IR 05000382/1987022

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Insp Rept 50-382/87-22 on 870916-1015.Violations Noted. Major Areas Inspected:Followup of Previously Identified Items,Ler Followup,Plant Status,Fire Protection,Tech Spec Adherence & Procedural Adherence to Attention to Detail
ML20236S504
Person / Time
Site: Waterford Entergy icon.png
Issue date: 10/27/1987
From: Jaudon J, Will Smith, Staker T
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20236S498 List:
References
50-382-87-22, NUDOCS 8711250215
Download: ML20236S504 (14)


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APPENDIX B l l

U. S. NUCLEAR REGULATORY COMMISSION  !

REGION IV

NRC Inspection Report: 50-382/87-22 License: NPF-38 Docket: 50-382

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Li censee: Louisiana Power & Light Company (LP&L)

142 Delaronde Street New Orleans, Louisiana 70174 Facility Name: Waterford Steam Electric Station, Unit 3 Inspection At: Taft, Louisiana Inspection Conducted: September 16 through October 15, 1987 Inspectors: /0!/6 97 W. F. Smith, Senior Resident Inspector Date

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Inspection Summary

. Inspection Conducted September'16 through October 15, 1987 (Report 50-382/87-22)

' Areas Inspected: Routine;: unannounced inspection consisting of: (1) followup of previously identified ' items. -(2)flicensee event report followup, (3) plant status, (4) monthly maintenance observation, (5)' monthly surveillance observation, (6) engineered safety feature system walkdown, (7) operational

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- safetyverification,(8)onsitefollowupofevents,(9)verificationof I containment' integrity, (10) fire protection, (11) Technical Specificatio adherence.and(12)proceduraladherenceattentiontodetai .

Results: .Within the areas inspected..three violations were identified (failure

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to adhere to procedural _ details,' paragraphs 5, 9.a.. and 10; failure to follow Technical Specification action statements, paragraph 9.c.; and failure to

- follow fire prevention procedures, paragraph 11).

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DETAILS l Persons Contacted Principal Licensee Employees )

J. G. Dewease, Senior Vice President, Nuclear Operations

  • R. P. Barkhurst, Vice President, Nuclear Operations
  • N. S. Carns, Plant Manager, Nuclear P. N. Backes, Corporate QA Manager S. A. Alleman, Assistant Plant Manager, Plant Technical Staff -

J. R. McGaha, Assistant Plant Manager, Operations and Maintenance l J. N. Woods, Quality Manager, Nuclear 1 A. S. Lockhart, Nuclear Operations Support and Assessments Manager R. F. Burski, Engineering Service Manager

  • E. Wuller, Onsite Licensing Coordinator D. W. Vinci, Maintenance Superintendent, Nuclear (Acting)

L. W. Myers, Operations Superintendent, Nuclear

  • Present at exit intervie i

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In addition to the above personnel, the NRC inspectors held discussions !

with various operations, engineering, technical support, maintenance, and administrative members of the licensee's staf . Followup of Previously Identified Items (0 pen) Violation 382/8602-03: Entry into operational Mode 3 (Hot Standby) I with containment spray (CS) system train "B" inoperable, contrary to i Technical Specification (TS) 3.6.2.1 and 3.0.4. This event was reported to the NRC in Licensee Event Report (LER)85-055, Revision 1, dated May 26, 1986, pursuant to 10 CFR 50.73. The LER is closed in paragraph 3 <

of this inspection report. The NRC inspectors verified that corrective )

actions had been taken as committed to in the licensee's response to the

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violation, LP&L letter W3P86-0077, dated May 16, 1987. The corrective actions described included procedure changes, changes to the appropriate annunciator windows to reflect the actual components affected, and attachment of warning tags to valve operators equipped with reach rods cautioning operators to ensure that the valves actually operate when reach rods are used. The NRC inspectors verified that the warning tags had been

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installed but noted that the tags for Valves CS-111A, SI-124A, SI-4178, CS-1178 and CVC-208 had apparently broken off. The licensee was notified and committed to replace them and check for other missing warning tag The NRC inspectors will verify replacement and maintenance of these tags during future routine tours. The implementation of the corrective actions taken serves to close this violation. Nevertheless, this item remains ,

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open pending NRC inspector verification that missing warning tags have been replace . .

