IR 05000263/1997018
| ML20202C772 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/04/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20202C752 | List: |
| References | |
| 50-263-97-18, NUDOCS 9802130021 | |
| Download: ML20202C772 (22) | |
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c U.S. NUCLEAR REGULATORY COMMl88 ton
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REGION lit i
l Docket No:
50 263
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License No:
DPR 22 Report No:
50 263/97018(DRP)
i Licensee:
Northern States Power Company
Facility:
Monticello Nuclear Generating Station
Location:
414 Nicollet Mall l
Minneapolis, MN 55401
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Detes:
November 25,1997 January 12,1998 t
inspectors:
A. M. Stone, Senior Resident inspector
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D. Wrona, Resident inspector Approved by:
J. McCormick Barger, Chief Reactor Projects Branch 7
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d 9902130021 980204
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PDR ADOCK 05000263 G
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EXECUTIVE SUMMARY Monticello Nuclear Generating Station, Unit i NRC Inspection Report No. 50 263/g7018(DRP)
This inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a seven week period of resident inspection.
Operations in general, the conduct of operations was acceptabie. Operator performance during e
routine operations and surveillance test activities was good. However, the inspectors noted inconsistent communication technlques used by operators during responses to annunciators. (Section 01.1)
Operations personnel responded promptly and appropriately to a less of both recombiner
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trains and subsequent manual reactor scram. Procedure adhertace and informative briefings were observed. The shift manager maintained excellent comrnand and control of activities during the event. The reactor startup was conducted without incident.
(Section 01.2)
Two reactor powcr reductions were conducted in a controlled manner. Operators
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responded appropriately 16 a low circulating water and intake structure basin level condition. Power ascensions to 100 percent were pe%med without problems.
(Section 01.3)
Main'enance inadequate planning of surveillance test and maintenance activities by planning, e
maintenance, and operations personnel resulted in a reactor scram signal for 13 control rods and a reactor water level transient. The licensee initiated a thorough investigation and implemented several corrective actions. (Section M1.1)
The material condition of equipment was acceptable. The operators interviewed were
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knowledgeable of degraded conditions, including an inoperable discharge canal radiation monitor and a difficult to move control rod drive. (Section M2.1)
The licensee's current testing program did not include verification that the operability of
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the control room inhibit switch associated with the *B" train of automatic depressurization system (ADS) initiation logic could be inhibited by control room operators. The licensee responded appropriately by initiating actions to promptly test the switch. (Section M3.1)
Enoineerina Engineering personnel demonstrated a good questioning attitude in the identification of a
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missed ADS surveillance test. (Section M3.1)
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The operations shift supervisor promptly evaluated operability of the ADS system who i a
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discrepancy between the USAR and as built plant conditions was identifed by the inspectors. Engineering personnel responded appropriately by confirming operability and evaluating safety significance. (Section E2.2)
Plant sunoort Radiation protection personnel displayed a strong concem for maintaining personnel i
e doses as low as reasonably achievable when radiation levels near the main steam pressure transmitter root valve were higher than expected. A conservative decision to reduce reactor power further to lower dose rates was made. (Section Ri)
The fire brigade respnnded promptly and appropriaiety during two routine fire drills. Good
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interaction between the fire brigade leader and the security force was evident.
(Section F5.1)
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Report Details Summary of Plant Status The Unit operated at power levels up to 100 percent power until November 25,1997, when the reactor was manually scrammed. The licensee manually scrammed the reactor in anticipation of a loss of condenser vacuum caused by an electrical fault on power esbles to the recombiner system. On December 1,1997, operators commenced startup activities and the Unit was synchronized to the grid. On December 13,1997, the licensee reduced power to 60 percent to adjust control rod positions and to repair a leaking main steam pressure transmitter root valve.
The Unit retumed to 100 percent power a few hours later. On January 10,1998, the licensee reduced power to 75 percent to perform main steam valve exercise tests and turbine generator l
operational tests. The Unit operated at or near 100 percent power for the remainder of the inspection period, l. Operations I
Conduct of Operations 01.1 Q1Deral Comments (71707)
Usin9 Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing
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plant operations, including observations of control room evolutions, shift turnovers, and l
operator rounds. The inspectors also reviewed control room logbooks and operability determinations, Updated Safety Analysis Report (USAR) Section 13," Plant Operations,"
was reviewed as part of the inspection.
In general, the conduct of operations was acceptable; specific events and noteworthy observations are detailed in the sections below. Operator performance during routine operations and suiveillance test activities was good. Operator and operations management turnovers were conducted in accordance with procedures and included detailed discussions of equipment status as necessary. However, the inspectors noted inconsistent operator responses to annunciators. Some operators did not announce alarms and used nonverbal communication techniques; however, the inspectors did not identify any instances where actions required in response to annunciators were not completed. These observations were discussed with the operations superintendent.
01.2 Reactor Manual Scram Due to an Electrical Fault a.
Inspection Scope (71707)
On November 25,1997, at about 3:00 p.m., the electrical power supply breakers to nonsafety related motor control centerni (MCC) 115,116,124, and 125 tripped. This resulted in a trip of the A and B recombiner trains and all offgas auxiliary systems.
