IR 05000458/1990034
| ML20029A306 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 01/29/1991 |
| From: | Harrell P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20029A305 | List: |
| References | |
| 50-458-90-34, NUDOCS 9102130101 | |
| Download: ML20029A306 (17) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION REGION IV-'
i NRC Inspection Report:
50-458/90-34 Operating License:
NPF-47:
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Docket: 50-458
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Licensee: GulfStatesUtilitiesCompany(GSU)
P.O. Box 220 4'
St. Francisville, Louisianc 70775 Facility Name:
River Bend Station (RBS)
Inspection At: RBS, St. Francisville, Louisiana inspection Conducted: November 28, 1990, through January _15,1991 Inspectors:
'E. J. Ford, Senior Resident Inspector D. P. Loveless, Resident. Inspector Approved:
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A,wL l-29-4l P
HerrQlyChief, Project Seltion C-Date Inspection Sumary Inspection Conducted November 28c 1990, through-January-8,- 1991-f (Report 50-458/90-34)
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Routine, unannounced inspection -of onsite followup of-even'ts,-
Areas Inspected:
operational safety verification, maintenance and surveillance observ;ations, and
licensee event report followup.
E Results:
E On November 30, 1990, during plant startup, the TS maximum heatup rate of 1 100*F per hour was exceeded due to an operator error. ~ However, while the'
. operator played a principal role in the event.the inspectors noted that; weak control room comunications and less then desirable instrumentation appeared to also have been contributing factors. An'open item o
(458/9034-01) mas issued pending completion of-the licensee's evaluation-.
of the observations made by the. inspectors (paragraph 3.a).-
On January 4, 1991, the licensee declare'd that'the ADS system may have-
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been inoperable for approximately 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> (which is' greater than the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed by the TS) when both SVV compressors were.out of service for unscheduled maintenance.
The details of' this issue are discussed in-NRC Inspection Report 50-458/91-04, issued' on January 17, 1991, as a special -report to document' this issue.(paragraph 3,b). -
9102130101 910201 DR ADOCK 050
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Two~temporeyy waivers'of compliance to the TS were issued. One involvedi
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the RCIC system and the other-the drywell air lock..The-bases for the waivers were well developed and presented by the licenseec(paragraphs 4.f and4.g).
The performance of maintenanceLand surveillance activitics appeared to be'
adequate _(paragraphs' 5: and 6).
On December:3 and-4,1990, the licensee conducted !a controlled, methodical-
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. increase in; plant power. LConsiderable' management, operator, 'and I
engineering resources were on. hand for the power increase and subsequent successful testing-off the' ADS /SRVs:(paragraph 6.a).
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Note:
Acronyms and initialisms used in this raport are identified'in
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an alphabetical listing in the attachment at.the end of this-
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inspection report.
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DETAILS 1.
Persons Contacted W.1. Beck, Supervisor, Balance of Plant Design
- E. M. Cergill, Director, Radiological Programs
- J. W. Cook, Technical Assistant
- T. C. G ou e, Manager, Administration
- W. L. Curren, Cajun Site _ Representative-J. C. Ueddens.-Senior Vice President,' River Bend Nuclear Group
- P. D.' Graham,. Plant Manager.
- J. R. Hamilton, Director, _ Design Engineering
- G. K. Henry, Director, Quality Assurance Operations
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- D. E. Jernigan, General Maintenance Supervisor-
- D. N. Lorfing, Supervisor, Nuclear Licensing J. F. Mead,: Supervisor. Electrical Design
- L. W.- Rougeux, Senior ISEG Engineer
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- J. P. Schippert, Assistant Plant Manager;.0perations Radwaste, and
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Chemistry
- J. E. Spivey, Senior Quality Assurance: Engineer s
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- K. E. Suhrke, General Manager, Engineering and Administration
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l S. L. Woody, Supervisor,- Nuclear _ Security '
l In addition to the above_ personnel, the inspectors contactediother J
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personnel during this inspection period.1
- Denotes attendance at the exit interview conducted on January 15, 1991, to discuss the overall results of. this inspection.=
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Plant Status l.
At the_ beginning of this inspection. period, the reactor was in cold:
shutdown (Mode _4) with the-new core loaded and preparations'in progress to;
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restart the~ unit..