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3. Licensee Event Report (LER) Followup  !

The following LERs were reviewed and closed. The NRC inspectors verified l that reporting requirements had been met, that causes had been identified, corrective actions appeared to be appropriate, generic applicability had been considered, and that the LER forms were complete. The NRC inspectors also confirmed that unreviewed safety questions and violations of Technical Specifications, license conditions, or other regulatory requirements had been adequately describe (Closed)LER 382/85-055: Disconnected reach rod operator resulted in the discharge valve of the containment spray pump being closed when mode change was mad Onsite verification of completed corrective actions was performed by the NRC inspectors and is documented under the discussion of Violation 382/8602-03 in paragraph 2 of this repor (Closed)LER 382/87-019: Hourly fire watch tour not performed due to personnel error, The NRC inspectors verified that Procedure FP-1-014,

" Duties of a Fire Watch," had been revised to improve the log sheets, to provide additional supervisory reviews and documentation, and to clarify instructions on the fire watch log change proces No violations or deviations were identifie . Plant Status The inspection period began with the plant at 100 percent power. On September 21, 1987, the unit was shut down and cooled below 200 F because of an excessively leaking shaft seal on Reactor Coolant Pump 2B. The outage took approximately 18 days, and the unit was restored to power on October 9, 1987. The unit returned to full power on October 11, 1987. On October 13, 1987, power was reduced to 60 percent to accommodate isolation of a portion of the main condenser water box for repair of possible tube or tubesheet leakage. As of the end of this inspection period, the unit was at 95 percent power, and the condenser leak appeared to have been repaire No violations or deviations were identifie . Monthly Maintenance Observation The station maintenance activities affecting safety-related systems and components listed below were observed and corresponding documentation was reviewed to ascertain whether the activities were conducted in accordance with approved procedures, Technical Specifications, and appropriate industry codes or standard .

Work Authorization 01003440, " Repair of Body to Bonnet Steam Leak on MS-1244, Main Steam Isolation Valve "A" Bypass Valve."

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. Condition Identification Work Authorization 032389, " Repair of Gasket Leak on FW-1763A, Emergency Feedwater Bypass Header "A" Check Valve."

. Work Authorization 01003521, " Rework of the "A/B" Emergency Feedwater Pump Mechanical Overspeed Trip Device."

. Work Authorization 01003978, " Disassembly of Charging Pump "B" Reduction Gear Box for Troubleshooting."

. Work Authorization 01003521 involved the replacement of the reduction gear shaft bearing for Charging Pump "B" and the reassembly of the gear box. The pump crankcase was also opened for inspection. During the reinstallation of the pump crankcase cover on October 1, 1987, the NRC inspectors noted that the quality control signoff for verification of cleanliness at closure was entered in the work authorization and cleanliness control fonn on September 30, 198 Procedure UNT-7-005, Revision 2, " Cleanliness Control" requires closure inspections to be performed "immediately prior to" system or component closure. This represents a failure to adhere to procedural requirements and is therefore a violation of the requirements of Criterion V of Appendix B to 10 CFR Part 50. (382/8722-01)

Personnel involved with this work stated that a cleanliness inspection was performed immediately prior to an attempt to reinstall the cover on September 30. The cover could not be reinstalled because the gasket did not fit. Work on the pump was secured until a second gasket could be obtained the next day, but the cleanliness inspection was not repeated at this time. After the NRC inspectors addressed this concern, additional work instructions were written to clarify cleanliness control . Monthly Surveillance Observation The NRC inspectors observed the surveillance testing listed below to verify that the activities were being performed in accordance with the Technical Specifications. Applicable procedures were reviewed for adequacy; test instrumentation was verified to be in calibration, and the test data was reviewed for accuracy and completeness. The NRC inspectors ascertained that any deficiencies identified during the surveillance were properly reviewed and resolve . The NRC inspectors witnessed performance of Surveill Oce ProcedureMI-3-362, Revision 5,"FuelHandlingBuildikgIsolation Radiation Monitor Safety Channel "A" Functional Test ARM-IR0 300.4,"

and reviewed the result No violations or deviations were identified during this portion of the inspection.