Control room operators immediately reduced reactor power from 100 percent to about 57 percent. With the recombiners inoperable, the operators were concemed with a loss of condenser vacuum. At about 3:10 p.m, the shift supervisor ordered the cperators to scram the reactor when it became apparent that the power supply breakers would not be restored irnmediately.
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The inspectors were in the control room shortly after the circuit breakers to the nonsafety-related MCCs tripped. The inspectors observed operators actions and reviewed the sequence-of events recorder and vatious strip charts, computer alarm printouts, control room log books, and procedures, b.
Qbservatiq.1Lgrd Findire r
i The inspectors noted thu when the MCC circuit breakers tripped, operations personnel respon6ed promptly by decreasing reactor power to 57 percent power. Operations and 9ngineering personnel were immediately dispatched to the breakers to assess the damsp. The s51" ms4eier orriered the operators to scram the reactor when it became 6pparo4 fom %s t rcabtt's could not be restored. Tho shih supervisor assigned roles and responsib,5 ties 19 peratbna personnel during a briefing conducted prior to the scram.
The inspectors noicd that nonessential personnel left the corn of room which minimized the background noise and pcter.tial operator distractions. The shift manager maintained excellant command and control during the event and ensured that proper notifications l
wwe made. Periodic briefings were informative.
All systems respondad as expected to the reactor scram. As anticipated, the reactor vene! water level decreased to about g inches and resulted in a Containment Group ll l
and Group 111 isolation. The shift supervisor executed the emergency operating l
procedures as required. Reactor water level quickly retumed to normal levels using the
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feedwater and conder. sate systems and the operators reset the reactor scram at about 3:20 p.m. The licensee decided to reduce reactor pressure to less than 600 pounds per square inch (psi) to prevent another reactor scram signal es condenser vacuum continued to decrease. However, at 4:49 p.m., the condenser vacuum approached the trip setpoint and the shift manager ordered the operators to insert another manual scram since reactor pretsuio was still greater than 600 psi. Operators maintained the Unit in hot shutdown and then proceeded to cold shutdown in the evening of November 26, 1997. The reactor startup was conducted without incident later on December 1,1997.
The inspectors reviewed the sequence of events recorder and various strip charts, coraputer alarm printouts, control room log books, and procedures The inspectors verified that plant parameters responded as expected and plant equipment functioned as required.
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Conclusions Operations personnel responded promptly and appropriately to a loss of both recombiner trains and subsequent manual reactor scram. Procedure adherence and informative briefings were observed. The shift manager maintained excellent command and control of activltler during the event. The reactor startup was conducted without incident.
09.3 Performance Durina Planned Power Red _u_gignt a.
inspection Sepne (71707)
The inspectors observed operators perform a planned reactor power reduction on December 13,1997, and January 10,1998.
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Observations aMDMingt n
On December 10,1997, operators initiaiad a power reduction to support control rod adjustments and to (cpsir a ;)aking nonsafety-related main steam pressure transmitter root valve. The operators conducted the power change in a professional manner and were knowledgeable of their assigned tasks. The inspectors noted that communications between operators and the nuclear engineer were effective. Initially the licensee planned a reduction to 75 percent power. However, radiation levels near the main steam pressure transmitter root valve were higher than expected. The radiation protection staff displayed conservative decision making and a strong concern for maintaining personnel doses as low as reasonably achievable by recommending a further power reduction.
Consequently, the operators lowered power to about 60 percent until the tasks were completed. Later that day and after the repairs, reactor power was ralsed to 100 percent without incident.
On January 10,1998, operators initlated a power reduction to 75 percent to support main steam isolation valve and turbine generator testing. During the power reduction, extremely low outside temperatures caused sudden icing of the Mississippi River upstream of the plant and resulted in a low levelin the circulating water and intake structure basins. Operators responded appropriately by lining up a cooling tower in accordance with procedures. The basin level remained above the circulating water pump trip setpoint and did not impact plant operation. Later that day, operators retumed reactor power to 100 percent without incident.
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Conclusions Both reactor power reductions were conducted in a controlled manner. Operators responded appropriately to a low circulating water and intake structure basin level condition. Power ascensions to 100 percent were performed without problems.
Operational Status of Facilities and Equipment O2.1 Enoineered Safety Feature System Walkqqwilt The inspectors used Inspection Procedure 71707 during a walk down of selected portions i
of the control rod drive (CRD), core spray (CS), and high pressure coolant injection (HPCI) systems. No operability concems were identified. Minor discrepancies were discussed with engineering personnel. The inspectors had no further concems.
Miscellaneous Operations issues 08.1 (Closed) Licensee Event Report (LER)(50-263/97013). Revision 0: Manual Scram Inserted Due to Loss of Condenser Vacuum Causeu by Electrical Fault Tripping Riscombiners. This event is discussed in Section 01.2. The licensee planned to evaluate the root cause for the failed cables.
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II. Maintenance, M1 Conduct of Maintenance M1.1 General Comments a.
Inspection Scope (62703 ard 61726)
The inspectors observed all or portions of selected maintenance and surveillance test t
activities, included in the inspection was a review of the surveillance test procedures or work orders (Wos) listed as well as the appropriate USAR Sections pertaining to activities, b.
