IR 05000220/1986018
| ML18038A226 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/31/1986 |
| From: | Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18038A225 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-2.E.4.2, TASK-TM 50-220-86-18, 50-410-86-42, IEB-73-05, IEB-73-5, IEC-78-05, IEC-78-5, IEC-80-08, IEC-80-19, IEC-80-8, NUDOCS 8611110576 | |
| Download: ML18038A226 (26) | |
Text
- %.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
86-18/86-42 Docket No.
50-220/50-410 License No.
DPR-63/CPPR-112 Category B
Licensee:
Niagara Mohawk Power Corporation 300 Erie Boulevard Syracuse, New York 12302 Faci 1 ity:
Location:
Dates:
Inspectors:
Approved by:
Nine Mile Point, Units 1-and
Scriba, New York September 1,
1986 to September 30, 1986 W.A. Cook, Senior Resident Inspector C.S. Marschall, Resident Inspector G.W. Meyer, Project Engineer R.L. Nimitz, Senior Radiation Specialist W.L. Schmidt, Resident Inspector Linville, C ef, Reactor ojects Sectid 2C, DRP Date Ins ection Summar
Ins ection on Se tember
1986 to Se tember
1986 Re ort No. 50-220/86-18 and 50/410/86-42 Areas Ins ected:
Routine'nspection of the Control Rooms and accessible plant areas, station operations and work activities, surveillance testing, mainte-nance, Licensee Event Reports (LERs) Reviews, IE Bulletin 5 Circular Reviews, Allegation followup, TMI Action Plan Item Review, and safety system walkdowns.
This inspection involved 296 hours0.00343 days <br />0.0822 hours <br />4.89418e-4 weeks <br />1.12628e-4 months <br /> by the inspectors.
No violations were noted.
Results:
A single control rod scram occurred at Unit 1 on September 8,
1986 due to a scram isolation valve diaphragm failure (Section 2).
The licensee continues to pursue resolution of MSIV problems at Unit 2 (Section 2).
An allegation (RI-86-A-101) concerning the falsification of professional credentials was reviewed and closed (Section 10).
8611110576 861103 PDR ADOCK 05000220 G
DETAILS Persons Contacted The inspectors interviewed and discussed stat,ion activities with various licensee representatives and contractor personnel.
Summar of Plant Activities UNIT 1 The plant operated at full power throughout the report period.
On September 8,
1986, a single control rod scram occurred when the rubber diaphragm in the air operator of the Scram Outlet Valve failed.
The licensee attributes the diaphragm failure to aging'he licensee replaced the diaphragm and restored the rod to its original position in accordance with reactor physics procedures.
The inspector identified that a similar failure of the Scram Inlet Valve could potentially cause damage to the control rod drive internals.
It has been determined that the diaphragm that failed was installed in 1976 and that approximately half of the diaphragms currently in service were installed prior to 1969.
The licensee has taken two courses of action to evaluate the consequences of the diaphragm fai lure.
An analysis will be performed to determine the consequences of a diaphragm fai lure of a Scram Inlet Valve, and a representative sample of diaphragms in service wi 11 be examined for indications of aging.
Although the licensee has drafted a proposal for an analysis of the diaphragm failure, the proposal was still being processed as of the end of the reporting period.
In addition, examination of in-service diaphragms has been delayed due to the unavailability of replacement diaphragms from the vendor.
The licensee is considering analysis by a materials science expert to determine whether replacement diaphragms currently in stock can be used despite being past expiration of shelf-life.
In view of the potential effect on plant operations, licensee progress toward determining the significance of diaphragm failures and the extent of the aging problem appears slow.
The inspector s will follow licensee progress.
Followup Item (220/86-18-01).
UNIT 2 As stated in the last report, the licensee commenced disassembly of all the inside containment Main Steam Isolation Valves (MSIVs) to evaluate the cause of the seat leakage test failures.
At several locations on the ball, patches of tungsten carbide overlay were removed with indication of galling.
The valve seat rings and ball seating areas showed evidence of scoring.
All the MSIVs showed varying degrees of the same cladding removal, galling and scorin The licensee has tested several possible solutions to the leakage problems.
