IR 05000280/1995019
| ML18153A550 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/01/1995 |
| From: | Belisle G, Branch M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18153A549 | List: |
| References | |
| 50-280-95-19, 50-281-95-19, NUDOCS 9512120070 | |
| Download: ML18153A550 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W., SUITE 2900 ATLANTA, GEORGIA 30323-0199 Report Nos.: 50-280/95-19 and 50-281/95-19 Licensee: Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060 Docket Nos.:
50-280 and 50-281 License Nos.:
DPR-32 and DPR-37 Facility Name:
Surry 1 and 2 Inspection Conducted:
October 15 through November 4, 1995 Lead Inspector: ;CW fbt M. W. Bra~esident Inspector Other Inspectors:
Approved by:
Scope:
D. M. Kern, Resident Inspector W. K. Poertner, Resident Inspecto L. W. Garner, Project Engineer
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- e C1'~e1s1e Reactor Projects Branch 5 Division of Reactor Projects SUMMARY
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This routine resident inspection was conducted on site in the areas of plant status, operational safety verification, maintenance and surveillance
. inspections, refueling activities, engineering review, plant support, self-assessment, and Licensee Event Report followu Inspections of backshift and weekend activities were conducted.
9512120070 951201 PDR ADOCK 05000280 G
Results:
Plant Operations Command and control during Unit 1 startup for physics testing was good (paragraph 5.2).
A non-cited violation was identified for failure to properly implement a tagout procedure. Manipulation of the wrong component resulted in offsite power being lost to the IH and 2J emergency busses (paragraph 9.1).
Maintenance A personnel error during Reactor Protection System (RPS) testing resulted in both Unit 1 source range nuclear instruments becoming inoperabl The personnel error was classified as a non-cited violation (paragraph 4.2).
A strength in procedure usage was identified when Instrument and Control technicians identified an omission in a RPS logic test procedure (paragraph 4.2).
Root Cause Evaluation 95-11 developed several sound recommendations which are intended to enhance control room annunciator reliability (paragraph 8.2).
Engineering Use of an additional reactivity chart recorder greatly improved control room communication between the Reactor Engineer and the reactor operators manipulating control rods during physics testing (paragraph 5.2).
Units 1 and 2 were uprated to 2546 MWT in October and August 1995, respectively. Engineers demonstrated a thorough understanding of the P-250 secondary calorimetric program and changes related to the core uprat The decision to limit reactor power based on the feedwater flowrate calorimetric during core uprate activities demonstrated a sound safety perspective (paragraph 6.3).
Plant Support Central alarm station operators were knowledgeable and performed physical security plan duties in an alert manner (paragraph 7).
REPORT DETAILS Persons Contacted Licensee Employees *W. Benthall, Supervisor, Procedures H. Blake, Jr., Superintendent of Nuclear Site Services R. Blount, Superintendent of Maintenance
- D. Christian, Station Manager J. Costello, Station Coordinator, Emergency Preparedness D. Erickson, Superintendent of Radiation Protection
- B. Garber, Licensing R. Garner, Outage and Planning
- D. Hayes, Supervisor of Administrative Services
- C. Lovett, Supervisor, Licensing
- C. Luffman, Superintendent, Security
- J. McCarthy, Assistant Station Manager
- F. McConell, Superintendent of Materials
- S. Sarver, Superintendent of Operations R. Saunders, Vice President, Nuclear Operations
- B. Shriver, Assistant Station Manager K. Sloane, Superintendent of Outage and Planning E. Smith, Site Quality Assurance Manager T. Sowers, Superintendent of Engineering
- B. Stanley, Station Director of Nuclear Oversight
- J. Swientoniewski, Supervisor, Station Nuclear Safety Other licensee employees contacted included plant managers and supervisors, operators, engineers, technicians, mechanics, security force members, and office personne NRC Personnel
- M. Branch, Senior Resident Inspector D. Kern, Resident Inspector
- K. Poertner, Resident Inspector L. Garner, Project Engineer
- Attended Exit Interview on November 8, 199 #Attended Exit Interview on November 29, 199 Acronyms used throughout this report are listed in the last paragrap Plant Status On October 19, criticality was achieved for Unit 1 low power physics testing following the RF The unit was placed on the electrical grid on October 21 and was operating at 100% power at the end of the inspection period. Unit 1 core uprate was implemented during the RFO and no unexpected operational problems resulted during the startup. The
unit is currently licensed to operate at 2546 MWT which represented an increase of approximately 30 megawatts electrica Unit 2 operated at or near 100% power throughout the inspection perio.