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The licensee began the refueling outage. on-September 29, 1990. The outage l
was scheduled for 58 days and-lasted.66 ' days. Major ' outage:. activities
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included fuel ' shuffle, DG inspections, Division 11' electrical board work,.
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I high pressure turbine inspection, control rod drive replacement,
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safety / relief valve replacements,1MSlY test / repair / retest ;and suppression:
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pool cleanup.
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j On November 30,.19% the reactor was taken critical.
However,'powerf
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escalation and the end of-the: outage..were delayed because the RCIC system failed to pass surveille.nce testing. The RCIC turbine was tripping on"an-
l overspeed condition caused by mechanical. binding of the governor-and an '
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-FRis limit swttch.
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-4-On December 3.-1990, the NRC granted the licensee a temporary waiver of
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compliance to allow the unit-to enter Mode. I with the RCIC turbine inoperable. The main generator was. synchronized to the grid:on -
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The plant experienced a scram from 80 percent power, on December 12, 1990, i
during main turbine valve testing.
The combined intermediate valves were.
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heing tested when an RPS-actuation signal was generated-on low EHC system"
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pressure. The licensee conducted troubleshooting and: repair-ectivities-on
the system that included the installation-of orifices to dampen pressure S
surges._ The licensee'successfully: performed postmaintenance testing _of-
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the EHC system prior to returning to power.
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On December 16, 1990, theunitwastiedtothegrdfollowingcriticaliO*
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on December 15. At the end of this inspection period, the reactor was
-operating 1 at 100 percent power.
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3.
Onsitr. followup of Events -(93702)
a.
Heatup Rate Exceeded
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On ' November 30,M90, during a plant startup, the licensee exceeded!
- the TS-specified heatup rate ofn100*F _ per hour. The heatup of the reactor, durin a.1-hour period, was ' calculated by the licensee.ta be
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117'F, as shown by the data recorded in' STP-050-0700,'"RCS Pressure /
Temperature Limits Verification."
l TS 3.4.6.1.A-requires"that the reactor temperature be liinited to a maximum beatup of 100'F in any 1-hour period. The associated action-statement requires the licensee, with this limit exceeded, to restore.
the ternperature to within the limits within 30 minutes, perform an
engineering evaluation to determine the effects Lof. the 'out-of-limit l
condition on the structural integrity of-the RCS, and determine that L
the RCS remains acceptable for continued operation. 'These-requirements were met by.the licensee..
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Throughout the startup, the licensee had three' licensed t bators at o
l the' panels.- The ATC operator was; performing the startup,% heatup.
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l An SRO was -assisting in controlling reactor water level and the C0F q
was supervising these indidduals. At one point during the startup,
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the C0F left the control-runfor a period of. approximately 10 minutes.
'Before_ leaving, he assessed plant status and determined that the plant was stable and told the ATC operator'to maintain the plant-
where it;was. LDuring interviews with the individu'als, the C0F told i
the inspectors that h'e had meant that the ATC operator should not-pull any more control' rods. -,However, the ATC_ operator believed that -
the C0F had told him to' keep the IRMs steady whereithey were, which may have required rod pulls.-
Following departure of the C0F, reactor water level fluctuations.
distracted the ATC operator, who' apparently began to use. control rods
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i to steady the fluctuations. During this evolution, three rods were pulled from Notch Position 12 to Position 48 in approximately
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l 5 minutes. Since previous rod pulls were one step at a time with a 10- to 15-minute wait between each pull, the operator should have
realized that these pulls would increase the heatup rate significantly. However, the ATC operator relied on monitoring of the
reactor temperature, performed every 30 minutes by the STA, to keep
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the heatup within the requirements. During interviews, the
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inspectors noted that the STA had notified the ATC operator that the heatup rate may have been exceeded; however, it did not appear that (
the operator understood the communic. ion.
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Shortly after return of the C0F to the control room, the STA noted that the heatup rate was excessively high.- Immediate corrective action was taken by reinserting the three control rods in the reverse i
sequence. This action stopped the heatup and w p completed within i
the 30 minutes, as required by TS. Duringthisperiod,thereactor temperature had increased frcm 213 to 332 F and the' vessel presst.re increased from 0 to 88 psig. -Shortly after the event, the plant
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manager had the ATC operator relieved from' licensed duties pending a
complete evaluation of the event.
