IR 05000400/1996005
| ML18012A278 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 06/25/1996 |
| From: | Brady J, Mark Miller, Darrell Roberts NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18012A277 | List: |
| References | |
| 50-400-96-05, 50-400-96-5, NUDOCS 9607020329 | |
| Download: ML18012A278 (17) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION II
Docket No:
License No:
50-400 NPF-63 Report No:
50-400/96-05 Licensee:
Carolina Power
& Light (CP8L)
Facility:
Shearon Harris Nuclear Power Plant, Unit
Location:
5413 Shearon Harris Road New Hill, NC 27562 Dates:
April 28 - June 8, 1996 Inspectors:
J.
Brady, Senior Resident Inspector D. Roberts, Resident Inspector M. Miller, Reactor Inspector (Section El. 1)
Approved by:
M. Shymlock, Chief, Projects Branch
Division of Reactor Projects 9607020329 960625 PDR ADOCK 05000400
EXECUTIVE SUHHARY Shearon Harris Nuclear Power Plant, Unit
NRC Inspection Report 50-400/96-05 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support.
The report covers a 6-week period of resident inspection; in addition, it includes the results of a reactive inspection by a regional reactor inspector.
~0er tiees
~
In general, the conduct of operations was professional and safety-conscious (Section 01. 1).
Operator performance during the startup was good (Section 01.3).
The post-trip review did not discuss the greater than expected Reactor Coolant System post-trip cooldown (Section 01.2).
Self-assessment activities were good, although two negative observations were noted in relation to the Plant Nuclear Safety Committee (Section 07.1).
a'e a ce Haintenance activities observed were completed professionally and thoroughly (Section Hl.1).
The surveillance performances observed were adequately conducted.
The licensee took a conservative approach to the resolution of the Reactor Auxiliary Building Emergency Exhaust System surveillance issues.
Oeficiency tags should have been more promptly placed on the associated control room indications that were inaccurate (Section H2. 1).
n ineerin The Engineering staff was implementing a thorough investigation to determine the cause of the post-trip equipment problems.
The problems that were still being investigated were not safety issues that would prohibit plant operations (Section El. 1).
Plant Su ort The general approach to the control of contamination and dose for the site was good.
Teamwork between the various departments was considered a major contributor to overall reduced dose (Section Rl.l).
The security and safeguards activities were performed well (Sections Sl.l).
Fire protection activities wer e acceptable (Section Fl. 1).
e o
eta S
m of Pl nt Stat s
Unit 1 began this inspection period recovering from a reactor trip that occurred on April 25, 1996.
The reactor was critical in Hode 1 at the start of the period, the generator was placed on-line at 12:16 a.m.
on April 28, 1996, and reached 100X'power on April 29, 1996.
Operator reduced power to approximately 50X on Hay 1, 1996, at 10:57 p.m.
due to high unit output breaker disconnect temperatures.
The unit was taken off-line on Hay 3, 1996, however, the reactor remained at low power generator during this evolution.
All unit output breaker disconnects were replaced.
The generator was returned to service at 12: 13 a.m.
on Hay 4, 1996 and the reactor reached 100X. power at 4:35 p.m. that same day.
The unit continued at 100X power through the end of the inspection period.
0 e at o s
Conduct of Operations 01.1 Ge e
l Co ments 1707 Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations.
In general, the conduct of operations was professional and safety-conscious; specific events and noteworthy observations are detailed in the sections below.
222.2 PPPli R2 i
a.
s ect'on Sco e
93702 The inspector reviewed the licensee's Post-Trip Review for the reactor trip that occurred on April 25, 1996 to determine whether all problems that occurred during the trip had been identified.
b.
Observ t'ons nd Findin s
Procedure OHM-004, Post trip/Safeguards Actuation Review, Revision 4, describes the licensee s post-trip review process.
The post-trip review package required by ONH-004 was completed on April 27, 1996.
The root cause of the trip was the failure of the supply side disconnect and subsequent fault to ground for generator output breaker 52-7 which resulted in a turbine trip/reactor trip due to generator lockout.
Several unexpected equipment actuations occurred after the reactor trip.
