IR 05000220/1986012
| ML18038A214 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/01/1986 |
| From: | Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18038A213 | List: |
| References | |
| TASK-2.B.2, TASK-2.K.1, TASK-2.K.3.25, TASK-2.K.3.28, TASK-TM 50-220-86-12, 50-410-86-39, IEB-86-001, IEB-86-1, IEIN-84-83, NUDOCS 8610140398 | |
| Download: ML18038A214 (32) | |
Text
- W.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
86-12/86-39 Docket No.
50"220/50-410 License No.
OPR-63/CPPR-112 Category B
Licensee:
Niagara Mohawk Power Corporation 300 Erie Boulevard Syracuse, New York 12302 Facility:
Nine Mile Point, Units 1 and
Location:
Scriba, New York Oates:
July 7, 1986 to August 31, 1986 Inspectors:
W.A. Cook, Senior Resident Inspector R.A.
Gramm, Senior Resident Inspector C.S. Marschall, Resident Inspector G.W. Meyer, Project Engineer W.L. Schmidt, Resident Inspector Approved by:
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Ins ection Summar
Ins ection on Jul
1986 to Au ust
1986 Re ort No.
50-220/86-12 and 50/410/86-39)
Areas Ins ected:
UNIT 1 - Routine inspection by resident inspectors of station activities, surveillance testing, maintenance, Licensee Event Reports (LER's),
and safety system walkdowns.
UNIT 2 - Routine inspection by resident and regional inspectors of work activi-ties, procedures and records relative to TMI Action Plan items, IE Bulletins, and Notices.
Two allegations were reviewed.
The inspectors also reviewed licensee action on previously identified items and performed plant inspection tours.
The inspection involved 265 hours0.00307 days <br />0.0736 hours <br />4.381614e-4 weeks <br />1.008325e-4 months <br /> by the inspectors.
Results:
No violations were identified.
A summary of the MSIV problems impacting fuel load at Unit 2 is documented in paragraph 2.
Twelve NRC open items and three TMI items (UNIT 2) were closed.
An allegation concerning welder qualifications (UNIT 2) is discussed in paragraph 9.
8610140398 861007 PDR ADOCK 05000220.
G PDR
DETAILS 1.
Persons Contacted The inspectors interviewed and discussed station activities with various licensee representatives and contractor personnel.
2.
Summar of Plant Activities UNIT 1 The plant operated at full power throughout the period with the following exceptions:
On July 7, 1986 a shutdown was conducted to make repairs to the ¹11 Control Rod Drive Pump.
Startup commenced July 13, 1986.
On August 1, 1986 a shutdown was conducted to make repairs to Emergency Condenser Condensate Return Isolation Valve 39-06, which was leaking excessively.
Startup began August 6, 1986.
On August 22, 1986, the unit was shutdown to check the operability of the control rods after making adjustments to scram isolation valve packing glands.
This event will be covered in detail in Inspection Report 50-220/86-17.
UNIT 2 The licensee's projected fuel load date continues to slip due to problems encountered with the Main Steam Isolation Valves (MSIVs).
One problem involves the failure of the valves to stroke in the 3 to 5 second time requirement.
The MSIV actuators do not operate fast enough due to a mechanical latch design deficiency.
The licensee is pursuing a design change to remove the mechanical latch mechanism.
The second problem involves the failure of the valves to meet Type C leakage requirements.
At the conclusion of this report period, the licensee was making plans to disassemble the inside containment MSIVs to inspect the ball and seats.
Both MSIV problems have been reported to the NRC in accordance with 10 CFR 50.55(e).
3.
Licensee Action on Previousl Identified Items (Cl osed)
CONSTRUCTION DEFICIENCY (410/83-00-23):
The inspector reviewed the following documents:
Inspection Report 50-410/86-09.
POT-200 - Secondary Containment Leak Tes SWEC upgrade(
the Reactor Building roof to a gA Category I structure.
The resident inspector reviewed and concurred with SWECs "accept-as-is" proposal, as discussed in Inspection Report 50-410/86-09.