The licensee performed an interim review of the impact on the RCS and determined that the reactor vessel was satisfactory for power operations pending a-formal-analysis to be performed by GE. This determination was bcsed on the reactor not being close to the limits dc the pressure / temperature curves in the TS, and that the RCS pressure remained less than 10 percent of the normal operating i
pressure throughout the event. These circumstances eliminated
brittle ftacture concerns and stress and fatigue impacts on the t
vessel according to the licensee's evalu'ation.
In'a' letter, dated November 30, 1990, GE stated that, since stress, fatigue, and brittle
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fracture impact of the heatup event were acceptable, continued operation was,iustified.
j During followup evaluation, the' licensee investigation team recommended that the following actions be taken:
Remove the ATC operator from licensed duties until further notice.
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Revise STP-050-0700, "RCS Pressure / Temperature Limits Verification," to incorporatt. the following:
Heatup/cooldown rates will be monitored and recorded every
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15 minutes. The rate was recorded every 30 minutes during the event.
Heatup/cooldown rates will be reviewed by the SR0s involved
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in the evolution promptly following recording of the data.
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Heatup/cooldown rates will be administratively limited to j
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80'F per hour.
This limit was previously 90'F per hour j
but was not prxedurally delineated.
- Provide a briefing on the incident to each crew prior to assuming duties on their next shift.
- Have the plant manager hol:
'ariefing with each crew to stressi significance of this event and safe plant operations.
Add a graphic display of the heatup/cooldown rate-to the process
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computer display monitor.
Train the STAS on this event and the changes 33 STP-050-0700._
The licensee has completed all the items reconmoded by the investigative team, except for the installatio of a graphic display of the heatup/cooldown rate.
The licensee stated that the root cause of the event was operator error.
It was determined by the licensee that sufficient licensed personnel were involved with the startup evolution. but that the ATC;
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operator was not focused on-controlling = the reactor _ vessel _ heatup rate.
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' Based on interviews nerformed by the' inspectors with the shift.
l personnelLinvolved, the inspectors were also concerned with an apparent weakness in control room conimunications between the ATC operator, STA, and C0F, Additionally, the-less than desirable heatu).
rate. tracking instrumentation was a possible contributing lcause of tie
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event.
At the end of this inspection period, the licensee was in;the process
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of evaluating the observatiops made by_ the inspectors, as discussed above. These. issues are cons 1dered open pending review of the licensee's evaluation of the observations (458/9034-01),
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Apparent ADS Inoperability-On January 4,1991, the licensee informed the inspectc" that the ADS
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l appeared to be inoperable for a period in excess of' the action time required uy TS 3.5.1;e.2. This TS. states, with two or more of-the requind-ADS /SRVs inoperable, be in ~at least hot shutdown'within
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12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. A detailed follovup of this issue was performed during this inspection period and-is docerxnted in NRC; Inspection Report-50-458/91-04, issued on January 17, 1991.
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4.
Operational Safety Verification (71703~
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Routine Plant Observations The ir.spectors observed plant opera'tions to verify that the facility vias oeing operated safely and-in conformance with regulatory requirements, the licensee's management controls were effectively discharging the 11cen?ee's responsibilit& for continued safe operation, the licensee's radiolngica'i_ p.ection program was
. implemented in compliance with regulatory raquirements, and the licensee was comolying with the approved security plan.
The inspectors conductM control room observations and plant'
inspection _ tours and reviewed logs and dncomentation'of equipment.
problems.
Routine observations of safety-system flow path-. alignments were performed from both control room indications and local position'-
checks. Through in-plant observations!and selected attendance of the
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licensee's daily meetings, the inspectors verified that the operations staff-maintained cognizance over plant status and TS LCO action.
stateinents in effect, b.
-Plant Tour of Electrical Equipment
On' December 31, 1990,-the inspector toured the Division III DG. room anc' its associated control room. __The a16sel was running at the-time and it showed no evidence of leaks or other abnormalities. The test-in progre.c', was d;scussed with the' operator who was taking data in j
compliaNe with the procedural requirements. The inspector also-toured the Division I and'II standby switchgear Rooms IA'and 1B, and-i noted carrect indications and breaker positions-for the;4160- and-480-Vac electrical boards. Similar correct equipment lineups were also noted in the Division III switchgear room. The inspector:then z
toured:the Class 1E battery rooms for all three divisions and noted that the electrolyte levels were within allowable limi'.s and ceneral
. 4 appearance of the batteries and bettery room to be acceptable. 1The=
switch positions for,the inverters'and chargers in the de equipment *
rooms of'all three divisions were observed to_be correctly ~ positioned.
