IR 05000315/1984015
| ML17321A348 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/12/1984 |
| From: | Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17321A347 | List: |
| References | |
| 50-315-84-15, 50-315-84-17, IEB-83-07, IEB-83-7, NUDOCS 8412130614 | |
| Download: ML17321A348 (14) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-315/84-15(DRP);
50-316/84-17(DRP)
Docket Nos.
50-315; 50-316 Licenses No.
American Electric Power Service Corporation Indiana and Michigan Electric Company 1 Riverside Plaza Columbus, OH 43216 Facility Name:
Donald C.
Cook Nuclear Power Plant, Units 1 and
Inspection At:
Donald C.
Cook Site, Bridgman, MI Inspection Conducted:
July 28, 1984 through August 31, 1984 Inspectors:
E.
R.
Swanson J.
K. Heller R. J.
Leemon Approved By:
G.
. Mri ht, Chief Projects Section 2A Date Ins ection Summar Ins ection on Jul
1984 throu h Au ust 31 1984 (Re orts No. 50-315/84-15(DRP).
50"316 84-17 DRP
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licensee actions on previous inspection findings; operational safety; surveil-lance; Licensee Event Reports; IE Bulletins; IE Circulars; plant trip review; Confirmatory Action Letter.
The inspection involved a total of'39 inspector-
.hours by three NRC inspectors including 34 inspector-hours off-shift.
Results:
No items of noncompliance or deviations were identified.
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Persons Contacted DETAILS 2.
M.
G. Smith, Jr., Plant Manager
"E. Townly, Assistant Plant Manager
"B. Svensson, Assistant Plant Manager T.
Kriesel, Technical Superintendent-Physical Science A. Blind, Technical Superintendent-Engineering K. Baker, Operations Superintendent D. Dudding, Maintenance Superintendent
~J. Stietzel, equality Control Superintendent T. Beilman, equality Assurance Supervisor The inspector also contacted a number of licensee and contract employees and informally interviewed operation, technical and maintenance personnel during this period.
"Denotes personnel attending exit interview on September 6, 1984.
Licensee Action on Previous Ins ection Findin s
(Closed)
Noncompliance (315/83-19-02):
Diesel Generator Day Tank valving error.
The licensee has implemented a revised method of implementing independent verification which includes.Technical Specification, safety related, important to safety and "sphere of jeopardy" equipment to ensure correct and positive equipment control (Plant Manager Standing Order.077).
'I In addition,.the procedure used to control the valve which was found mis-aligned was modified to require'hat the unique valve identification numbers be recorded.
These actions were verified by the inspector and this item is closed.
(Closed)
Noncompliance (316/83-20-02):
Inoperable Turbine Building Evacuation System Radiation Monitor.
The violation was apparently due to failure to recognize the inoperability of the radiation monitor.
Operations procedures now require a review of radiation monitor status every four hours.
This review is logged in the control room log.
This item is considered closed.
(Closed)
Noncompliance (315/83-19-05; 316/83-20-03):
Explosive mixture action statement was not met.
To resolve the communication problem that was considered to be at the root of the event, a communications agreement was established to coordinate responsibilities between the Chemistry and Operations group.
This arrangement has been relatively effective in ensuring compliance with Technical Specification action requirements.
The inspector also verified that procedure changes were made as described in the licensee response to prevent 'the accumulation of explosive mixtures.
This item is close L
(Closed)
Noncompliance (315/83-19-04; 316/83-20-04):
Corrective actions were not taken after the licensee identified that functional logic diagrams were incorrect.
The inspector verified that holders of the subject drawings were notified of the drawings questionable accuracy and that the drawings were removed from the control rooms.
Revised diagrams have since been issued.
This i'ssue is considered closed.
(Closed)
Open Item (315/83-19-06):
Following a reactor trip on November 22, 1983 a pressurizer code safety valve lifted prematurely.
Normal Operating pressure for the Reactor Coolant System (RCS) is approximately 2235 psig and was 2240 psig prior to the trip.
Following the trip pressure decreased due to cooling of the RCS.
When pressure recovered to 2235 psig one of the pressurizer safety valves opened and then closed, dropping RCS pressure to 2125 psig.
The NRC expressed concern over the premature lifting of the safety valve and it was agreed that a report of the results of the testing program would be provided to the NRC prior to the valve's reuse in the plant.
During the surveillance and maintenance outage which ended on August 10, 1984, all three safety valves were replaced and this included the ques-tionable valve which was previously installed as SV-45B.
When the inspector became aware that the valve was installed he reviewed main-tenance job orders, purchase orders for services and parts, available documentation of the investigation, and found that the valve had been refurbished prior to installation.
When the licensee was reminded of the overdue report it was signed and sent to the plant on the same day, August 30, 1984.