Operational Safety Verification (71707)
The inspectors conducted frequent tours of the control room to verify proper staffing, operator attentiveness and adherence to approved procedure The inspectors attended plant status meetings and reviewed operator logs on a daily basis to verify operational safety and compliance with TSs and to maintain overall facility operational awarenes Instrumentation and ECCS lineups were periodically reviewed from control room indications to assess operability. Frequent plant tours were conducted to observe equipment status, fire protection programs, radiological work practices, plant security programs and housekeeping. Deviation reports were reviewed to assure that potential safety concerns were properly addressed and reporte Spent Fuel Pool Cooling During the inspection period, the inspectors reviewed the design of the Spent Fuel Pool Cooling System with respect to core offload capabilit The spent fuel pool cooling system consists of two pumps and two heat exchanger The heat exchangers and pumps are arranged for cross-connected operation if require One pump and heat exchanger provide 100% heat removal capability. The original design requirements of the system stated tha~ ~he system has to maintain fuel pool temperature below 140° F, when one-third of a core is placed in the fuel pool 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown, and below 170° F when one and two thirds cores are placed in the fuel pool 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdow These temperatures are based on a maximum component cooling water temperature of 105° With two heat exchangers in service, spent fuel pool temperatures can be maintained less than 140° F assuming worst case conditions, i.e.,
maximum component cooling water_ temperature and full core offloa The inspectors determined that the original spent fuel pool design addressed a full core offload. The inspectors reviewed operating logs and determined that spent fuel pool temperature was maintained less than 106° F during the Unit 1 refueling outage with the entire core offloaded to the spent fuel poo The inspectors determined that the UFSAR and spent fuel pool design calculations assumed that the core offload was not placed in the spent fuel pool until 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdow The inspectors could find no administrative controls in place that would ensure that the 150 hour0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> criteria was met prior to core offload. The licensee is reviewing this item and has initiated a convnitment tracking item to follow its resolutio The recent Unit 1 core offload was not started until approximately 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> after
shutdow The inspectors will review this item further when the licensee has completed their evaluatio Within the areas inspected, no violations or deviations were identifie.
Maintenance and Surveillance Inspections (62703, 61726)
During the reporting period, the inspectors reviewed the following maintenance and surveillance activities to assure compliance with the appropriate procedures and TS requirement.1 Charging Pump Component Cooling Head Tank Sight Glass Replacement On November 2, the inspectors witnessed work activities associated with replacing the Unit 1 charging pump component cooling head tank sight glass. The work activity was accomplished in
, accordance with WO 0028786 The scope of the work order had been expanded to include replacing the inlet and outlet sightglass isolation valves due to leakage past the seat. This -activity was accomplished in accordance with skill of the craft criteri Since replacement of the isolation valves required that automatic level control for the head tank be defeated, an operator was stationed at the head tank to unisolate makeup if required. The inspectors reviewed the work package and monitored maintenance activities in progress. Activities observed were conducted appropriately and the sightglass was returned to service without inciden RPS Logic Testing On October 18, a personnel error during RPS logic testing.per l-PT-8.2, Reactor Protection Logic, revision 6P-l, resulted in both Unit 1 source range nuclear.instruments becoming inoperable for approximately one minut While the source range instrumentation was inoperable, the RCS boron concentration was 2402 PPM, all control rod assemblies remained fully inserted in the core, and no positive reactivity changes occurre A shutdown margin calculation was performe The results of the calculation determined the shutdown margin to be 8.9% CK-effective equal to 0.911). This item is the subject of LER 280/95-01 Failure to correctly perform a step in l-PT-8.2 constitues a violation, i.e., failure to follow procedure The inspectors reviewed the violation, the invnediate corrective actions, and the issues surrounding its root cause. This licensee identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section VII of the NRC Enforcement Policy. This NCV is identified as NCV 50-280/95-19-01, Failure To Follow RPS Logic Testing Procedur *