First, a
new ball was installed with the spring force on the seat ring redistributed.
The seat loading was changed to reduce the forces exerted during valve operation in the area where galling had occurred.
A second solution was to machine off the coating in the area of galling and reinstall the valve using the original seat ring spring package.
The purpose of this was to reduce the apparent high surface stresses in this area where the seat contacts the ball.
The third was to install a machined ball in a valve with a redistributed seat ring spring force.
After stroking and periodic leak rate testing of the three different valve configurations, the use of the new ball and the redistri-.
buted seat ring spring force proved the only successful alternative.
All eight MSIV balls are presently at the manufacturer for recoating.
The inside containment MSIVs will be assembled first and seat leak tested to ensure secondary containment integrity requirements are satisfied prior to fuel load.
The licensee is still pursuing resolution of the MSIV actuator problem.
A modification to the present actuator will hold the MSIV in the open posi-tion with hydraulic pressure rather than the presently installed mechanical latch.
When the hydraulic pressure is bled off, the existing spring assemblies will close the valve.
Testing and qualification of this modified actuator is in progress'
detailed review of the MSIV problems was conducted by a region based specialist inspector during the week of September 22, 1986.
The results of that inspection are documented in Inspection Report 50-410/86-53.
The licensee plans to meet with the NRC,staff to discuss MSIVs and their impact on the issuance of the fuel load license on October 15, 1986.
3.
Licensee Action on Previousl Identified Items (Closed)
FOLLOWUP ITEM (220/86-09-03):
Review revision and implemen-tation of Rod Worth Minimizer Operability Test, Nl-ST-V3.
Previously, 'this test could only be performed prior to startup.
It was revised so that Rod Worth Minimizer operability could be deter-mined under any plant operating condition.
The inspector reviewed, the additional Rod Worth Minimizer training committed to by the licensee.
The training session included a review of the applicable portions of Technical Specifications, Operations Procedures, the revised surveillance test procedure and events surr'ounding the Technical Specification violation.
The inspector also reviewed the simulator training on the Rod Worth Minimizer (RWM).
This training included identifying acceptable response of a RWM functional test, diagnosing failure of the RWM, performing Nl-ST-V3 and recognizing RWM malfunctions.
The licensee's programs were acceptable.
This item is close b.
(.Closed)
UNRESOLVED ITEM (410/85-04-02):
Accuracy of FSAR informa-tion.
To verify the accuracy of the FSAR, the licensee initiated numerous review processes coordinated by the NMPC Licensing Group.
These methods included:
a computer based FSAR text verification; engineering verifications by General Electric, Stone 8 Webster Engineering Corporation, NMPC site technical staff and corporate engineering; and an independent review by the United Energy Services Company.
In addition, the licensee's Nuclear Compliance and Verifi-cation group established an ongoing program for verification of the FSAR, the NRC Safety Evaluation Report and Technical Specifications.
Prior to and after identification of this item, a large portion of NRC inspection effort conducted at the Unit 2 construction site focused on the verification of as-built plant structures and systems as detailed in the FSAR.
A Region I team inspection was conducted in April 1986 to review the as-built condition of the plant.
The team concluded that the systems examined were in conformance with drawing requirements and constructed substantially in accordance with the description provided in the FSAR and in the NRC Safety Evaluation Report.
Based on a sampling review of the licensee's verification programs, previous NRC inspection results and the continuing review of the plant activities via the routine safety inspection program, this item is closed.
C.
(Closed)
FOLLOWUP ITEM (410/85-20-03):
Licensee to complete testing of safety related ventilation systems.
During an earlier review of the Standby Gas Treatment System, the inspector identified the following items:
System flow balancing was not completed.
Laboratory testing of representative samples of charcoal was not performed.
Bypass leakage testing of train B of the SBGT system has to be performed again because it did not meet Technical Specification requirements.
Work material and debris found on top of Train A of the SBGT system required removal and cleanup.
Bolts were found missing from a cover on Train A of the SBGT system.
Subsequent review by the inspector verified that the licensee satis-factorilyy resolved all of the above items.
