IR 05000280/1993020
| ML18153B317 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/02/1993 |
| From: | Belisle G, Branch M, Tingen S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18153B316 | List: |
| References | |
| 50-280-93-20, 50-281-93-20, NUDOCS 9309140283 | |
| Download: ML18153B317 (11) | |
Text
Report Nos.:
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W., SUITE 2900 ATLANTA, GEORGIA 30323-0199 50-280/93-20 and 50-281/93-20 Licensee: Virginia Electric and Power Company 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:
July 4 through August 8, 1993 Inspectors:
Accompanying Approved by:
Scope:
Inspector
- z;_ tc,l ~
/vi S. G. Ting~ent Inspector NRC Inspector: A. Ruff
- ~.I.~(#"
G. l\\. BeliSCt1on Chief Division of Reactor Projects SUMMARY 8-~3/- 1'3 Date Signed
~3/-93 Date Signed This routine resident inspection was conducted on site in the areas of plant status, operational safety verification, maintenance inspections, surveillance inspections, action on previous inspection items, and licensee event revie During the performance of this inspection, the resident inspectors conducted reviews of the licensee's backshifts, holiday or weekend operations on July 11, 15, 16, and August 3, 199 Results:
No violations or deviations were identified In the operations area, the following items were noted:
Command and control weaknesses were observed during the August 5 Unit 2 startup when the turbine was placed on the line and the feedwater regulating
- valves were placed in automatic control.- In contrast, the Unit 2 shutdown on August 6 was well controlled and command and control was evident.
REPORT DETAILS Persons Contacted Licensee Employees J. Artigas, Supervisor Quality Assurance
- W. Benthall, Supervisor, Licensing
- R. Bilyeu, Licensing Engineer
- H. Blake, Jr., Superintendent of Nuclear Site Services
- R. Blount, Superintendent of Engineering D. Christian, Assistant Station Manager
- M. Bowling, Manager, Nuclear Licensing J. Costello, Station Coordinator, Emergency Preparedness
- J. Downs, Superintendent of Outage and Planning
- D. Erickson, Superintendent of Radiation Protection R. Gwaltney, Superintendent of Maintenance
- L. Hartz, Manager, Corporate Quality Assurance
- A. Keagy, Supervisor, Nuclear Materials
- M. Kansler, Station Manager
- C. Luffman, Superintendent, Security
- J. McCarthy, Superintendent of Operations
- A. Price, Assistant Station Manager
- R. Saunders, Assistant Vice President, Operations
- E. Smith, Site Quality Assurance Manager
- J. Swienioniewski, Supervisor, Station Nuclear Safety
- G. Thompson, Acting Superintendent of Maintenance NRC Personnel
- M. Branch, Senior Resident Inspector
- S. Tingen, Resident Inspector
- Attended Exit Interview Other licensee employees contacted included control room operators, shift technical advisors, shift supervisors and other plant personne Acronyms and initialisms used throughout this report are listed in the last paragrap *
2 Pl ant Status Unit 1 operated at or about 100% power through the entire inspection period, day 178 of continuous operations. During this inspection period an increase in RCS leakage was noted and an investigation was conducte The details of the leakage and the licensee's actions are discussed in section 3.a of this repor Unit 2 began the reporting period in power operatio On August 3 the unit tripped from 97% power when the A MFRV failed shut. This trip is discussed in detail in paragraph 3.b. The unit was restarted on August 5; however, on August 6 it was shutdown from approximately 25%
power in order to repair leaking PZR safety valve The unit startup and subsequent shutdown is discussed in detail in paragraph 3.c and respectivel Operational Safety Verification (71707, 42700)
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 operations safety and compliance with TSs and to maintain awareness of the overall operation of the facilit Instrumentation and ECCS*lineups were periodically reviewed. from control room indication to assess operability. Frequent plant tours were conducted to observe equipment status, fire protection programs, radiological work practices, plant security programs and housekeepin Deviation reports were reviewed to assure that potential safety concerns were properly addressed and reporte Unit 1 RCS Leakage Early in the inspection period, an RO noted an increase in containment radiation as indicated on particulate radiation monitor l-RM-15 Investigation, including a containment walkdown, determined that the most likely source of the leakage was from the primary manway on the cold-leg side of the B S Two SROs that observed the leakage quantified it at 0.2 gpm and considered it identified leakag The normal 100% power radiation level in the area is approximately 10 REM/hr and to minimize exposure the licensee elected not to remove the mirror insulation for the inspectio The 16-inch SG manway opening closure consist of a gasketed mechanical joint with a backing diaphragm and cover plate. The cover plate is held in place with 16 torqued bolts. The concerns associated with continued operation with the noted leakage includes; 1) possibility of increase in leakage rate, 2) steam cutting the gasket surface, 3) boron buildup and bolt corrosion and 4) inability to accurately quantify and possible masking of