After the inadvertent source range de-energization, RPS testing was suspended until the event was investigated and the cause
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determined. Subsequent to the test resumption, the l&C crew that was to complete the test was counseled on management's expectations concerning procedure utilization and was briefed on the previous crew's error. The inspectors observed personnel at the protection racks satisfactorily complete l-PT-The inspectors verified that test switches were correctly aligned, test pushbuttons were properly actuated, and test lamp indications or relay actuations were received as specified by the procedur During the test, a technician observed, as he was closing a protection cabinet door, that test selector switch #1 had not been returned to its normal position. After verifying that no additional testing was to be performed in this protection cabinet and that the remaining test steps did not require this test switch be left selected to position 5, the technician returned the switch to its normal position. This was discussed with and concurred in by the other technician performing the test. A procedure review revealed that step 5.75 had failed to require the switch be returned to its normal positio A similar problem was encountered when step 5.80 was performe The technician submitted a procedure change request to correct this deficienc This procedure, as well as, the corresponding Unit 2 procedure were subsequently revise '
The inspectors reviewed RPS drawing number 113E244 sheet 19 and verified that leaving the test switch in a selected position would have no adverse effect on the RP Identification of this procedural deficiency by l&C personnel was considered as a strength, in that, it demonstrated proper procedure usage with attentiveness to expected equipment condition Within the areas inspected, one NCV was identifie.
Refueling Activities (71711) Hot Rod Drop Testing On October 18, the inspectors witnessed Shutdown Bank A and B hot rod drops being performed in accordance with 1-NPT-RX-014, Hot Rod Drops By Bank, revision Initial conditions were properly established and precautions were observe The inspectors independently verified that control rod designations marked on the chart recorder paper corresponded to the actual control rods being tested and that disconnected control rod position indicating signal cables were reconnected to their proper locations. The test data demonstrated that the control rod drop times were within TS criteria and that the control rods were latched to their drive mechanism On October 19, the test data for Control Banks A, B, C and D were reviewe The control rods performed as expected and the test procedure and TS criteria were met.
- Low Power Physics Testing The inspectors witnessed portions of the low power physics testing conducted on Unit 1 for fuel cycle 14. * The testing was conducted in accordance with procedure NPT-RX-008, Startup Physics Testing, revision The inspectors attended the evolution briefing conducted prior to pulling the control rods to criticality, observed the reactor startup to criticality, monitored low power physics testing activities in progress and reviewed the results obtaine Con111and and control during the unit startup was goo All activities observed were satisfactory and the test results were well within the predicted value The inspectors noted that an additional reactivity chart recorder was installed for use by the R This installation provided for better con111unication between the RO and reactor enginee In the past, the RO was provided verbal direction as to rod motion direction and duration from the test director. The additional chart recorder allowed.the RO to operate within the pre-established operational band controlled by the test directo Within the areas inspected, no violations or deviations were identifie.
Engineering Review (37551} Reactor Power Secondary Calorimetric Following Core Uprate Actual reactor power is calculated once per eight-hour shift using a P-250 process computer based secondary calorimetric program (CALCALC}.