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-6- Engineered Safety Feb.u; (ESF) System Walkdown The NRC inspectors condul 1 a walkdown of the accessible portions of the Emergency Feedwater (EFW) system to independently verify the operability of the system. A review was performed to confirm that the licensee's system operating procedure matched plant drawings and the as-built configuration. Equipment condition, valve and breaker position, housekeeping, labeling, permanent instrument indication and calibration, and apparent operability of support systems es Hntial to actuation of the ESF system were all noted as appropriat The NRC inspectors identified the following items to licensee management:

. The EFW system standby lineup of Procedure'0P-9-003, Revision 5,

" Operating Procedure Emergency Feedwater," described Valve EFW-1062A as being the EFW pump "A" bearing cooling water inlet valv However, Valve EFW-1062A is the outlet valv . No identification tag was installed on Valve EFW-229 l

. Valves MS-403A and MS-4038 are shown locked open on Drawing LOV-1564-G-151. These valves were found open but not locke They were not included in the locked valve list contained in Procedure OP-100-009, Revision 5, " Control of Valves and Breakers,"

which is the licensee's requirement document for locked valve !

. Drawing LOU-1564-G-153 shows valves EFW-215A and EFW-215B as open only but not locked. These valves were found open, locked, and are included in the locked valve list contained in Procedure OP-100-009, Revision 5, " Control of Valves and Breakers."

. Valves EFW-2263A, EFW-2262A, and EFW-2262B are shown on Drawing LOV-1564-G-153 as FW-2263A, FW-2262A, and FW-2262 This incorrect valve identification was previously described in NRC Inspection Report 50-382/86-29. Licensee action to correct the valve labeling had not yet been complete . A Condition Identification Work Authorization (CIWA) tag, dated September 11, 1984, and stating "Pending CIWA," was found installed upstream of flow element FE-FW-833 There was, however, no other evidence of a modification or work in progres .

The circuit tsreaker locations for Valves MS-401A and MS-401B (EFW ,

pump "A/B" steam supply) in the EFW breaker lineup contained in '

Procedure OP-9-003 were both listed incorrectly as EPDP-A-DC-S-1 The breaker at location EPDP-A-DC-S-14 was an unused spare. The

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correct breakers for MS-401A and MS-4018 are EPDP-AB-DC-5-21 and 2 These breakers were found in the correct positio . Circuit Breakers EPDP-A-DC-5-20 and EPDP-A-DC-S-24 were both labeled as the power supply for Valve EFW-228. However, the breaker lineup

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-7- l contained in Procedure OP-9-003 correctly lists EPDP-A-DC-S-24 as the  !

power supply breaker for Valve EFW-228. Circuit

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Breaker EPDP-A-DC-S-20 was apparently incorrectly labele . Plant housekeeping controls were not being maintained per Procedure UNT-7-006, Revision 3, " Housekeeping" as evidenced by the following:

A sheet of plywood and a can of insect spray were found in the main steam and main feedwater isolation valve room An extension cord, several types of pipecaps and fittings, a block of wood, a digital multimeter, and several tools were found in the vicinity of the "A/B" EFW pum The plastic sheath installed on the flexible conduit at }

Valve MS-401B was cracked and separate Also, the conduit i identification label was worn so that the numbers could not be i rea The cover at the elbow on the electrical conduit to Valve EFW-228B was loos j

. As previously discussed in NRC Inspection Report 50-382/86-29,the individual EFW pump recirculation line isolation valves are locked open while the common recirculation header isolation Valve (EFW-205)

is not locked. Any mispositioning of Valve EFW-205 would isolate the i recirculation flowpaths for all EFW pumps. The EFW reliability ,

analysis contained in Appendix 10.4.9B of the Waterford 3 Updated )

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Safety Analysis Report (USAR) considered Valve EFW-205 to be locked open as shown in Figure 10.4.9B- Correction of the deficiencies listed above will be tracked under Open Item 382/8722-0 No violations or deviations were identifie . Operational Safety Verification  !