This item is close d.
(Open)
FOLLQWUP ITEM (410/85-47-02):
Licensee to ensure that all Radiological Controls facilities and equipment needed to support fuel load, startup, and routine plant operation are in place and operable, as appropriately The licensee has adequate Radiological Control facilities and equipment to support fuel load.
Facilities and equipment to support startup and routine operation will be examined during a subsequent inspection.
Review of the following radiological control procedures for startup and power ascension has been conducted:
N2-SUT-02-00, Radiation Measurement Open vessel.
N2-SUT-02-HU, Radiation Measurement Heatup
~
N2-SUT-02-1, Radiation Measurement
- TC-1 (5-20% Reactor Power).
N2-SUT-02-2, Radiation Measurement - TC-2 (25-50% Reactor Power).
N2-SUT-02-3, Radiation Measurement
- TC-3 (45-75% Reactor Power).
N2-SUT-02-6, Radiation Measurement TC-6 (95-100% Reactor Power).
e.
These procedures control the performance of radiological surveys and specify when protected areas wi 11 be set-up and updated during reactor power level changes and normal plant evolutions.
Implementa-tion of these procedures will be reviewed in subsequent report periods.
This item remains open.
(Open)
CONSTRUCTION DEFICIENCY REPORT (410/86-00-21):
Failure of fire suppression deluge valve.
During an inadvertent actuation of the Reactor Building fire suppression system, the deluge valve failed due to the lack of proper drainage into the header to which the control lines drained.
During discussions with a licensee fire protection engineer, the inspector was informed that the drain line had been piped into the drain header after the preoperational testing of the deluge valve.
The fire protection engineer has walked down all other similar deluge valves to ensure that the control line drains are open.
The inspector verified that the drains for the deluge valves in the diesel generator area and the Turbine Building were open.
This item is open pending licensee action to reroute the drain line or establish administrative controls to assure a drain path for the control line drains for the Reactor Building deluge valve and licensee configuration verification of all similar deluge valves on statio ~
(Closed)
FObKWUP ITEM (410/86-02-03):
During a previous inspection, the load sequencing of Division I and II Emergency Diesel Generators for a simultaneous Loss of Coolant Accident (LOCA) and Loss of Offsite Power (LOOP),
as demonstrated by Interim Operating Procedure IOP-72,
"Standby and Emergency AC Distribution System",
did not appear to agree with the sequences described in FSAR Tables 8.3. 1 and 8.3.2.
The licensee verified that the sequencing of loads, as tested by IOP-72, is consistent with the diesel generator loading sequences in Tables 8.3-1 and 8.3.2 of the FSAR.
The inspector reviewed the verification process and results with the responsible engineers.
The inspector had no further questions.
This item is closed.
(Closed)
FOLLOWUP ITEM (410/86-18-01):
Review of Accident Monitoring Instrumentation SER.
Region I review of post accident monitoring instrumentation was completed and documented in Inspection Report 50-410/86-18.
At that time, the NRR licensing staff had not completed their review of this TMI Action Plan Item.
This inspector followup item was opened to identify any issues that required Region I action following issuance of the Safety Evaluation Report (SER).
The licensing staff review of the post accident monitoring instru-mentation was completed and documented in Supplement No.
4 to NUREG-1047, Safety Evaluation Report for NMP Unit 2, dated September 1986.
The inspector found that there were no additional issues.
This item is closed.
Plant Ins ection Tours During this reporting period, the inspectors made frequent tours of the Unit 1 and 2 control rooms and accessible plant areas to monitor station activities and to make an independent assessment of equipment status, radiological conditions, safety and adherence to regulatory requirements.
The following was observed:
Unit On September 17, 1986, while observing the plant's physical security system, the inspector noticed that a small table was placed on the far side of the metal detector from where the security guard was standing.
The table was intended for placement of small metal objects such as keys or change.
The guard was standing in a position which allowed viewing of the xray machine monitor, but view of the small table and contents placed on it was obstructed.
The inspector brought this to the attention of the security lieutenant on shift and the fixtures were rearranged to allow the guard to view objects placed on the table.