This item remained open pending the satisfactory completion of Preoperational Test Procedure (POT)-200.
The test was satisfactorily completed August 1, 1986.
This item is closed.
(Closed)
FOLLOWUP ITEM (410/83-18-113):
Document control problems.
During the I&E CAT Inspection numerous document control problems were found in the design control and construction documents.
This item addressed the issue generically, and six open items were assigned for specific findings.
The six specific findings have been closed, and recent inspections of this area have found acceptable document control.
Specifically, the As-Built Inspection (No. 50-410/86-13),
which reviewed many design and construction documents, found document control to be acceptable.
This item is closed.
(Closed)
VIOLATION (410/84-06-08):
Reactor building housekeeping.
A violation was issued for large quantities of untreated lumber and debris in the Reactor Building.
During numerous tours of the Reactor Building the inspector noted that a major effort to clean the area had taken place and that the lumber and debris had been removed.
Based on the good condition of the Reactor Building, this item is closed.
(Closed)
FOLLOWUP ITEM (410/84-18-04):
Site document control program.
Associated NRC open items 83-18-24, 83-18-67, 83-18-76, 83-18-77, and 83-18-87 have been previously closed in NRC inspection reports 85-44 and 86-09.
The licensee instituted the following corrective measures:
Consolidated the number of site drawing issue stations.
Performed additional surveillance to ensure necessary drawings were available at the issue stations.
Implemented a computerized posting system.
Revised the practice of posting design changes against only lead drawing sheets.
J Revised the Advance Change Notice program to provide for timely approval and incorporation into design drawings.
This item is closed.
(Closed)
CONSTRUCTION DEFICIENCY (410/85-00-19):
Control of valve disassembly and reassembly.
The following documents were reviewed by the inspector:
SWEC-t~e C IR QPSG0023 Deficiency Reports M04609, M04616, and M07767.
ITT-Grinnell Corrective Action Request 668.
ITT Quality Finding Report 109.
NMPC Surveillance M-84-407.
ITT Inspection Report 85-11-19982.
ITT Stop Work Order 84-05.
ITT procedure FQC-4.2-27-2,
"Valve Disassembly and Reassembly".
N 5 D IG-4855.
ITT operations on valve disassembly and reassembly were halted on November 6, 1984.
A procedure was written to control the field work and associated documentation for valve disassembly operations.
Quality Control inspections were required for visual inspection of the internals and seating surfaces, for proper segregation of valve parts, torque verification, and matchmark alignments.
ITT performed a review and determined that only valves provided for specification P800A or P304A would have to be disassembled in accordance with vendor instructions during installation.
Planner packages for the P304A Contromatics ball valves were reviewed by ITT.
Any valve, for which the associated documentation did not verify valve reassembly, was identified as potentially nonconforming to the Startup depart-ment.
The inspector was informed that during preliminary testing, the valves were manually stroked and determined operable.
Three service water system ball valves were disassembled and the heat numbers were verified for the internals.
Additional valve planner packages were reviewed by the licensee and the appropriate documen-tation associated with the valve disassembly was found.
Based upon the licensee investigation and resulting corrective actions, this item is closed.
f.
(Closed)
CONSTRUCTION DEFICIENCY (410/85-00-21):
RCIC Turbine Steamline Drain set point.
The licensee determined this item was nonreportable based on the following:
The primary concern was damage caused by water hammer to piping and the turbin The tuPS'ine will not be affected by water slugs (page 2-1 of Terry Turbine Tech.
Manual)
The piping was analyzed using a computer flow code and all stresses were less than the allowable stress.
The set point was recalculated.
The inspector concurred with licensee's position.
This item is closed.
g.
(Closed)
FOLLOWUP ITEM (220/85-10-01):
Upon completion of the analyses of water samples by the licensee and Brookhaven National Laboratory (BNL), a statistical evaluation was to be made.
During inspection number 50-220/85-10, the holding tank, reactor coolant and waste collector tank were sampled for analysis.
