IR 05000101/2002006
| ML17263A961 | |
| Person / Time | |
|---|---|
| Site: | Ginna, 05000101  |
| Issue date: | 02/27/1995 |
| From: | Lazarus W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17263A960 | List: |
| References | |
| 50-244-95-03, 50-244-95-3, NUDOCS 9503140006 | |
| Download: ML17263A961 (16) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION I
Inspection Report 50-244/95-03 License:
DPR-18 Facility.
Inspection:
Inspectors:
R.
E. Ginna Nuclear Power Plant Rochester Gas and Electric Corpor ation (RGSE)
January 1 through February 6, 1995 T. A. Moslak, Senior Resident Inspector, Ginna E.
C. Knutson, Resident Inspector, Ginna Approved by:
M.
rus, Chi f Rea r Projec ction 3B INSPECTION SCOPE Date Plant operations, maintenance, engineering, plant support, and safety assessment/quality verification.
eSoa>4000+
SSoaOS DR ADOCK 05000244 PQR
INSPECTION EXECUTIVE SNNLRY Operations At the beginning of the inspection period, the plant was operating at full power (approximately 98 percent).
On January 28, 1995, a planned power reduction to approximately 50 percent was performed to plug leaking main condenser circulating water tubes.
The plant was returned to full power on January 29, 1995, and stayed at full power for the remainder of the inspection period.
Naintenance During a routine monthly surveillance test on the B-emergency diesel generator, the supply breaker to one of the two associated class lE electrical buses failed to initially close.
The cause of the breaker failure was suspected to be an intermittent mechanical malfunction, associated either with the breaker itself or the synchroscope selector switch.
Subsequent inspection and testing could not positively identify the problem and the breaker was successfully returned to service.
However, the licensee concluded that the breaker should be replaced as soon as a spare could be modified and tested.
Work on the replacement breaker was in progress at the close of the inspection period.
The inspector considered that the licensee's response was comprehensive and that the licensee's decision to ultimately replace the breaker demonstrated concern for the reliability of safety equipment.
Engineering On January 28, 1995, an operator noted evidence of leakage from the B-steam generator (SG) secondary side blowdown piping.
The source of the leak could not be precisely determined due to insulation on the piping, however it appeared to be in a portion of the pipe that could not be isolated from the SG.
The Plant Operations Review Committee concluded that the leak did not present an immediate operability concern, but initiated several actions to evaluate the B-SG blowdown line leak.
The consequences of a circumferential break of the one-inch line were examined by engineering and found to be bounded by the design basis faulted SG accident.
The plant simulator was used to evaluate overall plant response to a blowdown line failure.
Plant management established that a rapid decrease in the containment sump automatic pumpdown interval to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> would serve as the point at which the plant would be shut down to repair the leak.
On February 6, 1995, the insulation was removed from the B-SG blowdown piping, and no leak could be located.
Instead, steam was noted coming from between the insulation and the SG, at the blowdown line penetration to the SG.
At the close of the inspection period, the source of this leakage was under investigatio (EXECUTIVE SUMNARY CONTINUED)
The inspector considered that the licensee had demonstrated excellent safety perspective in expediting a thorough evaluation of this condition.
Plant Support Routine observations in the areas of radiological controls, security, and fire protection indicated that these programs were effectively implemented.
Safety Assessment/
guality Verification The inspector attended a meeting of the Nuclear Safety Audit and Review Board.
Issues reviewed included proposed changes to Technical Specifications and the guality Assurance (gA) Plan, Licensee Event Reports, NRC Inspection Reports, results of gA audits, Performance Indicators, and recent technical issues.
Indepth discussions addressed the causes of various incidents and the status of implementing corrective actions.
Excellent assessments were made of incident safety significanc EXECUTIVE SUMMARY.