Qhiervations and FintilDat In general, the inspecttre observed that the work tasociated with these activities was conducted in a professkaal and thorough manner. All work observed was performed with the work package present ond in active use. Technicians were experienced and knowledgeable of their assigned tasks. The inspectors frequently observed supervisors and system engineers monitoring job progress, and quality control personnel were present whenever required by procedure. When applicable, appropriate radiation control measures were in place.
The following work was observed. Specific concoms or observations are provided where appropriate.
WO 9705244, *PM [ preventive maintenance) V EF 24A 18-Month" and
WO 9705245, 'PM V EF 24818 Month" for the reactor building exhaust fans.
The inspectors observed the pre-job briefing between maintenance, engineering, and operations personnel. The system engineer discussed the need to restore the fans if the main steamline tunnel temperatures approached 165 degrees Fahrenheit ('F). The inspectors had no concerns with the execution of this work
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and post maintenance testing.
Surveillance Test 0056, Revision 19, 'HPCI Hi Steam Flow Sensor Test and
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Calibration Procedure, and Group 4 isolation Valve Closure Test"
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Surveillance Test 1040 3, Revision 30 " Turbine Generator Semlannual
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Opertilonal Tests" Surveillance Test 0255-07 IA 1, Revision 10, Main Steam Valve Exercise Tests'
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Surveillance Test 0286, Revision 5. " Torus Water Level Instrument Semi Annual
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Calibration Procedure" The inspectors verified that the acceptance ce;teria specified in the surveillance
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test were consistent with Technical Specifications (TSs) 3.7.A.1 and 4.7.A.1; USAR, Section 5.2.2.3 and Table 5.2-1; and Calculation CA 97177, Revision 4,
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" Torus Narrow Range LevelIndication and Alarm Uncertainty Calculation." Also, the inspectors had nu concems with operator performance of the surveillance test.
Surveillance Test 025510-IA 1, Revision 17, * Primary Containment Quarteriy
Isolation Valve Test" The inspectors verified that tha acceptance criterion specified in the surveillance test was consistent with TS 4.7,0.1.c, USAR Table 5.2 3b, and the inservice testing program. Also, the inspectors had no concerns with operator performance of the surveillance test.
Surveillance Test 1339, Revision 9, *ECCS ( emergency core cooling system]
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Pump Motor Cooling Flush" Surveillance Test 0255-03 IA 1, Revision 25, ' Core Spray System Tests" and
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Surveillance Test 49001PM, Revision 12, *PM for Limitorque Motor Operated l
Valves" l
During this period, the licensee performed preventive maintenance on the CS system. The work activities included lubrication of the valve stems for several motor operated valves. The inspectors noted that the licensee did not perform valve stroke time testing prior to the maintenance work. The inspectors reviewed
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l stroke time test data since 1993 for seven CS valves (MO 1741, MO-1742, MO 1749, MO 1750, MO 1751, MO 1752, and MO 1753) to determine if the PM activity preconditioned the valves. The inspectors found that the stroke times for each valve did not appreciably change after stem lubrication. The inspector concluded that the FM activity did not precondition these valves.
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Qonclusioni in general, the performance of work and survkilance activities observed was acceptable.
M1.2 Improper Coordination of Wotk Resulted in Two Evepis a.
Inspection Scope (61726 and 71707)
On November 36,1997, a scram signal was generated for 13 inserted control rods during a surveillance 13t. Later the same day, a reactor water level transient occurred during calibration of a turbine pressure transmitter. The inspectors reviewed the licensee investigation into the events 6nd corrective actions. The following documents were also reviewed:
Condition Report (CR) 97002963, "lRM (intermediate range monitor)/SRM (source
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range monitor) testing coincident with scram pilot solenoid isolation during plant S/D (shutdown) resulted in opening scram valves" CR 97002978, " inadvertent operation of EPR (electrical pressure regulator)
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resulting from calibration of PT 1177 results in opening of bypass valves *
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Surveillance Test 0013, Revision 11, 'lRM Scram and Rod Block (and) SRM Rod
Block Calibration" Surveillance Test 1045, Revision 9, *EPR and MPR [ mechanical forced restored
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pressure regulator) Functional Test *
j Surveillance Test 7231, Ri 11, *EPR/MPR PM Procedure'
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WO 9705746, * Calibrate /Chc A turbine efficiency instruments listed"
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Operations Manual B.6.102.07
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Qbervations and Findinas As discussed in Section 01.2, on November 25, '997, control room operators manually scrammed the reactor due to 6n electrical fault o1 cables to the offgas building. The Unit remained in hot shutdown conditions while the licensee assessed the damage. On November 26,1997, with the reactor in hnt shutdown, operations personnel isolated 13 control rods for scram solenold pilot valve replacements. The isolations involved pulling fuses on the 'B" side of the reactor protection system logic. At about 8.00 a.m.,
instrument and control (l&C) personnel requested authorization from operations shift management to commence Surveillance Test 0013. During the surveillance test, a half scram from the "A" side of the reactor protection system was received as expected. Th!s resulted in a full scram signal to the 13 isolated control rods. No rod movement occurred since the control rods were already fully inserted at the time of the event. The licensee suspended the surveillance test until the scram solenoid replacement werk was con pleted.