The "B" train emergency electrical buses did not transfer to the startup transformer as designed which resulted in the "B" train emergency diesel generator starting and providing power to the "B" train emergency buses.
The "A" train diesel was found in the "Operational" mode with the
"Haintenance" mode indicating lights illuminated.
The licensee's investigation and resolution for these actuations and the corrective action for the disconnect failure are discussed in section El. 1 of this
repor t.
Also, the charging pump suction supply auto-transferred from the volume control tank to the refueling water storage tank.
The inspector reviewed the report and found that all problems that occurred during the trip were identified in the post-trip review.
The inspector found that the cause of the charging pump suction swapover to the refueling water storage tank was not fully addressed in the package.
The unexpected automatic swapover was attributed in the post-trip review package to the automatic make up for the volume control tank not having actuated.
Later the licensee determined that the rotometer to the boric acid makeup pump was powered from the "B" train safety bus which was lost.
With no boric acid flow indicated, the control circuit automatically secured makeup.
This automatic function worked as designed to prevent a dilution accident and was not an equipment malfunction as originally thought.
In addition to auto makeup not occurring, the reactor coolant temperature had decreased significantly below the no-load average temperature of 557 'F down to approximately 535 F.
The resulting reduction in coolant density caused pressurizer level to decrease.
When the automatic letdown isolation set point was reached (17X pressurizer level) letdown to the volume control tank was automatically isolated.
With automatic makeup to the volume control tank not occurring, letdown to the volume control tank isolated, and with charging flow high due to low pressurizer level, the volume control tank low level set point was rapidly reached.
The post-trip review did not mention the significant contribution of the cooldown to the charging pump suction swapover.
The inspector found that the cooldown was caused by operators not throttling back on auxiliary feedwater flow promptly.
Discussion with the Operations Hanager also revealed that he was aware of the cooldown and considered operator performance related to prompt throttling of auxiliary feedwater as needing attention.
The inspector also noted that the Plant Nuclear Safety Committee had not discussed the cooldown during their review of the trip prior to plant startup.
The Plant General Hanager/
Plant Nuclear Safety Committee Chairman confirmed that he was aware of the cooldown, but both he and the Operations Hanager had forgotten to bring up the issue at the Plant Nuclear Safety Committee meeting.
The cooldown was addressed in LER 50-400/96-008 which is discussed in section 08. 1.
Conc usions The post-trip review identified the root cause, evaluated the equipment actuations after the trip, and identified corrective actions for equipment deficiencies.
However, the greater-than-expected reactor coolant system cooldown was not addressed in the post-trip review or during the pre-startup Plant Nuclear Safety Committee meetin.3 St rt Observ t'o s
a ~
Ins ection Sco e
71707 b.
The inspector observed reactor startup on April 27, 1996, at the end of the Inspection Report 400/96-04 inspection period, and observed generator synchronization to the grid at the start of this period on April 28, 1996.
Observ t'o s
nd d'
C.
The inspector observed that operators were following procedures and ensuring that equipment worked properly.
The inspector observed that turbine governor control was improved from the previous startup.
Prior to generator synchronization, operators discussed problems that occurred during the last startup when an intermediate range rod block was received after generator synchronization.
Generator synchronization was accomplished with no problems.
The inspector observed that the controllers for loop "B" and
"C" main feedwater bypass control valves would not control in the automatic mode.
An operator was stationed to manually control feed flow using the controller.
The operators had a problem during the previous startup with these controllers and were only able to successfully place the "A" loop controller in automatic.
The inspector noted that the problem was identified in the operator work-around log.
The inspector discussed this problem with operations and engineering department personnel.
Both departments explained that the current tuning procedure was lengthy and required a significant amount of time to perform (6-12 hours).
During the forced outage on Nay 3, 1996, the plant was held at low power for an extended period of time with the unit generator off-line.
The inspector observed that the licensee was tuning the main feedwater bypass valve controllers during that time.
The tuning significantly improved the performance of the valves in automatic as indicated by improved feedwater flow characteristic.
Co c s'o s
02.1 The inspector concluded that operator performance during the startup was good+
Operational Status of Facilities and Equipment i ee d Sa t SF Feat re S ste o
s
The inspectors used Inspection Procedure 71707 to walk down accessible portions of the following ESF systems:
~
Auxiliary Feedwater as described in Final Safety Analysis Report (FSAR) Section 10. Main Steam (safety-related portion)
FSAR Section 10.3 Equipment operability, material condition, and housekeeping were acceptable in all cases.