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Tour of the Auxiliary Building On January 2,11991, the inspector toured 'all.levelssof the auxiliary building.and made the following observations:
All observed. radiological. monitors _were within their. calibration due date and operating properly.
- Fire Door AB 095-09 was held open by_two hoses running through it to a portable HEPA filter. :This' condition was being tracked in the main control room on tracking LCO 88-188 and was an item nn the roving firewatch lir.t.
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Lighting in the HPCS room was unsatisfactory in that only one of the eight lamps on the upper level was lit and only five of the eight lamps on the lower level were lit.
The LPCS room had less than 50 percent of the tvailable lairps lit. This was brought to management's attention for corrective action.
The licensee corrected the lighting problems in the rooms;
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however, the problem was identified to the licensee twice before
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action war taken. The sec'nd notification was approximately 1 week after the first.
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Partial Walkdown of ECCS On January 7, 1991, the inspector verified that the ECCS suction valves on Auxiliary Building Elevation 70 were appropriately positioned. The inspector also verified that the breakers on selected safety-related motor control centers were in the correct position, e.
Outage Startup Observations At 2:42 p.m. on November 30, 1990, the reactor was taken criticci following RF-3.
The inspector observed criticality and associated s
activities. Tht. main generator was synchronized to the grid at 10:39 a.m. on December 4: Ca December 3 and 4, the inspector perfonned extended cuntrol rwm observations of-reactor vessel heatup and low-power te', ting of the ADS /SRVs. After completion of satisfactory testing, the licensee proceeded, without incident, to 75 percent power and held the plant at this power level for further testing to conduct troubleshocting activities on a drifting condenser bypass valve.
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TS Temporary Waiver of-Compliance for-the RCIC Systeq On December 3,1910, a temporary waiver of compliance froc the provisions of TS 3.0.4 on' the requirements of TS 3.7.3.b,'" Reactor; Core Isolation Cooling System," was granted to allow transfer from Mode 2 to Mode 1 with the RCIC system inoperable.
The waiver allowed the licensee, for this single occurrence, to continue plant startup so the plant could be placed in a condition that is less sensitive to minor control system perturbations ~that-could result in undesirable transients or scrams.
In addition, the
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power increase would minimize tht thermal stresses on the feedwater nozzles and piping that results from low-power operation with low'
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feedwater heating and thermal stratification in the feedwater piping.
The waiver of compliance was documented in a letter, dated
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December 5,1990, to the licensee.
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TS Temporary Waiver of Compliance for-the Drywell Air locksc On December 12, 1990, a temporary waiver of compliance from the
. i i provisions of Action a. of TS 3.6.2.3, "Drywell Air Locks," was.
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granted to allow' entering the drywell with one of the two drywell; air. -
l lock doors inoperable.
The waiver allowed the licensee. for a period not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, i
to enter the drywell,Tiith the plant ?ressurized and in Mode 3,_ to identify and repair the source of lea < age from the RCS. This action-was documented in a letter, dated December 13, 1990, to the licensee.
The inspectors noted that the bases for the waivers discussed above were.
well prepared and presented by the-. licensee. -
5.
-Maintenance Observations (62703)
Or November 28 and 29,1990, the inspector observed and reviewed-activities associated with MWO R056700. This MWO was written to repair a weldicrack on the Division 11 Standby DG., This weld sec0res the 14-inch combustion.
air pipe adapter (f rom the turbocharger) to the.end plate of the-intercooler inlet pan. The crack was discovered during the-perfonnance of a 1-hour surveillance test of the-DG, The cause of the crack and the 111censee evaluation of this welding; problem are discussed;in' detaillin NRC Inspection Report 50-458/90-33.-
The. inspector determined that the welding activities were performed using
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an adequate procedure and the repair was successfule es determined by a licensee visual. examination.- The welders were qualified to perform the
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welds,'and the materials used in the reinstallation of the adapter were.
properly qualified.
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Surveillance Observations- (61726)-
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ADS Testing On December 3 and 4,1990.. the inspector observed a power increase
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evolution and the. testing of the ADS /SRVs.