A few hours later, l&C personnel requested authorization to calibrate various pressure and flow transmitters to support a future turbine efficiency test. The EPR pressure transmit'er, PT-1177, was included. The EPR was designed to control reactor steam pressurs oetween 900 and 1050 pounds per square inch - gauge (psig) by opening or closing the turbine control or bypass valves as necessary. The MPR functioned similarly with a control band of 150 to 1050 psig. At the time of the event, the MPR was controlling reactor steam pressure. When l&C personnelinserted a high pressure signal into PT 1177, the EPR sensed a high reactor steam pressure which caused the turbine bypass valves to fully open. This caused the indicated reactor water level to increase to 48 inches which resulted in a feedwater pump trip. Operators responded quickly and appropriately to this reactor water level transient. The inspectors noted that PT 1177 had previously been calibrated using Surveillance Test 1045 which required no pressure in the main steam lines (reacto,' vessel vented or main steam isolation valves closud).
Work order 9705746 did not specify conditions to perform the calibration.
The inspectors were concemed that both of these events resulted from inadequate planning of the surveillance and maintenance activities. Planning, maintenance, and operations personnel did not consider the compatibility of performing these activities simultaneously, The licensee Initiated several corrective actions including discussing these events with the appropriate personnel, adding notes to the scram solenoid pilot valve and PT 1177 equipment notebooks, and developing a forced shutdown process to coordinate known evolutions. Administrative work instruction (AWI) 04.01.01, " General
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Plant Operating Activities,' Revision 17, required the shift manaper and shift supervisor to maintain a broad perspective of conditions affecting plant safety The failure to follow AWI 04.01.01 and the failure to provide adequate instructions for calibrating PT 1177 constituted violations of 10 CFR Part 50, Appendix B, Criterion V, * instructions, Procedures, and Drawings." Howevar, because the licensee initiated a thorough investigation and implemented corrective act;ons, these violations are being treated as two examples of a non-cited violation (NCV), consistent with Section IV of the NRC Enforcement Policy (NCV 50 263/97018-01(DRP)).
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Conclusions inadequate planning of surveillance and maintenance activities resulted in a reactor scram signal for 13 control rods and a reactor water level transient. Planning, l
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l maintenance, and operations personnel dij not consider the compatibility of performing
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several activities simultaneously. The licensee initiated a thorough investigation and implemented several corrective actions. A non cited violation of 10 CFR Part 50, Appendix B, Criterion V, ' Instructions, Procedures, and Drawings," was identified.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Current Material Conditions and Impact on Operations Personnel a.
Inspection Scope (71707 and 62703)
The inspectors conducted control room and plant inspections and interviewed operations
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personnel to assess the material condition of plant equ!pment.
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Observations and Findinas
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During this inspection period, the following material conditions and equipment performance concerns were found:
Discharae Canal Radiation Monitor Samplina Pymo Problems: On numerous
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occc lions during this inspection period, operators declared the discharge canal radiation monitor inoperable due to the sample pump losing its prime. The operators appropriately entered a 30-day limiting condition for operation and initiated manual sampling as required. In each case, the operators were able to establish the prime and the pnp was declared operable within a short time period. The licensee perform 'd several actions including verification of no in-leakage at connections, replacement of the pump suction valves, and back flushing stand pipes to determine the cause and correct the problem.
However, no root cause was determined. The licensee noted that this problem occurred predominantly in November and December ano could be due to low river temperature and low dissolved oxygen. The licensee planned to continue troubleshooting activities to recolve this issue.
Control Rod Drive Difficult to Mon: During startup activities, control room
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operators noted that control rod 14-07 was difficult to withdraw from notch position 00 and also noted high stall flows of over 5 gallons per minute (gpm).
Operators vented the drive in accordance with procedure; however, the
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withdrawal problem continued. The licensee determined that a possible cause could be failure of the collet piston seal rings. The control rod drive was isolated in the fullinsert position and the licensee Manned to investigate the root cause during the next refueling outage, c.
Qonclusions The material condition of equipment was acceptable. The operators interviewed were knowledgeable of the degraded conditions.
M3 Maintenance Procedures and Documentation M3.1 Licensee-identified Fallure to Test the *B" Train Automatic Depressurization System (ADS) Inhibit Switch From B Loalc a.
Inspection Scope (71707. 35551. 62703)
Generic Letter 96-01, " Testing of Safety Related Logic Circuits,' required licensees to verify that electrical contacts within logic systems w' J tested in accordance with TSs.
On December 19,1997, during review of the ADS logic scheme, an engineer identified that the *B" train ADS automatic initiation inhibit switch was not tested in accordance with TS 4.5.A.4. The inspectors observed the licensee's immediate actions includi.ig the performance of a special test. Thr following docv 7ts were also reviewed:
TS 4.5.A.4
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USAR Sections 6.2, Revision 14
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Surveillance Test 01131, Revision 17, * ADS System Simulated Automatic
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Actuation" Temporary change to Surveillance Test 01131 to test the B train inhibit switch
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Safety Evaluation Report for Amendment No. 62 to Facility Operating License No.
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DPR 22 Drawings NX 7831 1431 through 3," Elementary Diagrams Automatic Blowdown
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Systems" b.