Unit 2 No discrepancies were noted'
5.
Surveillance TesbaObservations The inspector observed portions of the surveillance test procedures listed below to verify that the test instrumentation was properly calibrated, approved procedures were used, the work was performed by qualified personnel, limiting conditions for operation were met, and the system was correctly restored following the testing.
a.
UNIT
On September 7,
1986, the inspector witnessed the performance of N1-ST-M4, Emergency Diesel Generator Manual Start and 1 Hour Rated Load Test, Revision 17, dated May 8, 1986.
Although the diesel started normally, a trip on positive crankcase pressure occurred while the diesel was being manually synchronized to its bus.
When operators could find no reason for the trip, the pressure sensing device was reset, and the surveillance was then run successfully.
The shift supervisor declared the diesel inoperable and wrote a work request to investigate the cause of the diesel generator trip.
Trouble-shooting revealed no apparent cause, and post-maintenance testing was successfully performed.
b.
UNIT 2 (1)
The inspector reviewed N2-0SP-EGS-M002, Diesel Generator Operability Test Division 3, Revision 1, dated May 20, 1986, on September 2,
1986, and identified the following:
Procedure step 1. 1.4 had no specific requirement to verify that the diesel comes up to the correct RPM in 10 seconds and generator frequency within 13 seconds.
Procedure step 8.4 was marked as unsatisfactory.
This step ensures that the diesel starts and comes up to speed, voltage and generator frequency in the proper time periods.
There was no remark explaining why the data in Step 8.4 was unsatisfactory.
The inspector also noted that the surveil-lance test was not signed as completed satisfactorily.
These items were subsequently discussed with licensee represen-tatives'ith the following results:
Step 1. 1.4 has been changed to include the 10 seconds RPM and 13 seconds frequency requirements'he licensee may measure RPM in one of two ways.
When available, one channel of an eight channel recorder can be connected to the diesel RPM instrument.
The alternate method is by directly relating diesel speed to generator frequenc SKp 8.4 was unsatisfactory because the generator frequency was not in the specified frequency band within 10 seconds.
This requirement could not be satisfied using Technical Specification criteria at that time.
A change request was submitted and approved for the final draft Technical Specification to broaden the frequency band.
Remarks will be added to all previously performed proce-dures to state why data was unsatisfactory.
Procedure N2-OSP-EGS-M002 was listed on the Cold Functional Test List as being an exception.
This exception will be cleared by the Superintendent of Operations after all previously performed diesel surveillance test procedures are reviewed and found to comply with the revised Technical Specifica-tions.
The inspectors will review licensee action to revise the appli-cable diesel generator surveillance test procedures and the Cold Functional Test List in a subsequent reporting period.
Followup Item (410/86-42-01).
(2)
The inspectors reviewed completed surveillance tests of the High Pressure Core Spray (HPCS)
System to verify that the tests met the Technical Specifications and that the records were complete and accurate.
The inspector reviewed test procedures N2-OSP-CSH-M001,
-Q001, and -Q002, completed on September 18 and 19.
The inspector found that the tests met the requirements and were completed in a generally acceptable mannner.
However, the inspector noted a problem with the acceptance criteria on test procedure Q002 and its review.
Specifically, the acceptance criteria for the HPCS pump discharge pressure had been modified via a change notice to be 333
+ H, where H was the pressure due to the level of water in the Condensate Storage Tank (CST).
Due to confusion, this was filled out to be 669.76 psig, and then crossed out and entered as 188 psig, both grossly in error.
There was no technical problem, because the measured discharge pressure of 460 psig'xceeded the acceptance criteria of 368 psig.
However, the subsequent review by the shift supervisor and the operations supervisor did not identify and correct the calculation error.
As this was an isolated error, the inspector agreed that an acceptable action would be correction of the error and reemphasis to reviewers of the importance of an accurate review of the measured values versus the acceptance criteria in surveillance tests.
During a subsequent review of test procedure N2-0SP-CSH-Q002, the inspectors determined that a test change was issued which included a
new data/calculation sheet for determining pump head when recirculating back to the CST.