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The power increase was performed in a slow, well-coNrolled manner by the operations shift in accordance with G0P-001, " Plant Startup."
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The inspector noted that thel ATC operator was assisted by two:other licensed individuals and the ':S was assisted with his supervisory duties by the A05.- 00A coverage was.provided by a"previously; licensed individuai, anc, two reactor. engineers were on hand to' assess i
the rod pulls. Additionally, the STA and' operations engineer *
i provided technical support. : This coverage was' also provided during.
the testing of the ADS /SRVs and is typical:of. the resources the licensee commits to high-r,isk evolutions..
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1 The inspector observed the conduct of STP-202-0602, " ADS Safety Relief Valve Operability Test," Revision 6, in accordance with TS 4.5.1.c.2.
i All seven ADS /SRVs Acre successfully tested and the inyccior ptd a'
proper response of the acoustic monitors and SRV position indication.
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The bypass and feedwater system responded to the testing-imposed transients, as expected._ The inspector noted.that all: testing prerequisites were appropriately satisfied.
This included opening of the bypass valves to the required amount' and-an announcement of-containment access denialsprior to commencement-of' testing.= The operators took the conservative = action of' stopping the test to reduce
'i containment pressure. The. licensee's~ administrative limit is-O.30 psig and pressure had reached _0.28 psig.
Both surveillances and the power. increase evolution were carefully..
conducted with_ good attention to procedures and their. requirements.
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resources were applied.
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DG Testing
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On December 31, 1990, the insp'ector observed portions-of the-performance of STP-309-0203, Division =III Diesel Generator-l Operability Test,"l Revision-8, that was in progress.1 The inspector verified with station document control ~ that;the most. current revision of the procedure was utilized, comunications-were established
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between personnel conducting the test at the DG-control; panel and the.
main control room, and-' personnel were qualified. operators, as
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required by the proceoural' prerequisites. The inspector observeE that the operators were following thesprocedure and:were familiar;
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with its contents.
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7.
LER Followup '(92700)
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'The inspector reviewed the LERs listed below to verify that resortability;
requirements were fulfilled, corrective actions were accomplis 1ed, and'
actions were taken to prevent recurrence.
a.
(Closed)LER88-018:l Reactor scram due to main generator; exciter
brush failure.
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.i On August 2'i, 1988, with the unit at:100 percent power, the reactor.
automatically scramed on a turbine control valve: fast. closure signal caused by a loss of main generato'r field excitation,'resulting in:
automatic main generator and turbineitrips.. All-plant: equipment responded, as designed, to this event.
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This event was partially reviewed prior to the issuance of the LER, as documented in NRC; Inspection Report 50-458/88-19.
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completed the review of'the licensee's ' corrective actions that-
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included implementing preventive maintenance procedures to replace the exciter brushes prior to failure, checking the undervoltage.
relays, training of maintenance personnel, implementing appropriate postmaintenance testing, lysis of-the effects of reactor watermodification of HPCS and i
transmitters, and an ana
entering the HPCS line.
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The corrective actions implemented by the licensee appeared to.be adequate, i
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.(Closed)LER_88-021:
Grounding transformer fault caused a: generator trip, reactor scram, and HPCS and-RCIC system injections.
On September 6,1988, with the unit at 100 percent > power, the generator tripped, due to a fault on the neutral _ grounding for
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Transformer ISTX-XGNIA (normal 13.8-kV station service transformer),
and caused a reactor scram. The fault was caused by,a stray cat-
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shorting out the high side of the grounding transformer. The HPCS and RCIC systems were inadvertently initiated on spurious Level 2 differential: pressure signals. _A NOUE_was declared based on-an-ECCS injection into the reactor vessel.
This event was partially reviewed prior to the issuance of. the LER, as documented in NRC Inspection Report 50-458/88-19..- The inspector further reviewed the licensee's corrective _ actions'and determined that they were adequate, c.
'(Closed)LER88-022: -Autostart of_the fuel building ventilation-
treatment system due to a radiation monitor high signal.
This event was previously reviewed, as documented in NRC Inspection
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Report 50-458/89-26. The re identifiedtheroot:cause(s)portnotedthat-thelicenseehad
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_and.had implemented corrective actions to prevent recurrence, i
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(Closed)LER88-023:
Voluntary report due. to inoperable MSIVs.
