IR 05000315/1984013
| ML20133B523 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/31/1985 |
| From: | Paperiello C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Dolan J AMERICAN ELECTRIC POWER SERVICE CORP., INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| Shared Package | |
| ML20133B528 | List: |
| References | |
| NUDOCS 8508060198 | |
| Download: ML20133B523 (2) | |
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JUL 311985 Docket No. 50-315 Docket No. 50-316 American Electric Power Service Corporation Indiana and Michigan Electric Company ATTN:
Mr. John Dolan Vice Chairman Engineering and Construction 1 Riverside Plaza Columbus, OH 43216 Gentlemen:
This is to acknowledge your letters dated November 29, 1984, January 30, 1985, and April 12, 1985, informing us of the steps you have taken to correct the violations and address the unresolved items which we brought to your attention in Inspection Reports 50-315/84-13(DRS) and 50-316/84-15(DRS) forwarded by our letter dated October 12, 1984. Our position regarding your responses is as stated below.
With regard to your response to item 1, we have reviewed your letter to the Office of Nuclear Reactor Regulation (NRR) dated November 28, 1984, and NRR's subsequent response dated April 11, 1985. Although the practice of temperature compensation for the inservice testing of the turbine driven auxiliary feedwater pump (TDAFP) is technically acceptable for explaining differences in pump developed head, the inspection revealed that the temperature measured during the test was the pipe surface temperature, and not the fluid temperature. As no correlation between the temperature of the pumped fluid and the pipe surface temperature was defined, the temperature used for compensation was not valid. Review of previous test data showed that discharge pressures less than the TS limit of 1285 psig were accepted as being satisfactory; therefore, we do not agree with your position that this item is not a violation and the violation stands as written. We agree that a TS clarification is appropriate and that more direct means of determining the pumped fluid temperature is necessary as addressed in the NRR response to your inquiry.
With regard to your response to item 3.a. the stated intent of Section XI, per IWV-1100 and IWV-3410 of the 1974 Code (IWV-3413 and IWV-3417 of the 1980 Code) is to verify iridividual valve operational readiness; therefore, maximum stroke times must be indicative of component operability. Your choice of system response time for operability criteria did not meet the intent of the Code.
Consequently, we do not agree with your position that this item is not an example of a violation and item 3.a stands as written. An NRC memo, attached, clarifies our position regarding maximum stroke times.
Your response to item 3.b appears to appropriately address the identified concern and will be reviewed in subsequent inspections.
8500060198 350731
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[R ADOCK 05000315 PDR
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JUL 31 1985 American Electric Power Service
Corporation
'With regard to your response to item 3.c it is the position of the Comission, as stated in the attached memo from NRR, that exemption from the leak testing
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methods delineated in IWV-3420 does not implicitly include relief from the
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analysis and corrective action criteria defined in IWV-3420f and IWV-3420g.
In addition, these analysis and corrective action requirements apply to all valves which are leak tested. As stated in your November 29, 1984, letter, a relief request specifically addressing said requirements will be submitted to NRR.
This matter will be reviewed in subsequent inspections.
With regard to your response to Unresolved Item 315/84-13-07(DRS);
316/84-15-07(DRS), it is cur. position that ASME Code interpretations are clarifications of existing requirements of the Code and are applicable. As noted in the inspection report, accessible valves with remote position indicators have not been verified per Code requirenents. We therefore disagree with your position that you have been in literal compliance with the Code. As stated in your response dated April 13, 1985, remote position verification for Units 1 and 2 will be completed, per 1980 Code requirements, prior to the end of each unit's refueling outage.
Actions taken to address the remaining items identified in the subject
. inspection report will be reviewed during subsequent inspections.
Your cooperation with us is appreciated.
Sincerely,
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C. J. Papdriello, Director Division of Reactor Safety Enclosure: Memos As Stated cc w/ encl:
W. G. Smith, Jr., Plant Manager DMB/ Document Control Desk (RIDS)
Resident Inspector, RIII Ronald Callen, Michigan Public Service Commission EIS Coordinator, USEPA Region 5 Office Nuclear Facilities &
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