The new data sheet includes a signoff for the person computing the data and the reviewe P
'he in<ectors also verified that the surveillance test performed on September 19 was corrected and properly annotated.
The inspectors had no further questions.
No violations were noted.
6.
Ins ection of General Electric T e AK-F-2-25 Breakers The inspector reviewed the application and failure history of General Electric Type AK-F-2-25 Breakers at Nine Mile Point, Unit 1.
The results of the review were:
AK-F-2-25 Breakers are not used in safety-related applications at Unit 1.
Recirculation Pump field breakers in use are General Electric Type AK-F-1B-10 Breakers.
The Recirculation Pump Trip (RPT) function is accomplished by a second shunt trip coil on the field breakers for each of the five pumps.
Preventive maintenance is accomplished once every three cycles as required by Nine Mile Point Nuclear Station Unit 1 procedure Nl-EPM-C12, T
e AK Breaker/Motor Ins ection and Breaker Load Test.
A review of work requests from 1977 to the present revealed no problems pertaining to the AK-F-1B-10 field breaker s.
The inspector had no further questions.
7.
Review of Licensee Event Re orts LERs The LERs submitted to NRC, Region I were reviewed to determine whether the details were clearly reported, including accuracy of the description of the cause and adequacy of the corrective action.
The inspectors also determined whether the assessment of potential safety consequences had been properly evaluated, whether generic implications were indicated, whether the event warranted on site follow-up and whether the reporting requirements of 10 CFR 50.72, where applicable, and
CFR 50.73 had been
.
met.
UNIT
During this inspection period, the following LERs were reviewed:
LER No.
Event Berte Subject 86-22 Aug.
1, 1986
'ev.
Fire Watch Patrol Surveillance Requirement Exceeded 86-24 Aug. 3, 1986 Turbine Trip and Subsequent Scram During Testing with Reactor in the Shutdown Node 86-26 AUQ. 22, 1986 Reactor Shutdown Required by Technical Specifications with Followup Documented in NRC Inspection Report 50-220/86-17 No violations were noted.
8.
Safet S stem 0 erabilit Verification UNIT
On a sampling basis, the inspectors directly examined selected safety system trains to verify that the systems were properly aligned in the standby mode.
This examination included:
Emergency Diesel Generator System Core Spray System Containment Spray System No violations were noted.
9.
Licensee Action on IE Bulletins and Circulars:
The inspector reviewed licensee records related to the IE Bulletins and Circulars identified below to verify that:
the IE Bulletins and Circulars were received and reviewed for applicability; a written response was provided, if required; and the corrective action taken was adequate.
The following IE Bulletin and Circulars were reviewed:
UNIT 2 a.
IE Circular 78-05, Inadvertent Safety Injection During Cooldown, dated May 23, 1973.
Thi s Circular i s not applicable to BWRs.
Thi s Circular is closed.
b.
IE Circular 80-08, BWR Technical Specification Inconsistency -
RPS Response Time, dated April 18, 1980.
This Circular identified a
potential inconsistency between RPS response times specified in
Technical Sj@cifications'nd those used in the safety analysis for plants licensed prior to the development of GE Standard Technical Specifications.
The RPS response time used in the Unit 2 Technical Specifications is the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor unti 1 de-energization of the scram pilot valve solenoids.
This response time interval is consistent with the FSAR safety analysis interval and GE Standard Technical Specifications.
This Circular is closed.
IE Circular 80-19, Noncompliance with License Requirements for Medical Licensees, dated August 26, 1980.
This Circular is not applicable to commercial BWRs.
This Circular is closed.
d.
IE Bulletin 73-05, Manufacturing Defects in BWR Control Rods, dated October 4,
1973.
This Bulletin addressed BWR control rods which were assembled with absorber tubes inverted.
NMPC requested and received verification from GE (reference GE letter No.
NMP2-7067) that the NMP2 Control Rods were manufactured using absorber tubes designed to be inserted from either end, thereby eliminating the problem of potential inversion.
In addition, specific gA procedures were used to ensure this phase of control rod assembly was correct.
This Bulletin is closed.
ations During the inspection period, the inspectors conducted interviews and inspections in response to allegations presented to the NRC.
