IR 05000309/1986008
| ML20202E198 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/26/1986 |
| From: | Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20202E181 | List: |
| References | |
| 50-309-86-08, 50-309-86-8, NUDOCS 8607140304 | |
| Download: ML20202E198 (8) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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Docket / Report: 50-309/86-08 License:
DPR-36 Licensee:
Maine Yankee Atomic Power Company Inspection At: Wiscasset, Maine Dates:
May 11 - June 16, 1986
Inspectors:
Cornelius F. Holden, Senior Resident Inspector Jeffrey Robertson, Resident Inspector Norman J. Blumberg, Lead Reactor Engineer Righard truckmeyer, Radiation Specialist Approved:
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. '4(N 4 86 L/. TrippV Chief, Reactor Projects Section 3A Date Summary:
Inspection on May 11 - June 16, 1986 (Report No. 50-309/86-08)
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Areas Inspected:
Routine resident inspection (286 hours0.00331 days <br />0.0794 hours <br />4.728836e-4 weeks <br />1.08823e-4 months <br />) of the control room, accessible parts of plant structures, plant operations, radiation protection, physical security, fire protection, plant operating records, maintenance and
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surveillance.
Results:
No violations were identified.
The plant trip and subsequent review I
efforts are detailed in Section 7.
Several minor discrepancies between the ser-vice water system and controlled drawings were identified (Section 4).
Inservice
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Testing Program changes (Section 11) and the Annual Radiological Environmental j
Monitoring Program Report (Section 12) were reviewed during this inspection.
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DETAILS 1.
Persons Contacted i
Witnin this report period, interviews and discussions were conducted with
.various licensee personnel including plant operators, maintenance technicians i
and the licensee's management staff.
2.
Summary of Facility Activities At the beginning of the report period, the motor-driven feed pumps were in
service, limiting reactor power to 97%.
The turbine-driven feed pump was out of service due to discharge pressure control problems.
On May 28, power was reduced to 54% due to a leak on the "A" motor-driven feed pump discharge pip-i ing.
The turbine-driven feed pump was placed in service on May 29 and power
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j was increased to 75% and held there for valve testing and mussel control j
operations.
Power was increased to 97% on May 31, but was reduced to 90% when an electronic controller (HD-A-180) in the feed system failed.
Power was re-
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turned to 97% on May 31.
Because of steam generator level fluctuations, power was maintained at or below 99% until June 12 when the reactor tripped.
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rupture disk failure resulted in the inadvertent closure of #2 steam generator excess flow check valve (EFCV) and the reactor tripped on low steam generator
level (details in Section 7).
A startup was performed on June 13 and power j
was increased to 20% but reduced to 15% due to high chloride concentrations in the condensate and steam generators.
On June 14, power was increased to a
80% and then to 85% on June 16.
Power is being maintained at 85% while fur-ther investigations continue on the EFCV rupture disk failure.
3.
Followup on Previous Inspection Findings
a.
(Closed) IE Bulletin (IEB 50-309/79-80-14).
Seismic Analysis of as-built safety related piping systems.
The licensee is participating in a Seis-mic Design Margin Program sponsored by NRR which will review the as-built
seismic condition of the plant.
The review compares current seismic in-
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formation (Probabilistic Risk Assessments, seismic testing and earthquake experience) with as-built plant design.
Following the completion of this
research program, NRR will issue a Safety Evaluation Report which will document the resolution of the seismic issue for Maine Yankee.
b.
(Closed) IE Bulletin (IEB 50-309/79-BU-018).
Environmental Qualification
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of Electrical equipment.
Inspection of this area is being coordinated with IE and will result in a separate inspection report.
c.
(Closed) Unresolved Item (UNR 50-309/82-19-08).
Adequacy of loss of load safety analysis for all allowed operating conditions required formal i
documentation for steam generator and pressurizer safety valve operabil-ity.
Yankee Atomic performed an analysis to evaluate the limiting over-pressure event, a loss of load without a direct reactor trip to assure that peak pressure greater than 110% of design would not occur.
