IR 05000458/1987002

Insp Rept 50-458/87-02 on 870101-0215.Violation Noted: Inadequate Procedures for Control of Radioactivity

ML20212A202
Person / Time
Site:	River Bend
Issue date:	02/23/1987
From:	Chamberlain D, Jaudon J, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20212A140	List: IR 05000458/1987002 ML20212A136 ML20212A170
References
50-458-87-02, 50-458-87-2, NUDOCS 8703030307
Download: ML20212A202 (12)

Text

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APPENDIX B -

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U. S. NUCLEAR REGULATORY COMMISS10N

REGION IV

NRC Inspection Report: 50-458/87-02 Docke't: 50-458 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: January 1 through February 15, 1987

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Inspectors: o[

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.7 D." Chamberlain, Senior Resident Inspector 247-87 Date Project Section A, Reactor Projects Branch jJ E eno a 2- 17- Pil W. B. Jones, Resideys Inspector Date Project Section A, actor ojects Branch Approved: _ 424/ 1 8 J./P./audon, Chief,ProjectSectionA Dat*e /

geacitor Pr jects Branch

G703030307 B70224 PDR ADOCK 05000458

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Inspection Summary Inspection Conducted' January l'through February 15, 1987 (Report- 50-458/87-02)

Areas' Inspected: Routine, unannounced inspection of licensee action on previous ,

inspection findings, Licensee Event Reports'.(LERs)', IE Notices, IE Bulletin 86-01,

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l maintenance witnessing,. safety system walkdown.. surveillance test witnessing, operational safety verification and allegation followu Results: Within the areas inspected, one violation was identified'(inadequate procedures for control of radioactivity, paragraph 9).

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DETAILS-s Persons Contacted

• L. Andfews, Director, Nuclear Training

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• R.-J. Backen, Supervisor (Acting), Operations Quality

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Assurance (QA)- '

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• D.,H. Barrow,' Board:of Directors

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.. W. J. Beck,; Supervisor,~ Reactor Engineering

• J.NE. ~ Booker, Manager, Oversight

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_ *J..L. Burton, Supervisor, Independent Safety

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Engineering Grcdp *

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• W. HC Cahill', Jr. , ; Senior Vice President, Special 7 ' d Projects- v4

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^ < *E.!M. Cargill, Supervisor, Radiological Protection

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3*J.-; W. Cook,:-Lead Environmental Analyst

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"fT. C.i Crouse;' Manager,?QA f

, *J.~C.' Ddddens,' Senior,Vice_ President, River Bend

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--D.-R.".Derbonne,S0peNisor,GeneralMaintenance

• L. A.1 England, Su~pervisor, Licensing

• P. E. Freehill, Outage Manager A. O. Fredieu, Assistant Supervisor, Operations

• K. J. Gladrosich, Supervisor, Operations Quality Control (QC)

'*D. R.7Gipson, Director, Quality Services

• J. Do Gore,' Consultant .

. P. Graham, Assistant Plant Manager, Operations

• E. R. Grant, Director, Nuclear Licensing

• J. R. Hamilton, Director, Design' Engineering I

• Harris, Director, Joint Operations R. . W. Helmick, Director, Projects .

• K. C. Hodges, Supervisor, Quality Assurance Engineering

• B. Kimmell, Supervisor, Operations QA

• J. King, Supervisor, Nuclear Licensing A. D. Kowalczuk, Assistant Plant' Manager, Maintenance

• J. Krueger, Supervisor, Engineering Administration

• E. M. Lambremont, Consultant J.-W. Leavines,-Director, Field Engineering

• M. Malik, Supervisor, Quality Systems

• J. H. McQuirter, Licensing

• T. O. Moffitt, Engineer ,

• C. L. Nash, Supervisor, Technical Services / Chemistry

• V. J. Normand, Supervisor, Administrative Services

• H. Odell, Manager, Administration

• T. H. Pigford, Consultant

• T. F. Plunkett, Plant Manager

• J. J. Pruitt, Director, Management Systems

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i F. Sankovich, Manager, Engineering

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• *J. E. Smith, Consultant R. B. Stafford, Director, Operations QA

• K. E. Suhrke, Manager, Project Management

• D. W. Williamson, Supervisor, Operations

• H. H. Woodson, Consultant .

