IR 05000458/1986037
| ML20207E049 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 12/19/1986 |
| From: | Bennett W, Chamberlain D, Jaudon J, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20207E028 | List: |
| References | |
| 50-458-86-37, IEIN-85-063, IEIN-85-63, NUDOCS 8701020049 | |
| Download: ML20207E049 (11) | |
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APPENDIX
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U..S.
NUCLEAR REGULATORY COMMISSION
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REGION IV
. NRC Inspection Report: 50-458/86-37 Docket: 50-458
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Licensee: Gulf' States Utilities Company (GSU)
P. O. Box 220 St. Francisville, Louisiana 70775 Facility Name: River Bend Station (RBS)
Inspection At: River Bend Station, St. Francisville, Louisiana Inspection Conducted: November 1 through November 30, 1986 Inspectors:
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I 85 D. D. VChamberlain, Senior Resident Inspector Date (pars. 1, 2, 3, 4, 5, 7, 8 and 10)
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m 11 Ill 86 W. B. Johes, Resident Inspec r Date I (pars. 1,2,3,5,6,7 8)
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B t. Projdct Engineer Date U
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- Approved:
..Jpdon, hief, Projdet Section A Date '
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acto'r Pr ects Branch 8701020049 861219 PDR ADOCK 05000458 G
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Inspection Summary Inspection Conducted November 1-30, 1986 (Report 50-458/86-37)
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Areas Inspected:
Routine, unannounced inspection of licensee action on previous inspection findings, Licensee Event Reports (LERs), licensee action on IE Notices, operational safety verification, surveillance test witnessing, maintenance witnessing, safety system walkdown, and in-office review of written reports of nonroutine events at power reactor facilities.
Results: Within the areas inspected, no violations or deviations were identified.
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DETAILS 1.
Persons Contacted Principal Licensee Employees W. J. Beck, Supervisor, Reactor Engineering
- B. M. Burmeister, Licensing Engineer
- W. H. Cahill, Jr., Senior Vice President, River Bend Nuclear Group
- 8i. M. Cargill, Supervisor, Radiation Programs
- T.' C. Crouse, Manager, Quality Assurance (QA)
- J. C. Deddens, Vice President, River Bend Irclear Group P. E. Freehill, Superintendent, Startup ard Test A. O. Fredieu, Assistant Supervisor, Operations D. R. Gipson, Assistant Plant Manager, Operations
- P. Graham, Assistant Plant Manager, Operations
- E. R. Grant, Director, Nuclear Licensing
- B. R._ Hall, Assistant Superintendent, Field Quality Control (Stone and Webster)
- R. W. Helmick, Director, Projects G. K. Henry, Supervisor, Electrical Engineering K. C. Hodges, Supervisor, Quality Systems
- R. J. King, Supervisor, Nuclear Licensing
- A. D. Kowalczuk, Assistant Plant Manager, Maintenance
- J. W. Leavines, Director, Field Engineering
- I. M. Malik, Supervisor, Quality System
- J. H. McQuirter, Licensing Engineer
- V. J. Normand, Supervisor, Administrative Services
- W. H. Odell, Manager, Administration T. F. Plunkett, Plant Manager S. R. Radebaugh, Assistant Plant Manager, Services
- C R. Roberts, ' raining Coordinator
- M. F. Sankovich, Manager, Engineering
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R. R. Smith, Licensing Engineer
- R. L. Spence, Resident Quality Control Manager (Stone and Webster)
- R. B. Stafford, Director, Operations QA
- K. E. Suhrke, Manager, Projects D. Williamson, Supervisor, Operations The NRC senior resident inspector (SRI) and resident inspector (RI) also interviewed additional licensee personnel during the inspection period.
- Denotes those persons that attended the exit interview conducted on December 5, 1986.
NRC resident inspector (RI) W. B. Jones also attended the exit interview.
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- * ~ 2r : Licensee Action'on Previous Inspection' Findings
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a.
(Closed) Open Item _(458/8604-03): : Monitor licensee actions' relative '
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- to main feedwater regulating-' valves sticking problem.
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'Sincetthe initial problems.with feedwater regulating valves sticking
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.during initial startup;-improved valve disc. material has been
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. control room during the. latest outage.
