IR 05000458/1988012

Insp Rept 50-458/88-12 on 880401-30.No Violations Noted. Major Areas Inspected:Licensee Action on Previous Insp findings,LERs,10CFR21 Repts,Surveillance Test Observation & Operational Safety Verification

ML20154Q481
Person / Time
Site:	River Bend
Issue date:	05/24/1988
From:	Holler E, William Jones, Madsen G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20154Q479	List: IR 05000458/1988012 ML20154Q474
References
50-458-88-12, NUDOCS 8806070031
Download: ML20154Q481 (12)

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APPENDIX U. S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-458/88-12 Docket: 50-458 Licensee: Culf States Utilities Company (CSU)

P. O. Box 220 St. Francisville, Louisiana 70775 Facility Name: River Bend Station (RBS)

Inspection At: River Bend Station, St. Francisville, Louisiana Inspection Conducted: April 1-30, 1988 Inspectors: Db;a 3 p 5. u - it W. B. Jones, Resident [ Inspector Date Project Section C, Division of Reactor Projects A 0 & 5

'G. L. Madse6, Project Engineer Date'/

Project Section C, Division of Reactor Projects Approved: - -

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E . J .# Hofler, Chief, Project Section C D' ate '

Division of Reactor Projects 8806070031 880526 PDR ADOCK 05000458 O DCD

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Inspectien Sununary

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Inspection Conducted April- 1-30,1988 Areas Inspected:- Routine, unannounced inspection of licensee action on previous inspection findings, licensee event reports,10 CFR Part 21' Reports,

surveillance test observation, maintenance observation, and operational safety ,

verificatio Results: _ Within the areas inspected, no violations or deviations were

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DETAILS

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• Persons Contacted

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D. L.- Andrews, Director, Nuclear Training W.'J. Beck, Supervisor, Reactor Engineering J. E. Booker, Manager, Oversight

• J. G. Cadwallader, Supervisor, Emergency Planning

• E. M.- Cargill, Director, Radiation Programs

• J. W. Cook, Lead Environmental. Analyst,' Nuclear Licensing T. C. Crouse, Manager, Quality Assurance (QA)

• W. L. Curran, Site Representative, Cajun

• J.-C. Deddens, Senior Vice President, River Bend Nuclear Group _

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D. R. Derbonne,' Assistant ~ Plant Manager, Maintenance R. G. . Easlick, Supervisor, Radwaste

• L..A. England, Director, Nuclear Licensing A. 0. Fredieu, Supervisor, Operations

.P. D.-Graham, Assistant Plant Manager, Operations

• J. R. Hamilton, Director, Design Engineering W. C. Hardy, Supervisor, Radiation Protection

• G. K. Henry, Director, Quality Operations

• K. C. Hodges, Supervisor, Chemistry

• L. G. Johnson, Technical Operations Manager, Cajun G. R. Kimmell, Director, Quality Services R. J.. King, Supervisor, Nuclear Licensing

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A. D. ' Kowalczuk, Director, Oversight

• A..J. Kugler, Acting Director, Field Engineering I. M. Malik, Supervisor, Quality Systenis

• V. J. Normand, Supervisor, Administrative Services

• H. Odell, Manager, Administration

• T. F. Plunkett, Plant Mana'ger M.-F. Sankovich, Manager, E.,gineering

• K. E. Suhrke, Manager, Project Management R. G. West, Supervisor, General Maintenance NRC

• L. Madsen, Project Engineer, Project Section C, Division of Reactor Projects The NRC inspectors also interviewed additional licensee personnel during the inspection period

• Denotes tho a persons that attended the exit interview conducted on May 3, 198 _

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-4-2. Licensee Action on Previous Inspection Findings (Closed)OpenItem(458/8720-02): Backlog of training management reports (TMR) and training material discrepancy reports (TMCR) for review of events and inclusion in training material, respectivel The licensee has reviewed all the 1MR$ up to 1988 and initiated iMORs for those items which may need to be added to the existing training material During the review of TMR 86-0001 on fire protection, it was noted that the subject matter had not been appropriately addressed for applicability to the licensed operator requalification training program. The TMR response indicated that fire protection is addressed in the fire brigade training program, however, the subject of the event concerned the effect the deluge system may have on plant equipment if it actuates. This apparent inappropriate response was discussed with licensee management and a review of the THR is being perforre This item is close . Licensee Event Report Followup ihe following licensee event reports (LERs) were closed on the basis of the NRC inspector's review of licensee documentation and discussions with personnel:

86-024 ESF Actuation Due to an Electrical Protection Assembly (EPA)

Breaker Failure - All EPA logic cards were replaced with cards of the same design but containing hand selected components. Two additional spurious EPA breaker trips were reported for the River Bend Station in LER 87-033 indicating that the previous corrective action was not entirely effective. Therefore ,

investigations are continuing with General Electric. The results of this investigation will be provided in the followup of LER 87-03 Trip of EJS*SWG2A Breaker Due to a Faulty Output Transistor - The licensee replaced the output transistors in all the affected ITE 50D relays. During this replacement, it was determined that ITE 50H relays could be susceptible to a similar failure. Therefore, the transistors were replaced in nine ITE 50H relay Failure to Declare HPCS Inoperable - Human factor mimics were installed on the nuclear steam supply system panels to provide a visual relationship between the master trip units and their slaves. Licensed operator training en Rosemont + rip units and LER 86-054 was complete on April 10, 198 RCIC Isolation on Apparent High Steam Flow - This anomaly was first noted in November 1985. At that time, it was determined to be caused by a water column forming inside the

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5-sensing line during operation of the reactor core isolation cooling-(RCIC) system. As an interim measure, the instrument setpoints were lowered. An instrument piping modification was performed _in October 1986 and the setpoints were returned to-their normal values. As an interim action for LER 85-067, the licensee again lowered the instrument setpoints. During the

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1987 refueling outage, the sensing line routing was altered to correct the water bui', dup proble ,86-068 RCIC Isolation Due to a Failed Pressure Transmitter - The transmitter was replaced. The failed transmitter was returned to the manufacturer for a failure analysi The vendor found a broken seal between the sensor and the electronic head. Field engineering and the vendor agreed that moisture in the electronics would cause the reported condition. The licensee reviewed the plant maintenance and surveillance procedures to assure agreement with the vendor recommendations and to assure adequate instructions to prevent seal damag '87-003 Reactor Scram Due to a Feedwater Valve Misoperation - The operator involved was counseled and all shift crews reviewed the '

details of the event prior to assuming a shift assignmen Additionally, the administrative controls now include the assignment of an extra senior reactor operator (SR0) to the control room during startups. Licensed operators received requalification training on feedwater level control. A design modification was implemented which provides feedwater control valve position indications on the main control pane Division I Diesel Generator Output Breaker Failure - The licensee inspected'and tightened the bolts on safety-related '

breakers of the same type. Additionally, thread adhesive was applied to the charging motor bolts. To verify the automatic '

closing capibility of these breakers, the licensee has issued a standing orcer that includes daily visual inspection of the closing spr ing mechanical charge indicator on all 20 safety-related 4.16 kv circuit breaker .

,87-007 Annulus Preisure Outside of The Technical Specification  :

Limits - 7 e setpoints of the controllers were changed to allow l operation fa the automatic mode of operation and provide an l alarm prior to exceeding the technical specification limi Reactor Scram on High Level Setpoint Due to Feedwater Regula+.ing Valve Lockup - Station Operating Procedure S0P-0048 and corrective maintenance procedure CMP-1056 were revised to include placing the battery inverter in the manual bypass mode !

prior to trouble shootin Missed Gas Sarrple Due to Inadequate Communications - The volume l

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-6-of the gaitronics system in the chemistry laboratory was increased. The licensee reinforced operator training to assure that appropriate sections are notified following plant transients that significantly change plant condition,87-015 Scram Initiation While Transferring'RPS Power Due to an IRM Spike Performance of surveillance on August 18, 1987, revealed no abnormalities. During the startup of August 20, 1987, observation of intermediate range monitor operability revealed no noise problems. There was some indication of welding activities near the preamps that could have caused the IRM

. instabilit Mispositioned Instrument Valves Rendered Leak Detection System Inoperable - LER 87-017 involved a reactor core isolation cooling (RCIC) transmitter instrument valve and a scram discharge volume (SDV) isolation valve for a float switch being found in the closed position. . General maintenance procedure GMP-0042 was revised to clarify the definition of independent verifier. The verification is in addition to and independent of the system restorer. Instrument valve lineups servicing safety-related instruments and all other instruments contained in the technical specifications were verified prior to startu Deficient Tracking of Diesel Fuel Oil Surveillance - Fuel oil analysis requirements have been added into the River Bend

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Station surveillance test proctdure tracking program. The licensee initiated a ' review of chemistry procedures and procedure CSP-0100 was revised. The subject of deficient diesel oil surveillance was addressed as violation 458/8720-05 which was closed in NRC inspection report 88-0 Fire Seal Not Installed in Spent Pool Cooling Pump Cubicle Wall - These seals were not included on the construction drawings. The associated penetrations were sealed and the licensee added the penetrations to the applicable design drawing Residual Heat Removal System Isolation Due to Procedural Error - STP 051-4209, STP 051-4210, and STP 051-4211 were corrected to eliminate the improper sequence steps which caused this event.