This evaluation was done assuming only 2 of 3 pressurizer safety valves and
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15 of 18 steam generator safety valves were operable as allowed by Tech-nical Specifications.
The results of this analysis indicate that, for the current steam generator level and useable heat transfer area, a loss of load accident would not cause peak pressure to exceed 110% of design for the minimum operable safety valves.
d.
(Closed) Violation (NCS 50-309/85-34-02).
Failure to adequately imple-
ment a Measuring and Test Equipment (M&TE) Program.
As stated in the licensee's response letter dated March 4, 1986, the QA Department was i
requested to increase surveillance of the M&TE program to assure that new administrative controls provide the proper documentation of the use and issue of measuring and test equipment.
A review of the four audits performed by QA since May has determined that they have provided meaning-ful feedback and resulted in recommendations to improve administrative controls.
The corrective actions adequately address the deficiencies noted in the Notice of Violation.
e.
(Closed) Violation (NC5 50-309/85-34-04).
Uncontrolled drawings in use were not periodically verified.
The licensee has removed uncontrolled I
drawings from the control room and several other locations.
Those draw-ings that were considered to be necessary were replaced with controlled drawings.
A reinspection of this item found no discrepancies.
4.
Routine Periodic Inspections a.
Daily Inspection During routine facility tours, the following were checked: manning, ac-cess control, adherence to procedures and LCO's, instrumentation, recor-der traces, protective systems, control rod positions, control room an-nunciators, radiation monitors, radiation monitoring, emergency power l
source operability, control room logs, shift supervisor logs, and i
operating orders.
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b.
System Alignment Inspection i
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Operating confirmation was made of selected piping system trains.
Ac-
cessible valve positions and status were examined.
Power supply and breaker alignment was checked.
Visual inspection of major components was performed.
Operability of instruments essential to system perform-
ance was assessed.
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A walkdown was performed on the Service Water (SW) system comparing the mechanical system drawing with as-installed equipment.
The operating procedure was also used to verify proper system alignment.
f Various minor discrepancies were found between the procedure, drawing j
and system including the following:
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Several instrument isolation valves and two local pressure indica-tors shown on drawing not installed.
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Valve SW-168 incorrectly labeled as SW-22.
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Pressure switch downstream of RW-129 incorrectly identified on FM.
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Procedure 1-15-3, Service Water System, references Pan Alarm Re-sponse sheet #G-4-5 vice proper response sheet #G-4-6.
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The isolation valve (RW-129) for a pressure switch ir the raw water to
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the "B" SW pump bearing cooling supply was found isolated.
This pressure switch provides an alarm and a signal to a solenoid operated valve on low pressure in the raw water supply to the SW pump.
The solenoid valve repositions to provide bearing cooling from tne backup supply (discharge of the pump).
A redundant pressure switch on the "C" SW pump was avail-able to perform the:e functions.
The licensee was notified and investi-gated the cause of the isolated pressure switch, li The licensee took prompt action to correct the discrepancies and to look for further discrepancies not identified by this inspection.
c.
Biweekly Inspections During plant tours, the inspector observed shift turnovers, chemistry sample results and the use of radiation work permits and Health Physics
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procedures.
Area radiation and air monitor :se and operational status was reviewed.
Plant housekeeping and cleaniiness were evaluated.
d.
Plant Maintenance The inspector observed and reviewed the following maintenance and problem investigation activities to verify compliance with regulations, admini-strative and maintenance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualifications, radiological controls for worker protection, fire pro-tection, retest requirements, and reportability per Techni::al Specifica-tions:
EFCV troubleshooting
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RWST Temperature Control Maintenance d
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Post Maintenance Test of waste gas compressor C38.
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e.
Surveillance Testing
The inspector observed parts of the following tests to assess performance
in accordance with approved procedures and LCO's, test results, removal j
and restoration of equipment, and deficiency review and resolution:
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ECCS valve testing
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Steam line radiation monitor calibration
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EDG gage calibration.
5.
Observations of Physical Security Checks were made to determine whether security conditions met regulatory re-quirements, the physical security plan, and approved procedures.