E. J. Zoch, Supervisor, Design Engineering The NRC senior resident inspector (SRI) and resident inspector (RI) also interviewed additional licensee personnel during the inspection perio * Denotes those persons that attended the exit interview conducted on February 13, 1987. NRC regional inspectors J. B. Nicholas, G. A. Pick, R. C. Stewart, and R. Wise also attended the exit intervie . Licensee Action on Previous Inspection Findings (Ciosed) Open Item (458/8608-03): Monitor licensee actions to develop a consistent criteria for locked valve Subsequent to the identification of this open item, a violation (458/8620-02) was identified which included the lack of consistency for locked valves. The licensee response to the violation included development of a procedure to control locked valves and revisions to piping and instrument diagrams to specify locked valves required by regulatory requirements. Final verification of corrective action will be documented during closure of Violation 458/8620-0 This open item is close (Closed) Open Item (458/8608-04): Monitor licensee actions to correct piping and instrument diagram (P& ids) valve position designation discrepancie Subsequent to the identification of this open item, a violation (458/8620-02) was identified which included the valve position designation discrepancy problem. The licensee response to the violation stated that a note would be added to P& ids stating that valve positions shown on the P& ids are for the normal mode of operation and are for information onl The valve positions indicated by the P& ids are not to be used for actual valve lineup Station Operation Manual procedures will govern actual valve positions. Final verification of this corrective action will be documented during closure of Violation 458/8620-0 This open item is close . Licensee Event Reports (LERs) Review During this inspection period, the SRI reviewed LERs for compliance with requirements established in 10 CFR Part 50.73. Specifically, the LERs

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wererev'iewehfor'accuracyandclarityoftheeventdescription,thecause of each. component'or system failure or personnel error, the failure mode

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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. Completion of corrective actions for selected.significant events was also verifie The following LERs were reviewed and closed:

86-013 Technical Specification Noncompliance -

Procedure Deficiency 86-064 RHR Isolation Due to a Grounded Power Supply 86-069 Reactor Scram Due to High Stator Cooling Temperature The above listed LERs are close No violations or deviations were identified in this area of inspectio . Licensee Action on IE Notices This area of inspection was conducted to review licensee actions related to selected NRC Inspection and Enforcement (IE) Information Notices. The results of the review are documented below: IE Notice 86-82 Failures of Scram Discharge Volume Vent and Drain Valves: This notice identified a problem with Hammel-Dahl valves used as scram discharge volume vent and drain valves. The problems occurred when the valves were operated with the manual handwheels partially engaged. The licensee took corrective action to verify that the Hammel-Dahl vent and drain valves handwheels were locked in the neutral position. Associated drawings and operating procedures were revised to reflect the requirement to lock the handwheels in the neutral positio No other valves of this type were identified by the licensee at River Ben This IE Notice is close IE Notice 86-89 Uncontrolled Rod Withdrawal Because of a Single Failure: This IE Notice was issued to alert recipients of a potentially generic problem with a single failure that could cause both a single control rod drift to the full-out position and tnen failure to insert on demand. The licensee has not experienced this problem at River Bend but the alarm response procedure for a rod drift was revised to include similar actions as identified by the IE Notice with the exception of having someone manually scram the rod

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Tne licensee believes that this is outside of normal operating practices and should only be done at

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This IE Notice is closed, IE Notice'86-109 Diaphragm Failure In Scram Outlet Valve Causing Rod Insertion: - This IE Notice was issued to alert addressees to the potential for a failure of the diaphragm in the scram . inlet or outlet valve operator _on the control rod drive hydraulic control unit _

This could_ result in a single control rod scram while at_ powe Investigation at the affected facility revealed that the diaphragm failed because of an aging process that resulted in a radial crack in the rubber (buna-n and nylon material). -The licensee has not completed their investigation of this IE Notice, but preliminary review revealed that a preventive maintenance requirement exists at River Bend to replace the subject valve diaphragm on a seven year cycle as recommended by the manufacture This IE Notice will be closed when the licensee. investigation is complete IE Notice 86-106 Feedwater Line Break: This IE Notice was issued to alert addressees of an event at the Surry Power Station where the 18-inch suction line to the main feedwater pump A for Unit 2 failed catastrophically. The failure was attributed in part to pipe wall thinning erosion. The licensee has not completed their investigation of this IE Notice, but the preliminary review revealed that a pipe wall thinning inspection program was alrudy planned at River Bend for the first refueling outage on selected piping system This program did not include the feedwater pump suction lines. The licensee initial review indicates that certain factors contribute to erosion / corrosion on inner pipe surfaces. These include high velocity compounded with duration of service, piping configuration and water chemistr The River Bend design has been compared with the Surry design and the licensee has determined that there is no immediate need for inspection of the feedwater pump suction piping at River Bend. This determination is based on lower fluid velocity (10.5 versus 1 feet /second), greater piping wall thickness'(.59 versus .5 inches),