.87-023 Radiation Monitor Heat Exchangers Plugged With Corrosion Due to Service Water Chemistry - Chemical cleaning of the lines was only partially successful. The licensee replaced the carbon steel inlet and outlet piping with stainless steel pipe. To address the impact on other safety-related components associated with the service water system, the licensee has instituted a program to improve the overall control of the service water chemistry.

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-7-87-024 Reactor Protection Syst'em Initiation Due to_ Personnel Communications - Immediate action was-taken' to revise open job orders that pertained.to work involving more than one logic channel by adding a step requiring the performer to date and initial after ensuring applicable. work was completed on one channel _and the logic reset prior to starting work on another channel. The assistant plant manager for maintenance issued memorandum APM-M-88-28 which issued directions for all the plant

. staff regarding utilization of and modifications to the scram prevention shee Valid Failure of Diesel Generator Due to Output Breaker Failure - The site procedures for racking circuit breakers was compared with the vendor instructions. As a result, operating procedure' SOP-0046 was revised by TCH 87-1256 by adding a caution statement'to emphasize the importance of proper circuit breaker racking and te provide additional clarification to the procedur Missed Fire Door Surveillance Due to Error in Test Procedure - Surveillance procedure STP-0000-3001 was revised to add the subject fire door Additionally, a review of all other fire door STPs against design documents was performed. This revealed that STP-0000-3401, "Semi Annual Fire Door Operability check," also omitted the same fire door Procedure STP-0000-3401 was revise Residual Heat Removal System Isolation Due to High Differential Temperature - The initial corrective action included returning the RHR A pump room cooler to service, resetting the isolation signal, and returning RHR A system to service. SOP-0031,

"Residual Heat Removal System," was revised to state that the respective unit cooler (s) should be in service prior to starting-a RHR pump for shutdown coolin ~ Residual Heat Removal System Isolation Due to Inadvertent Jumper Grounding - The blown fuse was replaced and RHR A was returned to servic The licensee instituted shop training on the use of proper tools and equipment (especially jumpers) when performing testing and maintenance for instrumentation and control technician Manual Reactor Scram Due to Control Rod Drive Trip - Deficient Procedure - SOP-0002 was revised to clarify the step which was improperly performed. Other procedures used during the current startup were reviewed for similar deficiencie The licensee responded to Violation 458/8729-01, "Failure to Follow Procedures for Control of Locked Valves," which was closed in NRC Inspection Report 458/88-08. Operations personnel were also reminded, by memorandum, of their responsibility to control evolutions and follow procedures.

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- Followup on Part 21 Reports Part 21 reports are received by the licensee from a variety of sources, most often from the vendors and the resident inspectors. The following Part 21 reports are considered closed on the basis of the NRC inspector's review of licensee records.and discussion with personnel: B&B Promatec - Fire Protection Penetration Seals - dated February 17, 1986. Licensee tracking number PRC 86-003. The licensee determined that the seal configurations described in the Part 21 report are not utilized at the River Bend Station.-

.o 87-001 - Brown Boveri - Battery Ground Detector Relays - dated November 7, 1986. The licensee identified ten 125V DC ground detectors with ITE 278 relays and that the recommended interim actions had been previously implemented. MR 87-0294 was initiated for implementation of permanent corrective action which consists of the reconsnended replacement of two resistors of each circuit boar (Licensing track No. PRC 86-28)

o 87-012 - Atwood Morrill Company - Main Steam Isolation Valve Return Springs - dated June 13, 1986. The licensee inspected the MSIV return springs and no cracking was identified. Testing of the MSIV revealed no abnormal condition (PRC86-13)

o 87-014 - Transamerica Delaval - Lube Oil Sump Tank Foot Valve Liner - dated March 10, 1986. Subsequent evaluations by Transamerica Delaval revealed that the River Bend Station was not subject to the potential problem. River Bend has relief valves in the piping to protect the system from overpressurization. (PRC86-06)

o 88-003 - Limitorque Corporation - Defective Wire Lugs on Valve Motor Operators - All Limitorque SMB 4 Valves were inspected. Five of ten had oversized wire lugs. GSU reported this Part 21 reporting matter to the NRC. The licensee replaced the defective wire lugs and revised their receipt inspections for limitorque operator (PRC87-66)