Those checks included security staffing, protected and vital area barriers, vehicle searches and personnel identification, access control, badging, and compensatory meas-ures when required.
6.
Radiological Controls Radiological controls were observed on a routine basis during the reporting period.
Standard industry radiological wurk practices, conformance to radio-logical control procedures and 10 CFR Part 20 requirements were observed.
Independent surveys of radiological boundaries and random surveys of non-radiological points throughout the facility were taken by the inspector.
7.
Plant Trip At 1:19 p.m. on June 12, the plant tripped from 99% power on low steam genera-tor level.
The low level was a result of the inadvertent closing of #2 steam generator EFCV.
The EFCV closure caused steam generator pressure to increase, shrinking generator level which caused the low level trip, and opening the secondary safety valves on that steam line.
An Unusual Event was declared and terminated as required by the Emergency Plan for lifting the secondary safety valves.
The Post Trip Review found no conclusive reason for the EFCV to shut except for the possible failure of the rupture disk.
The excess flow check valve uses two air pistons to hold the valve open.
Air pressure is provided by two dedicated air compressors.
When the EFCV is open, a paddle remains in the steam flow.
If steam flow becomes excessive (as it would during a steam line break accident) the force exerted on the paddle pulls the disk of the valve into the steam flow.
The larger surface area of the disk results in more closing force on the valve disk.
As the valve begins to shut, the pistons in the air cylinders increase the pressure causing the rupture discs to burst against a set of knife edges and release the pressure of the air cylinders.
The EFCV then goes shut to stop steam flow.
Two previous plant trips have been attributed to rupture disk failure.
Following the replacement of the failed rupture disk, the plant cycled the
- 2 EFCV six times satisfactorily.
While checking for any leaks at the rupture disk joint, the newly installed rupture disk failed. Following the replacement of this rupture disk, a similar failure of the rupture disk caused the valve to shut.
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The plant held numerous discussions with the valve manufacturer and the rup-ture disk manufacturer.
The plant concluded that installation of the rupture disks is critical.
Storage, handling and cleanliness of installation all play a part in proper operation of the rupture disks.
The licensee replaced the failed rupture disks under controlled conditions and subsequent testing of the valve was satisfactory.
The licensee also decided to reduce the pressure of the rupture disk air system since there is some speculation that the normal operating pressure is causing premature failure.
The Plant Operations Review Committee (PORC) reviewed the information concern-ing the rupture disks.
They determined that the plant was ready to operate.
The plant manager decided to limit power to 85 percent until more data could be gathered concerning optimum pressure setting for the rupture disk air sys-tem.
The licensee also plans to participate in a testing program with the manufacturer.
Longer term actions include investigating the EFCV design for possible improvements that will increase system reliability.
8.
Biofouling of Cooling Water Heat Exchangers An inspection was performed in accordance with TI 2515/77 to determine actions taken to recognize and prevent biofouling of safety-related equipment cooled by open cycle service water systems. The Primary and Secondary Component Cool-ing (PCC & SCC) System heat exchangers are the only components cooled by the Service Water System.
The PCC and SCC systems provide cooling for safety-related equipment.
The first indication of heat exchanger fouling is an increase in SW header pressure.
Inlet and outlet temperatures are also available to monitor heat exchanger performance.
Both the SW header pressure and inlet and outlet tem-peratures are recorded and reviewed twice a shift by the operations personnel.
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Operating procedures for the Service Water System contain instructions for Mussel Control (thermal backwash of condenser circulating water) and the Amer-tap cleaning system. Mussel Control is performed when SW header pressure in-creases above 25 psig.
If SW header pressure exceeds 25 psig after mussel control operations, the heat exchanger is opened and cleaned.
The Amertap cleaning system cleans the heat exchanger tubes by circulating cleaning balls through the tubes.
This is done for a 24-hour period every other day on one of the two outboard heat exchangers.
Discussions with operations personnel indicate that they are aware of indica-tions and required actions if heat exchanger performance significantly de-grades.