On September 30, 1988, with the unit-at. 75 purcent power, a reactor.'
a shutdown was: initiated after two inboard lMSIVs:(1821*A0V-F022B and 1821*A0V-F0220) were found to be inoerable during testing Lin
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response to NRC Information Notice 88-43';" Solenoid Valve Problems."
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All remaining MSIVs were tested and each remained;in the full-closed i
- position, indicating proper operation of the fast-closure 50V and;the capability of the MSIVs to close on a valid isolation signal.
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This event was reviewed in detail prior to issuance of the LER, as documented in NRC Inspection Report 50-458/88-23. Additionally,ithis -
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LER was reviewed for corrective action adequacy and implementation,.
as documented in NRC Inspection Report 50-458/89-26.
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(Closed)LER88-024:
Spurious RWCU system isolation during a temperature reading as part of a surveillance.
This event was previously reviewed, as docurrented ih NRC Inspection Report 50-458/89-26. The re identified the root cause(s) port noted that the licensee had.-
and had implemented corrective actions to prevent recurrence, i
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(Closed)LER88-025:.RCIC system isolation due to procedural error and parsonnel oversight.
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On December 8,- 1988, with the unit at 100 percent power, an ' solation-of the RCIC system occurred. At the time of the isolation, tiu RCIC
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system was removed from service to perform preplanned maintenance.
%e isolation resulted from the Division II RCIC. steam supply ' leak
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, atection transmitter 11E31*PTN08'JB) being calibrated instead of the
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Division I transmitter (1E31*PTh083A), as required by the= STf. The-
STP had been recently re/ised and incorrectly-specified the -location
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of Transmitter 1E31*PTN083A. Although the technician read the transmitter identification, tag, the procedure error went undetected
and the technician began-calibration of the wrong transmitter.
This event was reviewed in detail prior to issuance of'the LER. as _.
documented in NRC Inspection Report 50-458/89-07 Additionally, this
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,m LER was reviewed for corrective action adequacy e.? implementation, as documented in NRC' Inspection-Report 50- G /89-26.
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(Closed)LER88-026:
Inadequate filter application lfor safety-related dampers due to a-. design error.
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This event was previously reviewed,- as documented in.NRC Inspection Report 50-458/89-26. The-report.noted that:the~ licensee had'
identified the -root cause(s) and had implemented corrective actions-to prevent recurrence, h.
(Closed)LER88-027: -InoperaMiity of the RCIC system due to an ircomplete constructior. r.% fication.
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On December 19, 1988, with the unit at approximately 95 percent power, the licensee determined that the installation of the~RCIC'
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system turbine had not been complett.d~according-to design
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requirements.
This condition was r.oted as part-of c program by the-
. design engineering group to review and prioritizel outstanding modification packages.
The RCIC system was declared inoperable, althouch it was:available and would have operated if required.
Proper inttallation was complated and, after satisfactory retest, thefRCIC system was-
-restored to an operable status,
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This event was the subject of Violation 458/8826-01, issued in NRC Inspection Report 50-458/88-26, because the turbine was not mounted according to seismic design. This violation, the licensee's response, and the event itself, were reviewed and closed in NRC Jnspection Report 50-458/90-26, 1.
(Closed)LER88-029:
Inadvertent autostart of annulus mixing and standby gas treatment systems due to a stuck check source in a.
radiation monito;*,
This event was previously reviewed, as documented in.NRC Inspection detennined the root cause(s) port stated that the licensee hadand had imple; Report 50-458/89-26.
The re to prevent recurrence.
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(Closed)LER89-004:
ESF actuation occurred when'I&C personnel took wrong voltage readings.
On February 10, 1989, with the unit at 80 percent power, an ESF actuati n occurred when an I&C. technician incorrectly took a voltage reading on an instrument trip unit. The instrument tripped as a result of this r.rror, causing the RCIC system to isolate due' to an inadvertent high steam flow sigrt1. The licensee verified that an actub' high steam flow condition did not exist and the isolution signe was promptly reset by operations personnel, allowing the RCIC system to be imediately restored to standby service.
The licensee classified the root cause as a personnel error. The inspector reviewed this event for adequate corrective actions that included co9nseling of the Vividual.and training for the I&C department.
No' problems wer noted, k.
(Closed)LER89-006:
RPS actuation due to downranging IRMs during insertion.