TABLE OF CONTENTS
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TABLE OF CONTENTS 1V 1.0 OPERATIONS (71707)
1. 1 Operational Experiences
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1.2 Control of Operations
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1.3 Main Condenser Tube Plugging
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1 2.0 MAINTENANCE (62703, 61726)
2. 1 Preventive Maintenance 2. 1. 1 Routine Observations 2. 1.2 Safety System Maintenance Outages
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2. 1.3 Emergency Diesel Generator B Bus 16 Supply Breaker Failure to Close During Periodic Testing 2.2 Surveillance Observations
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3.0 ENGINEERING (71707, 37551)
3. 1 Possible Leak From B-Steam Generator Secondary Side Blowd 4.0 PLA 4.1 4.2 NT SUPPORT (71750)
Radiological Controls
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4. 1. 1 Routine Observations Security 4.2. 1 Routine Observations 4.3 Fire Protection
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4.3. 1 Routine Observations
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5. 0 SAFETY ASSESSMENT/EQUALITY VERIFICATION (71707)
5. 1 Nuclear Safety Audit and Review Board (NSARB)
5.2 All Employee/Management Meeting
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5.3 Periodic Reports 6. 0 ADMINISTRATIVE 6. 1 Senior NRC Management Site Visit 6.2 Backshift and Deep Backshift Inspection 6.3 Exit Meetings
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DETAILS 1. 0 OPERATIONS (71707)
1.1 Operational Experiences At the beginning of the inspection period, the plant was oper ating at full power (approximately 98 percent).
On January 28, 1995, a planned power reduction to approximately 50 percent was performed to plug leaking main condenser circulating water tubes.
The plant was returned to full power on January 29, 1995, and remained at full power for the balance of the inspection period.
There were no other significant operational events or challenges during the inspection period.
1.2 Control of Operations Control room staffing was as required.
Operators exercised control over access to the control room.
Shift supervisors maintained authority over activities and provided detailed turnover briefings to relief crews.
Operators adhered to approved procedures and were knowledgeable of off-normal plant conditions.
The inspectors reviewed control room log books for activities and trends, observed recorder traces for abnormalities, assessed compliance with technical specifications, and verified equipment availability was consistent with the requirements for existing plant conditions.
During normal work hours and on backshifts, accessible areas of the plant were toured.
No operational inadequacies or concerns were identified.
1.3 Nain Condenser Tube Plugging Monitoring of steam generator chemistry over several weeks showed a slow but persistent increase in chloride concentration from less than one part per billion (ppb) to approximately six ppb on January 27, 1995.
This increase was directly attributed to an increase in main condenser tube leakage which rose to 0.66 gallon per minute (gpm) in the lAl waterbox, 0.86 gpm in the lA2 waterbox, and 0.85 gpm in the 1B2 waterbox.
A forth section of the main condenser, the 1B1 waterbox, showed no increase, remaining at its nominal leak rate of about 0. 18 gpm.
In response to the degrading feedwater chemistry specifications, licensee management decided to reduce power to less than 50 percent to permit isolation of the affected waterboxes and tube plugging.
At 12:53 a.m.
on January 28, 1995, power was smoothly reduced at 10 percent per hour until stable operation was achieved at 46 percent at 5:58 a.m.
Upon isolating the lA2 waterbox, a helium tracer gas technique was used to identify tube regions that were potentially leaking.
Intact tubes were sequentially eliminated, and suspected leaking tubes were marked, then plugged.
Five tubes were plugged in the 1A2 waterbox, and subsequently ten tubes were plugged in the lAl and eleven tubes plugged in the 1B2 waterboxes.
Following completion of tube plugging, power was smoothly escalated at
percent per hour beginning at 3:42 a.m.
on January 29, 1994, achieving a
maximum power of 98 percent at 11:22 The inspector observed aspects of the tube inspection/plugging and related activities.
Power reduction/escalation operations were found to be effectively carried out.
Activities were well coordinated with the control room.
A larger, more conservative tube population was plugged to assure that potentially leaking tubes were addressed.
Review of feedwater chemistry results show that chloride concentration returned to less than one ppb, indicating that plugging was successful.
To correct the long standing problem of leaking main condenser tubes, RGLE will replace the existing Admiralty Brass tubes with more resilient Stainless Steel 316 tubes during the next refueling outage, scheduled to begin Narch 27, 1995.
2.0 MAINTENANCE (62703, 61726)
2.1 Preventive Naintenance 2.1.1 Routine Observations The inspector observed portions of maintenance activities to verify that correct parts and tools were utilized, applicable industry code and technical specification (TS) requirements were satisfied, adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon completion.