Equipment and systems were as described in the FSAR.
The inspectors identified no substantive concerns as a result of these walkdowns.
guality Assurance in Operations 07. 1
'ce see Se f-Assessment Acti 't'
4 500 During the inspection period, the inspectors reviewed multiple licensee self-assessment activities, including:
two Plant Nuclear Safety Committee (PNSC) meetings; the May 28 Nuclear Safety Review Committee meeting; Nuclear Assessment Section Audits on Security (HNAS96-115)
and Material Control (HNAS96-124);
b.
Observations and Findin s
The inspector observed two instances where the PNSC did not fully address issues during their meetings.
The first was in relation to the post-trip cooldown discussed in section 01.2 above.
A second was observed in relation to discussion of NRC Violation 96-04-01.
The PNSC did not discuss whether it was acceptable to test reactor protection system equipment using installed test equipment with identified deficiencies.
Other PNSC discussions and Nuclear Safety Review Committee discussions were thorough.
The Nuclear Assessment Section Audits were thorough.
c.
Co c usions The licensee's self-assessment activities were good, although two negative observations were noted in relation to the PNSC.
Miscellaneous Operations Issues (92700)
08.1 0 en L
R 50-400 96-008:
Reactor Trip Due to a Failure of an Output Breaker Disconnect Link.
This LER reported the April 25, 1996 reactor trip which was previously discussed in Inspection Report 50-400/96-04.
Paragraphs 01.2 and El. 1 of this report discuss the post-trip review and some of the corrective actions taken for equipment problems that occurred after the trip.
The LER also discussed the post-startup disconnect problems and subsequent action to take the unit off-line to replace the remaining generator breaker disconnects.
This activity is discussed in section El.l of this report.
In addition to the corrective actions taken, the LER described ten planned corrective actions.
These
included actions to address the cooldown issue discussed in section 01.2 of this report.
This item will remain open until after'he planned corrective actions are sufficiently completed to allow evaluation.
a te a ce Hl Conduct of Naintenance Mi.i ~GC t
a.
s ec io c
e
3 The inspectors observed all or portions of the following work activities:
~
WR/JO ADZP001 Inspection of Limitorque Operator for Refueling Water Storage Tank Suction Valve to Containment Spray (1CT-71).
This implemented preventive maintenance procedure PH I0020, Limitorque Operator Inspection, Revision 8.
~
WR/JO 96-ADRHI Spent Fuel Cask.
This implemented support for procedure CH-H0300, Spent Fuel Cask Handling, Revision 13.
~
WR/JO 96-ALFL002 Lubricate B Motor-driven Auxiliary Feedwater pump coupling.
This implemented PH-HOOll, Equipment Lubrication Schedule, Revision 10.
~
WR/JO 95-AERWl - Lubrication Schedule, Revision 10, Replace Rosemount Transmitter on B HDAFW Discharge Pressure Sensor (PT-01AF-2150BSB) in reference to NRC Bulletin 90-01.
This included loop calibration per procedure LP-P-2150B, Auxiliary Feedwater Pump B Discharge Pressure, Revision 5.
~
WR/JO 96-AEPB001 Loop calibration of Motor-driven Auxiliary Feedwater pump B Suction Pressure Transmitter per procedure LP-P-2250B, Auxiliary Feedwater Pump B Suction Pressure, Revision 4.
b.
Obse v tions and Findin s
The inspectors found the work performed under these activities to be professional and thorough.
All work observed was performed with the work package and procedures present and in active use.
Technicians were experienced and knowledgeable of their assigned tasks.
The inspector observed quality control personnel involved whenever required by procedure.
Independent verification was performed where required.
When applicable, appropriate radiation control, measures were in plac Co sio s
Maintenance activities observed were completed professionally and thoroughly.