- shorter duration of service (1 year versus 13 years) and a more favorable piping configuration (45 dagree elbow transitions versus 90 degree elbow transitions). While no immediate need for inspection is deemed necessary, River Bend design engineering will recommend a pipe wall thinning inspeci,!qn of the feedwater pump suction piping at the second refueling outage. The licensee is continuing their evaluation of this IE Notice for other fluid systems to add to the pipe wall thinning inspection program. The specific criteria or inspection

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locations have not been developed at this time. This IE Notice will remain open pending licensee finalization of the pipe wall thinning inspection program at River Ben No violations or deviations were identified in this area of inspectio . Licensee Action on IE Bulletin 86-01 The NRC issued I&E Bulletin 86-01 (IEB 86-01), " Minimum Flow Logic Problems That Could Disable RHR Pumps." It was discovered at the Pilgrim Nuclear Power Plant that a single failure under certain accident sequences could result in all the residual heat removal (RHR) minimum flow bypass valves being signaled to close while all other pump discharge valves are I

still close This condition could result in no flow through the RHR pump with subsequent damage to the pump within several minutes if not detecte The licensee has completed their review of IEB 86-01 and determined that the single failure vulnerability problem described in the bulletin does not exist at River Bend Station (RBS). The design at RBS provides for separate control of each RHR pump minimum flow by pass valve so that all three low pressure coolant injection loops are completely independen This-response was submitted to the NRC within the required 7 days from receipt of'this bulleti No violations or deviations were identified in this area of the inspection. This completes the inspection requirements of Temporary Instruction (TI) 2515/8 This IE Bulletin (86-01) is close " Maintenance Witness

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On January 14, 1987, the RI witnessed corrective maintenance activities for the replacement of the RHR, A Suppression Pool Suction Valve 1E12*MOVF004A motor operator. This valve failed to open during attempts to realign the RHR system from the shutdown cooling (SDC) mode to

'the low pressure-coolant injection (LPCI) mode. The licensee has

. postulated that the motor operator failed because of the temperature differentialthatexistsacrossthe1E12*MOVF004Avalvediscswhenfnthe SDC mode. Thistemperaturedifferentialexisgedbecauseofthe165F

' reactor water on the outboard disc and the 75 F suppression pool water on

,the inboard disc. The licensee is presently evaluating this condition to

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determine what corrective actions can be taken to alleviate this condition. The'NRC resident inspectors will continue to monitor the licensee's actions in this are The licensee conducted the above maintenance activity in accordance with approved Prompt Maintenance Work Order (PMW0) Request 87-5507 The RI noted that the appropriate limiting condition of operation (LCO) and clearance was initiated prior to beginning wor During the performance of the maintenance activity, a QC inspector was present to observe the

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work and also observed the subsequent testing of the valve prior to restoring the valve to an operable status. The RI reviewed the PMWO close out documentation and no problems were identifie No violations or deviations were identified in this area of the inspectio . Safety System Walkdown During this inspection period, the SRI and RI walked down the Division II and Division III emergency diesel generators to verify proper system alignment in accordance with station operating procedures 50P-053,

" Standby Diesel Generator and Auxiliaries," and 50P-052, "HPCS Diesel Generator." The above two diesel generators are required to be operable by Technical Specification (TS) with the plant in Operational Condition During the walkdown, it was observed that:

. System valves were properly aligned

. No abnormal control room or local panel instrumentation readings or alarms were present which would affect the diesels operability

. Accessible hangers, supports and snubbers were intac The NRC inspectors also observed deficiency tags on the Division II and Division III diesel generators which identified leaks on the air start and lube oil systems. Although these defic' ncies would not prevent the diesel generators from operating as rc; . red, several of these de_f.iciencies have.been allowed to exist for about a year. The licensee has undergone at least two extended outages over the past year during which time the above deficiencies could have been repaired. Several alarms,were also present on the Division II diesel generator local alarm panel during this walkdown. Although the alarms that were present did not indicate that the diesel generator was inoperable, many of the alarms have been present since the end of the startup program. The licensee has initiated design changes to clear the alarms; however, this work has not been completed. During the NRC exit meeting conducted with the licensee's staff, the SRI reemphasized the importance the Coalmission places on maintaining safety-related equipment and the need to identify and resolve alarms in a timely manner. The NRC inspectors will continue to monitor licensee's actions in these areas to assure tt.at identified deficiencies and alarms are being addressed and resolved in a timely manne No violations or deviations were identified in this area of inspectio . Surveillance Test Witness During this inspection period, the RI observed the performance of the following surveillance tests and reviewed the data packages for each of