Subsequent operation at
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various. power levels has revealed no' additional valve problems.
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ThisLopen item-is' closed.
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(Closed) Violation (458/8581-02):. Inadequate retest following a
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-modification or repair.
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The licensee:has modified maintenance and operations administrative-
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procedures ~to clarify. requirements and responsibilities for
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establishing. retest requirements following modifications or repairs Land for. initiating tracking. limiting' conditions for operations.
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Documented training has been provided for the maintenance and u
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- operations departments.
The'NRC resident inspectors' monthly
' maintenance' witnesses has revealed;.no post maintenance retest n-;
'2 deficiencies.
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This: violation is closed.~
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c. ' (Closed) Violation (458/8614-01):. Inadequate corrective action for
. licensee identified problem with misaligned diesel generator fuel filter stop valve.
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iNuclear plant engineering (NUPE) personnel have been instructed that
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- the evaluation of-identified problems or condition reports (CRs)
should identify the root cause of the event and fully describe
'f remedial-and/or. generic corrective action taken to prevent
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-recurrence. The administrative procedure for CRs has been revised to
. standardize the form used for identification of problems and to
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provide more detailed, instructions for problem evaluations.
The procedure revisicn also provides for quality assurance (QA)
involvement in the evaluation and closeout stages to provide added assurance that the root cause has been identified and that the
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. corrective. action is adequate.
The licensee compliance group
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expedited the closure of approximately 400 backlogged CRs.
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'This violation is closed.
d.
(Closed) Violation (458/8614-02):
Failure to follow procedure for P
' issue of maintenance work requests for performance of maintenance on e
plant emergency diesel generator.
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The administrative procedure for conduct of maintenance has been revised to delete the use of " Shop Work Orders" for maintenance activities. Maintenance personnel were instructed that all work on permanent plant equipment is to be performed only under a valid maintenance work request (MWR) or preventive maintenance (PM)
document.
An independent review of maintenance department activities has been completed and corrective actions implemented.
This violation is closed.
e.
(Closed) Violation (458/8614-03):
Inoperable division II emergency diesel generator due to a'mispositioned fuel oil system filter stop valve.
The licensee initiated corrective actions for this problem included:
match marking of the fuel oil valve and valve handle for proper
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alignment on the diesel generators; demonstrating operability of the emergency diesel generators;
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revising the operations daily log report to verify proper fuel
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oil strainer valve alignment during each shift; inspecting of other plant equipment for similar duplex strainer
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configurations and taking action as appropriate;
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assigning security the responsibility for key control of door,s;
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and initiating of programs for operations and quality assurance to
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perform independent reviews of selected systems for proper valve alignment.
This violation is closed.
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(Closed) Open Item (458/8602-01):
Investigation of bolt failure and verification of torque values for motor operated valves (MOV).
The licensee has completed their investigation of the bolt failure on the feedwater isolation valve 1FWS*MOV7B which occurred on January 5, 1986.
Licensee Event Report 86-30, " Inadequate Bolting on Motor Operated Valves," was issued to provide an analysis of the event and the corrective actions being taken to prevent reoccurrence.
The licensee has determined that the bolt failure which occurred between the valve' actuator (SMB-4 Limitorque) and the valve body (Velan) on 1FWS*M0V7B was because of inadequate bolt thread engagement and an improperly applied torque.
After discussions with Limitorque and Velan, the licensee determined that the bolt engagement should have been 1.5 diameters of the bolt as opposed to
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ithe 1. diameter realized by the short bolts that were installed on m
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~this1 valve.. Based on this' finding, all Category 1 MOVs: have been
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reviewed'and' bolts of_ length sufficient to give a 115 di.ameter.
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- engagementhinstalled. ' Lock washers have also;been installed on all
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SM-4.and 6-HPLand 1arger motors where they were found to be missing.
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lThel torque values for the SMB-4. valves have_been revised to
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. 700;ft.-lbs.
The torquing procedure haszalso been revised to only-
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ensure the full torque value is realized between the valve actuator '
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The' licensee also found that several other Category I MOVs-had
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< experienced 11essening of.the required torque, values'and concluded M
that this had been caused by~the-dissimilar. tensile strengths of the hy actuator casting and bolting materials,. vibration, thermal cycling
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and normal bolt relaxation.