Dupli-cate samples were sent to BNL for independent verification of analysi s.
The analyses were completed and an evaluation was performed:
SPLIT SAMPLE COMPARISON Sample Source Chemical BNL NMP-1 Parameter Value Value Com arison Coolant chloride(ppb)
25 N/A Waste Tank si 1 ica(ppb)
615 666 N/A copper(ppb)
363 +49.1 639 + 51.6 disagree iron(ppm)
2.262 +.42 2.499 +.54 agree Holding Tank boron(ppm)
18,600 21,650 N/A The copper disagreement was due to insufficient acid in the BNL sample and the copper hydrolyzed.
The boron difference could be a sampling problem.
Overall, the analytical comparisons for the analyses were acceptable.
This item is closed.
h.
(Open)
FOLLOWUP ITEM (410/85-10-03):
Nitrogen inerting system opera-tion.
The inspector reviewed FSK-14-1E and found that valve AOV143 had been corrected to show that it fails closed.
The inspector was informed that during preoperational test N2-POT-201 a bypass leakage-test was performed.
The drywell was maintained at 3 PSI above the wetwell pressure and the measured leakage was acceptable.
The
inspector-retiewed operating procedure N2-IOP-61A, "Primary Contain-ment Vent Purge
& Nitrogen System" and ascertained that a precaution had been incorporated to preclude the injection of nitrogen below 55 degrees F.
This item remains open pending completion of preopera-tional test N2POT-88.
i.
(Closed)
CONSTRUCTION DEFICIENCY (410/86-00-08):
Fuse block damage in the Uninterrupted Power Supply (UPS).
The inspector reviewed the following documents:
SWEC Problem Report E-57 documenting problems encountered with Elgar inverters.
Correspondence from Elgar Division of Cooper Industries to SWEC identifying the fuse block connection as a potential problem.
NMPC Deficiency Reports (DR) 18699 and 18700 authorizing work on 2VBA"UPS2A 8( 2VBA*UPS2B.
NMPC Quality Control Inspection Reports(QCIR)
2-86-2366 and 2-86-2545 which document completion of satisfactory QC inspec-tion.
The problem of fuse block damage was corrected by a design change made by the Elgar Division.
The modification involved changing the bus bar to a flexible metallic braid, eliminating the high stress placed on the plastic fuse block when the bus bar was torqued to the attachment stud.
A visual inspection of reworked fuse blocks was performed; no damage was identified.
This item is closed.
j.
(Closed)
CONSTRUCTION DEFICIENCY (410/86-00-09):
Oil leakage from Borg-Warner electro-hydraulic actuated valves.
The inspector reviewed the following documents:
SWEC letter 9N2-18,864 Correspondence between Copes-Vulcan and SWEC outlining the reorientation of the pump-motor assembly to be performed on the actuator.
Engineering and Design Coordination Report (ELDCR) Z81071.
Work completion signed for on 6/6/86.
Inter-Office Communication from lead engineer P.
Conte to N.
Cardone dated 6/12/86, stating that all valves were reinstalled, and preliminary testing was conducted under a Borg-Warner representative's supervision.
A visual inspection was conducted on all seven repaired valve actuators; no oil leakage was observed.
This item is close k.
(Closed)
FONWUP ITEM (410/86-02-04):
SRV maintenance and testing.
The inspector reviewed the following:
FSAR Section 5.2.2. 10.
N2-MPM-R35 Main Steam Safety Relief Valves Preventive Main-tenance.
N2-NMP-Removal and Installation of Main Steam Safety Valves.
N2-MSP-MSS-Rl - Main Steam Safety Relief Valve Verification.
NMPC ASME Section XI Inservice Testing Program.
As identified in an earlier report, the licensee did not address the FSAR requirements for pneumatic power actuator opening and closing, testing of bolted closures, and testing of pneumatic actuator leakage as specified in Section 5.2.2. 10.
The licensee has since updated the IST program, and maintenance and surveillance procedures to comply with testing specified in the FSAR.
This item is closed.
l.