Observationi and Findinas During a review to verify full logic system testing, an engineer identified that the current testing for ADS did not verify the operability of the control room inhibit switch associated with the *B" train of actuation logic. The purpose of the inhibit function was to allow operators to perform a manually controlled depressurization if conditions did not warrant an automatic depressurization. The engineer immediately notified station management of this concern and initiated a condition report to evaluate operability. The licensee reported the event in accordance with 10 CFR 50.73 for the missed TS required surveillance test.
Technical Specification 4.5.A.4 specified that the switch be tested once per operating cycle, but did not specify a limiting condition for operation if the switch was inoperable.
The ADS valves remained operable since the automatic depressurization function was not affected. Revision 14 of USAR Section 6.2.5 stated that an operator could manually
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block an automatic depressurization by utilizing the !nhibit switch in the control room. The licensee declared the inhibit switch inoperable and initiated actions to immediately test the switch.
The licensee conducted several meetings to discuss switch operability, reportability requirements, and the proposed test. The inspectors noted good interaction between maintenance, operations, and engineering personnel during discussions concoming the proposed testing methods. Engineering personnel developed the new test as a temporary change of Surveillance Test 0113 2. The operations committee members reviewed and approved this temporary change prior to the performance of the test.
Within a few hours of identification of the problem, the licensee successfully tested the j
- B" automatic initiation inhibit switch.
The failure to test the ADS switcilin accordance with TS 4.5.A.4 is a violation. However,
since this issue was identified by the licensee and corrective actions including testing the switch were promptly initiated, no violation will be cited in accordance with Section Vil of the NRC Enforcement Policy (NCV 50 263/g7018-02(DRP)).
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Conclusions u
Engineering personnel demonstrated a good questioning attitude in the identification of a missed TS surveillance test. The engineer iden'ified that current testing methods did not verify that the *B" train of ADS initiation logic could be inhibited by control room operators.
The licensee re ponded appropriately by initiating actions to promptly test the switch.
M8 Miscellaneous Maintenance lasues M8.1 (Closed) LER (50-263/96005). Revision 00: Unexpected Voltage Transient During Maintenance Causes Partial Containment isolation. This LER pertained to an inadvertent closure of the primary containment atmospheric sample valves, closure of the nitrogen make-up valves, initiation of the standby gas treatment system, and isolation of the reactor building ventilation caused by a false division I fuel pool radiation moaitor high radiation signal during maintenance. The inspectors verified the completion of the corrective actions which included revising several procedures to ensure the upscale trips on either the fuel pool or the reactor building exhaust plenum radiation monitors were bypassed when an activity could generate a trip signal on either radiation monitor, The following procedures were reviewed:
Surveillance Test 0067A, " Spent Fuel Pool & Reactor Building Exhaust Plenum
Monitor Functional Test," Revision 13 Surveillance Test 0068, " Spent Fuel Poci & Reactor Building Exhaust Plenum
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Monitor Calibration," Revision 15 M8.2 (Closed) LER (50-263/96010). Revision 00: Panet Fire Barrier Not in Correct Position.
This LER pertained to a floor to-panel fire board barrier found out of position in main control board panel C03. The inspectors verified Procedure 02751, * Fire Barrier Penetration Seal VisualInspection," Revision 8, included steps to periodically inspect and verify proper placement of the control room floor to panel fire barriers.
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M8.3 (Closed) Violation (VIO) (50 263/96003-02AfDRP)): This violation pertained to personnel failing to follow the established temporary change process for procedure changes during control rod drive changeouts. The inspectors verified that the licensee provided event-related discussions during operator continuing training with an emphasis on a questioning attitude. The inspectors also verified that Gurveillance Test 9019. Revision 15,
- Changeout Selected CRD's Operations,' included prerequisites te ensure the procedure could not be initiated unless shutdown margin requirements were met for all rods to be withdrawn and the fuelloading was verified correct. The mspectors also verified that Form 9051, Revision O. *Changeout Selected CRD's Operations," included a signoff for a nuclear engineer to verify that the prerequisites were satisfied for each centrol rod drive changeout. These procedure changes eliminated the need for the original temporary change.
M8.4 (Closed) VIO (50-263/96003-02Bf DR P)): This violation pertained to personnel failing to follow drawing requirements during inttallation of speed sensing panels in an emergency diesel generator panel. The inspectors verified that the licensee provided training discussed the importance of following installation requirements and the necessity for
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attention to detall. The inspectors also verified that Form 4263, Revision 3, * Electrical Maintenance and Construction Pre Job Briefing Checklist," included items that cover construction notes and other special requirements.
M8.5 (Closed) LER (50-263/97014). Revis[pn.QQ: Division il ADS Inhibit Switch Was Not Tested as Required by TS. This issue was the subject of a non cited violation as discussed in Section M3.1.