actual leakag The licensee did not specifically address each concern but stated that they-had the same concern The inspectors reviewed the licensee's evaluation of the leakag The inspectors determined that it may have been more prudent to classify the leakage as unidentified since the insulation had not been remove This would have absolutely determined the leakage source. Additionally, there is a 3/8 inch welded channel head drain valve assembly in the vicinity of the manway cover that could also be the source. A similar drain valve on the unit 2 A SG experienced a weld joint leak in the 1984 time fram The inspectors discussed the leakage classification with plant managemen Even though there is another possible leakage path, the licensee was comfortable with their determination that the leakage was from the mechanical joint on the SG manway and was acceptable based on TS 3.1. The inspectors observed a close circuit TV monitor that was installed to trend the leakage and determined that without removing the mirror insulation (i.e., a high radiation dose effort) the most probable leakage source is from the SG manway cove The licensee's evaluation also concluded that the 3/8 inch drain line was sized such that in the unlikely event of a failure, the leakage would be limited to the capacity of a single charging pum The inspectors will continue to monitor the licensee's a~tions in this are Unit 2, August 3, Reactor Trip At 8:05 p.m., on August 3, Surry Unit 2 tripped from 98% reactor powe The trip was caused by the A MFRV failing closed. This resulted in a loss of feedwater flow and a reduction of inventory level in the A S The actual signal that tripped the reactor was SG Low Level coincident with a Steam-Flow/Feed-Flow mismatc The plant response to the transient was normal; however, several equipment related problems were experience The IRPI indication for rod M-10 was sluggish and the rod bottom light did not illuminate. Additionally, a problem was noted with the operation of the C secondary POR It lifted before its set pressure was note The inspectors went into the plant when notified of the trip. A review of control room charts and discussions with the RO and SRO were conducted. The inspectors verified that RCS pressure response was as expected and that even with RCS pressure starting at 2135 psig versus the normal 2235 psig, no low pressure problems were experience Material problems noted above were evaluated by the licensee. The actual cause of the A MFRV going closed was determined after extensive troubleshooting. The licensee has attributed the problem to an erratic manual/auto station -15 volt DC power supply
- which was replaced and satisfactorily tested. The slow response of rod M-10 !RPI is a recurring proble The licensee has contributed the problem to residual magnetism in the !RPI coil and has documented this in the pas To ensure that there were no problems with rod M-10, the licensee performed prior to the subsequent startup a hot rod drop test with satisfactory result Unit 2 Startup on August 5 The inspectors monitored portions of the unit startup and main turbine generator loading as outlined in procedures 2-GOP-revision. 6, Unit Startup, HSD To 2% Reactor Power and 2-GOP-revision 6, Unit Startup, 2% Reactor Power to Max Allowed Powe During the turbine loading and transfer from the MFRV bypass valves to the main MFRVs the inspectors observed weakness that could be attributed to a lack of effective command and contro The situation arose when SG levels were swinging in 2 of the 3 SGs and the 5 ROs assigned to the unit in the control room were not working in concert. There were 4 licensed SROs in the area and it was not clear to the inspectors who, if anyone, was directing/supervising the evolution. The evolution was further complicated by sporadic SG PORVs openin The licensee's operations observers, part of the self assessment process, was also critical of this performanc The observers wrote a report of their observations and submitted this to management for their evaluation. Operations and station management discussed with the inspectors corrective actions to prevent future occurrences. This operator performance was not typical of previous performance observation Improved operating performance was observed during the August 6 Unit 2 shutdow Unit 2 Shutdown on August 6 due to PZR Safety Valve Leakage The inspectors monitored the control room portion of the Unit 2 shutdown to repair leaking PZR safety valve The shutdown was conducted in accordance with procedures 2-GOP-2.2 revision 5, Unit Shutdown 25%-30% Reactor Power to 2% Reactor Power and 2-GOP-revision 4, Unit Shutdown 2% Reactor Power to HS The pre-evolution brief was detailed and management involvement was eviden Command and control problems that were noted during the August 5 Unit 2 startup were not observed during this plant transient. The evolution was handled smoothly with no errors and good procedure adherence was note Within the area inspected, no violations were identifie.
