ML17317A759
| ML17317A759 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 11/07/1978 |
| From: | Shaller D INDIANA MICHIGAN POWER CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 7811290136 | |
| Download: ML17317A759 (7) | |
Text
('0 DONALD C. COOK NUCLEAR PL'ANT P.O. Box 458, Bridgman, Michigan 49106 November 7, 1978 Mr. J.G.
- Keppler, Regional Director Office of Inspection and Enforcement United States Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137
Dear Mr. Keppler:
Operating License JFR-74 Docket No. 50-316+
Special Reports SI-01/SI-02 The purpose of,this letter is to forward to you the attached Special Reports required by Appendix A Techni cal Specification 3.5.2, Emergency Core Cooling Subsystems.
Special Report SI-01 should have been submitted prior to May 29, 1978.
Special Report SI-02 should have been reported prior to November 2, l978.
These reports are late because the tickler system for reports established in the Operations Department tracked the reporting of Licensee
.Event Reports and Abn'ormal Environmental Occurrence Reports but failed to include Special Reports.
The tickler system has been modified to include Special Reports.
Sincerely, D.V. Shaller Plant Manager
/bab CC:
J.E.
Dolan R.H. Jurgensen R.F.
Kroeger R.J. Vollen BPI'.R.
Baker RO:III R.C. Callen MPSC P.W. Steketee, Esq.
R. Halsh, Esq.
M.H. Judkis PNSRC J.F. Stietzel R.S. Keith
.".'ov l 3 )978
INDIANA 5 MICHIGAN POklER COMPANY DONALD C.
COOK NUCLEAR PLANT Operating License DPR-74 Docket No. 50-316 Special Report SI-01 SAFETY INJECTION ACTUATION -
MARCH 1, 1978 CONDITIONS PRIOR TO OCCURRENCE The reactor plant was in Mode 3, Hot Standby, with the Reactor Coolant Sys-tem average temperature 547oF and 2235 psig pressure.
All 4 Reactor Coolant Pumps were in operation.
Pre-operational test 2-PO-050-543 Steam Dump System was being conducted.
DESCRIPTION OF OCCURRENCE Pre-operational test 2-PO-050-543 was being conducted to determine the capacity of the Steam Generator Atmospheric Steam Dump Valves.
The portion of the test being conducted called for closing the Steam Generator Stop Valve on one steam generator and then opening'he Steam Generator Atmos-pheric Steam Dump Valve and monitoring the level decrease to determine the valve's capacity.
With the Steam Generator Stop Valve shut and the Atmos-pheric Steam Dump Valve open, the Steam Generator pressure decreased until the differential pressure between that and the other Steam Generators reached 100 psid and caused the initiation of safety injection.
DESIGNATION OF CAUSE OF OCCURRENCE The safety injection was caused by opening the Steam Generator Atmospheric Dump Valve with the Steam Generator Stop Valve shut causing the Steam Generator pressure to drop in that Steam. Generator.
This allowed the 100 psi bP on 2 out of 3 pressure channels and on 1 out of 4 Steam Generator logic to complete the safety injection actuation as designed.
ANALYSIS OF OCCURRENCE The following is a list of major items that were reviewed for.their safety implications:
a)
Reactor Coolant Pumps Cooldown Rate.
There was approximately a
10o change in Reactor Coolant System temperature from 547oF to 537oF over approximately a minute and a half.
The cooldown was terminated by shutting the Steam Generator Atmospheric Steam Dump Valve The overall cooldown rate was,less severe than that experienced during a reactor trip without safety injection.
Report PSI-Ql November 6, 1978 Page Two (2) b)
Thermal Affects of'Safety Injection.
c)
It is estimated that during this occurrence the centrifugal charging pumps ran approximately one minute and injected 300 gallons of borated water'nitially at 165oF into the Reactor Coolant System.
A complete safety injection cycle would deliver over 350,000 gallons of borated water a'- a tempera-ture significantly lower than the 300 gallons delivered
- during this inadvertant safety injection.
This is the first inadvertant safety injection in which water was injected into the Reactor Coolant System and would conservatively constitute less than 1.5/10;000 of allowable cycles.
Effects on the Emergency Core Cooling System Piping (ECCS)
The piping'and supports in the ECCS were given a thorough visual inspection to determine.if any mechanical damage was experienced during the'safety injection.,
There was no evidence of any mechanical damage or abnormal movements of the piping.
0 INDIANA 5 MICHIGAN POCR COMPANY DONALD C.
COOK NUCLEAR PLANT Operating License DPR-74 Docket No.
50-,316 Special Report SI-02 SAFETY INJECTION ACTUATION - AUGUST 4, 1978 CONDITIONS PRIOR TO OCCURRENCE The reactor was in Mode 3, Hot Standby, with the Reactor Coolant System average temperature at approximately 530oF and the system pressure at approximately 2200 psig.
Reactor Coolant Pump number 3 and.4 were running to provide flow.
All four Main Steam Stop Valves were closed.
The unit was in a.
scheduled shutdown for repairs to the Moisture
=
Separator Reheaters and was being cooled down to.repair RC-123 (bypass around Pressur izer Spray Valve NRV-164).
4 DESCRIPTION OF OCCURRENCE The leak in RC-123 had been found while conducting a leak rate check of the reactor coolant system.
RC-123 is unisolable from the reactor coolant system.
A cooldown was in progress to take the unit down to less than 210 F and 50 psig.
The cooldown was being made using the number 2 and 3 steam generator atmospheric steam dump valves to reduce temper ature.
Number 2 and 3 steam generator steam pressure was reduced to the point where the differential pressure between these steam
'generators and number 1 and 4 steam generators reached 100 psid and caused initiation of safety injection.
DESIGNATION OF CAUSE OF OCCURRENCE The safety injection was caused opening the steam generator atmospheric dump valves.on number 2, and 3 steam generators too far and causing the steam generators to depressurize faster than number 1 and 4 steam generators.
This allowed the 100 psi AP on 2/3 pressure channels on 1/4 steam generators logic to complete and safety injection actuated as designed.
ANALYSIS OF OCCURRENCE The following.is a list of major items that were reviewed for their safety implications:
a}
Reactor Coolant System Cooldown Rate There was approximately a 10oF change from 539oF to 529oF over approximately two minutes.
The cooldown was termi-nated by shutting number 2 and 3 steam generator atmos-pheric steam dump valves.
The overall cooldown rate was
I Report p SI-02
~
November 3, 1978 Page Two (2)
. ~
less severe than that experienced during a reactor trip without safety injection.
b)
Thermal Effects of Safety Injection It is estimated that dur ing this occurrence the Centrifugal
- Charging Pumps ran approximately 1.5 minutes and injected 450 gallons of borated water. initially at 165 F into the Reactor Coolant System.
A complete safety injection cycle would deliver over 350,000 gal-lons of borated water at a temperature significantly lower than the 450 gallons deli-vered during the inadvertant safety injection.
This is the second inadvertant safety. injection in which water was injected into the Reactor Coolant System and conservatively would constitute 1'ess than 1.8/10,000 of allowable cycles.
c)
Effects on the Emergency Core Cooling System Piping,(ECCS)
The piping and supports in the ECCS were given a thorough usual inspection to determine if any mechanical damage was experienced during the safety injection.
There was no evidence of any mechanical damage or abnormal movements of the piping.
t g
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