(Closed)
FOLLOWUP ITEM (410/86-09-45):
A 10 CFR 21 report, filed on March 13, 1986, described a problem concerning galvanic corrosion between the carbon bearings and stainless steel shaft of Clow Corporation butterfly valves.
The corrosion caused excessive stroke time for these valves.
NMP2 has two valves which are equipped with the subject
"METCAR GRADE M-10" carbon steel bearings.
(2RHS'MOV 1A 8 1B).
These valves are located outside primary containment on RHR suction lines from the suppression pool.
These valves are normally key locked open, except during the Shutdown Cooling mode of plant operation, in which the valves are closed to prove a flow path from the RPV to RHR pumps.
These valves are manually controlled.
There-fore, stroke time is not a critical parameter.
These two valves are exercised every three months in accordance with the licensee's IST program.
The inspector also determined that the control logic switch is bypassed through the first 5% of valve travel.
The stall torque-is approximately six times the value of the breakaway/running torque, and is sufficien't to overcome the mechanical binding.
This item is closed.
(Closed)
FOLLOWUP ITEM (410/86-09-47):
Standby Gas Treatment System Preoperational Test Review.
The inspector reviewed the following:
FSAR Section 6.5. POT-61-2, Standby Gas Treatment System (SGTS) Preoperational Test Procedure POT-83, Primary Containment Isolation POT-200, Secondary Containment Leak Test
In Inspectiaa Report 410/86-09, inspectors identified that the following functions of the SGTS were not tested in POT-61-2:
system auto-initiation for low reactor water level and high drywell pressure.
-
FSAR section 6.5. 1.2 specifies that the system be capable of providing 3500 cfm flow within 25 seconds.
The first item was successfully demonstrated during performance of POT-83.
The second item was added as an acceptance criteria in POT-200.
During the performance of the latter test, the criteria could not be met.
This was documented in NMPC Deficiency Report 19705.
The mechanical backdraft dampers were reset and the system was retested satisfactorily during a subsequent performance of POT-200.
This item is closed.
4.
Survei llances Observations The inspector observed portions of the surveillance 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.
UNIT 1 N1-RSP-12/,
Routine Calibration of Reactor Building Ventilation Monitor, Revision 0, dated May 15, 1986.
Nl-ST-g3, HPCI Pump Operability Test, Revision 2, dated April 9, 1985.
N1-ST-W1, Control Rod Exercising, Revision 6, dated January 26, 1984.
N1-ST-IC2, Emergency Cooling Surveillance with an Inoperable Component Test, Revision 2, dated January 2,
1985.
No violations were noted.
5.
Maintenance Observations The inspector observed portions of various safety-related maintenance activities to determine that redundant components were operable, that these activities did not violate the limiting conditions for operation, that required administrative approvals and tagouts were obtained prior to initiating the work, that approved procedures were used or the activity was within the "skills of the trade", that appropriate radiological controls were properly implemented, that ignition/fire prevention controls
0
were properly implemented, and that equipment was properly tested prior to returning it to service.
During this inspection period the following activities were observed:
UNIT 1 WR 100187, 11 Control Rod Drive Pump rebuild.
WR 100168, 11 Motor Driven Feedwater Pump rebuild.
WR 100192, Trouble-shoot and Replace Liquid Poison Selector Switch.
WR 100665, Emergency Condenser 12 Isolation Valve 39-06 Inspection and Repair.
WR 100279 and WR 100280,LPRM 20-09A and 20-09C defective connection, replaced connector.
Inspector followup of events related to LPRM connector installation and repair wi 1 1 be documented in Inspection Report 50-220/86-17.
6.
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 1 During this inspection period, the following LERs were reviewed:
LER No.
Event Date Subject 86"02 Rev.
86-14 Feb.
1, 1986 Jun.
3, 1986 Inoperable Stack Gas Sample Pump Reactor Scram While Performing Surveillance Test
86-16 May 23, 1986 Potentially Inoperable Feedwater Pump High Level Trip System
86-17 Jun..1'986 Reactor Scram Resulting From Surveillance Testing 86-18 Jun.