Maintenance Inspections (62703) (42700)
During the reporting period, the inspectors reviewed the following maintenance activities to assure compliance with the appropriate procedure Troubleshooting of Unit 2 A MFRV Failure The inspectors monitored the licensee's troubleshooting of the A MFRV that resulted in the Unit 2 reactor trip. The RO at the time of the failure indicated that the control board A MFRV indication was "O" position with a "O" deman The troubleshooting concentrated on the electrical portion of the control system that could cause that type of a proble The troubleshooting consisted of performing component level checks followed by loop functional verification. The troubleshooting did not immediately identify why the valve failed closed. The licensee replaced the manual/aµto station and the flow controller. Subsequent test stand testing of the removed manual/auto station identified that the internal -15 volt DC power supply output was erratic and this had caused the A MFRV to fail close Within the areas inspected, no violations were identifie.
Surveillance Inspections (61726, 42700)
During the reporting period, the inspectors reviewed surveillance activities to assure compliance with the appropriate procedure and TS requirement Containment Air Partial Pressure/SW Inlet Temperature Calibrations TS'3.8.D required that containment air partial pressure be maintained within the acceptable operation range as specified in Figure 3.8-1 whenever the RCS temperature and pressure exceed 450 degrees Fahrenheit and 350 psi Two parameters of Figure 3.8-1 were allowable containment air partial pressure and maximum SW temperature limi The inspectors reviewed procedures 1 and 2-PT-2.12, Reactor Containment Pressure, dated June 24, 199 The purpose of these procedures was to calibrate the reactor containment air partial pressure channels to ensure compliance with TS Table 4.1-1, Item 2 Item 21 required that the containment air partial pressure instrumentation be calibrated each refueling. The inspectors verified that all components in the channels were properly teste The system was classified as safety-relate No deficiencies were note The inspectors reviewed the licensee's method for calibrating the SW inlet temperature indicatio In each unit there were four RTDs, one at each main CW inlet. The output of each RTD provided input into the control room computer and was displayed in the control roo The average of four inlet SW temperatures was used for determining allowable containment air partial pressure. The inspectors reviewed Unit 1 WOs 138178, 138179, 138180, and 13818 These WOs were utilized to calibrate the Unit 1 inlet SW inlet temperature indication channels on April 4, 199 The RTDs were
considered non-safety related and brief instructions for calibrating the channels were provided on the WO SW inlet indication channels were required to be calibrated every three year The licensee routinely compared the four SW inlet indications in each unit and verified that the readings were similar which provided assurance that the RTDs were indicating properl During the inspection period, the licensee submitted a TS change to broaden the band of SW inlet temperatures specified on Figure 3.8-Within the area inspected, no violations were identifie.
Action on Previous Inspection Items (92701) Feedwater Isolation Function (URI 50-280,281/93-15-0l)
IR 50-280,281/93-15 described a possible unreviewed safety question associated with the feedwater isolation metho URI 50-280,281/93-15-01 was identified to address the appropriateness of the use of reliability and Probabilistic Risk Assessment methodologies in the 10 CFR 50.59 SE that resolved this issu As discussed in IR 50-280,281/93-15, the feedwater isolation function was described in UFSAR section 14.3.2. A detailed description was not provided, but the system consisted of a safety related protective system that is actuated by an SI signal which causes the automatic closure of all three pairs of air operated MFRVs and bypass MFRVs, and the tripping of the MFWP With any operating unit MFRV disabled in the full open position on its jack to facilitate trouble shooting/repairs, there are electrical system failure scenarios that would require considerable manual operator actions to accomplish feedwater isolatio As described in IR 50-280,281/93-15 the licensee's CNS group challenged previous unit operations with the MFRV on its jac Consequently, SE 93-142 was developed to justify an eight-hour AOT with operator action within 30 seconds. This was subsequently superseded by SE 93-15 SE 93-155 and a TS interpretation were developed to allow plant operations to continue with an administrated 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AOT with 220 seconds for operator manual action for an inoperable MFRV (i.e., jacked open).
The new SEs were based on reliability and probability considerations that the licensee considered acceptable by NSAC-125, Guidelines for 10 CFR 50.59 Safety Evaluation The licensee also considered that manual operator action to trip the MFRV off the jack and allow it to close within a 220 second time limit would be achievable and acceptabl The inspector's review of SE 93-155 identified a concern with the extent of the licensee's analysis of the impact of operating with
- the MFRV jacked ope Section 18 of the SE indicated that a transient assessment of MSLB-using a 220 second isolation time shows that excessive cooldown {as measured by the reactor vessel integrity criteria) does not occur. Through an independent review of supporting calculation SM-899 and discussions with the licensee's Nuclear Fuels Analysis Group, the inspectors determined that the licensee only re-evaluated the PTS steam break analysis for the 220 second feedwater isolation time for PT Assumptions in the calculation excluded the need to re-analyze DNBR effects and eliminated the inclusion of a stuck control rod based on probability. Also it appears that containment pressurization from a MSLB was considered by the licensee to be outside the current licensing basis for Surry and has not been analyze Additionally, the use of non-safety related equipment, which is not covered in the TS, to perform the isolation function complicates this issue as wel Since the use of NSAC-125 has not been endorsed by the NRC, the inspectors requested technical assistance in the resolution and evaluation of the UR Of particular concern was whether this issue or similar issues can be evaluated on reliability and probability considerations based on NSAC-125, Guidelines for 10 CFR 50.59 Safety Evaluations, which is currently being generically utilized by the industry in the evaluation of safety issue Additionally, the inspectors questioned whether PTS was the only event that would be sensitive to delays or failure to isolate feedwater and is it acceptable to use non-safety equipment for the feedwater isolation functio This URI remains open pending further review by the NR It should be noted that the licensee has not operated with the MFRV on the jack since this issue was raised and the licensee's TS interpretation requires NRC notification prior to implementation if possibl Within the areas inspected, no violations or deviations were identifie.