Maintenance and Material Condition of Facilities and Equipment Survei l nce Observation s ect'o c
e 6 726 The inspectors observed all or portions of the following surveillance tests:
OST-1004; Power Range Heat Balance Computer Calculation, Daily Interval, Mode
(Above 15X Power); Revision ll OST-1052; RAB Emergency Exhaust System Operability, 18 Month Interval, All Modes; Revision 5/1 Obse t o s
d
The inspector found that the testing for OST-1004 was adequately performed per the test procedure.
OST-1052 tested the ability of the RAB Emergency Exhaust System to achieve a negative pressure of greater than or equal to 1/8" water gauge in the RABEES area.
The test was being run because the charging pump rooms were not required to be tested by the previous revisions of the procedure.
The plant had entered Technical Specification Limiting Condition for Operation 4.0.3 until the testing could be performed.
The licensee found that the initial conditions for the test did not specify whether other ventilation systems other than the normal Reactor Auxiliary Building system should be secured.
The licensee took a conservative approach and secured all ventilation systems that would not be running in a Loss of Coolant Accident situation.
A permanent procedure change was being processed to better define the test setup conditions.
The test for "A" train failed.
During the testing of the "B" train, initial data indicated that the test would also fail, however exhaust flow indication was high while RAB pressure differential indication to the outside atmosphere was slightly below the 1/8" criteria.
Since TS LCO 3.0.3 was applicable, a decision was made to take a local manometer and check the control room indication readings at the transmitter to ensure that entry into TS LCO 3.0.3 and resulting plant shutdown was appropriate.
The control room reading did not agree with the local indication.
Additional local manometer comparisons independent of the transmitter confirmed that the control room indication was inaccurate.
In addition, local pitot tube readings also indicated that control room flow indications were also inaccurate.
The local readings were taken with calibrated test equipment and indicated that the "B" train flow and
differential pressure met the TS requirements.
An engineering evaluation of the data, contained in Engineering Service Request 9600321, concluded that the local data proved operability of the RABEES.
TS LCO 4.0.3 was exited and TS LCO 3.7.7 was entered for one train of RABEES inoperable.
The inspector noted that the control room indication provided only indication and that the RABEES control functions were provided from other transmitters.
The inspector observed that the licensee was slow in placing deficiency tags on the control room indicators even though the ESR identified the indicators as providing inaccurate readings.
The licensee went back to "A" train and used local manometer checks to verify the control room indications.
The control room indications for
"A" train were also inaccurate.
However,
"A" train would not pass with local indications.
Both flow and differential pressure were low.
After maintenance was performed on the "A" train suction vortex damper control circuit and on the "A" train discharge back-draft damper, acceptable readings were obtained by local instrumentation.
The engineering evaluation of the local calibrated test equipment and readings was contained in ESR 9600331.
The inspector observed good management attention and team work between the on-site organizations.
Concl sions The surveillance performances were adequately conducted.
The licensee took a conservative approach to the resolution of the RABEES surveillance issues.
Operation personnel were slow in placing deficiency tags on inaccurate control room indications.
Niscellaneous Naintenance Issues (92700)
e LER 50-400 96-002:
Failure to Properly Perform Technical Specification Surveillance Testing.
This LER was discussed in Inspection Reports 50-400/96-02 and 50-400/96-04 and is a result of the licensee's ongoing Technical Specification Surveillance Review related to Generic Letter 96-01.
Supplement 4 was issued during this inspection period and identified one additional item related to containment fan cooler post accident dampers.
The dampers receive an open signal from the emergency safeguards sequencer.
The licensee's surveillance program had not verified all start paths to open the dampers.
Supplement 5 was also issued which included items on Essential Services Chilled Mater Chillers, Containment Spray Pumps, Computer Room and Control Room dampers.
This LER and its supplements will remain open until the licensee completes the surveillance review program later this year.
e LER 50-400 96-00
Failure to Perform Technical Specification Surveillance Testing in Accordance with Specification 4.7.6.d.3.
This LER was issued to report a problem in how the licensee was testing the control room ventilation system.
This problem was initially discussed
in IR 50-400/96-04 in relation to the conflict between the control room ventilation system and the computer room ventilation system.
The licensee had been testing the control room system in relation to only one adjacent area.
The Technical Specification does not use the word
"all" with adjacent areas, but the licensee determined that it is implied and chose the conservative approach by declaring the surveillance as having been missed.