The fire protection system serves as a backup heatsink for some safety-related
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equipment.
The water supply for the fire protection system is a man-made pond supplied by city water.
A flow test is performed annually to verify the sys-
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tem capabilities have not degraded due to fouling or reduced pump performance.
No concerns were identifie,
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4 9.
Electric Driven Lube Oil Pump for the Charging Pump On May 8, 1986, the plant shifted the operating charging pumps from the B pump to the S pump.
Approximately two minutes after the B pump was secured, the low lube oil pressure alarm sounded in the control room for the B pump.
An investigation by the operating crew revedled that the breaker for the motor-driven pump was energized and therefore no reason for the alarm could be de-termined.
Plant Engineering was called to assist in the determination of the
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cause of the low lube oil pressure.
The charging pumps are provided with two lube oil pumps.
One is electric driven and runs while the charging pump is shutdown to provide sufficient prelube to the pump.
The second lube oil pump is shaft driven and provides the lube oil needs while the charging pump is operating.
Further investigation and disassembly of the electric lube oil pump revealed that the idler arms of the pump were worn.
No immediate explanation of this wearing was available.
The electric driven lube oil pumps were part of the
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original installation.
Approximately one year ago, the normal mode of opera-
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tion of the electric pumps was changed from an intermittent operation (only l
required prior to startup and shutdown of the pump) to continuous shutdown operation.
This change was necessitated by power cable proximity and the requirements of 10 CFR 50, Appendix R.
The licensee continues to evaluate the failure of the electric driven lube oil pump.
10.
Radiological Environmental Monitoring Program The inspector reviewed the licensee's Radiological Environmental Monitoring l
Program annual report for 1985.
This report summarizes the results of the sampling and analyses of environmental media to determine the radiological impact of station operations.
These environmental media include air, water, vegetation, and aquatic plants and animals.
In addition, direct radiation is monitored by placement of thermoluminescent dosimeters at various locations around the station.
As a result of this review, the inspector determined that the licensee has complied with its Technical Specification requirements for sampling frequen-cies, types of measurements, analytical sensitivities, and reporting schedules.
The report included summaries of the laboratory quality assurance program and of the land use survey.
The analyses of environmental samples indicated that doses to humans from radionuclides of station origin were negligible.
11.
Inservice Testing Pumps and Valves (IST)
In a letter dated August 7, 1985 (MN-85-139), the licensee submitted to the NRC a request for relief from certain inservice tests required by the 1980 Section XI of the ASME Code as impractical to perform for certain Containment Spray System valves and the High Pressure Safety Injection System pumps suc-j
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tion gauges.
This relief request is currently under evaluation by NRC Region I.
As part of this evaluation, on May 12 and 13, 1986, a region-based reactor engineer performed an onsite review of the relief request and held discussions with licensee personnel responsible for the inservice test program at Maine Yankee.
During this onsite evaluation, applicable piping and instrument drawings and IST procedures were reviewed and the locations of the High Pressure Safety Injection pumps suction gauges were physically inspected.
The results of this review were presented to licensee representatives at a meeting on May 13, 1986.
On June 6, 1986, the licensee was requested by telephone to submit, in writing, a revised relief request based on observations noted during the onsite review.
The results of this review will be documented in separate correspondence and in the issuance of a formal NRC Safety Evaluation Report at a future date concerning this relief request.
12.
Review of License Event Report (LER)
The inspector reviewed the following LER to verify that the details were clearly reported, including accuracy of the description of cause and adequacy of corrective action.
The inspector had previously verified that appropriate corrective action was taken or responsibility assigned and that continued operation of the facility was conducted in accordance with Technical Specifi-cations and did not constitute an unreviewed safety question as defined in 10 CFR 50.59.
No discrepancies were identified.
LER #
SUBJECT 86-02 Controller Malfunction caused a Feedwater Flow Transient and Subsequent Plant Trip on High Steam Generator Level.
13.
Exit Interview Meetings were periodically held with senior facility management to discuss the inspection scope and findings.
A summary of findings for the report period was also discussed at the conclusion of the inspection.
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