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On February 17, 1989, with the unit in hot shutdown, the RPS actuated from upscale trip signals in the intermediate range of the neutron monitoring system.
All control rods were inserted and no additional rod notion occurred. The cause of the RPS actuation was a result of operator error. The RPS responded as designed and the control rods'
.were fully inserted prior to the actuation.
Corrective actions included procedure clarifications and required reading or onshif t briefings for licensed operators.
The inspector reviewed this event for adequate corrective actions that included counseling of the operator and training of the operations staff on the event. Additionally, a caution ctatement was added to GOP-00P, " Power Decrease / Plant Shutdown,* and A0P-001,
" Reactor Scram," stating the impact on the RPS if IRMs are downranged prior to the detectors being fully inserted.
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1 (Closed).LER 89-007: Reactor scram due to an IRM upscale trip.
On February 20,1989, with.the reactor. mode switch in startup and
power in the intermediate range, a reactor scram occurred as a result c
of an IRM upscale trip.
The IPd upscale was caused by a sudden-
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increase in feedwater flow rate, resulting in a power increase.
This event was reviewed in' detail in NRC Inspection
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Report 50-458/89-07 prior to the issuance of the LER.' Alli corrective actions were determined to be adequate.
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(Closed).LER89-008:
Relay failure causing a generator trip,-reactor
scram, and HPCS and RCIC system injections, i
On February 25, 1989, with the unit at 78 percent power, the reai: tor _
a automatically scranad while 'perfonning a routine upper thrust
bearing wear detector test in accordance with OSP-0101'. The scram!
occurred as a result of a turbine trip caused by-a defective bypass -
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relay. The relay failed to open the trip-bus circuit, as designed,
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to prevent a turbine trip while testing the thrust bearing wear 1-detector.
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This evei:t-was partially reviewed as documented inlNRC Inspection Report 50-458/89-07. The-only item.left open was to review the _
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licensee's actions-to modify the' Rosemount 1154 transmittersiwith a
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dampening circuit. The modifications were completed.by June 1989.
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(Closed)LER89-012:
RHR shutdown cooling isolation due to a?losslof.
j power while' taking a breaker out of service.
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On March 25, 1969, with the unit in refueling, an ESF isolation.
occurred for the RWCU system main steam line drains and the RHR shutdown cooling systems. The ESF isolation occurred due to a' loss of power to the Division 11 isolation logic caused by an operator
opening the breaker supplying the-logic system while hanging a clearance tag for maintenance work.
This event was reviewed prior to the issuance of'the LER, as.
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-i documented in NRC Inspection Report 50-458/89-11. The. report noted
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that corrective actions implemented by the licensee were' adequate.
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(Closed) LER.89-015:
ESF-( :tuation due to isolation of an RHR
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shutdown c( ling suction valve.
On March 29, 1989, with the unit in.refuelingLand the refueling' pool water level greater than 23 feet-above the top of the reactor- -
a pressure vessel flange, a half-scram signal;on the RPS occurred and the RHR shutdown cooling suction valve (E12*HOV-F008) isol' ted.
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a technician incorrectly replaced.a jumper that hadLinadvertently fallen
'g off its terminal, causing' Fuse C71-F30 to blow.. This action deenergized the associated control logic.
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documented in NRC Inspection Report 50-458/89-11. The report notedi
that the corrective actions taken b,, the licensee were adequate,
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(Closed)LER89-020: Loss ofl shutdown cooling'when containment.
1 solation valves actuated due to a power loss when electrical equipment was flooded by a SWS freeze seal loss.-
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On April 19, 1989, with the unit in: refueling and thel refueling pool water level greater than 23 feet above theitop of the' reactor i
aressure vessel flange, a freeze plug, on an:SWS line11nithe auxiliary
)uilding, failed.
1 This resulted in leakage of service water into the t
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auxiliary building and selective power outages throughout the plant.:
The power outage included the Division II RPS-bus and deenergization:
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of a vital 120-volt power supply, resulting'in-the closure ofz
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containment isolation valves. As a result of these isolations, shutdown cooling was' lost.
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This event was reviewed, as documented in.NRC Inspection.
- Report 50-458/89-11. Additionally, Region IV dispatched an.AIT to
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study the background and consequences.of-this event.
The All review
was documented in NRC Inspection: Report =50-458/89-20.