Corrective actions were to test relative to all adjacent areas and to fail the computer room return damper open to prevent computer room pressurization.
Several corrective actions were not completed including review of the Reactor Auxiliary Building Emergency Exhaust System (RABEES) surveillance since it has similar wording.
The inspector was aware that after the issuance of the LER the licensee's review found similar problems with RABEES (TS 4.7.7)
and the Fuel Handling Building Emergency Exhaust System (TS 4.9. 12).
The test of RABEES was discussed in section H2. 1 above.
This LER will remain open pending review of the resulting overall corrective actions for the ventilation systems.
III. E i eer n
Conduct of Engineering Post r'view I
e t'on Sco e
9370 d 3755 The inspector evaluated the significance of the plant trip on April 25, 1996, including performance of the electrical systems, and the action taken by the Engineering Department to address this event.
These evaluations were performed to verify that the safety concerns were adequately addressed by the licensee.
The inspector reviewed Chapter 8,
Electrical Power, of the FSAR to determine if any discrepancies existed between it and the plant.
The inspector made onsite assessment of the associated electrical equipment that malfunctioned.
Malkdown inspections of the switchyard and the switchgear were performed to observe their material condition.
Activities conducted by the electrical engineering group were observed to determine if appropriate corrective action was being implemented and adequate analysis was being performed.
The inspector reviewed three engineering evaluations concerning electrical equipment that misoperated or failed during the reactor trip event.
In addition, testing of the 6.9kV switchgear breaker by the electrical engineering group was observed.
Observations a
d Findin s The licensee initiated three engineering evaluations,
"Engineering Service Request" (ESR).
These three ESRs discussed components and
systems that failed or misoperated during the trip and engineering's response or resolution.
ES o.
960023
- a's o
me The purpose of this ESR was for engineering to evaluate the failure of generator switchyard breaker 52-7, unit generator side "disconnect switch".
The normal switchyard line up was to have two breakers closed, 52-7 and 52-9, which shared the current between them.
Each breaker has a disconnect switch on each side (unit generator or line side and transmission lines or load side) to allow removing the breaker from service.
Breaker 52-9 was taken out of service for maintenance.
All the current was passed through breaker 52-7 and the two disconnect switches.
The "A" phase line side disconnect switch failed catastrophically due to excessive heat buildup that melted the contacts.
The melting of the switch contacts resulted in a ground fault.
The engineering evaluation for this event was that "the initial cause of the failure of the unit side, phase A, disconnect switch on the 52-7 circuit breaker was due to a high resistance connection.
This high resistance connection'was caused by the blade contacts not fully rotating to the proper horizontal position."
The disconnect switches are manually operated by a hand crank for both opening and closing the switches.
The mis-alignment of the disconnect switch linkages of "A" phase indicated an out-of-adjustment condition.
The engineering evaluation further stated that "the adjustment issue coupled with the possibility that the disconnect crank may not have been fully rotated are the most probable causes of the event."
The inspector walked down the switchyard and examined the replaced disconnect switch to verify appropriate corrective action was implemented.
In addition, the damaged disconnect switch was also examined.
The inspector concluded that the licensee had adequately addressed the failed disconnect switch issue and was in the process of investigation to determine if further corrective action was needed.
The licensee continued to monitor the switchyard disconnect switches for "hot spots" by using thermography instrumentation.
After the startup the disconnect switches were fully loaded and thermography taken.
Hot spots were observed on the other switches.
Power was reduced to approximately 50X and the two output breakers were opened one at a time to permit cleaning of the disconnects.
The inspector observed that the cleaning eliminated all but one hot spot.
The licensee decided to take the unit off-line and replace the remaining three generator output breaker disconnect switches.
After bringing the unit back on-line and during the power ascension, the disconnect temperatures were again monitored using thermography.
Thermography indicated that the problems were resolve S o.
960233 Diesel Gene at r tern The purpose of this ESR was for engineering to provide an explanation for the emergency diesel generator (EDG)
1A-SA alarms that annunciated when the plant trip occurred.
When the transient occurred, an operator was at the EDG 1A-SA engine control panel.
He observed that the lighting went out, the horn sounded, and the "shutdown lockout relay" (86DG) tripped.