No violations or deviations were identified on this area of the inspectio . Surveillance Test Observation During this inspection period, the resident inspector observed the performance of Surveillance Test Procedure STP-05-4247, "ECCS Reactor Vessel Pressure Low /SRV Actuation Instrumentation Monthly CHFUNCT, 18 Month CHCAL, 18 Month LSFT (B21-N068A, B21-N668A, B21-N669A, B21-N670A, B21-N617A, B21-N618A, B21-N697A, B21-N699A)," and STP-051-4505, "RPS/RHR Reactor Vessel Level-Low, level 3, High, level 8, Monthly CHFUNCT, (B21-N080A,B21-N680A,B21-N683A)."

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o STP-051-4247: This surveillance test procedure was performed on April 19,1988, to verify operability of the reactor vessel low pressure - low pressure core spray (LPCS)/ low pressure coolant inspections (LPCI) injection valve permission and the relief valve / low-low set function pressure actuation instrumentation. These monthly channel functional test requirements are established in the River Bend Station Technical Specification Section 4.3.3.1, Table 4.3.3.1.A.1.d and in Sections 4.4.2.1.2.a and 4.4.2. Prior tobeginningthisSTP,theinstrumentationandcontrol(I&C)

technicians obtain permission from the control operating foreman m

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(C0F) to perform the surveillance. The maintenance and test ,

equipment was verified to have been calibrated within the required time period. The required jumpers to prevent inadvertent safety relief actuation were installed and controlled in accordance with General Maintenance Procedure GMP-042, "Circuit Testing and lifted Leads and Jumpers." All as-left values were within the specified setpoint range, o STP-051-4505: This surveillance test procedure was performed on April 30,1988, to verify operability of the reactor protection system (RPS)/ residual heat removal (RHR), reactor vessel level 3 and level -8 instrumentation. This monthly channel functional test requirement is specified in the River Bend Station Technical Specifications Sections 4.3.1.1, 4.3.2.1, Tables 4.3.1.1-1. and 4.3.2.1-1.6.c. The I&C technicians received permission from the C0F prior to beginning the STP, The at-the-controls (ATC) was notified prior to the I&C technicians inserting the RPS half-scram as required by the procedure. The ATC operator was immediately notified when the half-scram could be reset. The licensee determined that all the acceptance criteria had been met and the test results were properly reviewed by the C0 No violations or deviations were identified in this area of inspectio . Maintenance Observation During this inspection period, the resident inspector observed corrective maintenance activities for the repair of narroo range level instrument 1821*N680D which failed the channel check on April 19, 1988, and again on April 27, 1988. Prompt maintenance work orders (PMW0s) R056151 and R056173 were initiated, respectively, to evaluate and correct the indicated level deviation on the "0" channel, o PMWO-R056151: The narrow range level instrument 1821-N680D was noted to have devfated from the other three channels by nine inches during the performance of Surveillance Test Procedure STP-000-0001, "Daily Operating Log." The channel was placed in the trip condition within one hour as required by the River Bend Station Technical Specifications. Calibration of the level transmitter 1B21-LTN080D was checked and found to be acceptable. The "D" channel reference leg was then backfilled in accordance with Maintenance Corrective ,

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-10-Procedure MCP-4191, "Reactor Vessel Reference, Log 1821-TKD004D Isolation, tilling, and Purge." This restored the level indication-to within the acceptance band for the narrow range level instrumen The instrument was declared operable following successful completion of STP-051-4208, "RPS/RHR Reactor Vessel Level-Low Level 3; High Level 8, Monthly CHFUNCT,18 Month LSFT (821-N080D, B21-N680D, B21-N6830)."

o PMWO-R056173: The narrow range level instrument 1821-N680D again was noted to have deviated from the other three channels by approximately nine inches during the performance of STP-000-0001. The licensee conducted a complete walkdown of the "0" level reference leg and identified a root valve with a small packing leak. The licensee was able to stop the leak by tightening the packing. The reference leg was backfilled in accordance with MCP-4191 and the channel returned to operable status following successful completion of STP-000-000 All required Technical Specification actions were initiated during the performance of this PMW0 within the required time period No violations or deviations were identified in this inspection are . Operational Safety Verification The resident inspector observed operational activities throughout the inspection period and closely monitored operational events. Control room activities and conduct were generally observed to be well controlle Proper control room staffing was maintained and access to the control room operational areas was controlled. Select shift turnover meetings were observed and it was found that information concerning plant status was being covered in each of these meetings. System walkdowns of the "A" and

