston Palenske Hemorial Library, 500 Harket Street, St.
- Joseph, Hichigan
- 49085, and at the Commission's Public Document Room, the Gelman Building, 2120 L t eet, NW, Washington, DC 20555.
0FF IGE LA PD31 CJamerson DATE PH ckman OGC OPA*
BHayden 2
25 94 PD: PD31 LBHarsh OFFICIAL RECORD CO Y
F ILENAHE: G: IhWPDOCSiDCCOOKiC085971. PA
- Previously concurred on notice
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Mr.
E.
E. Fitzpatrick Indiana Michigan Power Company CC:
Regional Administrator, Region III U,S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street
- Lansing, Michigan 48913 Township Supervisor Lake Township Hall Post Office Box 818 Bridgman, Michigan 49106 Al Blind, Plant Manager Donald C.
Cook Nuclear Plant Post Office Box 458 Bridgman, Michigan 49106 U.S. Nuclear Regulatory Commission Resident Inspector Office 7700 Red Arrow Highway Stevensville, Michigan 49127 Gerald Charnoff, Esquire
- Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.
W.
Washington, DC 20037 Mayor, City of Bridgman Post Office Box 366 Bridgman, Michigan 49106 Special Assistant to the Governor Room 1 - State Capitol
- Lansing, Michigan 48909 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N. Logan Street P. 0.
Box 30195
- Lansing, Michigan 48909 Donald C.
Cook Nuclear Plant Mr. S.
Brewer American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43215 Dcccmber 1993
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c PUBLIC NOTICE NRC STAFF CONSIDERING LICENSE AMENDMENT REQUESTS FOR D.
C.
COOK UNIT 1
The U.S.
Nuclear Regulatory Commission (NRC) has received two applications dated February 15,
- 1994, from Indiana Michigan Power Company (the licensee) for exigent amendments to Facility Operating License No.
DPR-58 for the Donald C.
Cook Nuclear Plant, Unit No.
1, located in Berrien County, Michigan.
If approved, the amendments would implement interim steam generator (SG) tube plugging criteria for the tube support plate elevation outer diameter stress corrosion cracking for Cycle 14 (AEP:NRC: 1166L) and would incorporate an acceptance criterion for SG tube degradation in the tubesheet region known as F*, taking into account the reinforcing effect of the tube sheet on the external surface of the tube in the roll expansion region (AEP:NRC: 1166K).
Amendment request AEP:NRC: 1166L also supplements a previous amendment request dated December 15, 1993, which was noticed in the Federal Re ister on January 5,
1994 (59 FR 621).
The February 15,
- 1994, submittal revises the acceptance criteria from 1.0 volt to 2.0 volt.
Other changes to the proposed amendment are made consistent with the voltage change in acceptance criteria.
The NRC has determined that the licensee used its best efforts to make timely applications for the proposed changes and that exigent circumstances do exist and were not the result of any fault of the licensee.
The exigent circumstances result from a recent change in NRC staff acceptance of higher interim voltage limits (i.e., 2.0 volts).
This change was made known to the licensee during a meeting on February 9, 1994.
As a result of this meeting, the licensee has requested to incorporate the 2.0 volt criterion into the Unit 1
SG inservice inspection and repair program during the current Unit 1
refueling outage.
The repairs are currently scheduled to begin March 6, 1994.
The required technical analyses for the F* amendment were pursued by the licensee when it became aware that F* could be coupled with a tube rolling technique, making F* applicable to the licensee's SG design.
When the analyses were completed the request was promptly proposed to the NRC and submitted.
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:
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Incor orate 2.0 Volt Criterion AEP:NRC:1166L (1)
Based on the existing data
- base, the Regulatory Guide 1.121 criterion requiring maintenance of a
safety factor of 1.43 times the steamline break pressure differential on tube burst is satisfied by 7/8 inch diameter tubing with bobbin coil indications with signal amplitudes less than 9.6 volts, regardless of the indicated depth measurement.
A 2.0 volt plugging criteria compares favorably with the 9.6 volt structural limit considering the previously calculated growth rates for outer diameter stress corrosion cracking (ODSCC) within the Cook Nuclear Plant Unit 1
SGs.
Considering a voltage growth component of 0.8 volt (40% voltage growth based on 2.0 volts beginning of cycle (BOC))
and an NDE
[nondestructive examination] uncertainty of 0.40 volt (20% voltage uncertainty based on 2.0 volts BOC),
when added to the BOC interim plugging criteria of 2.0 volts results in a bounding end of cycle (EOC) voltage of approximately 3.2 volts for Cycle 14 operation.
A 6.4 volt safety margin exists (9.6 structural limit 3.2 volt EOC
= 6.4 volt margin).
Conservatively, an upper limit of 3.6 volts will be used to assess tube integrity for those bobbin indications which are above 2.0 volts but do not have confirming rotating pancake coil (PRC) calls.
Relative to the expected leakage during accident condition loadings, it has been previously established that a
postulated main steamline break outside of containment but upstream of the main steam isolation valve represents the most limiting radiological condition relative to the IPC.