Licensee Event Review (92700)
The inspectors reviewed the LER listed below and evaluated the adequacy of the corrective actio The inspectors' review also included followup of the licensee's corrective action implementatio (Closed) LER 50-281/92-01, Less Than One Operable Charging Pump Due to Mechanical Failure and Personnel Error. This issue involved all three Unit 2 charging pumps not being available. The A charging pump was declared inoperable because of an oil leak, the 8 charging pump was declared inoperable because its ventilation damper failed to automatically open when the pump was started, and the C charging pump control switch was in the pull to lock position in accordance with operating procedure The oil leak on the A charging pump was attributed to a degraded gasket which was replace The failure of the
,
B charging pump MOD to automatically open was attributed to personnel error in that the hand wheel was not properly disengaged following manual operation of the MO As corrective action operators were trained on the proper operation of the charging pump ventilation MODs, labels were placed adjacent to the MODs instructing operators to pull out on the MOD handwheel after manual operation, and the ventilation system test procedure was revised to instruct operators to pull out on the MOD handwheel after manual operation. The inspectors walked down the charging pump MODs and verified proper labeling, reviewed PT-32.lA, Refueling Frequency Auxiliary Ventilation Filter Train Test, dated April 24, 1992 and verified that there were instructions for disengaging the MODs following manual operation, and reviewed Training Synopsis (RQ-92.3-TS-2) Charging Pump Ventilation Dampers and verified that operators were trained on MOD operatio.
Exit Interview The results were summarized on August 13, 1993, with those individuals identified by an asterisk in Paragraph The following summary of inspection activity was discussed by the inspectors during this, exit:
Item Number Status URI 50-280,281/93-15-0l Open LER 50-281/92-001 Closed Description (Paragraph No.)
Use of PRA for Unreviewed Safety Question Determination (paragraph 6.a).
Less than one operable charging pump due to mechanical failure and personnel error (paragraph 7).
Concerns with the unit 1 B steam generator leak were discussed with plant management and are documented in paragraph Proprietary information is not contained in this report. Dissenting comments were not received from the license.
Index of Acronyms and Initialisms AOT CFR CNS cw DC DNBR -
ECCS -
F GPM HSD IR ALLOWED OUTAGE TIME CODE OF FEDERAL REGULATIONS CORPORATE NUCLEAR SAFETY COOLING WATER DIRECT CURRENT DEPARTURE FROM NUCLEAR BOILING RATIO EMERGENCY CORE COOLING SYSTEM FAHRENHEIT GALLONS PER MINUTE HOT SHUTDOWN INSPECTION REPORT
'
IRPI -
LER MFRV -
MFWP -
NRC NSAC -
PORV -
PTS PZR RCS RO RM RTD SE SG SI SRO SW TS UFSAR -
INDIVIDUAL ROD POSITION INDICATION LICENSEE EVENT REPOR MAIN FEEDWATER REGULATING VALVE MAIN FEEDWATER PUMP MOTOR OPERATED DAMPER MAIN STEAM LINE BREAK NUCLEAR REGULATORY COMMISSION NUCLEAR SAFETY ANALYSIS CENTER POWER OPERATED RELIEF VALVE PROBALISTIC RISK ASSESSMENT POUNDS PER SQUARE INCH GAUGE PRESSURIZED THERMAL SHOCK PRESSURIZER REACTOR COOLANT SYSTEM REACTOR OPERATOR RADIATOR MONITOR RESISTANCE TEMPERATURE DETECTOR SAFETY EVALUATION STEAM GENERATOR SAFETY INJECTION SENIOR REACTOR OPERATOR SERVICE WATER TECHNICAL SPECIFICATION UPDATED FINAL SAFETY ANALYSIS REPORT UNRESOLVED ITEM WORK ORDER