The NRC inspectors reviewed, on a daily basis, overall plant statu Significant safety matters related to plant operations were also reviewed by discussions with plant management and various members of the plant operating staf The NRC inspectors made visits to the control room at least daily when an NRC inspector was on site. Observations included instrument readings, setpoints and recordings, status of operating systems, tags and clearances on equipment controls and switches, annunciator alarms, adherence to Technical Specification Limiting Conditions for Operation (LCOs),

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i temporary alterations in effect, daily journal and data sheet entries, control room manning,' shift turnover, and access controls. This inspection activity included. numerous informal discussions with operators-and their supervisor Selected. Engineered Safety Feature (ESF) systems were confirmed operabl Confirmation was made by verifying, on a sampling basis, the accessible flow path valve alignment, power supply breaker and fuse status, no major component leakage or damage, proper lubrication, adequate cooling, general'

housekeeping conditions, and reasonable instrument indication General plant tours were conducte Accessible portions of the reactor auxiliary building, fuel handling building, turbine building, containment

' building, and outside areas were visited. Observations included safety-related tagout verifications, sampling program, housekeeping and general plant conditions, fire protection equipment, control of activities in progress, and problem identification. Periodically, the licensee's onsite emergency response facilities were toured to determine facility readines The NRC inspectors reviewed selected Radiation Work Permits (RWP),

observed health physics management involvement and verified awareness of significant plant activities, and observed controls over radioactive material and personnel radiation exposure. The NRC inspectors reviewed the licensee's program to limit personnel radiation exposure As Low As Reasonably Achievable (ALARA).

The NRC inspectors verified licensee compliance with physical security manning and access control requirements. The NRC inspectors verified the adequacy of physical security detection and assessment aid No violations or deviations were identifie . Onsite Followup of Events Reactor Coolant Pump Shaft Seal Leak On September 21, 1987, the licensee commenced a normal shutdown of the plant from full power because the Reactor Coolant Pump (RCP) 28 shaft seal was leaking excessively. Normally, RCP seals have a leak rate of about 1 gallon per minute (GPM). The RCP 2B seal was leaking l at approximately 5 GPM. This degraded condition warranted seal replacement; thus, the plant was shutdown and cooled to about 125 F to facilitate partial draining of the affected loop and subsequent seal replacemen The NRC inspectors reviewed the procedures used to establish the plant conditions required for the work and also reviewed the seal removal and replacement procedure The NRC inspectors reviewed Procedure OP-10-001, Revision 9, " General Plant Operations," used in the control room during the cooldown of i the plant. Random checks were made to verify that the procedure was being followed. The NRC inspectors noted that Step 8.9.11 had been i

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I signed off indicating that the last RCP was secured; however, the i control room operating log showed that the RCP was secured several hours prior to degassing the reactor coolant system (RCS) to 5 cc/Kg hydrogen as required by procedure. A note on the previous page in the procedure prohibited securing the last RCP until the hydrogen content in the RCS was reduced to 5 cc/Kg. The procedure was not annotated to show that the note was disregarded because of plant l conditions as permitted by procedure. The. decision to proceed'with securing the last RCP prior to reaching the 5 cc/Kg hydrogen limit appeared to be correct because by the time the pressurizer was purged i andtakensolid,RCShydrogencontentwaswellbelowthe5cc/Kg i requirement. The licensee s representative explained that such notes are not considered procedure steps, but rather are guidelines to )

ensure that.important requirements or good operating practices are l met. Plant administrative procedures did not appear to support this position. The NRC inspectors concluded that this represented an l example of failure to adhere to the detail of procedural requirements j and is therefore considered another example of the violation of the i requirements of Criterion V of Appendix B of 10 CFR Part 50  !