The following maintenance activity was observed:
"Calibration of Containment Pressure Transmitter PT-947," observed January 24, 1995 2.1.2 Safety System Naintenance Outages During this inspection period, the following safety systems were taken out of service for corrective/preventive maintenance, as identified by the associated work orders (MOs):
B-Notor Driven Auxiliary Feedwater (AFM) Pump Taken out of service (OOS) 5:30 a.m. January ll, 1995 Returned to service (RTS) 10: 12 p.m. January ll, 1995 Allowed outage time (AOT) 7 days per TS 3.4.2. l.a WO 19404324,
"Ninor Nechanical Inspection per N-11.5C and Oil Change and Analysis" MO 19401616,
"Replace Lube Oil Flow Indication Sight Glass with Gage in accordance with Technical Staff Request (TSR)-056" MO 19500077,
"Electrical Diagnostic Notor Tests" C-Standby Auxiliary Feedwater (SAFW)
Pump OOS 5:22 a.m. January 12, 1995 RTS 9:36 a.m. January 13, 1995 AOT 14 days per TS 3.4.2.3 WO 19404330,
"Change and Analyze Lube Oil"
"Replace right hand rail latch in breaker cubicle (bus 14)"
"Electrical Diagnostic Notor Tests" MO 19404116,
"Calibration of Flow Loop 4084".
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A-Residual Heat Removal (RHR)
Pump OOS 6:34 a.m. January 17, 1995 RTS 2:56 p.m. January 17, 1995 AOT 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per TS 3.3.1.5.a WO 19404319, WO 19403232, MO 19404112, WO 19404207, pump run"
"Ninor Naintenance (Lube Oil Change)"
"Calibration of Flow Transmitter (FT) FT-626"
"Calibration of Air Operated Valve (AOV) AOY-625"
"Inspect flow element (FE) FE-673 for leakage during
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A-Containment Spray (CS)
Pump OOS 6:35 a.m.
January y 19, 1995 RTS 3:14 p.m. January 19, 1995 AOT 3 days per T.S. 3.3.2.2.b WO 19500080,
"Electrical Diagnostic Notor Tests" WO 19404290,
"Replace right hand rail latch in breaker cubicle (bus 14)"
The inspector verified that the duration of these system outages was less than the allowed outage time as permitted by technical specifications, that the redundant train was operable, and that acceptance testing adequately verified system operability.
2.1.3 Emergency Diesel Generator B Bus 16 Supply Breaker Failure to Close During Periodic Testing At 9: 16 a.m.,
on January 23, 1995, while attempting to perform a routine monthly surveillance test on the B-emergency diesel generator (EDG), the EDG supply breaker to bus 16 (one of the two class IE electrical buses supplied by the B-EDG) failed to initially close.
Mhen the operator placed the breaker control switch in the "close" position, the breaker position disagreement light energized, and main control board (NCB) annunciator J-9,
"Safeguard Breaker Trip," alarmed.
As part of the test, the B-EDG had earlier been electrically connected to bus 17.
Testing was secured, and the B-EDG was electrically unloaded and shut down.
Following operability verification of the A-EDG, the B-EDG was declared inoperable at 9:35 a.m.,
January 23, 1995.
Technical specification 3.7.2.2 allows one EDG to be inoperable for up to seven days before requiring that the plant be shut down.
Troubleshooting of the B-EDG supply breaker to bus 16 was unsuccessful in identifying the cause of the failure.
Visual inspection of the breaker revealed no problems.
The breaker operated correctly when test-cycled remotely, and test readings indicated that the breaker control circuitry was
operating properly.
After multiple cyclings in the test position, the breaker was reinstalled and operationally tested satisfactorily.
The B-EDG was declared operable on January 23, 1995.
The cause of the breaker failure was suspected to be an intermittent mechanical malfunction, associated either with the breaker itself or the synchroscope selector switch on the NCB.
The licensee concluded that the breaker should be replaced as soon as a spare could be modified and tested; at that time, the synchroscope selector switch would also be examined.
Wor k on the replacement breaker was in progress at the close of the inspection period.
The inspector reviewed completed work order 19500269,
"B-Diesel Generator Breaker Did Not Close During PT-12.2."
Troubleshooting was accomplished in accordance with procedures CNE-50-02-52/EG1Bl,
"Westinghouse 480Y Air Circuit Breaker, Type DB-75, Emergency Diesel Generator B Bus 16, Position llC, Maintenance For 52/EGlBl," and GHE-00-99-01, "Electrical Troubleshooting."
The inspector noted several minor pen-and-ink change discrepancies.
A concern that pen-and-ink changes were being overused was discussed with the appropriate plant management, who plan to discuss the item with maintenance personnel during upcoming training.
The inspector considered that these changes had not invalidated the test results, and considered that the troubleshooting approach had been appropriate.
In conclusion, the inspector considered that the licensee's response to failure of the B-EDG supply breaker to bus 16 was appropriate.