In support of implementation of the interim plugging criteria, it will be determined whether the distribution of crack indications at the tube support plate intersections at the end of Cycle 14 are projected to be such that primary to secondary leakage would result in site boundary doses within a
small fraction of the 10 CFR 100 guidelines.
A separate calculation has determined this allowable steamline break leakage limit to be 12.6 gpm.
Tube pull results from Cook Nuclear Plant Unit 1 indicate that tube wall degradation of greater than 40% throughwall was detectable either by the bobbin or RPC probe.
The tube with maximum throughwall penetration of 56%
(43% average) had a voltage of 2.02 volts.
This indication also was the largest recorded bobbin voltage from the EOC 12 leakage of 2.81 volts, inclusion of all IPC intersections in the leakage model is quite conservative.
(2)
Implementation of the proposed SG 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 of the region of the tube support plate elevations; no ODSCC is occurring outside the thickness of the tube support plates.
Neither a single or multiple tube rupture event would be expected in a
SG in which the plugging criteria has been applied (during all plant conditions).
Specifically, the licensee will continue to implement a maximum leakage rate limit of 150 gpd (0. 1 gpm) per 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
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based upon leak-before-break considerations to detect a free span crack before potential tube rupture during faulted plant conditions.
The 150 gpd limit should 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 crack length.
(3)
The use of the voltage based bobbin probe interim tube support plate elevation plugging criteria at Cook Nuclear Plant Unit 1
is demonstrated to maintain SG tube integrity commensurate with the criteria of RG 1. 121.
This is accomplished by determining the limiting conditions of degradation of SG tubing, as established by inservice inspection, for which tubes with unacceptable cracking should be removed from service.
Upon implementation of the criteria, even under the worst case conditions, the occurrence of ODSCC at the tube support plate elevations is not expected to lead to a SG tube rupture event during normal or faulted plant conditions.
As noted previously, implementation of the tube support plate elevation plugging criteria will decrease the number of tubes which must be repaired.
The installation of SG tube plugs reduces the reactor coolant system (RCS) flow margin.
- Thus, implementation of the interim plugging criteria will maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging.
Incor orate F* SG Tube Plu in Criterion AEP:NRC: 1166K (1)
The supporting technical and safety evaluations of the subject criterion demonstrate that the presence of the tubesheet will enhance the tube integrity in the region of the hardroll by precluding tube deformation beyond its initial expanded outside diameter.
The resistance to both tube rupture and tube collapse is strengthened by the presence of the tubesheet in that region.
The F* length of roll expansion is sufficient to preclude tube pullout from tube degradation located below the F* distance, regardless of the extent of the tube degradation.
Any leakage out of the tube from within the tubesheet at any elevation in the tubesheet is fully bounded by the existing SG tube rupture analysis included in the Cook Nuclear Plant FSAR [Final Safety Analysis Report].
(2)
Implementation of the proposed F" criterion does not introduce any significant changes to the plant design basis.
Use of the criterion does not provide a mechanism to initiate an accident outside of the region of the expanded portion of the tube.
Any hypothetical accident as a result of any tube degradation in the expanded portion of the tube would be bounded by the existing tube rupture accident analysis.
(3)
The use of the F* criterion has been demonstrated to maintain the integrity of the tube bundle commensurate with the requirements of RG 1. 121 (intended for indications in the free span of tubes) and the primary to secondary pressure boundary under normal and postulated accident conditions.
Acceptable tube degradation for the F* criterion is any degradation indication in the tubesheet
- region, more than the F* distance below the
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bottom of the transition between the roll expansion and the unexpanded tube.
The safety factors used in the verification of the strength of the degraded tube are consistent with the safety factors in the ASHE Boiler and Pressure Vessel Code used in SG design.
The F* distance has been verified by testing to be greater than the length of roll expansion required to preclude both tube pullout and significant leakage during normal and postulated accident conditions.
Implementation of the tubesheet plugging criterion will decrease the number of tubes which must be taken out of service with tube plugs or repaired with sleeves.
Both plugs and sleeves reduce the RCS flow margin;
- thus, implementation of the F* criterion will maintain the margin of flow that would otherwise be reduced in the event of increased plugging or sleeving.
If the proposed determinations that the requested license'mendments involve no significant hazards considerations become final, the NRC will issue the amendments without first offering an opportunity for a public hearing.
An opportunity for a hearing will be published in the Federal
~Re ister at a later date and any hearing request will not delay the effective date of the amendments.
If the NRC staff decides in its final determinations that the amendments do involve a significant hazards consideration, a notice of opportunity for a prior hearing will be published in the Federal Re ister and, if a hearing is granted, it will be held before the amendment is issued.
Comments on the proposed determinations of no significant hazards considerations may be telephoned to Ledyard (Tad) Harsh, Director, Project Directorate, III-1, by collect call to 1-(301)-504-1991.
or submitted in writing to the Rules and Directives Review Branch, Division of Freedom of Information and Publication Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
All comments received by close of business on Harch 7,
- 1994, will be considered in reaching a final determination.
Copies of the applications may be examined at the NRC's Local Public Document Room located at the Maud Preston Palenske Memorial Library, 500 Market Street, St.
- Joseph, Michigan
- 49085, and at the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW, Washington, DC 20555.
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