The inspectors and licensee actions resulting from these allegations are noted below:
UNIT 2 (RI-86-A-101) The NRC received information concerning a
NMPC contractor employee that had falsified his professional credentials at another nuclear facility.-
During the previous inspection period, (reference Inspection Report 50-410/86-39),
the inspector determined that the individual had been employed by the licensee as an engineer.
The inspector requested that the licensee evaluate this individual's work and assess the contract firm's background verification program.
The licensee determined that this engineer was employed between April and August 1986.
During this time, the individual performed technical reviews of two procedures and reviews of numerous responses to NRC IE Notices and GE SILS.
The inspector verified that each of the items reviewed by this individual was also reviewed by his immediate supervisor (a NMPC employee).
In addition, these documents
received muMple routine reviews by the various station committees specified in Administrative Procedure AP-2.0, "Production and Control of Procedures".
The licensee contacted the consulting firm, which employed the individual, to verify all the firm's employee credentials who have worked at or are currently working at NMP Units 1 and 2.
A copy of the procedure or policy by which the firm verifies their employee backgrounds and resumes was also requested.
A contractor represen-tative met with the licensee on September 10, 1986 to review their employee credentials and to provide the licensee with a copy of their employment procedure.
It was determined that the Professional Engineer license for the engineer identified could not be verified by the contractor with the state PE Board.
The professional credentials of all other employees contracted to NMPC were properly verified.
The inspector will review the licensee program for assuring adequate background checks prior to permitting site access.
This item will be reviewed during a subsequent inspections 11.
Three Mile Island Action Plan Items UNIT 2 As a result of the Three Mile Island (TMI) plant accident, generic reactor enhancements were developed by the NRC.
NUREG-0737 documents the specific action requirements.
The following TMI Action Plan Items were reviewed during this inspection period:
a o
(Closed)
TMI ITEM II.B.1, REACTOR COOLANT SYSTEM VENTS, (410/86-29-01).
NUREG-0737 required that licensees provide a means to vent noncondensible gases from the reactor coolant system.
The licensee has endorsed the position of the BWR Owners Group, which stated that the main venting capability is provided by the safety/
relief valves.
In addition, venting can be accomplished via the reactor core isolation cooling system, the reactor head vent line, and the shellside vent line on the RHR heat exchanger.
This position meets the requirements and is consistent with the SER and the FSAR.
This TMI ITEM is closed.
b.
(Open)
TMI ITEM II.E.4.2, CONTAINMENT ISOLATION DEPENDABILITY, (410/86-29-03).
NUREG-0737 Item II.E. 4. 2 lists seven requirements to improve the reliability of containment isolation.
The licensee provided its response to the NUREG positions in 'FSAR Section 1.10, paragraph II.E.4.2.
The NRC staff concluded, in SER Supplement 3,
Section 6.2.4.1, that the FSAR meets the requirements of this NUREG-0737 Action Plan Item.
The following documents were reviewed by the inspector to ensure that the FSAR requirements were adhered to.
POT-83 Primary Containment Isolation POT-61-1 - Preop Test of Containment Purge System
J
POT-94~ Preop Test of Transversing Incore Probe System N2-RSP-RMS-R103 - Op Check of Containment Purge System Isolation on High Rad Levels The inspector reviewed the test results of the above mentioned pre-operational test procedures to verify compliance with the containment isolation requirements specified in the FSAR.
Surveillance procedure N2-RSP-RMS-R103 has not been performed to date.
The inspector will review the completed surveillance procedure in a subsequent inspec-tion period.
This TMI ITEM remains open.
12.
Site Visits On September 22 and 23, 1986, Commissioner Kenneth Carr toured the Unit
and Unit 2 facilities and discussed the readiness for fuel load of Unit 2 with the licensee.
I
.
~Ei At periodic intervals and at the conclusion of the inspection, meetings were held with senior plant management to discuss the scope and findings of this inspection.
Based on the NRC Region I review of this report and discussions held with licensee representatives, it was determined that this report does not contain information subject to
CFR 2.790 restric-tion 'c