111. Ennineerina E2 Engineering Support of Facilities and Equipment E2.1 Results of USAR Review a.
inspection Scope (37551)
While performing inspections associated with the HPCI system, the inspectors reviewed the following documents:
USAR Section 7.0, * Plant Protection System"
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TS Section 3.2, " Protective Instrumentation" a
General Electric (GE) Design Specification 22 A1380, ' Nuclear Boller Leak
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Detection" GE Design Specification 257HA354, "High Pressure Coolant in,iection System"
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Surveillance Test 0056, Revision 19, "HPCI Hi Steam Flow Sensor Test and
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Calibration Procedure, and Group 4 Isolation Valve Closure Test" Surveillance Test 0058, Revision 11,'HPCI Steam Line Area Temperature Test
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and Calibration Procedure, and Group 4 Isolation Valve Closure Test"
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Observations and Findinas
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The HPCI high steam line flow, HPCI high area temperature, and HPCI steam supply low pressure instruments provide signals for primary containment isolation of the MPCI steam supply valves' The inspectors identified that TS Table 3.2.1, " Instrumentation That
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initiates Primary Containment isolation Fun:tions," did not list the HPCI steam line flow pressure trip as described in USAR Table 7.6-2, " Primary Containment isolation System."
l The inspectors also identified that Surveillanta Test 0056 and Surveillance Test 0058 i
functionally tested and calibrated the HPCI higt. steam line flow and HPCI high area
temperature instruments monthly; however, the HPCI steam supply low pressure Instrument was functionally tested and ca9braud yearly. The inspectors identified that
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this issue also applied to the reactor core ir,olation cooling (RCIC) system The
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inspector: questioned whether the low pressure instruments for HPCI and RCIC should be included in TS Table 3.2.1. This is considered an Unresolved llam (URI 50 263/97018-03(DRP)) pending further review by the Office of Nuclear Reactor Regulation (NRR),
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The inspectors also identified the following concems:
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The USAR description did not match the description contained in GE Design
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Specification 22A1380, Section 4,3 of that specification stated that the primary reasons for the HPCI steam supply low pressure trip are: (1) to detect and annunciate at low pressure, which may be caused by a gross leak between the steam line elbows (supplying HPCI and RCIC systems) and reactor vessel and (2) to orovide a turbine isolation signal if the HPCI or RCIC systems should be in operation during the loss of steam pressure. Sections 7.6,3.2.4.11 and 7,6.3.2.4,8 of the USAR describes the purpose of the HPCI and RCiC low steam
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pressure trips as signals to close the associated isolation valves so that steam and possible accompanying radioactive gases do not escape from the turbine shaft seals when the reactor pressure has decreased below the pressure at which the systems effectively operate. The USAR does not mention that the pressure
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l switches can be used to detect a low pressure condition that might be caused by a steam line leak.
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Contrary to GE Design Specification 22A1380, there was no annunciator
associated with a HPCI or RCIC steam supply low pressure condition.
Surveillance Test 0056 used the HPCI low steam pressure trip setting of 100 psig
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as a nominal setpoint value. Table 7.6-2 of the USAR and GE Design
Specification 22A1380 specifically listed the trip setting as 100 psig, implying that the 100 psig may be a limit vice a nominal value.
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The inspectors discussed these issues with the licensee and the licensee initiated Condition Report 98000006, "USAR Table 7.6-2 (PCIS) [ Primary Containment isolation System) Inconsistencies with CML [ Component Master List] Instrument Setpoint i
laformation." The issues described above are considered an Unresolved item (URI 50-263/97018-04(DRP)) pending further NRC review of the licensee's evaluation.
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Conclusions The inspectors identified discrepancies between the TS and USAR, and the USAR and design bases associated with the HPCI and RCIC systems. These issues are being reviewed by the licensee and/or NRR to determine what actions may be necessary to address the problems.
E2.2 Incorrect Description of ADS loaic in USAR a.
inspection Scope (37551)
During review of the ADS surveillance test concem discussed in Section M3.1, the inspectors identified a discrepancy between the USAR description and as built logic circuitry. The following documents were also reviewed:
TS Table 3.2.2 (required trip sdpoint of less than or equal to 120 seconds)
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Operations Manual B.3.3.-06
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USAR Section 6.2.5.2, Revision 14
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NX 7831 143-1 through 3, " Elementary Diagrams - Automatic Blowdown Systems"
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b.
Observatiom, and Findinas During review of the ADS surveillance test concem discussed in Section M3.1, the inspectors identified that Section 6.2.5.2 of the USAR stated that automatic actuation of ADS required indication of low-low reactor water level and that the circuit was connected in a one-out-of-two-twice logic arrangement. The inspectors reviewed drawing NX 7831-143-1 through 3 and determined that the logic was actually a two-out-of two-once logic. The inspectors reviewed the TS bases which stated that each ADS trip system consisted of two trip logic channels such thnt both trip logic channels were required to permit a system trip. The inspectors notified the opeations shift supervisor of this discrepancy. The shift supervisor concluded that AUS was operable since the as-built circuitry was consistent with the TS bases.
The licensee also identified three additional errors in the description of ADS. The USAR stated that the re:.idual heat removal and core spray pumps discharge pressure switches were in a one-out-of-one arrangement, the permissive signal from the time delay circuit was a one-out-of-two network, and two circuit breakers were provided for each valve, As-built, the pressure s vitches are in a two-of-two arrangement for each pump, the time delay circuit is in a one-out-of-one arrangement, and a relay and fuse provide protection from single failure.