Based on the acoustic monitoring devices it appeared that valve SV-45A was the valve that had opened.
Based on downstream temperature readings it also appeared that SV-45A had lifted.
Subsequent evaluation including thermal-hydraulic analysis determined that there is no way to determine which safety valve lifts during a pressure relief utilizing the currently installed monitors.
Following some initial testing onsite, the licensee sent two valves, SV-45A and SV-45B to Wyle Laboratories where, with an AEPSC engineer and the valve manufacturer's (Crosby) representative, the two valves were tested and inspected.
Results of the testing were inconclusive, with SV-45B demonstrating somewhat erratic liftpoint behavior.
Some minor mechanical deficiencies were found in this valve, but none that adequately explained the premature lift event at 90K of the pre-service set pressure.
The licensee's summary evaluation dated August 30, 1984 draws no conclusion as to what caused the premature safety valve lift.
At the inspectors'equest the licensee agreed to submit a supplementary report to their Annual Operating Report.
During the review of this issue, the licensee has again demonstrated how ineffective their corrective action tracking and commitment followup systems are (this is an RPIP Action Item).
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3.
0 erational Safet Verification 4,
The inspector observed control room operation including manning, shift turnover, approved procedures and LCO adherence; reviewed applicable logs and conducted discussions with control room operators during the inspection period of July 28, 1984 through August 31, 1984.
Observa-tions of control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation monitoring systems, and nuclear and reactor protection systems.
Reviews of surveillance, equipment condition, and tagout logs were conducted.
Proper return to service of selected components were verified.
Tours of the auxiliary building, turbine building, and screenhouse were made to observe accessible equipment conditions, including fluid leaks, potential fire hazards, and control of activities in progress.
During a tour of the Unit 1 lower containment annulus the inspector found two open doors with sign affixed "Danger-High Radiation Area".
The inspector was informed that the doors are locked during fuel movement and the area is not currently controlled as a high radiation area.
The inspector discussed this with the Radiation Protection Supervisor and asked why the signs were not removed while the area was no longer controlled as a high radiation area.
The Radiation Protection Supervisor agreed that the sign should have been removed and agreed to review the matter.
By observation and direct interview it was verified that physical security was being implemented in accordance with the station security plan.
The inspector performed a review of the Control Room Emergency Core Cooling System instrumentation and valve lineup to verify that:
each accessible flow path valve was in its correct position; power (visual breakers and fuses)
was aligned.to actuate on automatic signal; essential instrumentation was operable; and, no condition existed that degraded the system.
Observations of the plant housekeeping/cleanliness conditions and the implementation of the radiation protection program and controls were made.
These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established per Technical Specifications,
CFR and Administrative Procedures.
No items of noncompliance or deviations were identified.
Honthl Survei.llance Observation The inspector reviewed Technical Specifications required surveillance testing on the systems listed below and verified that testing was per-formed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that'est results conformed with Technical Specifications and procedure
requirements and were reviewed by personnel other than the individual directing the test, and that deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
OHP 4030 STP.022 2 THP 4030 STP. 154 2 THP 4030 STP.187 East Essential Service Water System Operability Test Control Room area monitor surveillance test Pressurizer Power operated relief valve emergency and pressure alarms surveillance test
OHP 4021.001.002 2 OHP 4021.013.005 Reactor Start-up Visual Audio Count Rate Channel 12 THP 4030 STP.225 Low pressure Coq fire suppression No items of noncompliance or deviations were identified.
5.
Licensee Event Re orts Through direct observation, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportabi lity requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accom-plished in accordance with Technical Specifications.
Unit 1 RO 83-102/01T-0 RO 83-104/03L-0 RO 83-108/03L-0 RO 83-109/03L-0 RO 83-111/03L-0 RO 83-112/03L-0 RO 83-115/03L"0 Non-conservatism in Barton Transmitter (yodel 763, Lot 2) used for pressurizer pressure instruments Diesel generator output breakers failed to close Steam generator pressure transmitter inoperable Containment sump flow monitor inoperable Steam generator pressure transmitter inoperable Steam generator blowdown isolation valve declared inoperable when position indicator malfunctioned Fire doors inoperable to perform corrective maintenance RO 83-116/03L-0 Fire door inoperable due to hardware malfunction
RO 83-117/03L-0 and Rev 1 Steam generator feedwater flow instrument inoperable due to transmitter drift RO 83-118/03L-0 RO 83-119/03L-0 Unit 2 Fire door inoperable due to hardware malfunction Hydraulic snubber located on the pressurizer safety relief line found inoperable RO 83-104/03L-0 RO 83-106/03L-0 RO 83-107/03L-0 RO 83"109/03L-0 RO 83-111/03L-0 and Rev 1 Fire door inoperable due to hardware malfunction Rod B-6 position 'indication inoperable CD diesel generator inoperable for oil leak repair f
Debris left in upper containment Fire doors inoperable to perform corrective maintenance RO 83"112/03L-0 RO 83-113/03L"0 Incomplete start for AB diesel generator Cardox system inoperable for diesel generator room RO 83"114/03L-0 Cardox system inoperable for diesel generator room No items of noncompliance or deviations were identified.