18, 1986 Reactor Scram Due To Operator Error 86-19 Jun.
18, 1986 Loss of RWM During Start-UP With Less Than 12 Rods Withdrawn And Less Than 20% Power 86-20 Jun.
18, 1986 Leak From In-Core Water Sample Line 86-21 Jun.
18, 1986 Reactor Scram and HPCI Initiation Due to IRM Spike LER 86-19 contains two errors in the description of the events which occurred during the start-up of June 18, 1986.
The licensee committed to revise the LER to correct the errors.
The inspector discussed the reporting requirements of 10CFR21 and 10CFR50.73 with the licensee.
The licensee believed that an LER can be used to meet the requirements of 10CFR21.
The inspector pointed out the differences in reporting requ'irements with respect to timeliness.
The licensee has committed to treat the reporting requirements of 10CFR21 separately when an LER cannot be prepared within the time requirements of 10CFR21.
The inspectors will continue to monitor the licensee's events analysis and compliance with reportability requirements.
Inspector follow item (220/86-12-01).
No violations were noted.
7.
Safet S stem 0 erabilit Verification UNIT 1 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 note r
8.
Licensee Action on IE Bulletins and Information Notices The inspector reviewed licensee records related to the IE Bulletins and Notices'identified below to verify that: the IE Bulletins and Notices 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 Notice were reviewed:
UNIT 2 a.
(Closed)
IE Bulletin No. 86-01:
Minimum flow logic problems that could disable RHR pumps.
The licensee responded by letter dated June 9, 1986 (NMP2L-0738) that the potential problem did not exist at Nine Mile Point 2.
The inspector reviewed the following documents:
FSAR sections 6.3, 5.4.7, and 9'. 1.
FSAR figures 6.3-6, 6.3-7, and 5.4-13.
Residual Heat Removal System flow diagrams FSK-27-7.0, 7A, 7B, 7C, 7D, and 7E.
SWEC P&IDs," PID-31F, 31C, 31B, 31G and 31D.
GE Elementary diagram 807E170TY, sheets 14, 22 and 23.
SWEC ESK-6RHS26 and 6RHS27.
The design provides for electrical, mechanical, and spatial separa-tion of the three trains of Emergency Core Cooling Systems (ECCS).
Each pump discharge line is provided with a flow element to assure that the associated minimum flow valve is open to avoid pump dead head conditions.
The logic for the minimum flow valves is completely independent.
The inspector concurs with the licensee determination that a single failure will result in the loss of a single ECCS train, at worst.
The Nine Mile accident analyses have demonstrated that only two ECCS trains are necessary to safely shut the plant down.
This item is closed.
b.
(Closed)
Information Notice No. 84-83: Various battery problems.
The inspector reviewed the following documents:
GE letter NMP2-7047.
NMPC procedures:
N2-ERM&EN-V628, "Battery Equalizing Char ge" N2-EMP-73.5, "Station Batteries Cell and Intercell Connector Replacement" N2-EPM-GEN-W665,
"OC/UPS Weekly Checks" N2-EPM-BWS-Q667,
"24/48 VOC Specific Gravities and Battery Voltage N2-EPM-BYS-Q670,
"125 VOC Specific Gravities and Battery Voltage" N2-ESP-BYS-W675,
"125 Volts OC Weekly Battery Surveillance Test" N2-ESP-BYS-Q676,
"Quarterly Battery Surveillance Test" N2-ESP-BYS-R677,
"Div I/II/IIIBattery Intercell Resistance Test" N2-ESP-BYS-5Y683,
"Div I/II/IIIBattery Performance Discharge Test" N2-EPM-GEN-R671,
"125 VDC Batteries-Plates, Terminals and Rack'ntegrity".
GE and SWEC have evaluated the installed battery capacity and loads for the Class IE battery systems and found the batteries adequate.
Preoperational testing has been performed to confirm the adequacy of the installed batteries.