The inspectors reviewed the original final safety analysis section and the original TS bases and found that the description of the logic system in these documents wers similar to current versions. The licensee initiated CR 97003216, "USAR and control prints for ADS logic appears to be inconsistent," to evaluate operability and determine corrective actions. The inspectors had no concerns with respect to operability of the ADS system.
The licensee had previously initiated a program to review the USAR for accuracy and this discrepancy may have been identified through this program. The discrepancy between the USAR description and as-built plant conditions is an inspection Followup Item (IFl
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50-263/97018-05(DRP)) pending review of the licensee's 10 CFR 50.59 ovEluation and correction to the USAR.
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Conclusions i
The operations shift supervisor promptly evaluated operability of the ADS system when a
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evaluating safety significance.
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. E2.3. Current Material Conditions and Enaineerina Operability Evaluations a.
Insoection Scope (71707 and 37551)
The inspectors conducted control room anc plant inspections and reviewed condition reports to evaluate current material conditions and operability decisions.
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Observations and Findinas
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During this inspection period, the following material conditions and equipment -
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performance and design concams were found:
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No.14 Emeroency Service Water Pumo
Dearaded Flow:
During a routine test,
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the measured flow rate for the pump was 46.6 gpm. The minimum acceptable flow rate per the procedure was 47 gpm, assuming a river temperature of 90 ?F.
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The engineer determined that the pump was operable based on a current river temperature of 40 *F, The licensee planned to continue efforts to identify the root
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cause and take appropriate corrective actions, Low Flow in the No.12 CS Pumo Motor Coolor: During a routine surveillance test j
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on December 17,1997, operators found that the flow through the No.12 CS:
pump motor cooler did not meet the acceptance criteria; however, the as-found
results were bounded by an existing operability evaluationi The operators flushed the piping in 'accordance with the surveillanca test and were able to obtain the desired flow ratei The inspectors reviewed previous test data and determined that the flow rates measured in the previous three quarterly surveillance tests t
' were low. The inspectors discussed this observation with the system engineer j
who later initiated an accelerated testing schedule.
I Low Flow in tho' No.11 Residual Heat Removal Service Water (RHRSW) and
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No.13 RHRSW Pumo Motor Coolers: As discussed in Inspection Report No; 50-263/97015(DRP), the licensee identified low flow to the RHRSW motor
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coolers. At that time, engineering personnel performed an evaluation to justify continued operation with these degraded flow rates, noting that the
- No.11 RHRSW pump had been operating for 80 days during the forced outage
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from May through July 1997 without performance problems. Engineering personnel developed a special test to trend thrust bearing oil temperatures with the pump operating.- Based on the results of the special test, river temperature at or below $3*F, and performance observations during the 80 days of pump operation, the licensee declared the RHRSW pumps operable. During this
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i inspection period, the licensee measured the flow through the Nos.11 and 13 RHRGW motor coolers and found that the flows were less than those assumed in the previous operability evaluation. The engineer performed a new evaluation and determined that the pumps were operable with this new flow rate based on a river temperature of 40*F. This issue is the subject of IFl 50 263/97015-03(DRP).
Discrepancy in Documentation of Hioh Enerov Line Break Forces on Masonry
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Walls: The licensee identified that formal documentation of the effects of a high energy line break on 78 safety-related masonry walls may not exist. The walls were evaluated in response to NRC Bulletin 80-11, " Masonry Wall Design';
however, this evaluation did not include an evaluation of the effects of area pressurization due to a high energy line break. The licensee planned to determine the effects on the walls and take appropriate corrective actions. This is an Inspection Followup Item (IFl 50-263/97018-06(DRP)) pending licensee completion of evaluation and inspectors review.
c.
Conclusions Engineering evaluations reviewed during this period were acceptable. The licensee initiated actions to resolve the degraded conditions.
E8 Miscellaneous Engineering issues E8.1 (Closed) Unresolved item (50-263/97012-03(DRP)): This issue was opened to evaluate the applicability of 10 CFR Part 21 for the non-stainless steel parts contained in the emergency core cooling system suction strainers. The inspectors discussed this issue with the NRR project manager who forwarded it to the NRR vendor inspection branch.
The inspectors determined that the licensee was not responsible for reporting the discrepancy; therefore, this issue is closed.
IV. Plant Support R1 Conduct of Radiological Protection and Chemistry Controls During normal resident inspection activities, routine observations were conducted in the t
area of radiation protection. The inspectors noted that effective radiological controls were ostablished and that technicians provided support during maintenance and surveillance l
activities. Radiation protection personnel displayed a strong commitment to maintaining personne; doses as low as reasonably achievable when radiation levels near the main steam pressure transmitter root valve were higher than expected. Based on the request of radiation protection personnel, a conservative decision to reduce reactor power further
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to lower dose rates was made.
S1 Conduct of Security and Safeguards Activities During normal resident inspection activities, routine observations were conducted in the areas of security and safeguards activities. No concerns were noted.
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Fire Protection Staff Training and Quallfloation F5.1 ' Observations of Fire Brioado durina fire drills l
a. -
Inspection Scope (71750)
The inspectors observed two routine fire drills.
0 b.