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these tests. The following surveillance tests were observed:

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9 Surveillance Test STP-509-0101 - The RI observed the performance of STP-509-0101, " Main Turbine Bypass System Valve Cycle Test,"

conducted on January 30, 1987. This surveillance test implemented the 7 day surveillance requirements of TS 3/4.7.9.a. which demonstrates the operability of each of the turbine bypass valves by cycling each valve open then closed. This surveillance was performed with reactor thermal power and turbine generator at 80 percent of rated power. During the performance of this test, the RI noted that each operator was made aware of each step as it was performed which could affect other balance of plant systems. After each step was completed, the plant was allowed sufficient time to stabilize before proceeding with the surveillance. Review of the completed STP documentation did not reveal any problem Surveillance Test STP-110-0101 - The RI observed the performance of STP-110-0101, " Turbine Overspeed Protection System Weekly Operability Test,"' conducted on January 31, 1987. This surveillance test implemented the requirements of TS 3/4.3.8.2.a. by demonstrating the operability of each high pressure turbine stop valve and control valve and each low pressure turbine interinediate stop valve and intercept valve. During the performance of this test, the control operatingforeman(C0F)coordinatedtheactivitiesbetweentheatthe controls operator (ATC) and the unit operator (U0) to assure that no unerpected plant transients were induced. The RI noted that the ATC operator reviewed each section of the STP and was then able to proceed rapidly through each section, reducing the time the plant was in an off-normal configuration. No probleir.s were identified during the review of the completed STP packag Surveillance Tes t STP-0209-3302 - The RI observed the performance of STP-0239-3302, "RCIC Pump Operability and Flow Test," on February 5,1987. This surveillance test satisfies the requirement of TS 4.7.3.6 for (RCIC) pump flow greater than or equal to 600.gpm with a reactor steam pressure of 1025 (+ 25,-100) psig. During the initial RCIC pump start, the RCIC turbine tripped on overspee After resetting the mechanical overspeed trip unit at the turbine, the RCIC pump was restarted without further incident. This event has been identified on Condition Report (CR) 87-0169. The NRC resident inspectors will review the closecut of this CR during subsequent inspections. During operation of the RCIC turbine, the RI r.cted that the RCIC Turbine Exhaust Check Valvy 1E51*F040 was chattertug during low flow steam flow conditions. Di m ussion with the licensee revealed that they were aware of the chatter condition and had incorporated precaut!on statements into this STP and the applicable station operating procedure cautioning against running the RCIC pump at low rp flo violations or deviations were identified in this crea of inspection.

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9 .' Operational Safety Verification The resident inspectors observed operational activities throughout the inspection period and closely monitored operational events. Control room activities-and conduct were generally observed to be well controlled and efficien Proper control room staffing was maintained. While access to the control room operational areas was generally controlled, some congestion during a unit startup and during other high traffic time periods was noted by the licensee. The licensee has reemphasized control room access control and additional methods to control personnel traffic and access in the control room are being investigated. The licensee initiated a new schedule for shift technical advisor (STA) coverage during this inspection period. Each STA now covers a 24-hour period on site every 20 days. Special quarters have been provided on site for the assigned STA where he remains on call when he is not present in the control room or in other plant area A walkdown of the Division II and III emergency diesel generators was conducted and the results are documented in paragraph 7 of this repor Plant tours were conducted and overall plant cleanliness was good. During these plant tours general radiation protection and security activities were observed. It was noted during a walk thru of the RHR pump rooms that certain piping runs measured greater than a 100 MR radiation dose rate on contac This was discussed with radiation protection supervision and iminediate action was taken to post the areas where dose rates were significantly higher than the general area. The areas noted did not meet the licensee's criteria for posting hot spots, but it has been their general practice to post such areas. Also, while exiting the site on February 9,1987, the SRI observed at least two personnel who exited the protected area through the portal monitors without resetting the portal monitor when it alarmed the first time and without repeating the walk through. The security officer on duty did not stop the personnel and require them to rcpeat the walk through, and the exit turnstiles were not locked. Upon further investigation by the SRI, it was found that neither security procedures nor radiation protection procedures required resetting the portal monitor, repeating the walk through, and/or notifying radiation protection when the protected area exit portal monitors alar The apparent inadequacy of procedures for control of radioactivity was identified as an apparent violation (458/8702-01). The licensee took immediate action to initiate strict controls on personnel exiting the site and procedures are being revised to include necessary controls. The SRI recalled that the licensee's general employee training (GET) lectures had covered this type of even During this inspection period, the SRI monitored any potential effects of the licensee financial condition on the safe operation of River Ben In a December 15, 1986, letter from GSU to the Director of the Office of Nuclear Reactor Regulation, US NRC, GSU emphasized that safety would not be compromised to control expenditures. They cited staffing, training and maintenance as specific areas that are not being compromised. The SRI also monitored these general areas and the only potential effect noted was

