ML17329A122
| ML17329A122 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/07/1988 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:09800, AEP:NRC:9800, NUDOCS 9108160002 | |
| Download: ML17329A122 (11) | |
Text
ACCELERATED Dl'RIBUTION DEMONSTRATION SYSTEM 1
l 1'EGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9108160002 DOC.DATE: 88/09/07 NOTARIZED: NO DOCKET g
FACIL:50-316 Donald C.
Cook Nuclear Power Plant, Unit 2, Indiana 05000316 AUTH.NAME AUTHOR AFFXLXATION ALEXICH,M.P.
(formerly Indiana S Michigan Ele RECIP.NAME RECIPIENT AFFILIATION MURLEY,T.E.
Document Control Branch (Document Control Desk)
SUBJECT:
Suppls 880901 telcon re integrated startup testing program includinq,testing re vibration measurements of RCS,natural circulation of RCS 6 steam generator level control instrumentation.
DISTRIBUTION CODE:
DF01D COPIES RECEIVED:LTR ENCL SIZE:
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50 Docket (PD Avail)
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NOTE TO ALL"RIDS" RECIPIENTS:
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PLEASE HELP US TO REDUCE WASTE! CONTACI'HE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAMEFROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
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Indiana Michigan Po~ Cen pany P.O, Box 16631 Co'.nfnbus, OH 43216 AVPMKA kllCHIQAN POWER AEP:NRC:09800 Donald C.
Cook Nuclear Plant Unit 2 Docket No. 50-316 I.icense No.
DPR-74 INTEGRATED STARTUP TESTING PROGRA 1
U.S. Nuclear Regulatoxy Commission Attn:
Document Control Desk Washington, D.
C.
20555 Attn:
T.
E. Murley September 7,
1988
Dear Dr,
Murley:
This letter is being sent as a follow-up to a September 1,
1988 telephone conversation between American Electric Powex Service Corporation and Nucleax Regulatory Commission personnel.
The phone eall was made to respond co three questions regarding the Unit 2 Integrated Startup Testing Program.
The three areas questioned concerned testing relative to.vibration measurements of the xeactor coolant system, natural circulation of the reactor coolant system, and sceam generator level control instrumentation.
The accachment to this letter pxovides a discussion of each of che items listed above.
This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to ensure its accuxacy and completeness prior to signature by the undersigned, Sincerely, M.
P, Alevich Vice President MPA/eh Attachment 08.
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<<e PAGE.883 Dr. T.
E. Murley AEP:NRC:09800 cc'.
D.
H. Williams, Jr.
W.
G. Smith, Jr.
- Bridgman R.
C. Callen G. Charnoff G.
Bruchmann NRC Resident inspector
- Bridgman A. B. Davis
- Region III 08.
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'1 l5: 88 FR AEPSC NOD PAGE F 884 Attachment 1 to AEP:NRG:09800 Page 1
UESTION 1 What procedure(s) wi11 be used to conduct vibration measurements of the Unit 2 reactor coolant system components?
RESPONSE
Reactor internals vibration measurements are deemed unnecessary since no changes weze made to the internals package.
Changes in replacement; steam generators have slightly reduced the reactor coolant system resistance, with only a negligible impact on system flow velocities.
Reactor coolant system vibration measurements are also deemed unnecessary.
The replacement steam generators offer only a slight increase in total system mass and are expected to have negligible impact on system vibration response.
During start-up and operation the reactor coolant system is routinely monitored by operator observations and monitoring instrumentation.
These same activities axe deemed sufficient to notice any unusual system vibration which might occur as a result of Steam Generator Repair Project activities.
QUESTION 2 What procedure(s) will be used to conduct natuxal circulation tests of the reactor coolant system to confirm that the design heat removal capability exists or to verify that flow (without pumps) or temperature data are comparable to prototype designs for which equivalent tests have been successfully completed?
RESPONSE
Ramp down testing was conducted during initial start-up, and the overa11 reactor coolant system flow xesistance was compared to that of our protype plane, Trojan Nuclear Plant.
System resistance was lower than Trojan's and therefore a natural circulation test was not performed.
The replacement steam generators have a lower pressure dzop than the original steam generators.
Specifically the tube ends are flush with the tube sheet surface, rounded and smooth which result in lower inlet pressure losses Zn addition the replacement steam generators have an increased number of tubes.
This will reduce the system resistance to a value loss than during initial start-up.
Therefore, no natural circulation test is required.
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Discussions have been held with Westinghouse regarding the licensee's response to Questions 1 and 2 and Westinghouse concurs with both responses.
W
. What procedure(s) will be used to calibrate and verify the performance of the steam generator level control system?
RESPONSE
Steam generator level control will be calibrated, using Plant Procedures THP 6030 IMP.204,
.205,
,206, and.20?.
The individual inputs to the steam generator le~el control/feedflow (steam flow, steam pressure, steam generator level) will be calibrated using appropriate plant procedures.
No additional testing or calibration is planned as a result of the Steam Generator Repair Prop ect.
TOTAL PAGE.885 08.
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