(382/8722-01).

Uptn replacing the RCP 2B seal, initial operation of the pump  !

revealed pressure staging problems with the new seal. The licensee again proceeded to cool down, depressurize, and partially drain the RCS to install a new design of seal provided by the RCP vendor, which is the Byron-Jackson Pump Company. The previous seals used (and l currently installed in RCPs 1A, 1B, and 2A) were Type SU, The new seal is Type N, which the licensee explained is made of superior seal surface materials, has less moving parts and is expected to last longer. On October 6, 1987, the new type seal was placed in service but, at low RCS pressure (about 380 PSIA), it exhibited abnormally high leak-off water temperatures. Normally, the temperature rise across the sealing assembly should be about 30 F. The new seal temperature rise was initially about 90 F. As the plant heated up l and approached normal operating pressure, the seal appeared to stabilize at a normal temperature rise but with a controlled leak off rate of about 2 GPM, or twice what was expected. The vendor representatives, with whom this abnormal temperature rise was discussed, did not appear concerned. On October 9, 1987, the reactor was started up and the unit placed on the grid. As of the end of this reporting period, the seal appeared to be stable and performing satisfactorily with the 2 GPM leak of Fuel Pool Draining Incident at River Bend In response to an upper fuel pool draining incident at River Bend Nuclear Power Station, the NRC inspectors reviewed the applicable j i

Waterford 3 USAR, system drawings, and operating procedures. The NRC  ;

inspectors determined that the use of siphon breakers and the

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i l elevation of piping connected to the spent fuel storage pool, in conjunction with the elevation of the opening of the transfer canal, i I

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l would prevent uncovering spent fuel assemblies by an inadvertent draining of the pool. However, after review of Procedure OP-2-006, l Revision 8. " Operating Procedure Fuel Pool Cooling and Purification System," the NRC inspectors identified the following apparent inadequacies to licensee management: ,

. Procedure OP-2-006 requires an operator to be present when adding make up water to the spent fuel pool; however, an operator is not required to be present when draining the refueling canal to the fuel handling building sum . Procedure OP-2-006 does not require installation of the fuel pool gates prior to draining the transfer canal; thus, the fuel storage pool may, by procedure, be lowered to well below the minimum level required by the applicable technical specificatio .

. If the pool gates are installed, there is no requirement to verify proper sealing while draining down the transfer cana Correction of the above weaknesses in order to prevent an inadvertent drain down will be tracked under Open Item 382/8722-0 . Followup of Other Events Attion 37 of TS 3.3.3.11, requires, in part, that when the main j Nndenser evacuation noble gas radiation monitor is out of service,  !

grab samples for gross activity analysis be taken every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. On October 14, 1987, the licensee identified a failure to meet this requirement when a sample was taken 3 1/2 hours late because of a I personnel error. Another example of a failure to meet Technical Specification surveillance requirements involved Actions 38 and 40 of TS 3.3.3.11 which require, in part, that when the waste gas hold up l system explosive gas monitoring system is out of service, the

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on-service gas decay tank shall be grab sampled for hydrogen and oxygen every 8 and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, respectively. On October 5, 1987, the licensee identified a failure to meet this requirement when samples were taken late because of a miscommunication on which gas decay tank was in service. This resulted in the "A" tank being sampled in error while the "C" tank was in service. Previous failures to meet Technical Specification surveillance requirements, because of personnel errors, were identified in LERs87-014, 87-017, and 87-01 Fire-hose house and computer halon system surveillance intervals were a exceeded because of a personnel error as identified in LER 87-01 LER 87-017 reported a failure to obtain a diesel fuel oil analysis within the required time due to a personnel error. Four fire watch tours were not conducted because of a misunderstanding of verbal communications as reported in LER 87-019. Failure to comply with the i

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above TS action statements for an inoperable main condenser i

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l evacuation noble gas activity monitor and inoperable waste gas

. collection system explosive gas monitoring system j 3.3.3.11. Actions 37, 38, and 40) is an apparent violation

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382/8722-04).