EDG lA-SA was not required to start since a "fast'us" transfer from the auxiliary transformer to the startup transformer took place as designed.
The engineering investigation determined that the most likely explanation was that a power line voltage transient caused a very sensitive 86DG relay to trip.
The 86DG relay as discussed in Electroswitch Vendor Technical Publication LOR-1 has a Type E coil.
The Type E coil has a threshold voltage of 23VDC.
LOR-1 also stated that these relays are sensitive to stray voltages causing nuisance trips.
The inspector conducted a walkdown inspection of the EDG 1A-SA with the system engineer to examine the 86DG relay and its associated wiring.
The electrical drawings were reviewed with the system engineer.
There were no concern in this area.
The inspector concluded that the ESR theory that a stray transient voltage was most likely to have caused the sensitive 86DG relay to trip was reasonable.
The inspector verified by review of the electrical schematics that the tripped relay would not have prevented EDG lA-SA from starting had a safety injection or undervoltage signal been received.
ESR No.
9600240 - 6.9kV AC Distribution S ste The purpose of this ESR was for engineering to determine why the 1B-SB undervoltage scheme activated and started the EDG 1B-SB during the plant trip..
When the switchyard breaker 52-7 tripped, it caused the generator lockout relay to trip sending a signal to trip the reactor.
During reactor power operation, electrical power for the 6.9kV system is fed from the two 22kV/6.9 kV unit auxiliary transformers (UAT) "lA" and
"1B" which are connected to the 22kV generator.
During plant outages, electrical power for the 6.9kV system is fed from the two 230kV/6.9kV startup transformers (SUT) "1A" and
"1B" which are connected to the 230kV switchyard.
When the turbine/reactor tripped, the source of power for the 6.9kV system needed to switch from the auxiliary transformers to the startup transformer in less than one second.
This transfer of power is called a "fast bus transfer".
The "fast bus transfer" signal from the generator lockout relay was sent simultaneously to both "1A" and
"1B" transformer breakers.
The fast bus transfer for the "lA" transformers breakers (101, 102, 107, and 108)
was confirmed since there was no loss of power on the 6.9kV "A" train buses.
Therefore, there was no undervoltage activation identified
and the lA-SA emergency diesel generator (EDG) was not required to start.
The fast bus transfer occurred for only half of the "B" train transformer breakers.
The "1B" auxiliary transformer breaker 128 and its associated
"1B" startup transformer breaker 127 that power auxiliary bus 1B made a successful fast transfer as evidenced by the fact that the associated reactor coolant pump continued to run.
However,
"1B" auxiliary transformer breaker 122 and "1B" startup transformer breaker 121 that powered Bus 1E did not make a successful fast transfer prior to the undervoltage relays tripping for both busses 1E and 1B-SB, which caused EDG 1B-SB to start.
Bus 1E provides power to the safety-related bus 1B-SB on the "B" train side of the 6.9kV system.
Breakers 121 and 122 were found to be in their required position which indicated that they had operated.
Engineering had not determined why there was an undervoltage actuation of the lE bus.
Corporate engineering was investigating this problem through analysis using a computer model.
On-site engineering was investigating breaker 121 through inspection and testing.
One of the main problems in determining the undervoltage activation was to determine the timing sequence of the breakers.
Unfortunately, the computer input card for timing signals from breakers 121 and 122 was defective due to a blown fuse.
Therefore, the time recording of the breaker sequence of operation was lost. If this timing data were available the problem would be more easily resolved.
On April 28, 1996, after the plant started up with the 6.9kV system receiving power from the startup transformers, breaker 122 failed to operate.
Breaker 122 was removed from the switchgear cubicle and several breaker parts were found on the cubicle floor.
Inspection of the breaker revealed a spring clip was missing from the shaft of the four bar linkage.
This allowed a roller and a bearing to fall from the shaft.
In addition, a support bar to the "A" phase switch blade was not properly connected.
Engineering and maintenance ran a sequence of tests with the breaker fully restored and then with the parts removed to determine their effect on breaker operation.
Regardless of the breaker condition (parts removed) during the testing, the breaker successfully operated within its timing specification every time.
The testing observed by the inspector was adequate to determine breaker timing and mechanical operation.
The inspection and testing did not provide engineering with data to resolve the undervoltage trip activation.
The resolution of these issues will be tracked through the corrective action review of LER 50-400/96-008.
No discrepancies were observed with Chapter 8 of the FSAR related to these issues.
Conclusion The Engineering staff was implementing a thorough investigation to determine the cause of the post-trip equipment problems.
The problems
E7 E7.1
that were still being investigated were not safety issues that would prohibit plant operations.
EQUALITY ASSURANCE IN ENGINEERING ACTIVITIES S ECIAL FSAR R VI W
Fl A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the FSAR descriptions.
While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the FSAR that related to the areas inspected.
The inspectors verified that the FSAR wording was consistent with the observed plant practices, procedures and/or parameters.
Control of Fire'rotection Activities Fl.l General Comments 71750 R1 Rl.l The inspector observed fire protection equipment and activities during the conduct of tours and observation of maintenance activities and found them to be acceptable.
The inspector observed the conduct of a fire drill from the main control room.
A smoke generator was used to provide a sense of reality for the fire team.
Communication between the control room and the fire team was acceptable.
V.
la t Su o t Radiological Protection and Chemistry (RPSC) Controls General Comments The inspector observed radiological controls during the conduct of tours and observation of maintenance activities and found them to be acceptable.
The inspector attended an As Low As Reasonably Achievable committee meeting and found the committee closely tracking dose for all site organizations.
The general approach to the control of contamination and dose for the site was good.
Teamwork between the various departments was considered
"a major contributor to the overall reduced dose.
S1 Sl.l Conduct of Security and Safeguards Activities Ge eral Comme ts 7 750 The inspector observed security and safeguards activities during the conduct of tours and observation of maintenance activities and found them to be good.
Compensatory measures were posted when necessary and properly conducte aaemet eet s
Xl Exit Neeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on June 11, 1996.
An interim exit was conducted on Hay 2, 1996.
The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any material examined during the inspection should be considered proprietary.
No proprietary information was identified.
X2 Nanagement Meeting to Discuss NRC SALP On May 3, 1996, a management meeting was conducted to discuss the NRC's Systematic Assessment of Licensee Performance (SALP) for the Shearon Harris facility.
A summary and list of attendees at that meeting was contained in a letter to the licensee, dated Hay 14, 199 PARTIAL LIST OF PERSONS CONTACTED
~egsee D. Alexander, Supervisor, Licensing and Regulatory Programs D. Batton, Superintendent, On-Line Scheduling D. Braund, Superintendent, Security A. Cockerill, Superintendent, I&C Electrical Systems J. Collins, Hanager, Training J.
Dobbs, Hanager, Outage and Scheduling J.
Donahue, General Hanager, Harris Plant W. Gautier, Hanager, Haintenance M. -Gurganious, Superintendent, Chemistry H. Hamby, Supervisor, Regulatory Compliance H. Hill, Hanager, Nuclear Assessment D. HcCarthy, Superintendent, Outage Hanagement K. Neuschaefer, Acting Hanager, Environmental and Radiation Control W. Peevyhouse, Acting Superintendent, Hechanical Systems W. Robinson, Vice President, Harris Plant G. Rolfson, Hanager, Harris Engineering Support Services S. Sewell, Superintendent, Design Control T. Malt, Hanager, Performance Evaluation and Regulatory Affairs A. Williams, Manager, Operations NRC T. Le, Harris Project Manager, NRR H. Shymlock, Chief,,Reactor Projects Branch
IP 37551:
IP 40500:
IP 61726:
IP 62703:
IP 71707:
IP 71750:
IP 92700:
IP 93702:
INSPECTION PROCEDURES USED Onsite Engineering Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Surveillance Observations Maintenance Observation Plant Operations Plant Support Activities Onsite Followup of Events Onsite Response to Events
~0e ed None Closed None Discussed 50-400/96002 50-.400/96007 50-400/96008 ITEMS OPENED, CLOSED, AND DISCUSSED LER Failure to properly perform technical specification surveillance testing LER Failure to perform technical specification surveillance testing in accordance with specification 4.7.6.d.3 LER Reactor trip due to a failure of an output breaker disconnect link