No other-bolt.. thread engagement problems
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- The licensee has established a program to monitor all Category I MOVs s
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up through the first refueling outage. The torque seal applied to
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the SMB-4 valve's'duringthe initial inspection is being monitored
during each scheduled outage, for. breakage which would indicate loss
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of the required torque. ' Twenty-five' percent of the remaining a'
Category I MOVs are being inspected in the same manner as the SMB-4 valves every 3 months to determine if the bolts have relaxed.
At the
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.first refueling outage all Category I valves will be rechecked for b'reaking torques.
Comparison of this data with the "as-left"
-information will be used to set future inspection intervals.
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-licensee has implemented temporary procedures for reviewing the
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Category I valves up through the first refueling outage.
Based on
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.the results of the inspection to be performed during the initial l
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the future inspection program for Category I motor operated valves.
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This' item is closed.
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3.
Lic'ensee Event-Reports-(LERs) Review
- During this-inspection period, the SRI and RI reviewed LERs for compliance
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system." Specifically, the LERs were reviewed for accuracy and clarity of the event description, the cause of each component or system failure or
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personnel error, the failure mode and effect each event had on plant operation, operator actions that affected the course of the event, and the corrective actions taken to prevent recurrence of the event.
j The following LERs were reviewed and closed:
",85-025 Reactor Water Conductivity Measurement i
85-026 Inadvertent ECCS Initiation
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..- 86-030 Inadequate Bolting on Motor Operated Valves86-037 Reactor Scram on High Pressure Due To Low Setpoint 86-042 Reactor Scram on IRM Upscale 86-043 Standby Gas Treatment Start Due to Opened Breaker 86-044 Reactor Trip Due to MSR Drain Tank Gasket Failure 86-045 Reactor Scram Due to High Steam Seal Evaporator Tank Level 86-046 Missed Technical Specification Action on Fuel Building Radiation Monitor 86-048 Missed Control Rod Block Surveillance 86-049 Fuel Building Filtration Start on Spurious Radiation Monitor Spike 86-050 RWCU Failure to Isolate Due to a Faulty Bypass Switch 86-051 Spurious RWCU Isolation During Temperature Reading Surveillance 86-052 Spurious Control Room Ventilation Isolation 86-053 Thermolagging Material Removed for Maintenance Without a Fire Watch 86-055 Reactor Scram Due to a Faulted Transformer 86-056 Reactor Scram Due to Entrapped Air in Instrument Sensing Line 86-057 Reactor Water Cleanup Isolation From Grounded Jumper
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The above listed LERs are closed.
During this review of LERs, it was noted by the SRI that some initial reports and supplemental reports were submitted late.
During a November 1986 meeting at the NRC Region IV office with GSU senior management, the general subject of content and timeliness of LERs was discussed.
GSU management acknowledged a problem with LER submittals and outlined some of the steps they are taking to assure timely and quality submittal of LERs.
The timeliness of LER submittals was also documented as an open item in NRC Inspection Report 50-458/86-33.
This area will be monitored during future NRC inspections.
4.
Licensee Action on IE Notices-This area of inspection was conducted to review licensee action relative to IE Information Notice No. 85-63, " Potential for Common-Mode Failure of
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~ Standby Gas Treatment (SGTS) System on Loss of Off-Site Power." IE
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Information Notice 85-63 was issued to inform licensees of a facility
~ design error which would have prevented the automatic initiation of SGTS if_offsite power was not available. The specific design error noted was that'the instrument that senses the SGTS heater temperature deenergizes a relay on loss of offsite power and the relay does not automatically reset when power-is restored.' This condition would prevent the automatic initiation'of SGTS until the instrument is manually reset.
GSU reviewed the design at River Bend (RB) and found that the temperature switches.in the SGTS heater circuits reset automatically on restoration from power outages..Therefore, the SGTS logic design at RB would preclude the problem identified in IE Information Notice 85-63.
This IE Notice ~is closed.
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- Operational Safety Verification River Bend remained in an outage for the majority of this inspection
' period.' The SRI and RI monitored outage activities and witnessed the
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reactor startup at the conclusion of the outage.