. Observations and Findinas
The first routine fire drill consisted of a simulated fire in the No.11 emergency diesel
. generator fuel oil day tank room. The fire brigade members promptly responded to the ready room and donned the required protective clothing, including self-contained
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breathing apparatuses. The fire brigade members also discussed what equipment they would take to the scene of the simulated fire. A member of the security force responded to the ready room and followed the fire brigade to assist them as needed.
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Communications were established between the control room and the fire brigade leader.
At the simulated fire site, the fire brigade members discussed the fire fighting tactics they -
would ust'. The inspectors had no concems with the brigade's performance..
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- The second routine fire drill consisted of a simulated fire in the HPCI room witt,
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simulated personnelinjury in the same room. At the simulated fire site, the fire origade.
l members discussed the fire fighting tactics they would use. Communications were L
established between the control room and the fire brigade leader. The simulated injured
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person was extracted and brought to a safe area.' The inspectors had no concems with -
the brigade's performance.
The inspectors noted good discussions during the post-drill critiques. Fire brigade leaders asked probing questions to ensure brigade members understood the potential hazards and brigade members themselves were self-critical.-
c.
Conclusions The fire brigade responded promptly and appropriately during the fire drills. The -
inspectors noted good interaction between the fire brigade leader and the security force.
V. Manaaement Meetinas'
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Exit Meeting Summary -
On January 13/1998, the inspectors presented the inspNtion results to members of licensee management. The licensee acknowledged the findings presented. The
' inspectors asked the licensee whether any materials examined during the inspection
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should be considered proprietary. - No proprietary information was identified:
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PARTIAL LIST OF PERSONS CONTACTED Lloonsee M. Wadley, Vice President Nuclear Generation
- M. Hammer, Plant Manager B. Day, Training Manager
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K. Jepson, Superintendent, Chemistry & Environmental Protection L. Nolan, General Superintendent Safety Assessment
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- M. Onnen, General Superintendent Operations -
- E. Rolliy, General Superintendent Maintenance C. Schibonski, General Superintendent Engineering
- A. Ward, Manager Quality Services L'. Wilkerson, Superintendent Security _
J. Windschill, General Superintendent, Radiation Protection
- Indicates those present at exit meeting on January 13,1998, i
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J INSPECTION PROCEDURE 8 USED
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IP 37551F Onsite Engineering IP 61726:
Surveillance Observations IP 62703:
Maintenance Observatior,s IP 71707:
Plant Operations
- IP 71750:
' Plant Support '
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. IP 92700:
Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities
. IP 93702:
Prompt Onsite Response to Events at Operating Power Reactors
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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-263/97018-03(DRP)
URI -. Questions concoming whether the low pressure
instruments for HPCI and RCIC should be included in TS Table 3.2.1
- 50-263/97018-04(DRP)
URI USAR Table 7.6-2 inconsistencies with instrument setpoint
- information-
= 50-263/97018-05(DRP) -
IFl Discrepancy between the USAR description and as built ADS logic circuitry
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50 263/97018-06(DRP)-
- IFl Effects of high energy line break forces on masonry walls -
not documented -
Closed 50-263/96003-02A(DRP).
VIO Failure to follow the temporary change process for-procedure changes during control rod drive changeouts 50-263/96003-02B(DRP) -
VIO Failure to follow drawing requirements during installation of.
speed sensing panels in an emergency diesel generator.
panel 50-263/97018-01(DRP)
-NCV Failure to follow or provide procedures results in two events 50-263/97018-02(DRP)
NCV Failure to test the ADS switch in accordance with
' TS 4.5.A.4 50-263/97012-03(DRP)'
URI Applicability of 10 CFR Part 21 for the non-stainless steel parts contained in the emergency core cooling system suction strainers
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50 263/96005 LER Unexpected Voltage Transient During Maintenance Causes -
Partial Containment isolation 50 263/96010 LER Panel Fire Barrier Not in Correct Position -
50-263/97013.
LER Manual Scram inserted Due to Loss of Condenser Vacuum.
Caused by Electrical Fault Tripping Recombiners 50 263/97014 LER Division 11 ADS Inhibit Switch Was Not Tested as Required by TS Discussed 50-263/97015-03 IFl Actions to resolve low flow conditions in RHRSW
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1 LIST OF ACRONYMS USED
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ADS Automatic Depressurization System AWI Administrative Work Instruction CFR Code of Federal Regulations CML Component Master List CR Condition Report-CRD Control Rod Drive CS Core Spray ECCS Emergency Core Cooling System EPR.
Electric Pressure Regulator ESW-Emergency Service Water GE General Electric gpm Gallons per Minute HPCI High Pressure Coolant injection l&C Instrument and Control IFl Inspection Followup Item IRM -
Intermediate Range Monitor LER Licensee Event Report MCC Motor Control Center MPR Mechanical Forced Restored Pressure Regulator NCV Non-Cited Violation NRC Nuclear Regulatory Commission PM Preventive Maintenance psi Pounds per Square Inch psig Pounds per Square Inch - Gauge RCIC Reactor Core Isolation Cooling RHRSW Residual Heat Removal Service Water S/D__
. Shutdown.
SRM Source Range Monitor TS Technical Specification URI Unresolved item
- USAR Updated Safety Analysis Report VIO Violation WO Work Order -
'F Degrees Fahrenheit
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