Operators compare all four PRNI channel indications
.with the CALCALC generated power level to verify PRNI accurac The PRNis are then adjusted, as necessary, to maintain their input to RPS overpower protection circuitry within prescribed tolerance The CALCALC program can be run based upon either SFR or FF Un~ts 1 & 2 were uprated to 2546 MWT in October and August 1995, respectively. Prior to core uprate, the licensee used CALCALC based on SFR to verify that reactor power was maintained within TS limits. The FFR based CALCALC was considered less accurate, but conservative, due to gradual feedwater nozzle foulin Engineers anticipated that steam moisture content would increase following the core uprate, which would in turn affect SFR based CALCALC accurac Feedwater based CALCALC accuracy would not be affecte FFR based CALCALC power was higher than SFR based CALCALC computed power prior to core uprate on both units. The inspectors questioned how the licensee intended to verify that reactor power did not exceed the license limit during and after the core uprat ET No. CEE-95-065, Steam Moisture Content Increase, Core Uprate Project Effect on Steam Flow Derived Power by P-250 Computer SPS, Unit 2, revision 1, documented the anticipated core uprate effects on CALCALC computed powe Steam moisture content was postulated
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to increase from.25% to 1.0%.
Increased moisture content would directly affect two CALCALC parameters; steam enthalpy and steam density. Actual steam enthalpy decreases below that used in the CALCALC program, making CALCALC computed power conservative relative to actual powe Steam density increases which causes indicated SFR to be less than the actual SF This second factor influences CALCALC in the nonconservative direction. Engineers determined that the net result would cause the SFR based CALCALC computed power to be slightly more conservative than before the core uprat The inspectors reviewed ET No. CEE-95-065, verified the current P-250 CALCALC computer program used the enthalpy
- conservatism assumed by engineers, and independently calculated the affects of increased moisture content on the SFR based CALCALC progra The inspectors concluded that ET No. CEE-95-065 was technically soun Engineers demonstrated a thorough understanding of the P-250 CALCALC progra During power ascension for core uprate, operators monitored both FFR based and SFR based CALCALC reactor power The FFR based CALCALC remained higher than SFR based CALCALC computed power by approximately 0.5% to 1.0% on both units. Although the anticipated affects of higher moisture carryover on SFR based CALCALC were well understood, they could not be quantified until moisture carryover was quantified. The licensee decided to use the FFR based CALCALC value as the controlling reactor power indication to ensure the license limit was not exceeded following core uprat The inspectors determined that this action demonstrated a sound safety perspectiv Engineers informed the inspectors that several post core uprate tests were scheduled to support return to the SFR CALCALC standard. Special tests l-ST-0317 and 2-ST-0315, on Units 1 and 2, respectively, were in progress at the close of this report period to recalibrate feedwater flow instruments. Steam moisture carryover measurement tests are scheduled for December 199 Engineers plan to use data from the above tests to rescale the flow instruments and update the P-250 CALCALC program to restore desired program accuracy. Criteria for rescaling feedwater and steam flow instruments were under developmen The inspectors reviewed the P-250 CALCALC computer program with system engineers and discussed specif~c program changes which require evaluation in conjunction with flow instrument rescaling. The i.nspectors concluded that planned actions to restore SFR based CALCALC accuracy were reasonabl Within the areas inspected, no violations or deviations were identifie Plant Support (71707, 71750)
The inspectors conducted facility tours, work activity observations, personnel interviews, and documentation reviews to determine whether programs were effectively implemented to comply with regulatory
requirements in the areas of radiological protection, security, emergency preparedness, and fire protectio The inspectors observed radiological control practices and radiological conditions throughout the plant. Portal and handheld monitors were observed to be in good condition and within proper calibration periodicities. Workers complied with radiation work permits and properly used required personnel monitoring device Radiological postings and control of contaminated areas were generally goo The licensee determined that on one occasion, contamination surveys had failed to detect fixed contamination on tools released from the RC On November 3, the licensee established additional controls including an independent verification survey by a second HP technician prior to releasing tools and equipment from the RC The inspectors determined that this interim action was adequat Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accordance with the physical security plan and implementing procedures. This review included security measures, vital and protected area barrier integrity, maintenance of isolation zones, personnel access control, searches of personnel packages and vehicles, and visitor escor No discrepancies were note The inspectors observed security force practices from the CA The CAS operator was attentive to assigned duties and properly monitored perimeter surveillance devices required by the PS Vehicle access authorization was properly verified and*controlled. Security force personnel were knowledgeable and performed their duties in an alert manne ;
Within the areas inspected, no violations or deviations were identifie.