Troubleshooting was timely and of sufficient scope to bound the potential causes of the problem.
The operability determination was well considered and was adequately supported by the test results.
The licensee's decision to ultimately replace the breaker demonstrated concern for the reliability of safety equipment.
2.2 Surveillance Observations 2.2.1 Routine Observations Inspectors observed portions of surveillances to verify proper calibration of test instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to limiting conditions for operation (LCOs),
and correct system restoration following testing.
The following surveillance was observed:
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Periodic Test (PT)-2.2g,
"Residual Heat Removal System quarterly,"
revision 6, effective date November 8, 1994, observed January 18, 1995 The inspector determined through observing this testing that operations and test personnel adhered to procedures, corrective action was promptly initiated if test results and equipment operating parameters did not meet acceptance criteria, and redundant equipment was available for emergency operatio.0 ENGINEERING (71707, 37551)
3. 1 Possible Leak From B-Steam Generator Secondary Side Blowdown Piping r
The A-containment sump automatic pumpdown interval provides one indication of water leakage from systems inside the containment building.
During this operating cycle, the sump pumpdown interval has trended downward, from a historic value on the order of 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />, to the current value of approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.
Facility sampling has demonstrated that the leakage is from secondary, non-radioactive systems.
Periodic containment entries have identified and repaired several water leaks from these systems, the most significant of which was the A-steam generator (SG) secondary side manway leak (discussed in inspection report 50-244/94-25).
Despite an apparently successful repair of this leak in October 1994, the sump pumpdown interval was not significantly affected; it was later determined that the leak had not actually been stopped.
During the main condenser tube leak repairs on January 28, 1995, the licensee utilized the period of low power operation to again inspect inside the containment building for water leaks.
During this inspection, an operator noted water droplets on the side of the B-SG, in the vicinity of the secondary side blowdown piping.
Upon further examination, the operator observed wisps of steam coming from the connection of this one-inch diameter pipe to the SG.
The source of the steam could not be precisely determined due to insulation on the piping, however it appeared to be in a portion of the pipe that could not be isolated from the SG.
Insulation removal was not immediately pursued due to the high general area radiation level (4-5 rem/hour).
The Plant Operations Review Committee (PORC) met to discuss the possible B-SG blowdown line leak.
The leak was considered not to present an immediate operability concern, based on:
1) the leakage rate was small, as evidenced by no significant change in the containment sump automatic pumpdown interval over the preceding several weeks; 2) the applicable TS limit of less than 0.5 gallons per minute from a closed loop in containment (which translates to a containment sump automatic pumpdown frequency of less than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />)
was satisfied; 3) degrading conditions would be promptly identified by a decrease in the containment sump automatic pumpdown interval; and, 4) recent radiographs (obtained during the 1994 refueling outage) of most of the affected piping had shown no problems.
As a result of the PORC meeting, several actions were undertaken to evaluate the B-SG blowdown line leak.
The consequences of a total line failure (circumferential break)
were examined by engineering.
Results indicated.that such a fault would cause safety injection to initiate due to containment pressurization; however, the scenario was bounded by the design basis faulted SG accident.
Additionally, failure of the B-SG blowdown line concurrent with a main steam line break in the A-SG was evaluated.
Although such a scenario is beyond the design basis, the analysis showed that containment pressure would remain below the design maximum value of 60 psig.
Finally, the plant simulator was used to evaluate overall plant response to a blowdown line failure.
Plant management established a minimum containment sump automatic pumpdown interval of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> as the point at which the plant would be shut down to repair the lea Over the weekend of February 4-5, 1995, the containment sump automatic pumpdown interval decreased to approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
On February 6, 1995, preparations were made to enter containment to remove the insulation from the B-SG blowdown piping.
Although this would involve work.in a high radiation area (approximately 10 rem/hour),
management considered the potential safety significance to justify the dose expenditure.
At the time, power reduction was considered unadvisable due to cold weather conditions.
An ALARA briefing was conducted prior to the entry to evaluate the expected dose and examine how dos'e reduction could most effectively be achieved.
Insulation removal from the B-SG blowdown piping was performed on February 6, 1995.
Approximately 1.2 person-rem of exposure was expended in this effort.
The blowdown piping was found not to be leaking.
Steam was still noted to be coming from between the insulation and the SG, but it was not associated with the blowdown line penetration to the SG.
At the close of the inspection period, the source of this leakage was under investigation.