These reports i
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noted that adequate corrective ~ actions had been taken by the licensee.;
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(Closed)-LER89-021: -Loss of sh'utdown cooling and RPS actuation'duei to power transient from test lead grounding.and a blown:lfuse.
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On April 27 -1989, with the unit in refueling, an unplanned ESF actuation occurred as a result of a power transient to several trip-
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i cards associated with the RPS and the RHR shutdown-cooling isolation logic. The power transient _ occurred as= a result of a test connection shorting against protective control wires widle' performing
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surveillance testing.-
This event was reviewed and cited as Violation 1458/8911-01Lin NRC'
Inspection' Report 50-458/89-11. :A review of the licensee's
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corrective actions will be perfonned during followup-of the
violation,
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(Closed)LER89-029: Two ESF:actuations: occurred due;to shorted l leads while r_eplacing a transformer.
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On June 13, 1989, with the uni! in cold shutdown, an unplanned ESF~
actuation occurred as a result of: technicians. shorting two leads
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a together while. installing w spare transfonner. lThis action:resulted
in a trip.of PreferredNrnsformer D and 'deenergization-of.
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safety-related buses;
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l-While contract electricians were preparing leads for tennination,. the leads came in contact with one another, resulting in aitransforner
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trip and subsequent ESF actuations.. The root cause of: this event was
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determined by the licensee to be a breakdown in comunications
~between the contractor foreman and'his work crrw. The corrective-
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action taken was to isolate the short until the leads were properly
terminated on the replacement transformer.
A second ESF actuation occurred when a relay technician operated contacts contrary to procedure guidance and tripped the main generatori feeder breakers. The root cause of the'second event was determined:
to be a personnel error and failure to follow a procedure.
These events were previously reviewed.:as documented in NRC Inspection Report 50-458/89-28. The inspector-performed further-review to verify that the licensee had implemented adequate
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corrective actions. Both these events were caused by comunication problems with outside organizations. These problems-have apparently '
ben corrected as= evidenced by the~1ack of problems with contractor er offsite comunications during the most recent outage.
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(Clortd)LER89-030: Pressure transmitter isolation valve found misaligned causing inability to-sense drywell pressure.
On June 17, 1989, with the unit in cold shutdown, a pressure transmitter root valve for the PVLCS was found closed, while.
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performing _a safety system valve lineup, causing one division to be'
Investigation determined that this valve had probably
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been mispositioned since.the conclusion of=the primary containment integrated leak rate test on May 30, 1989, q
This event was previously reviewed,-as documented in NRC -Inspection Report 50-458/89-28. The report noted that the licensee had
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implemented corrective actions to' prevent recurrence.,
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Exit Interview An exit interview was conducted with licensee representatives identified in paragraph 1 on January 15. 1991. During this interview, the inspectors.
reviewed the scope and findings of-this-inspection.- The licensee did not'
identify, as proprietary. any infonnation provided to : or reviewed by, the inspectors.
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ATTACHMENT i
Acronyms and Initialisms ADS Automatic depressurization system
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AIT Augmented inspection team
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AOP Abnormal operating procedure
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AOS Assistant operations supervisor
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At-the-controls
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C0F Control operating foreman
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Direct current dc
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DG Diesel generator
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ECC3 Emergency core cooling system
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ERIS -- Energency response information system ESF Engineered safety feature
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F Fahrenheit
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GOP General operating procedure
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GSU Gulf States Utilities
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HEPA High-officiency particulate air
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liPCS High-pressure core spray
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I&C instrumentation and controls
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IRM Intermediate range monitor
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ISEG Independent safety review group
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kV Kilovolt
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LCO Limiting condition for operation
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LER Licensee event report
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LPCS - Low-pressure core spray MSIV Main steam isolation valve
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MWO Maintenance work cruer
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NOUE Notice of unusual event
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NRC_
Nuclear Regulatory Comission
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00A '
Operations quality assurance
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OSP Operations section procedure
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psig Pounds per square inch, gauge
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PVLCS - Penetration. valve leakage control system RBS River Bend Station
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RCIC Reactor core isolation cooling
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Refueling outage RF
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Solenoid-operated valve SOV
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SR0 Senior reactor operator
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SRV Safety-relief valve
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SS Shift supervisor
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STP Surveillance test procedure
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Main steam safety / relief valve air system SVV
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SWS Service water system
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Technical Specification TS
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