Self Assessment (40500) Institute of Nuclear Power Operations Evaluation Review On October 16, the inspectors reviewed the facility's INPO Evaluation Report, dated September 18, 1995, and discussed the report's findings with NRC managemen No NRC follow up was planned for items identified in the repor Control Room Annunciator Power Supplies Problems were experienced with the power supplies to the Unit 1 A through E annunciator panels resulting in degraded operation of the annunciators in July 1995 and failure of these annunciators in August 199 The events were discussed_ in NRC Inspection Report Nos. 50-280, 281/95-14 and 95-1 The licensee initiated RCE 95-11, RIS Power Supplies, to determine the cause of the power supply failures. The RCE was completed and approved on October 25, 1995.
- The RCE determined that the power supply failures resulted from a ground loop created in 1994 by replacing an annunciator light bul The ground loop increased current flow and degraded the A through E panel power supplies. The RCE determined that the July event was initiated by a short circuit across a failed power supply light socket. This placed the failed power supply in parallel with the other power supplies and resulted in two other power supplies failing and the degradation of the remaining power supplie The RCE determined that the August event was the result of two years of accumulative power supply degradatio The RCE recolllll8ndations included evaluating replacement of the present system with a microprocessor based system, developing a procedure to monitor for power supply degradation, implementing a design change to install an annunciator ground detector alarm, and checking the condition of the Unit 2 annunciator power supplies during the next scheduled refueling outage. All these recommendations were accepted by licensee managemen During the Unit 1 refueling outage, the licensee checked all the Unit 1 annunciator power supplies (panels A through K) and repaired/replaced any defective power supplie The inspectors consider that implementation of the above recommendations should greatly enhance the future reliability of the control room annunciators. The inspectors also verified that tracking items had been assigned where appropriat Four Inoperable Component Cooling Heat Exchangers On October 7, 1995, the licensee determined that all 4 CCHxs had been inoperable from 10:50 p.m. on October 6, 1995, to 12:55 on October 7, 1995, due to inadequate service water flo This event is discussed in NRC Inspection Report Nos. 280, 281/95-1 The licensee initiated RCE 95-13, Four Inoperable CCHxs, to determine the cause of the inadequate service water flo The RCE determined that the inadequate service water flow resulted from tubesheet blockage by hydroids and other marine material. A contributing cause may have been air binding since the vacuum priming system was removed from service for maintenance during the event. The RCE determined that the debris resulted from returning the B high level intake structure to service after the bay had been drained for nine days allowing hydroids to slough off due to the changing environment and water velocity. Engineering analysis determined that the total heat removal capacity of the Component Cooling System was adequate for the required heat load The RCE recommendations included revising procedures to require that the associated condenser waterboxes be flowed for a minimum of ten minutes prior to opening the associated service water supply valve if maintenance at the high level intake structure has been performed. Also, procedures were revised to require that one CCHx be isolated prior to returning associated condenser
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waterboxes or service water supply lines to servic RCE enhancements included establishing a PM program on the vacuum priming check valves to ensure that they function properl The licensee has issued LER 50-280/95-10 describing this even Completion of the corrective actions to prevent recurrence will be reviewed during closure of the licensee's LE Within the areas inspected, no violations or deviations were identifie.
Licensee Event Report Followup (90712)
The inspectors reviewed LERs submitted to the NRC to verify accuracy, description of cause, previous similar occurrences, and effectiveness of corrective actions. The inspectors considered the need for further information, possible generic implications, and whether the events warranted further on-site followu The LERs were also reviewed with respect to the requirements of 10 CFR 50.73 and the guidance provided in NUREG 1022, Licensee Event Report System, and its associated supplement.1 (Closed) LER 50-280, 281/95-09, Personnel Error Results in Loss of 4160 V Transfer Bus and Start of Emergency Diesel Generators.