The plant surveillance and maintenance procedures were revised to include prohibitions from using organic solvents or hydrocarbon based grease on the battery cases which can induce cracks in the case.
This item is closed.
9.
~A>>
During the inspection period, the inspectors conducted interviews and inspections in response to allegations presented to the NRC.
The inspector and licensee actions resulting from these allegations are noted below:
UNIT 2 a.
(RI-86-A-47) The NRC was informed that welder qualifications had been improperly extended for a,craft foreman.
The inspector reviewed the
'following documents:
ASME Section IX QW-319.
NMPC Quality First File Q1P-86-0117.
SWEC Type C Inspection Report M6S03022.
SWEC Nonconformance and Disposition Report 1637 SWEC A9WE Section III guality Assurance and Control Manual Section 10.
gS-9.31 NM, "Welder/Welding Operator gualifications" CMP-6.9-11.82,
"Performance qualifications for Welders".
b.
SWEC welding engineering and quality control personnel reviewed the maintenance of welding qualifications for the individual in question.
SWEC identified that in several cases Welding Material Requisitions (WMRs) for another welder with the same last name had been incorrectly relied upon to extend the qualifications of the craft foreman'WEC further identified that on two occasions the craft foreman had returned all weld rod without actually performing any welding as documented by the WMRs utilized to extend the qualifications.
SWEC determined that, effective March 22, 1983, the foreman was not qualified to perform ASME welding.
The items welded under ASME by the foreman, after March 22, 1983, were reviewed and the welds were all nonsafety-related.
The welding engineering department and gC additionally reviewed the weld documentation for other nonproduction type welders and found proper extension oftheir qualifications.
The SWEC ASME guality Assurance and Control Manual has been revised to require that a supervisor witness welding for maintenance of qualifications.
The ASME welding qualification for the foreman was revoked on May 9, 1986.
The inspector additionally reviewed the welding qualifications maintenance log and associated WMRs for three welder s for the 1983 time frame; no disc-repancies were identified.
Based upon the licensee corrective actions this allegation is closed.
(RI-86-A-101): The NRC was informed of a contractor employee that had falsified his educational credentials at another nuclear plant.
The inspector was informed that the individual in question had been employed as an engineer by NMPC, but had left the NMP2 site.
The licensee was asked to further investigate the individual in question, to evaluate his work product and to assess the consultant's background verification program.
This allegation is open pending completion of the licensee's review.
10.
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 issues were reviewed during this inspection period:
a
~
(Closed)
ITEM II.B.2., Shielding Design Review, (410/86-29-02).
NUREG-0737 required each licensee to perform radiation and shielding design reviews of the spaces around systems that may, as a result of an accident, contain highly radioactive materials.
The design reviews should identify the location of vital areas and equipment
V
spaces in.w~h personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post accident operations of these systems.
The inspector verified that the licensee has performed dose calculation and shielding design reviews which demonstrate that the less than 15 mrem/hr criterion has been met for all vital areas requiring extended or continuous occupancy.
This item is closed.
(Closed) II.K.1.5, ESF Valve Positions, (410/86-09-20).
NUREG-0737 required that licensees review their controls for positioning of safety-related valves to assure proper operation of engineered safety features.
The licensee controls valve positions by means of valve lineup checklists in the operating procedures.
These lineups form the basis for the valve position controls utilized in maintenance and testing operations.
The valve lineups were derived from the valve positions contained in the design drawings and are being verified as part of the preoperational testing program.'he inspector reviewed the valve lineups for the Standby Gas Treatment System, the High Pressure Core Spray (HPCS)
System, and Loop A of the Residual Heat Removal (RHR) System in the operating procedures and compared the valve positions to those in the design drawings, i.e.,
the FSKs.
The inspector found the valve lineups to be.accurate and technically acceptable.
Also, the inspector reviewed the technical specification requirements concerning the monthly verification of flow path valves in the Low Pressure Core Spray ( LPCS),
Low Pressure Coolant Injection (LPCI)
Mode of RHR, HPCS, and Service Water Systems.
The inspector compared the design drawings, the Locked Valve List from procedure N2-0P-101A, and the surveillance procedures for the verification of LPCS, LPCI, and HPCS.
During the initial review the inspector found that valves in the minimum flow lines of LPCS and LPCI would be neither locked in position nor verified monthly.
The licensee had written the surveillance procedures only to verify the valves in the direct flow path from the suction source to the reactor.
The licensee agreed that proper functioning of the minimum flow line was an integral part of the flow path for an operable system, and the procedures were revised to verify these valves.
The inspector reviewed the, revised procedures later in the inspection period and found them acceptable.
The licensee's position acceptably met the requirements and was consistent with the FSAR and the SER.
This item is closed.
(Closed)
II.K.3.25, Power to Pump Seal Coolers (410/86-09-29).
NUREG-0737 required that licensees review the consequences of a loss of cooling water to the recirculation system pump seal coolers to ensure that seal failure and the resulting loss of coolant would not occur.
The licensee has endorsed the BWR Owners Group evaluation, which demonstrated that seal failure would not result in coolant
n
leakage io >@cess of existing reactor level control systems.
Further, the plant design includes the capability to connect the control rod drive pumps and the closed loop cooling system (the sources of cooling water to the seals)
to the stub bus, which can be supplied with power from the emergency diesel generators.
The inspector reviewed procedures N2-IOP-29, Reactor Recirculation System, which describes the seal temperature alarm and actions to respond to it, and N2-IOP-72, Standby and Emergency AC Distribution System, which provides the steps to re-energize the stub bus after loss of offsite power.
The inspector noted that the licensee has removed the automatic intertie between service water and closed loop cooling systems described in earlier versions of the FSAR and on which the SER was based.
The current licensee position is described in FSAR Amendment 27.
The licensee's position acceptably met the requirements and was consistent with the FSAR.
This item is closed.
d.
(Open) II.K.3.28, gualification of ADS Accumulators (410/86-09-31).
NUREG-0737 required that licensees ensure that the accumulators for the Automatic Depressurization System (ADS) provide the ability to cycle the ADS valves over a
100 day period in a hostile environment.
The design includes accumulators for each ADS valve, two receiver tanks for nitrogen in the Reactor Building,. and a bottled nitrogen supply system.
The inspector reviewed the design drawings ( FSK 12-1 series),
procedure N2-IOP-34 for the ADS valves, procedure N2-IOP-61A for the nitrogen supply system, and preop test N2-POT-34.
The inspector verified the accumulator leak test data from the preop test and the control room annunciators for the nitrogen supply system.
The inspector found that the leak test of the nitrogen supply system had not been performed during the preop test.
The licensee stated that the test would be performed prior to declaring the ADS System operable, as the test is required as part of the Technical Specifications.
Pending completion of the leak test of the nitrogen supply system for the ADS accumulators, the inspector concluded that the licensee's position acceptably met the requirements and was consistent with the FSAR and the SER.
This item remains open.
ll.
Plant Ins ection Tours During tours of the Unit 2 control room early in the inspection period, the inspector observed the use of miscellaneous hand written tags affixed to various control panel switches and gauges to identify abnormal system or component conditions.
It was determined that these tags were used to aid operators in identifying system anomolies but, that these tags were not controlled by any specific procedure or written guidance.
The inspector discussed the use of these tags with the licensee and the licensee resolved to establish formal controls on the use of operator aid tags.
The inspector will review the licensee's resolution of this item in a subsequent inspection period.
Inspector follow item (410/86-39-02).
12.
Site Visi ts On August 6, 1986, Region I management visited the Unit 2 site to tour the facility and to discuss with the licensee the plant's readiness for fuel load.
On August 7, 1986, Commissioner Frederick Bernthal toured Unit 2 and received a readiness for fuel load briefing from the licensee.
NRR management toured Unit 2 on August 12, 1986, and received a readiness for fuel load briefing from licensee management.
1. ~E<<
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 restrictions.