The reactor startup commenced on. November:29,~ 1986, at 12:16 p.m. CST and the reactor was
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'decla' red critical at 2:22 p.m. CST.
Control room activities and conduct were observed to be~well controlled and efficient during the reactor
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startup. ' Activities ' completed during the outage included removal of main turbine. screens, local leak rate testing, surveillance testing and completion of selected maintenance items and modifications.
Several plant tours were conducted during the outage including entries into the containment and drywell.
During the drywell entries, emergency core
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cooling systems valve positions were verified for proper lineup.
The post outage plant cleanliness level was generally good. A walkdown of the
. standby gas treatment (SBGT) system was conducted and the results are documented in paragraph 8 of this report.
The resident inspectors reviewed licensee actions on several potential
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problems during this inspection period.
The results of reviews of selected items are described below:
a.
Reactor' Protection System (RPS) Response Time Testing:
On November 7, 1986, the licensee discovered that incorrect diesel generator start times has been used during RPS response time surveillance testing for obtaining total channel response times.
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problem apparently occurred during transfer of response time data from one surveillance procedure to another.
The licensee took immediate action to obtain the correct response time values from the latest surveillance tests performed and recalculated the total channel response times.
A review of the data for approximately 51 surveillance tests revealed that all response times were well within acceptance' criteria and Technical Specification limits.
The licensee is continuing the investigation of the root cause and possible corrective actions to prevent recurrence.
The SRI will continue to monitor licensee actions in this area.
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' Division'I Emergency Diesel Generator Start Failure:' On November 24, m _' s-1986, the Division I emergency diesel generator. failed to start for an operability ~run'following maintenance on.the air start solenoids.
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The licensee, with assistance from a Transamerica De Laval vendor W
representative,; initiated an immediate investigation of the cause'.for the start failure.
It was determined that an excessive amount of
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. lubricating oil was. accumulating.in the air start distributor, the.
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air lines to at least two cylinder air start valves and two cylinder air 1 start valves. The start failure was attributed to the oil in the
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air lines because the difference in the viscosity.and density of air
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-and oil did not allow the' cylinder air start valves to respond to a pneumatic signal to open.
A certain amount of lubricating oil is injected into the air start distributor-to provide valve lubrication.
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L The excessive lubricating oil problem appears to be aggravated by the design of the rear air start distributor (e.g. no low point external
drain) and by required post engine run air rolling methods. The a
licensee discovered that engine air rolls were being performed by depressing the air roll button and releasing it when the tachometer started to move, thus not allowing the engine to achieve its normal
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air roll speed. -This air roll method would possibly tend to collect oil in the air distributor and air lines by not allowing time for excessive oil to blow out.
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Licensee' corrective ~ actions for this air start failure included:
replacing of the air start valves in cylinders 3 and 8 on
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Division I diesel (no actual problem with'the air start valves was identified);
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replacing of the Airmatic oil injection. valve on the Division I
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diesel.
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removing of the air start valve on cylinder 5 and draining
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accumulated oil on Division I diesel;
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draining of excessive oil from the air start distributor on
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' Divisions I and II diesels (very little oil was found in the Division II diesel);
initiating preventive maintenance to drain accumulated oil on a
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monthly basis for Divisions I and II' diesels; initiating a design change request to investigate ways to
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improve lubricating oil drainage from the rear air start distributor;
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i revising of surveillance test and station operating procedures I
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so that all air rolls of Division I and II diesel engines are performed for at least 5 seconds and that the engine speed exceeds 50 RPM during the air roll;
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performing.several) air; starts'ontheDivisionLIdieselwithout
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- any problems after the oil problem was corrected and the N
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. Division II diesel,was'also successfully operated; and
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The SRI considers-the licensee 1 response to this problem to be prompt
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'and' thorough and he will continue to monitor'the diesel generators
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reliability.during future inspections.
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-No violations'or deviations were identified in this? area of the d
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iSurveillance Test Witness'
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tThe RI observed the performance of surveillance' test STP-050-0700, "RCS
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? Pressure / Temperature Limits Verification," on November 29,~1986.