During tagout activities, an operator erroneously removed the F transfer bus potential fuses resulting in a loss of power to the F transfer bus and the lH and 2J emergency busses. This resulted in the #1 and #3 EDGs starting and energizing the lH and 2J emergency busse This event was discussed in NRC Inspection Report Nos. 50-280, 281/95-1 The inspectors reviewed the LER and the root cause evaluation performed by the licensee after the even The root cause determined that the error was due to the operator failing to perform an adequate self chec The operator was counseled and operations personnel were briefed on the details of the event and the lessons learne The root cause evaluation reco11111ended that a self check simulator be constructed and used during monthly and quarterly training to practice self checking and that simultaneous verification be performed when removing safety related equipment from service. Licensee management accepted the reco11111endation to construct a self check simulator but did not accept the simultaneous verification reco11111endatio TS 6.4 requires written procedures and instructions for preventive or corrective maintenance operation Instructions for tagging out equipment for maintenance activities were provided in VPAP 2002, Work Request and Work Order Task, revision Failure to properly perform a tagout procedure constitutes a failure to comply with VPAP 2002 and TS 6.4. This licensee identified and corrected violation is being treated as a Non-Cited Violation,
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consistent with Section VII of the NRC Enforcement Policy. This NCV is identified as NCV 50-280, 281/95-19-02:
Failure To Properly Perform Tagout Results In Loss Of Offsite Power To The lH And 2J Emergency Busse.2 (Closed} LER 50-280/95-011, Both Source Range Nuclear Instruments De-Energized due to Personnel Error. This item is discussed in paragraph 4.2 of this repor Within the areas inspected, one NCV was identifie.
Exit Interview The inspection scope and findings were summarized on November 8 and 29, 1995, with those persons indicated in paragraph l-The inspectors described the areas inspected and discussed in detail the inspection results addressed in the Summary.section and those listed belo Item Number NCV 50-280/95-19-01 Status Closed Description/(Paraqraph No.)
Failure To Follow RPS Logic Testing Procedure (paragraph 4.2).
NCV 50-280, 281/95-19-02 Closed Failure To Properly Perform Tagout Results In Loss Of Offsite Power To The lH And 2J Emergency Busses (paragraph 9.1)
LER 50-280, 281/95-09 Closed Personnel Error Results in Loss of 4160 V Transfer Bus and Start of Emergency Diesel Generators (paragraph 9.1).
LER 50-280/95-011 Closed Both Source Range Nuclear Instruments De-Energized due to Personnel Error (paragraph 9.2).
Proprietary information is not contained in this report. Dissenting comments were not received from the license.
Index of Acronyms CAS CCHx CFR ECCS EOG ET F
CENTRAL ALARM STATION COMPONENT COOLING HEAT EXCHANGER CODE OF FEDERAL REGULATIONS EMERGENCY CORE COOLING SYSTEM EMERGENCY DIESEL GENERATOR ENGINEERING TRANSMITTAL FAHRENHEIT
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FFR FEEDWATER FLOWRATE INPO INSTITUTE OF NUCLEAR POWER OPERATION LER LICENSEE EVENT REPORT MWT MEGAWATTS THERMAL NCV NON-CITED VIOLATION NRC NUCLEAR REGULATORY C01*tISSION PM PREVENTIVE MAINTENANCE PPM PARTS PER MILLION PRNI POWER RANGE NUCLEAR INSTRUMENT PSP PHYSICAL SECURITY PLAN RCA RADIOLOGICAL CONTROL AREA RCE ROOT CAUSE EVALUATION RCS REACTOR COOLANT SYSTEM RFO REFUELING OUTAGE RIS ROCHESTER INSTRUMENTATION SYSTEM RO REACTOR OPERATOR RPS REACTOR PROTECTION SYSTEM SFR STEAM FLOWRATE SPS SURRY POWER STATION TS TECHNICAL SPECIFICATION UFSAR UPDATED FINAL SAFETY ANALYSIS REPORT VPAP VIRGINIA POWER ADMINISTRATIVE PROCEDURE V
VOLTS
WO WORK ORDER