ML17324B057
| ML17324B057 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/19/1986 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| AEP:NRC:1007, NUDOCS 8609250205 | |
| Download: ML17324B057 (7) | |
Text
INDIANA& MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 September 19, 1986 AEP:NRC:1007 Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2
Docket Nos.
50-315 and 50-316 License Nos.
DPR-58 and DPR-74 NRC INSPECTION REPORT NOS.
50-315/86027 (DRS)
AND 50-316/86027 (DRS)
Mr. James G. Keppler, Regional Director U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137
Dear Mr. Keppler:
This letter is in response to Mr. C.
W. Hehl's letter dated August 21, 1986, which forwarded the report on the routine safety inspection conducted by members of your staff.
This inspection was conducted from July 7 through July 11, 1986 on activities at the D.
C.
Cook Nuclear Plant Units 1 and 2.
The Notice of Violat'ion attached to Mr. Hehl's letter identified one violation, which is addressed in Attachment 1 to this letter.
This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.
Very truly yours, v'.
P.
exich q i(j Vice President cm Attachment cc:
John E. Dolan W.
G. Smith, Jr.
- Bridgman R.
C. Callen G. Bruchmann G. Charnoff NRC Resident Inspector
- Bridgman 8gpcyp50205 86090193>I5 PDR ADOCN, 05000 8
SEP 22 198&
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Mr. James G. Keppler AEP:NRC:1007 bc:
J.
G. Feinstein/M.
W. Evarts S.
H. Horowitz/T. 0. Argenta/R.
C. Carruth J. J. Markowsky/S.
H. Steinhart/J.
A. Kobyra R.
W. Jurgensen R.
F. Kroeger M. L. Horvath
- Bridgman E. A. Morse
- Bridgman C. A. Erikson J.
B. Shinnock D. L. Wigginton, NRC - Washington, D.C.
AEP:NRC:1007 DC-N-6015.4
~ ~
I
ATTACHMENT 1 TO AEP:NRC:1007 NRC INSPECTION REPORT 50-315/86027; 50-316/86027
Attachment 1 to AEP:NRC:1007 NRC Violation "Unit 1 Technical Specification 3.3.2.1, specifies that the Engine'ered Safety Feature Actuation System instrumentation channels shall be response"-timed as shown in Table 3.3-5.
TS Table 3.3-5 requires that the turbine trip reactor trip ESF activation on steam generator high-high water level have an overall response time of less than or equal to 2.5 seconds.
TS definition 1.23,
'Engineered Safety Feature Response Time,'tates in part that the response time 'shall be that time interval from when, the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function.'ontrary to the above, prior to July 10,
- 1986, the licensee did not adequately response time test the Unit 1 steam generator water level high-high level turbine trip reactor trip, in that, response time was not measured through the turbine trip logic (i.e.,
the licensee did not fully test the time interval from the channel sensor to the ESF equipment).
The licensee demonstrated that the overall response
- time, was less than TS Table 3.3-5 requirement of 2.5 seconds on July 10, 1986."
Res onse to NRC Violation Corrective Action Taken and Results Achieved On July 10, 1986, following the sequence of events described in LER 315/86-007 (see Attachment 2), it was determined that the time-.response test procedure did not fully test the complete circuitry for the Unit 1 steam generator high-high water level turbine trip-reactor trip.
Therefore, a
special procedure was written and a time response test was conducted on that portion of the circuitry not included in the above surveillance procedure.
The test included that portion of the circuit from the output of the turbine trip control master relay to the closure of the steam stop valves.
In
- addition, the time response from the steam stop valve closure to reactor trip was determined from an actual trip. It was found that the total response time for the turbine trip-reactor trip from a steam generator high-high water level was within technical specification limits.
Corrective Action To Be Taken To Avoid Further Violation We will revise the appropriate time response testing procedures prior to the end of the Unit 1 1987 refueling outage to accomplish the testing of this circuitry as required by the Technical Specifications.
Date When Full Com liance Will Be Achieved Full compliance was achieved on July 10,
- 1986, when the response time testing was completed.
ATTACHMENT 2 TO AEP:NRC:1007 LER 315/86-007 REV.
1