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csurvH 11ance. test implements the' requirements of Technical 7 __g
! Spec ication-(TS) 4.4.6.1 and 4.4.6.2 which limits the reactor _ system
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. heat up rate.toiless then or equal to 100*F in any one hour period and_-
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,; establishes the minimum _ vessel metal temperature for a given vessel top (e m w
. head pressureirespectively.
'On November ~29,11986, the reactor achieved criticality following a planned
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= outage.
During the nuclear. heat up from cold shutdown'(less than 200*F)
to ratedireactor.-temperature and pressure, the 1 ensee maintained the heat 'up rate less than 100'F in any one hour period as required.
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- addition,~the:RI observed that the reactor vessel metal: temperatures for (6 ;
Lthe respective vessel top head pressure exceeded the minimum operating NP Jtemperature limit curves'of TS Figure 3.4.6.1-1 as required.
These curves g
,1-have been established to ensure the requirements of-10 CFR Part 50,
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Appendix G, for fracture toughness of ferritic materials in pressure
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retaining components are met.
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No violations or. deviations were identified in this area of the
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inspection.
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Maintenance Witness
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'_.During:this' inspection period, the SRI observed a maintenance activity for disabling the. shunt trip on electrical power line conditioner 1SCM*XRC14A1.
?The maintenance activity was conducted in accordance with approved prompt m
' modification request (PMR) No. 86-0118.
The post maintenance retest was
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- also conducted in accordance with the. instructions provided in the PMR.
A-quality control (QC) inspector was present to witness certain portions of
- the maintenance.
During this maintenance activity and subsequent retesting,
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the maintenance personnel identified an improperly lugged slip on connector
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on a rectifier, and they found some rectifiers that were not functioning properly. These identified problems were documented and rubsequently
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, resolved.
The~ SRI subsequently reviewed the maintenance work order close out documentation and no problems were identified.
No violations or deviations were identified in this area of the
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inspection.
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.SafetESystem'Wa'kdown
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MW lDuAing thislinspectio'n period, the ' SRI and RI. performeId a.walkdown 'of the
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- Division-I.and'II standby gas treatment' system (SGTS)..This system, which N-s,
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(is. required to beloperable during,0perational Conditions-l', 2, and 3 is
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designed lto process potentially radioactive: exhaust air from the annulus,
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Jauxiliary building and containment /drywell purge system, and to maintain a W -,
negative.5.w.g. pressure.in the' annulus area.to ensure that' any shield
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.buildingeleakagelis.into the annulus..Although this walkdown was~
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l performed with the plant.in Operational Condition:4, no anomalies were
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identified which would prevent the;SGTS from performing its intended p
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' function;when the' plant entered Operational Conditions -l',- 2, or 3.
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- However, Lone error concerning the mimic layout en,the control room panel g("W was-identified.' The SGTS mimic shows.a' passive mini ~ flow duct from the
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SGTS filterctrain discharge back to the suction duct on the filter train
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?for both divisions.3;The as-built system though does-not contain this
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passive mini _ flow duct...The duct work on the Division I train has a blank
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M inserted into the' duct to prevent return air flow and the Division II
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- itrain has had the. duct work: removed and the two ends blanked off.
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system layoutiis appropriately reflected on the respective piping and
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instrumentation drawing..This incorrect mimic was brought to the nif'
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. attention'of the on duty nuclear control operator and discussed-in the NRC
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, exit.meetingc The licensee will initiate-action;to correct the mimic.
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_ Prior to the reactor entering Operational. Condition 2, the RI verified
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Lperformed.
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No violations or deviations were identified in this area of the
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i-inspection.
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cIn-Office Review'of Written Reports of Nonroutine Events at Power Reactor (
. Facilities The purpose of this portion of the inspection was to ascertain whether corrective actions discussed in the licensee event reports appeared to be Eappropriate; and whether information reported satisfied reporting requirements.
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The NRC inspector reviewed LERs 86-45 through.86-52.
All reporting
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' requirements, were found to have been met and all reports were adequate to assess 1the events reported.
Corrective actions specified in the reports
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- i to be adequate to identify the root causes and to correct these Causes.
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No violations or deviations were identified in this area of the inspection.
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Exit'and Inspection Interview An exit interview was conducted on December 5, 1986, with licensee representatives (identified in paragraph 1).
During this interview, the SRI reviewed the scope and findings of the inspection.
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