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an increase in personnel turnover which possibly could be indirectly attributed to the financial condition of GSU. However, this increased turnover of personnel has not had any apparent effect on the safe operation of River Bend. The personnel turnover at River Bend will be monitored during future NRC inspection The resident inspectors also reviewed licensee actions on operational events and potential problems and observed a 1icensee conducted emergency dril The results of selected items are described below: Reactor Scram Caused by Feedwater Level Transient: During a reactor startup on January 16, 1987, with reactor power at approximately 20 percent, a reactor scram occurred on high reactor water leve The high reactor water level occurred when the reactor operator attempted to place the 'A' main feedwater regulating valve in servic The operator opened the feedwater regulating isolation valve with the regulating valve approximately 47 percent open. This caused a rapid increase in water level and the subsequent reactor scram. The -

operator apparently did not follow procedures for local verification i of regulating valve position nor did he acknowledge control room indication prior to opening the isolation valv The licensee has reviewed the potential causes of this event and plans to require tighter controls during future reactor startups including a more direct involvement of the C0F. This area will be monitored by the SRI and final closure of this event will be documented during routine Licensee Event Report revie Emergency Drill: On January 28, 1987, the licensee conducted an in-house emergency drill in preparation for the upcoming, February 25, 1987, emergency drill which will include local, state, and federal participation. The drill included simulated notifications as the severity of the accident escalated and an evacuation of nonessential site personnel from the protected are Selected individuals were sent to the off-site assembly area as a personnel accountability exercise. The local fire department participated in the drill by dispatching a fire truck and ambulance when called. Our Lady of the Lake Regional Hospital also participated by treating an individual with a simulated injury. The RI observed this drill and the drill was deemed effective to prepare for the upcoming drill in February 198 . Allegation Followup (4-86-A-092)

This area of inspection was conducted to review a concern received by NRC Region IV relating to potential temperature problems where high temperature lines penetrate the containment and shield building. The specific concern stated that a nonconformance had been issued in October or November of 1985 to address the temperatu- 'ffect on rubber boot seals attached to the guard pipe at the main steam sel, but the nonconformance failed to address the potential mperature effect on the concrete structural integrit The SRI revieweu copy of Nonconformance i

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Report (NR) 85NR0582-00 initiated on November 20, 1985. The NR referenced Calculation 219.710-EBC-2031 and also included certain temperature readings taken in the annulus area'on December 6, 1985. The conclusion reached based on the calculation and the. temperature readings was that neither the boot material nor the concrete would be affected by the seal design. The SRI review of the NR and referenced calculation did r.ot clearly indicate that adequate testing or analysis was documented to support the stated conclusion. The referenced calculation was completed in 1977 and addressed the temperature effect on the drywell concrete only-for different types of penetration It did not address the specific shield building penetrations or the potential effect on the rubber boot materia The temperature readings taken on December 6, 1985, in the annulus area were taken one and one half hours after a plant shutdown and during that time ventilation systems had been running which would have cooled the annulus area. Licensee representatives stated that actual temperature readings of the boot material had been taken with the plant at power, but no documentation was presented to the SRI during this inspection perio The adequacy of the testing and analysis for closure of NR 85NR0582-00 will be unresolved (458/8702-03) pending further review by the SRI. This allegation will remain open pending resolution of the unresolved item

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1 Unresolved Item m An unresolved item is"one about which additional information is required

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in order to determine if it is acceptable, or a deviation, or a violatio There is one unresolved item in this repor '

Paragraph ' Item N Subject

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10- '458/8702-03 Adequacy of Testing and Analysis for Closure of Nonconformance

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, 85NR0582-00

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. 12$ Exit and Inspection Interview

'An exit interview was conducted on February 13, 1987, with licensee

' representatives 1(identified in paragraph 1). During this interview, the SRI reviewed the, scope and findings of the inspectio _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ - - _ _ _ _ . _ - _ _ .