IR 05000220/1992016
| ML17056C000 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/27/1992 |
| From: | Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17056B999 | List: |
| References | |
| 50-220-92-16, 50-410-92-18, NUDOCS 9209170123 | |
| Download: ML17056C000 (40) | |
Text
Report Nos.:
Docket Nos.:
License Nos.:
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
92-16; 92-18 50-220; 50-410 DPR-63; NPF-69 Licensee:
Niagara Mohawk Power Corporation 301 Plainfield Road Syracuse, New York 13212 Facility:
Location:
Dates:
Nine Mile Point, Units 1 and 2 Scriba, New York July 5 through August 15, 1992 Inspectors:
W. L. Schmidt, Senior Resident Inspector R. A. Laura, Resident Inspector W. F. Mattingly, Resident Inspector (in training)
C. D.
dslee, Reactor En 'neer, Intern Approved by:
Larry
. Nicholson, Chief Reactor Projects Section No. 1A Division of Reactor Projects Dae
?~i:
Yli i p
i p
d
'
i i
p i
f plant operations, radiological controls, maintenance, surveillance, emergency planning, security, and safety assessment/quality verification activities.
R~resul:
See Executive Summary.
'7209170i23 920827 PDR ADOCK 05000220
EXE TIVE ARY Nine Mile Point Units 1 and 2 NRC Region I Inspection Report Nos. 50-220/92-16 & 50-410/92-18 July 5, 1992 - August 15, 1992 Plant eration The Niagara Mohawk Power Corporation (NMPC) conducted outage activities, restarted Unit 1 safely and properly responded to an inadvertent reactor scram. Unit 2 was operated safely and operators properly responded to a loss ofone off-site power line followingthe opening of an off-site breaker.
An NMPC investigation conducted in response to NRC Information Notice 92-30 preliminarily identified that seven operators at Unit 1 and four at Unit 2 falsified entry into plant areas.
The issue of falsified operator rounds was considered unresolved pending further NRC review.
R iolo ical C ntrol A review of the non-radiological water chemistry program at both units identified that the program was effectively implemented.
A previous Unit 2 violation, dealing with the failure of personnel to conduct whole body frisking when exiting a restricted area, was closed.
Maint n nc an rv illance Unit 1 maintenance personnel performed well during repairs to an emergency condenser condensate return line valve, Unit 2 personnel performed excellently during repair to a high pressure core spray suppression pool suction valve. In addition, the Unit2 staff took appropriate actions to address conditions identified during surveillance testing of Division III emergency diesel, generator (EDG) cooling water check valves and to address the failure of the Division 'I
.-EDG to start for a required operability verification.
~ecurit NMPC effectively implemented initial reconfiguration changes to the protected area to accommodate demolition of a site building and construction of a new building.
n in rin an T hnical ort An unresolved item was identified concerning the adequacy of the EDG cooling water pump testing at Unit 1.
During surveillance testing at Unit 2, NMPC had difficulty achieving acceptable local leak rate tests on the resilient seat suppression pool vent valves.
NMPC committed to keep these valves shut to maintain containment integrity, and to limituse of these valves to use ifdirected by the emergency operating procedures.
Further, NMPC committed to
Executive Summary (Continued)
test the other resilient seat containment purge vent and purge valves following use to verify containment integrity.
f A
ent alit V rifi ti n Effective August 1, 1992 Mr. Neil Cams became the new Vice President - Nuclear Generatio SUMMARYOF FACILITYACTIVITIF8
'2 1.1 Nia ara Mohawk P wer ti n A 'vi i The Niagara Mohawk Power Corporation (NMPC) safely conducted outage activities an'd power operation at Nine MilePoint Unit 1 (Unit 1) over the period.
Repairs to the emergency cooling system were completed and a reactor vessel hydrostatic test was successfully performed.
The unit started up on August 2 and operated at power until August 7 when an automatic reactor scram occurred due to a spurious local power range monitor spike.
The unit was restarted on August 8 and operated at power'hrough the end ofthis inspection period.
On August 13 reactor
.
power was lowered and the generator removed from the grid to allow the correction of a steam leak in a feedwater heater.
Following the repair of the leak, the generator was placed onto the grid and reactor power increased.
NMPC safely operated Nine Mile Point Unit 2 (Unit 2) at near rated power throughout the period.
On July 28 one of the off-site 115kV power lines was lost as a result of an unexpected off-site breaker opening.
On August 1, 1992, Mr. Neil Cams became the new Vice President - Nuclear Generation, replacing Mr. Joseph Firlit who became the Vice President
-
Nuclear Support.
NMPC documented that Mr. Cams met the ANSI 3.1 qualifications for a plant manager in an internal memorandum, dated July 20, 1992.
The inspector reviewed this memorandum and found Mr.
Cams's qualifications acceptable.
1.2 I~II A
Resident inspectors conducted inspection activities during normal, backshift and weekend hours over this period.
There were 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> of backshift (evening shift) and 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> of deep backshift (weekend, holiday, and midnight shift) inspection during this period.
During the week of July 27 a regional inspection of NMPC's response to Generic Letter 88-14 at Unit 1 and service water issues at Unit 2 was conducted; The findings of this inspection willbe documented in Combined Inspection Report 220/92-18 and 410/92-20.
On July 24, 1992, the NRC staff issued a schedular exemption to the 10 CFR 50, Appendix J, type C leak test requirements for the emergency condenser condensate return containment isolation check valves (39-03 and 39-04).
NMPC tested the outboard containment isolation valve (39-05 and 39-06) in accordance with Appendix J, during the most recent outage.
Based on the information provided by NMPC the schedular exception was granted to the 1995 refueling outag.0 PLANT OPERATIONS (71707,93702)
2.1 Plant rations Review -
ni
The operations department performed well during the unplanned outage, reactor vessel hydrostatic test, plant startups and power maneuvering, and in response to the automatic reactor scram which occurred on August 7. Operators conducted control room activities well, including panel manipulations an'd operator response to alarms.
Two markups were reviewed and found to be well written and provided proper isolation for plant maintenance.
A commendable initiative this period was the removal of a glass wall located between the station shift supervisors.
(SSS) office and the rest of the control room.
This change facilitated incieased oversight and awareness of control room activities. NMPC identified several minor equipment configuration control issues over the period. While each instance was of low safety significance, collectively the events appeared to indicate a negative trend.
NMPC issued deviation event reports (DERs)
for each event and was conducting an overall evaluation.
2.1.1 A
matic Rea t r cram An automatic reactor scram occurred on August 7 from 76% reactor power while operators were performing manual half scram testing.
A half scram signal on the redundant channel occurred due to an average power range monitor (APRM) trip, from a single spurious spike on a local power range monitor (LPRM). Allautomatic systems performed as designed and no safety or electromatic relief valves lifted.
Operators responded properly, maintaining the unit in a safe condition.
Reactor vessel water level decreased to a minimum of 21 inches (107 inches above the active fuel) before being restored by the feedwater system, which initiated in its high pressure coolant injection (HPCI) mode.
The post scram review package showed that the LPRM spiked about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the scram'ut did not cause its associated APRM to trip. Operators monitored the LPRM and verified it was reading correctly.
The SSS discussed whether the LPRM should be bypassed with the on-shift reactor engineer.
Following past operating practice regarding isolated cases of detector spiking which did not cause its APRM to trip, the LPRM was not bypassed.
The LPRM was subsequently placed in the bypass condition prior to start-up of the unit on August 9. NMPC implemented the use of an LPRM log to more closely track spiking LPRMs and allow compensatory measures.
2.1.2 Reac r V 1 H r
tatic Te A reactor vessel hydrostatic test was successfully conducted as a post maintenance test of the emergency cooling system valve repairs.
A hydrostatic test pressure of 1150 psig was attained with a control rod drive pump as the pressure source.
The test director and the operations staff controlled the test well. Inspection of the welds at test pressure identified no leakage.
Leakage was identified coming from; three control rod drive stub tubes, the packing of various valves,
and several control rod drive flanges.
The inspector attended a NMPC management meeting conducted to evaluate the leakage and observed good open discussion concerning the repair plan for the identified leakage.
Following the hydrostatic test, the maintenance department rolled the three leaking control rod drive stub tubes to stop the water leakage;
. The other leaking equipment was repaired or evaluated and a reactor vessel leak check was conducted to prove the effectiveness of the repairs.
The inspector assessed that the decision to perform another leak check prior to restart of the unit demonstrated an excellent safety perspective.
In summary, very good performance was evident during the conduct of the reactor vessel hydrostatic test, the management meeting held to evaluate the results, and the decision to perform a vessel leak check following the stub tube rolling.
2.1.3 Observation of ntr 1 R m Activities On August 3, the inspector observed very good control ofplant operations during bypass valve fluctuations. The SSS,.Chief Shift Operator (CSO), and plant management adequately discussed the plan to stabilize pressure control when the bypass valves began to hunt during a special test of the turbine control system.
Good coordination of actions to take control 'of reactor pressure with the manual bypass valve jack was observed.
The operations department had adequate staff-to monitor the control systems at the turbine front standard and used excellent communications to coordinate the evolution.
On August 13, the inspector observed the operation ofthe unit with bypass valve pressure control during the warming of the turbine following the repairs to a steam leak in the second stage feedwater heater.
Operators controlled plant conditions well and were sensitive to plant pressure changes.
Communications between the SSS, the assistant SSS, the CSO, and the other operators on shift were excellent.
In summary the inspector observed very good communications and control of plant conditions by the operating crews during these two control room observations, 2.2 Pl nt ration Review -
nit 2 NMPC safely operated the unit at near rated power in conformance with approved procedures and regulatory requirements.
Control room activities, including shift turnovers and crew briefings, panel manipulations, and operator response to alarms, were observed.
Regular tours of the 'plant were conducted to assess equipment conditions, radiological conditions, fire protection, security, general housekeeping practices and personnel safety.
In general, the inspector observed an acceptable level ofperformance and generally good conditions throughout the plan.2.1 i l f ff-ieP wr Control room operators properly responded to an inadvertent partial loss of off-site power.
On July 28, the 115kV substation feed breaker R-60 to the Division II 4160 V emergency bus inadvertently tripped open on low nitrogen accumulator pressure.
This caused the loss of Line 6 (one of two off-site power sources for Unit 2) and the subsequent deenergization of the Division IIemergency bus.
The Division IIEDG automatically started and loaded as designed.
Several other expected engineered safety features actuations occurred, including a reactor building ventilation system isolation and the automatic start ofboth standby gas treatment trains.
Line 6 was restored and the Division IIEDG secured within nine hours.
Throughout the event-Unit 2 operated at 100% power.
Twelve hours prior to the event, regional power control (RPC) notified the control room of a problem they were investigating with the charging pump for the breaker R-60 nitrogen accumulator.
Breaker R-60 is located in the Scriba switchyar) about one-half mile from the site.
The Scriba switchyard is operated and maintained by Fulton based traveling operators dispatched by RPC.
RPC notified the control room of the following shortly before the event:
charging pump fuses were removed to prevent pump motor damage, since the motor was operating continuously but the pump was not raising accumulator pressure; to correct the problem, Line-6 needed to be removed from service (scheduled to start on day shift, in approximately six hours); if accumulator pressure was lost,-breaker R-60 would remain shut and continue to provide bus fault protection, but would be unable to reclose ifthe breaker did trip open.
After the event, the control room learned that nitrogen accumulator pressure was required to both open and close the breaker R-60, and, contrary to the information provided by RPC, breaker R-60 did trip open on a loss of accumulator pressure.
The inspector concluded that the control room staff responded properly to the event, made timely NRC notifications, took the appropriate technical specifications actions, and properly restored
'he electric plant to a normal condition.
Also, prior to the event, the SSS properly questioned RPC regarding the plant impact of the R-60 troubleshooting and maintenance plan.
NMPC management stated that their corrective actions would include an evaluation of the interface and coordination with the off-site RPC.
The inspector had no further questions at this time.
2.3 Tem ra In t cti n 2-01 In ti n f Pl n t r Activitie -
nit 1 and 2 A review of operator rounds at both units was conducted by the inspectors as follow-up to NRC Information Notice 92-30 which discussed falsification of plant records.
The inspectors accompanied various operators on the different rounds, including one backshift tour.
Shift supervisors, training staff personnel, and various managers were interviewed on their expectations oftheir staff forcompleting rounds. The operator round procedures were controlled as preventative maintenance procedures, which contain implementing instructions.
The operators conducted rounds in a professional manner and in accordance with the implementation instructions located in the rounds guide. No evidence ofany type offalsification
or misrepresentation was identified by the inspectors.,
Some minor problems of low safety significance were identified at both units.
There were several examples where operators were not sensitive to the causes for long term alarming annunciators on non-safety related panels in the plant.,Some operators entered contaminated areas to check that high radiation area gates were locked, while others performed a visual inspection from outside the contaminated areas in front of the gates.
Verification of the operability ofmotor heaters in safety related pump motors was inconsistent.
These observations were discussed with operations management who took appropriate corrective actions.
Changes to the rounds guide procedure were made as needed.
In response to,NRC Information Notice 92-30, an NMPC investigation of operator rounds at both units identified that seven operators at Unit 1, including one licensed reactor operator, and four non-licensed operators at Unit 2 had falsified that they had entered areas during the conduct of operator rounds.
The investigation focused on a five week time period and utilized the security computer logs.
Some individuals were involved in a single case and others in multiple cases.
Initially, these individuals were suspended from site access.
No technical specification required logs were missed.
NMPC took disciplinary action against the individuals involved.
The operators involved in one instance, including the reactor operator at Unit l, were returned to their duties.
The inspectors assessed that 'there was no safety consequence to the missed operator rounds at.
either unit. However, the inspector was concerned that the missed rounds by the operators and the minor problems identified during inspector observation ofrounds indicated weak supervisory and management oversight and unclear expectations.
This item willremain an unresolved item (220/92-16-01 and 410/92-18-01) pending further NRC review and assessment of the operator round falsifications.
3.0 RADIOLOGICALAND CHEMISTRY CONTROLS (71707, 84750)
3.1 W
r hemist
-
ni
n nit 2 NMPC implemented its water chemistry'ontrol program according to the applicable site procedures.
The inspector accompanied chemistry technicians on their shift rounds and during the collection of reactor water for later analysis.
Sampling was conducted in a professional manner, according to radiation protection requirements and site specific procedures.
The technicians were very knowledgeable of the basis for performing the reactor water sampling and of managements expectations.
The chemistry program required technicians to effectively trend and predict plant chemistry conditions.'.2 Closed Vi 1 tion 41 / 1-2 -
F il re to M R
i
ical P tin R
uirem nt-Unit 2 This violation involved the failure of several individuals to perform the required two-minute whole body frisk with a hand-held frisker when exiting a radiologically controlled area.
NMPC admitted to the violation in their March 5, 1992, response.
For corrective actions, two of four
restricted area egress points were closed to provide more positive control over personnel exit.
Also, the restricted area boundary was adjusted to preclude exit via the auxiliary boiler room, thus eliminating the need for personnel to frisk when exiting that area.
Personnel directly involved in the event were disciplined and counseled by the plant manager on appropriate radiological conduct while working within the restricted area.
The plant manager issued a
memorandum to plant personnel emphasizing the requirement to perform a proper whole body frisk and supervisors discussed the event with plant personnel.
Furthermore, a series ofNuclear Generation personnel meetings were held to discuss the issue of personnel errors and reinforce management expectations for performance.
NMPC's response to the violation was adequate, therefore, this violation was closed.
4.0 MAINTENANCE(62703)
4.1 ervation f. Maintenance Activiti
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nit 1 Mechani
rrective Maintenance Mechanical maintenance technicians successfully lapped the seating surfaces of the emergency cooling system condensate return containment isolation valve 39-05.
The work resulted from a failed leak rate test performed according to 10 CFR 50 Appendix J and technical specification requirements.
During disassembly, the inspector examined the valve seat which exhibited discoloration caused by water leakage past the disc and seat.
The disc was removed and machined in the shop.
The valve seat was lapped in place and a blue check performed to prove proper mating of the disc and seat.
The maintenance workers followed the requirements of the radiation work request.
A vendor representative was present to offer technical advice.
Following refurbishment of the valve, a seat leak rate test was conducted with unsatisfactory results.
The valve was disassembled again and the seating surface angle altered slightly to provide a narrower area of contact between the disc and seat.
Subsequently, the valve successfully passed the Appendix J seat leak rate test.
Supervisory and management oversight of this work activity was evident.
In summary, mechanical maintenance performed well during this maintenance activity.
4.2 erv tion of Mainten nce A iviti
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nit 2 4,2.1 n e ed Maintenance Effort n Hi h Pr r
ore S ra Limi r e A tu r
Unit 2 personnel performed very well during repair activities associated with the high pressure core spray (HPCS) suppression pool suction valve (2CSH*MOV118) limitorque actuator. During a HPCS system surveillance test the valve motor breaker tripped on thermal overload while the valve was being stroked closed. 'he operations staff declared the valve inoperable and deactivated it in the closed position, per the technical specifications (TS),
to maintain containment integrity. Electrical maintenance staff inspection of the actuator determined that the torque switch roll pin had sheared off, disabling the torque switch.
This allowed the valve to
close without the protection of the torque switch to limit seating thrust, and caused the valve motor trip on thermal overload.
During subsequent discussions, licensing, engineering, and operations management questioned the operability of the valve as a primary containment isolation valve and a TS limitingcondition for operation (LCO) was entered.
This decision was based on a lack of confidence in the valve seat integrity because of possible actuator and/or valve seat damage from actuator over-thrust.
Instrumentation and controls staff performed a valve seat leak rate test, reverified containment integrity, and the LCO was promptly exited.
Restoration of 2CSH*MOV118 to an operable condition required additional close coordination between various maintenance shops and the operations department.
Several additional leak rate tests were perfornted, a special stem locking device was installed and removed, and TS LCOs were entered and promptly exited as appropriate.
Engineering analysis and investigation determined that the roll pin was undersized for this application. Alargerpinwasinstalledpriortoreturning thisvalvetoservice.
AnindustryPart 21 exists for the same roll pin in a different size actuator.
NMPC was performing a Part 21 review for this event and planned to increase its inspection/replacement schedule for the roll pin.
The inspector concluded that Unit 2 personnel properly performed the maintenance and testing, appropriately interpreted the TSs, and exhibited an excellent safety perspective. 'he close coordination between the operations and maintenance staffs and their integration with the licensing and engineering staffs demonstrated an outstanding effort to solve a maintenance issue.
5.0 5.1 SURVEILLANCE(61726, 61707)
Observation f urveillance Activiti
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nit 2 The inspectors observed and reviewed portions of the following surveillance tests to assess technical specification and procedural conformance, 'system/equipment removal and restoration from service, test results review, and deficiency resolution.
Operations personnel properly conducted the primary containment penetration verification test for valves outside containment.
The primary containment penetrations required to be closed during accident conditions, but not capable of being closed by operable containment automatic isolation valves, were closed by valves, blind flanges, or deactivated automatic valves. 'he operator was knowledgeable of the surveillance procedure, the affected systems, and used appropriate radiological safety precautions.
Electrical maintenance technicians properly conducted the quarterly battery surveillance on emergency DC battery 2B. This surveillance confirmed battery operability by: visual inspection of the battery and the battery room; measuring overall battery and individual cell voltages; measuring cell specific gravities; and cleaning flash arrestors.
The electricians were experienced, knowledgeable of the surveillance procedure, followed appropriate personnel safety precautions, and used properly calibrated test equipmen The inspector independently reviewed the surveillance data and confirmed battery operability.
Instrumentation and controls technicians demonstrated an excellent level ofperformance during the functional testing on a reactor core isolation cooling system high and low vessel level trip device.
Throughout the surveillance, technicians showed a high level of experience, procedural knowledge, and communication skills.
The channel was
'efficiently removed from service, tested, and returned to service, thus minimizing the time that the. channel was inoperable.
The inspector independently reviewed the surveillance data and concluded that the channel was functional.
Operators properly conducted the Division II EDG operability test.
The surveillance
"'emonstrated operability of the EDG, including the air start and fuel oil transfer systems, by a local start and remote loading to 4400kW for one hour.
EDG loading was delayed due to difficulties in obtaining a computer printout of the technical specification required EDG starting parameters (engine speed, and generator frequency and voltage versus time data).
The surveillance procedure initially required verification of the computer data prior to loading the EDG, but was subsequently changed to allow verification of the starting parameters at any time following the EDG start.
The inspector concluded that this action was appropriate.
5.2 mer enc Diesel enerator Pr lem
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ni 2 NMPC took adequate action in accordance with the technical specification to address problems on the Division IIIand Division I emergency diesel generators.
On August 13, NMPC found through surveillance testing that one of the cooling water inlet check valves for the Division III emergency diesel generator did not pass its reverse flow test and declared the EDG inoperable.
The technical specifica'tion for this condition required that the other two EDGs (Division I and II) be tested for operability within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and that the Division III EDG be returned to
'operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or that the high pressure core spray system be declared inoperable.
When the Division I EDG was started for its operability run it did not start.
The
'DG came to approximately 300 RPM and shut itselfdown with an annunciator for high bearing temperatures alarming.
Based on this test failure the Division I EDG was declared inoperable.
This placed the plant in the situation of having only one operable EDG (Division H) and the two off-site power supplies.
On August 14 the inspector observed management preparation for and the actual work and troubleshooting ongoing in the field to correct these two problems.
Management discussions surrounded the fact that the Division II EDG still needed to be tested. to prove its operability within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of declaring the Division IIIEDG inoperable, while work continued to find the cause for the Division I EDG failure to start.
Management had confidence that the Division II EDG would start and would be operable; however, they adequately discussed actions to be taken ifthe machine did not start.
The inspector observed the Division IIEDG testing and found that it was adequately conducted an'd proved the operability of that ED NMPC review of the Division I EDG failure to start centered around troubleshooting of the speed sensing circuit and the high bearing temperature detection system.
Troubleshooting proved that an actual high bearing temperature did not occur, but did not identify any clear cause of the
- failure to start.
This high bearing temperature trip would not have caused the EDG to trip ifit had been called upon to start in its emergency mode (loss of coolant accident or loss of off-site power), since the trip would be bypassed under those conditions.
NMPC started the Division I EDG several times successfully and then performed the normal operability surveillance to declare the EDG operable.
Although the troubleshooting did not identify the deficiency, the plan was well developed and appropriate.
The most probable cause was a problem in the control air supply to the bearing high temperature detection system.
NMPC considered this a valid failure to start and increased the frequency of testing the Division I EDG.
The DivisionIIIEDG work included disassembly and cleaning of the check valve. The inspector observed that the mechanics were knowledgeable and completed the work properly. Following completion of this work the operations department surveillance test conducted as a retest proved that the work corrected the leakage problem. However, when the other divisional cooling water supply check valve was tested, it too failed its reverse flow check. At this point NMPC had not declared the Division IIIEDG operable so work was begun to correct this condition.
NMPC completed the work and successfully performed retest on this second check valve on August 15, and declared the EDG operable.
The check valve failure mechanism was determined to be normal leakage from grit and debris on the seat.
6.0 SECURITY AND SAFEGUARDS (71707)
6.1 Pr tected Area han e
On July 8, NMPC met with the NRC regional office staff to discuss security plan changes developed for the construction of a new building within the plant protected area (PA).
The changes included the reconfiguration of the protected area barrier, intrusion detection systems and assessment aids. Additionally, NMPC discussed the tentative schedule for completion of the project.
Based on the presentation, the staff concluded that the proposed plan changes were well planned and that the changes would not decrease the effectiveness of the security program, 7.0 ENGINEERING AND TECHNICALSUPPORT (71707, 92703, 37700)
7.1 gni~l 7.1.1 ED oolin W er P m T in NMPC took adequate actions to ensure that the emergency diesel generators would be supplied with sufficient cooling water flow. The inspector did identify a concern with the lack oftest data concerning the operation ofboth EDGs and their cooling pumps at the same time. In December 1991, NMPC found that one of the EDG cooling water pumps performance was degraded.
This was documented on a DER and NMPC replaced the pump.
Following replacement, the flow
from this pump was limited by system resistance.
Atthat time, the flowfrom each of the pumps was less than that required (296 gpm) to cool the EDGs up to the lake water temperature of 81'F.
Based on this, NMPC established a lower lake temperature limit for EDG operability.
NMPC reviewed the reasons for the lower flow and determined that system resistance was high possibly due to a flowblockage in the common discharge line after the water passes through the diesel coolers. This common line then flows into the large service water discharge header in the turbine building.
The inspector reviewed the calculation used to determine the flowrequirements.
This calculation showed that a flow rate of 296 gpm was necessary when the EDG was loaded to its seven-day rating (4100 BHP). The inspector found this to be a very conservative calculation, since based on the actual worst case LOCA load study, discussed below, the maximum diesel load would be lower than the seven-day rating.
The heat removal rate was directly proportional to the brake horsepower generated by the engine.
NMPC adequately addressed the concerns of the pumps providing adequate cooling water flow individually. This was done by an engineering evaluation of the back pressure on the cooling water system due to its discharge into a normally pressurized non-safety related service water-discharge line.
NMPC installed a pressure gauge at the connection point and determined that the back pressure was about 18 psig.
Then it was assu'med that the back pressure would not be present when the EDGs would be called upon to supply power in an emergency since the normal service water pumps would not be receiving power.
Using this assumption, NMPC lowered the system resistance curve by the 18 psig and determined that the point on the system curve corresponding to a flow rate of 296 gpm was below the 10% degraded inservice testing (IST)
pump curve.
Thus, ifthe pump is operating at, or above (i.e., better than), the 10% degraded condition, flow through the pump should be greater than 296 gpm (without service water flow).
Inspector review of the normal testing conducted on these cooling water pumps suggested that NMPC did not test them in the most limitingcondition. Specifically, in a loss ofoff-site power situation both EDGs would start and power their busses.
This would require that each cooling water pump supply its EDG. However, the piping downstream of the EDG coolers is common to each system.
This means that the system resistance for the two pump configuration willbe higher than the resistance for a single pump run.
As such, each pump would produce lower flows than in the single pump configuration.
This was a concern because the pump with the lower head may pump at a lower flow rate than the design value. NMPC engineering suggested that testing be performed in the single and two'pump configuration, with and without service water flow, but considered that the systems are currently operable. 'MPC planned to conduct this testing during the 1993 refueling outage.
The inspec'tor considered this issue unresolved (50-220/92-16-02) pending further review of NMPC's engineering disposition of this issue.
7.1.2 n
nre lv I m22 2-12-1'MP R
n e oNR B
in This unresolved item dealt with the adequacy of the NMPC response to NRC Bulletin 88-08 and its three supplements concerning thermal cycling fatigue in unisolable sections of piping.
The
inspector performed a review of the information used by NMPC to submit their December 16, 1991, response letter, to decide ifNMPC should have been aware of the thermal cycling in the emergency cooling system.
Temperature data on the emergency cooling loop 12 condensate return isolation valve, 39-06, was obtained during the week of December 9, 1991.
The data showed evidence ofthermal cycling ofapproximately 35'F over six-hour time intervals. NMPC reviewed this temperature cycling data and provided itto a contractor who was preparing a study on the fatigue phenomena.
Valve 39-06 was isolable from the reactor coolant pressure boundary and was located outside the drywell.
NMPC concluded, based on the completed contractor reviews, in their December 16, 1991 response to the bulletin, that the unisolable sections of piping were not subject to temperature distributions that would cause unacceptable thermal stresses during normal plant operation,
. This item remains open pending NRR review of the NMPC decision not to install thermocouples in the unisolable sections of emergency cooling piping, under Action 3 of the bulletin, prior to the identification of the cracking.
7.1.3 n Vi I i n 1- -1Ind at T
in fEmerenc Di
ne r
nd nre lv I
m 1-
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Eff t fEmr n
Bi Vl e nA
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m ir i
Following the electrical design system functional inspection (EDSFI), NMPC committed to completing two tests of the emergency diesel generator electrical system during a shutdown after July 31, 1992.,
These tests involved the verification of the pump automatic timing circuit at under-voltage and under-frequency conditions which would be present during an EDG start initiated by a LOCA.
The other test was a load test to prove that the EDGs can carry the necessary design load at their rated power factor.
Following the August 7, 1992, unit scram, NMPC decided that they did not have sufficient time before restart to conduct these tests.
Both tests can only be completed with the unit off-lineand would take about 5 days to complete.
The test procedures have been prepared, and NMPC willconduct the testing before the 1993 refueling outage ifa shutdown of sufficient length occurs, such that the testing is not the critical path to restart.
7.1.4 n Unr Ived I em 1- 0-14 ED d
lculation The inspector reviewed the EDG load study calculations. completed following the EDSFI follow-up. These calculations were generally well prepared and contained adequate information and assessment to ensure that the load calculations were correct.
However, when reviewing the calculations, NMPC's response to the EDSFI, and the update to the USAR, the inspector found that a safety evaluation had not been conducted to ensure that the calculations were appropriate.
Further, because of this, the USAR was not updated to reflect the new information.
The inspector discussed this with the corporate engineering supervisor who stated that the safety evaluation should have been completed, but was not done because the engineer involved knew that changes were required to the calculations.
Safety evaluation 91-33 was performed prior to the calculation being completed, and gave general information as to the method and design basis assumptions used in the calculation, but did not. reflect the actual load profiles and results of the later calculation.
NMPC plans to have the safety evaluation of the load study completed by August 31, 1992.
The inspector left this issue unresolved pending review of the completed
safety evaluation.
7:2
@nit 2 7.2.1 mer enc Diesel Gen ra r Tin m
r 1 FR P 21 R view The inspector found that NMPC had taken adequate actions to review the reportability of the tin smearing on the Unit 2 Division I and II emergency diesel generator cylinder liners under 10 CFR Part 21.
The DER process functioned properly to identify that a Part 21 evaluation was needed.
The final Part 21 review was completed on July 15, 1992, which identified that this phenomena was reportable under 10 CFR 21; however, because the NRC was previously made aware of the issue by other utilities and the, diesel manufacturers, NMPC appropriately determined that a separate report was not necessary.
7.2.2 n
inmen V n n
P r Vlv Pr l
During routine containment vent and purge valve testing, NMPC determined that there were deficiencies associated with the suppression pool vent resilient seat valves.
Specifically, during the leak rate testing, the valves failed to meet the acceptance criteria. NMPC took appropriate actions on two instances to enter the applicable technical specification action statements for these valves being inoperable.
In both cases, NMPC was able to get the valves to meet their required leak rate before the LCO was exceeded.
On August 5, NMPC committed to NRC management to keeping the valves closed following verification that they passed their leak rate test, because the valve could not be relied upon to close and meet their leakage requirement (ifopen) during a loss of coolant accident.
Further, NMPC committed to NRC management to test the other valves in the containment purge and vent system after use to verify containment integrity. The inspector observed portions ofthe leak rate testing conducted on August 10 on the drywell containment vent valves, and found that it was properly conducted.
The valves had been used to lower containment pressures on August 9.
The root cause for the valve difficulties was under review.
7.3 T
rm -
i eal
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ni n
NMPC took adequate actions to address potential concerns involving the operability of small conduit fire seals made of thermo-lag 330 material.
The potential concerns were addressed by NRC Bulletin 92-01, dated June 26, 1992. The inspector discussed the installation of these seals with the fire protection engineer and walked down the seal areas.
NMPC identified where this type of material was installed and took appropriate compensatory measures.
This included verifying the operability of fire detection systems and the establishment of roving fire watches.
NMPC also identified that there was another type of thermo-lag material installation at both units, In these installations, sheets of the material were cut and installed in a box fashion to protect duct work, valves, and other components.
This type of installation has not been the subject of failure. NMPC submitted their response to the bulletin on July 24, 1992, stating that
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they were pursuing resolution ofthe problems by coordinating with NUMARC. By letters dated August 4, 1992, the NRC staff found that NMPC's implementation of actions to address this issue were acceptable.
7.4 Tem ra In ction 251 112 - Li n
Ev luati n f han Ar un Licen Reac or Faciliti the Envir n This temporary instruction was issued to allow the inspector to determine if the licensee's programs were adequate in evaluating public health and safety issues resulting from changes in population distribution, or from military, industrial, and transportation hazards that could arise on or near reactor sites.
The inspector found that the updated safety analysis reports (USARs) for both units contained information commensurate. with that required. However, the'process used gather the information discussed above is not clearly identified.
The inspector found that there were many ways that this type ofinformation can be obtained.
The emergency planning organization routinely tracks the population density in the emergency planning zone (EPZ).
Further, NMPC and the New York Power Authority recently completed a study to determine evacuation travel time estimates for the EPZ. This plan was reviewed and found to include estimates for bad weather conditions and special events which occur during the year.
This information was based on the new 1990 census information.
As a member of the local emergency preparedness committee, the on-site environmental coordinator was aware of new industries or new chemicals introduced into the EPZ. The local emergency preparedness committee receives this type ofinformation from local industries as part of existing protocols.
Overall, the inspector found that NMPC was effective at evaluating the effect of population density changes and hazards near the site.
7.5 Fir B rri rP netrati n V ntil i nD m r -
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The inspector reviewed firebarrier penetration ventilation damper installation at both units. Unit specific vendor drawings require a thermal expansion clearance of about 0.125 inches per linear foot of damper between the fire barrier wall opening and the fire damper ducting. If this clearance is not properly incorporated during initial installation and/or not maintained after installation, it is possible that a fire of sufficient intensity could cause the dampers not to close, due to thermal expansion, before the fusible link separates.
This could potentially render a fire damper inoperable and degrade its associated fire barrier.
Both units have periodic procedures to perform visual inspections and prove operation of the dampers.
However, after damper installation, it is not possible to verify thermal expansion clearances without removal of the permanently installed retaining angle which secures the damper sleeve to the penetration, and establishes the thermal expansion clearance.
As part of the 1990 fire barrier review at Unit 1, all of the fire dampers were disassembled, inspected, and reinstalled.
No deficiencies with thermal expansion clearances were identified. At Unit 2, following initial installation, several dampers were disassembled for various reasons and in no instance did-an as-found inspection
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identify any thermal expansion clearance discrepancies.
Based on this information, the inspector concluded that the thermal expansion clearances were acceptable, and the fire barrier penetration ventilation dampers would perform their intended function.
8.0 SAFETY ASSESSMENT AND QUALITYVERIFICATION(71707, 92700)
8.1 vi w f Licen Event Re ER i 1Re 8. 1.1 iinit2 The inspector found the information provided in the following LERs satisfactory:
Special Report, dated June 23, 1992.
Reactor building ventilation'gaseous effluent monitoring system inoperable for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> due to a faulty flow transmitter circuit card.
LER 92-09, dated June 1, 1992.
Reactor building ventilation system isolation caused by low air flow signals due to a momentary interruption in the flow switch power supply.
LER 92-10, dated May 21, 1992.
Secondary containment isolation caused by personnel error during electrical. bus restoration following maintenance.
See IR 50-410/92-13 for further
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LER 92-11, supplement 01, dated June 12, 1992.
Reactor building ventilation system isolation due to a poor electrical connection withina radiation monitor microcomputer.
See IR 50-410/92-13 for further discussion.
LER 92-16, dated August, 1992.
Reactor building ventilation system isolation caused by a false high radiation trip signal due to equipment failure.
8.2 R vi w f ti n Meein The inspector attended several station morning meetings, which were an initiative of the new Vice President-Nuclear Generation.
These meetings were attended by the plant and department managers from both sites and provided a good overall status ofboth plants.
These meetings were seen as a good initiative at coordinating and understanding the different activities at each unit.
8.3
Vi 1 tions220/ 2-2-
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r in Fir e Pr sure I u
fthe Itim Heat ink The inspector found acceptable the NMPC response, dated June 22, 1992, to the Notice of Violations and Civil Penalties dated May 21, 1992.
In both cases, NMPC agreed with the violations. The NMPC identified root cause ofthe events was consistent with the NRC identified root caus ~
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The loss of the ultimate heat sink event occurred on February 21, 1992.
A subsequent augmented inspection team identified a violation with multiple examples offailure to followwork control procedures.
The root cause was determined to be ineffective management oversight and supervisory control over the implementation of procedures which govern the work control program.
A follow-up NRC team inspection inspection Report 50-220/92-81),
conducted between February 29 to March 4 and April 10 to April 18, 1992, concluded that the immediate and short term corrective actions taken to address the causes of the event were acceptable.
Long term corrective actions, as specified in the response to the violation, included:
replacement of the operations department general supervisor, implementation of a work control monitoring program,-"Back to Basics Training" for plant workers, revisions to the work control procedures, and assist visits from an industry sponsored audit group. In general, the inspectors observed that the quality of work packages and procedural adherence has improved.
Plant operation with less than the minimum required number of operable instrument channels, per trip system, for the turbine stop valve closure scram function and the generator load rejection scram function, occurred between December 9, 1991 and January 22, 1992. NMPC found that there were two root causes of this event.
There was inadequate understanding by operations personnel of these reactor protection system features, and there was a lack of configuration control of an instrument root valve position. Corrective actions taken, and in progress, include:
enhancement of the annunciator response procedures related to bypassing RPS functions, verification ofRPS and TS root valve positions, operator training on enabling and bypassing RPS circuitry, simplification of the DER process, selection of a new operations general supervisor, and enhanced operator re-qualification training in this area.
Overall, these corrective actions were assessed as effective in the short term.
The longer term actions were assessed as being appropriate to prevent recurrence.
Based on these assessments, these violations are considered closed.
9.0 MANAGEMENTMEETINGS At periodic intervals and at the conclusion of the inspection, meetings were held with senior station management to discuss the scope and findings of this inspection.
Based on the NRC Region I review ofthis report, and discussions held with Niagara Mohawk representatives, it was determined that this report does not contain safeguards or proprietary informatio ~
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