The inspector considered that the licensee had taken prudent action to address the possible leak in the B-SG blowdown line.
The basis for continued operation was thoroughly examined and the consequences of component failure were appropriately analyzed.
Although the line was ultimately determined not to be leaking, the licensee demonstrated excellent safety perspective in not delaying critical evaluation of the condition until after this determination was made.
4.0 PlANT SUPPORT (71750)
4.1 Radiological Controls 4. 1. 1 Routine Observations The inspectors periodically confirmed that radiation work permits were effectively implemented, dosimetry was correctly worn in controlled areas and dosimeter readings were accurately recorded, access to high radiation areas was adequately controlled, survey information was kept current, and postings and labeling were in compliance with regulatory requirements.
Through observations of ongoing activities and discussions with plant personnel, the inspectors concluded that the licensee's radiological controls were effective.
Among the items reviewed was the inspection of secondary systems inside containment such as the 1.2 person-rem expended during insulation removal and inspection of the B-SG blowdown piping, discussed in Section 3. 1.
4.2 Security 4.2.1 Routine Observations During this inspection period, the inspectors verified that x-ray machines and metal and explosive detectors were operable, protected area and vital area barriers were well maintained, personnel were properly badged for unescorted or escorted access, and compensatory measures were implemented when necessary.
No unacceptable conditions were identifie.3 Fire Protection 4.3.1 Routine Observations The inspectors periodically verified the adequacy of combustible material controls and storage in safety-related areas of the plant, monitored transient fire loads, verified the operability of fire detection and suppression systems, assessed the condition of fire barriers, and verified the adequacy of required compensatory measures.
No discrepancies were noted.
5.0 SAFETY ASSESSMENT/EQUALITY VERIFICATION (71707)
5.1 Nuclear Safety Audit and Review Board (NSARB) Meeting On January 25-26, 1995, the NSARB met to review issues relevant to the safe operation of the Ginna facility.
Issues reviewed included proposed changes to Technical Specifications and the guality Assurance (gA) Plan, Licensee Event Reports, NRC Inspection Reports, results of gA audits, Performance Indicators, and recent technical issues.
Additionally, the board closed out the actions requested by the NSARB Chairman during NSARB Meeting 207, regarding LER 940-009 "'B'afety Injection Pump Root Cause Analysis."
The NSARB review required that facility management reinforce the responsibilities of all personnel involved with safeguard systems maintenance and operation and recommended increased operator awareness of safeguard system configuration during equipment operation.
5.2 All Employee/Management Meeting On February 1,
1995, a plant meeting was held by RGSE management.
Performance indicators for 1994 were reviewed and management expectations regarding safety, attention to detail, and maintaining a questioning attitude were presented.
5.3 Periodic Reports Periodic reports submitted by the licensee pursuant to Technical Specification 6.9. 1 were reviewed.
Inspectors verified that the reports contained information required by the NRC, that test results and/or supporting information were consistent with design predictions and performance specifications, and that reported information was accurate.
The following reports were reviewed:
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Monthly Operating Report for December 1994 No unacceptable conditions were identifie. 0 ADMINISTRATIVE 6.1 Senior NRC Nanagement Site Visit During this inspection period, one senior NRC manager visited Ginna Station.
On February 3, 1995, Nr. Charles W. Hehl, Director of the Region I Division of Radiation Safety and Safeguards, toured the site and met with senior licensee management.
6.2 Backshift and Deep Backshift Inspection During this inspection period, deep backshift inspections were conducted on January 2, 21, 28, and 29, 1995.
6.3 Exit Neetings At periodic intervals and at the conclusion of the inspection, meetings were held with senior station management to discuss the scope and findings of inspections.
The exit meeting for inspection report 50-244/95-01 (Maintenance, conducted January 17 February 3, 1995)
was held by Nr. Thomas Kenny on February 3, 1995.
The exit meeting for inspection report 50-244/95-02 (Engineering, conducted January 9-13, 1995)
was held by Nr. Leonard Prividy on January 13, 1995.
The exit meeting for inspection report 50-244/95-04 (Security, conducted January 17-20, 1995)
was held by Nr. Edward King on January 20, 1995.
The exit meeting for inspection report 50-244/95-05 (Radioactive materials transportation, conducted January 30 - February 2, 1995)
was held by Nr. James Noggle on February 2, 1995.
The exit meeting for the current resident inspection report 50-244/95-03 was held on February 9, 1995.