6.
IE Bulletin Followu For the IE Bulletins listed below the inspector verified that the Bulletin was received by licensee management and reviewed for its applicability to the facility. If the Bulletin was applicable the inspector verified that the written response was within the time period stated in the Bulletin, that the written response included the information required to be reported, that the written response included adequate corrective action commitments based on information presented in the Bulletin and the licensee's response, that licensee management forwarded copies of the written response to the appropriate onsite management representatives, that information discussed in the licensee's written response was accurate, and that corrective action taken by the licensee was as described in the written respons ~
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IE Bulletin 83-07 Apparently Fraudulent Products Sold by Ray Miller, Inc.
IE Bulletin 83-07, Supplement
Apparently Fraudulent Products Sold by Ray Miller, Inc.
No items of noncompliance or deviations were identified.
IE Circular Followu For the IE Circular listed below, the inspector ver Hied that the Circular was received by the licensee management, that a review for applicability was performed and that the circular was determined to not be applicable.
IE Circular 80-16 Operational Deficiencies in Rosemont Trip Units and Pressure Transmitters No items of noncompliance or deviation were identified.
Plant Tri s On August 5, 1984 at 1414 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.38027e-4 months <br /> while operating at lOOX power a Unit 2 reactor trip occurred from loss of Control Room Instrument Distribution (CRID) No. II inverter.
The failure of GRID II resulted in an indication of low reactor coolant flow which, coincided with power greater than P-8 caused a reactor trip.
The cause of the GRID failure was unknown but attributed to a failed rectifier(s) and/or diode(s),
All were replaced and the inverter was returned to service.
The licensee identified several, pieces of equipment that did not function properly and required repair prior to returning the unit to service.
These included calibration of a Nuclear Instrument source range, Calibration of a steam flow meter and adjustment of the limit switches for two motor operated feedwater flow control valves.
Following the plant trip the inspector ascertained the status of the reactor and safety systems by observation of control room indicators and discussions with licensee personnel concerning plant parameters and emergency system status.
Except as described above all systems operated as designed.
The plant was made critical on August 7, 1984 at 1210 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.60405e-4 months <br />.
On August 14, 1984 at 1529 hours0.0177 days <br />0.425 hours <br />0.00253 weeks <br />5.817845e-4 months <br /> while operating at 100K power, Unit 1 experienced a reactor trip and single train safety injection as a result of failure of the GRID I inverter.
The reactor tripped when the loss of power caused an indication of low reactor coolant system flow with reactor power above the P-8 setpoint.
The safety injection occurred due to an indication of low steam line pressure (false signal from power low)
coincident with high steam flow (setpoint drops back on a reactor trip).
Only a single train of safety injection actuated because a deenergized relay did not pass the signal to the other train.
The cause of the GRID failure was water leaking from the nearby eyewash station being blown
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into the GRID housing and causing short circuiting.
Smoke in the area resulted in the fire alarm being sounded.
One control rod did not indicate fully inserted although the rod bottom light was on.
The operators emergency borated as required by the Technical Specifications.
Rod N-17 on 'Bank C later drifted from the indicated 10 steps (228 is fully withdrawn)
to the bottom.
No additional action was taken by the licensee.
The licensee isolated the water leak and replaced the entire GRID.
The inspector reviewed the status of the plant following the trip by record review and discussion with licensee personnel.
Except as noted, systems operated as designed.
The plant was returned to service on August 15, at 1841 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.005005e-4 months <br />.
No items of noncompliance or deviations were identified.
9.
Confirmator Action Letter
- A licensed operator annual requalification examination was administered the week of June 18, 1984.
Three of the four sections of the written exam were provided by the NRC.
After NRC grading, it was determined that six Senior Reactor Operators had not met the 80K overall and 70K in each section minimum scores specified in the licensee requalification program.
After the-completion of the shift on August 21, 1984 (the date of the Confirmatory Action Letter) the licensee removed the operators from performing licensed duties and placed them in an accelerated training program.
The NRC will approve the written re-exam, monitor its administra-tion, and review the grading of the exams.
The licensee is complying with the commitments made during the August 21, 1984 conference call.
On August 28, 1984, further review of the examination grading resulted in a passing score for 'one individual who was restored to licensed, duties on that date.
10.
Exit Interview The inspector met with the licensee representatives (denoted in Paragraph 1)
throughout the inspection and on September 6,
1984 and summarized the scope of the inspectio e Pt