"C" low pressure coolant injection and the standby liquid control systems were conducted to verify major flow path alignments, proper breaker alignment and control board status for operability. Plant tours were conducted and overall plant cleanliness was goo General radiation protection practices were observed and no problems were notad. Personnel exiting the radiation control area were observed and radiation monitors were being properly utilized to check for contaminatio Contaminated areas are being well controlled and the total area that is considered contaminated is being reduce Security activities at the primary access point were observed on several occasion In each case, security officers verified that each individual entering the protected area (PA) had adequately cleared the security detection aides and had access to the protected area. Packages to be brought into the PA were also cleared prior to being release ,_--

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-11-The resident inspector also reviewed licensee actions on operational events and potential problems. The results of reviews of selected items are described below: Inadvertent Activation of Emergency Sirens: At 11:15 am CDT on April 6, 1988, the emergency sirens activated in West Feliciana Parish. 'The siren activation was caused by a personnel error during electro-magnetic interference (EMI) testing of the system at the West Feliciana Parish Emergency Operations Center (WFE0C). A siren "All Call" was initiated when a radio frequency (RF) choke came loose and grounded control. leads on an adjacent terminal board. The Louisiana Office of Emergency Preparedness was notified within one minute of the event and a message was broadcast over the emergency broadcast system at 11:18 am CDT stating that the siren activation had been inadverten On April 21, 1988, the resident inspector met with licensee emergency preparedness personnel to review corrective actions that have been and that are to be taken to prevent future inadvertent activation of the emergency siren (1) The following corrective actions have been completed:

o formalize Emergency Preparedness Procedures .EPP-2-701,

"Prompt Notification System Parish E0C Control Disable Procadure," and EPP-2-702, "Prompt Notification System Individual Siren Disable Procedure," to require sign offs as each step is performed; o require arming the sirens at the River Bend Station Emergency Operating Facility (E0F) before they can be actuated at the different parishes' E0Cs; o physically separated the base station and microprocessor at WFE0C to reduce the effects of EMI; and o improved grounding and RF chokes at the WFE0 (2) The following corrective actions are expected to be completed by August 1988:

o improve shielding and grounding at the remaining parishes'

E0Cs; a a modify the 92 sirens to reduce their susceptibility to individual activatioris because of lightening strikes l naarby.

(3) The following corrective actions are expected to be completed by

, October 1988:

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-12-o . install an upgraded computer with the hardware capability to provide a fully redundant siren control and monitoring system;'and o complete testing of the associated software which will provide for "safe" failure modes to prevent inadvertent siren activatio It should be noted that the monthly siren test will still be conducted on the first Wednesday of each month at 10 am central tim The licensee's actions to prevent future inadvertent siren activations will _be an open item pending review of the completed corrective actions by the NRC inspector. (458/8812-01) Pressure Regulator Induced Transient: On April 29, 1988, at approximately 1:30 pm (CDT), with the plant at 100 percent power, the Pressure Regulating System (PR) automatically switched from process circuitry Channel A to Channel B. The switch was initiated by a momentary decrease of the steam pressure reference signal. This resulted in a decreased error signal when compared to the pressure set point, causing the main turbine throttle valves to begin closin The steam reference signal was completely rettored within one second, however, the processing circuitry sensed the decrease in the steam -

flow signal from Channel A and initiated the automatic switch to Channel B. The steam pressure reference signal is common to both channels. The reactor responded as expected with a maximum neutron flux reaching 116 percent caused by the increase in moderator inventory in the reactor core as voids collapsed with the increase in reactor vessel pressure. Calculated maximum core thermal power was 104 percent for less than one second. The licensee reviewed the reactor protection system (RPS) response to the transient and concluded that no RPS activation should have occurred. The plant response was also determined to be within the boundaries identified in Chapter 15 of the Updated Safety Analysis Report for. a pressure regulator failure. The licensee is examining the potential causes for the sudden loss, then restoration of the steam pressure reference signal. General Electric electro-hydraulic control specialist personnel have been contacted to help in the licensee's review of this even No violations or deviations were identified in this area of the inspectio . Exit Interview An exit interview was conducted with licensee representatives (identified in paragraph 1). During this interview, the resident inspector reviewed the scope and findings of the inspectio J