No other violations or deviations were identifie . Verification of Containment Integrity The objective of this inspection was to verify that the licensee had established containment integrity as required by the Technical Specifications, USAR, and station operating procedures prior to commencing l heatup of the RCS above 200* The NRC inspectors conducted the verification by: personal observation of the proper positioning of all electrical or mechanical barriers and isolation valves associated with selected containment penetrations, witnessing the air lock local leak rate test which was performed after containment closure, and walking down a system designed to maintain containment integrity or to mitigate contamination release in the event of a loss of coolant acciden Prior to heatup of the RCS above 200 F, following the outage caused by the leaking seal on Reactor Coolant Pump 2B, the NRC inspectors checked the integrity of Containment Penetration Nos. 14, 36, 37, 49, 51, 55, 56, 60, 66, and 71 and found no problem On September 29, 1987, the NRC inspectors witnessed the performance of Surveillance Procedure OP-903-111, Revision 0,." Containment Air Lock Door '

Seal Leakage Test." The test was successfully accomplished; however, the NRC inspectors noted that.the test engineer had signed off Step 8.2.3, which required him to perform the test in accordance with an attachment to the procedure when the attachment was not yet completed. Though he had every intention of completing the test, this practice diminishes the credibility of step signoffs in completed documentation. Step 8. requires the test engineer to notify security that the outer door is ready to be secured and that the door alarm tested satisfactorily. The test engineer skipped that step by placing an "N/A" (Not Applicable) in the signature blank without obtaining the approval of the Shift Supervisor, as required by Procedure OP-100-001, Revision 4 "Outies and Responsibilities of Operators on Duty," paragraph 5.12.5. The test engineer explained that Step 8.2.8 was simply a reminder and was not appropriate for execution at that particular time. This is another example of failure to comply with a procedure and is an apparent violation of Criterion V of 10 CFR Part 50 l (382/8722-01).

On September 16 and 21, 1987, the NRC inspectors performed a walkdown of the containment hydrogen analyzer system to verify operabilit The NRC inspectors were accompanied by an auxiliary operator on September 2 _

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The NRC inspectors used the containment hydrogen analyzer system lineup ,

specified on Attachment 8.1 of Procedure OP-8-010 Revision 5, " Operating !

Procedure Containment Hydrogen Analyzer," in conjunction with )

Drawings LOV-1564-B-430 and LOV-1564-167 The following items were identified to licensee management:

i . Identification tags were found to be installed on the wrong valves.

l Valve HRA-1103B was labeled as HRA-1257B, HRA-1101A as HRA-1261A, and :

HRA-1102A as HRA-1252A and vice-versa for each pair of valves.

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, . Sample cylinder isolation Valves HRA-117A and HRA-118A, as well as the removable sample cylinder as shown in Drawing LOV-1564-1674, were not installe . Component cooling water supply and return Valves CC-8032A, CC-8032B, CC-8033A, and CC-8033B were not controlled by component cooling water I

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Operating Procedure OP-2-003, Revision 5. " Component Cooling Water System," nor by Procedure OP-8-010. They should be controlled by one or the othe . The hydrogen analyzer system standby valve lineup of Procedure OP-8-010 required analyzer bypass Valve HRA-114B to be shut, but it did not address Valve HRA-114A, a similar valve in the other train. These valves were controlled, however, by Procedure MI-3-431, Revision 3, " Hydrogen Analyzer Channel Calibration." This could result in confusion over the proper positioning of either or both valve I

. Pressure instrument isolation Valves HRA-115A and HRA-115B were found throttled. The hydrogen analyzer system standby valve lineup of l Procedure OP-8-010 requires these valves to be ope ;

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The identification numbers for several valves on the hydrogen analyzer panels were handwritten on the face of the panels. Although valve identification tags were installed on the inside of the panels, ,

they would not be visible when operating the valves. The licensee has, since the time of this walkdown, corrected this ite Correction of the remaining deficiencies above will be tracked under Open Item 382/8722-0 No violations or deviations were identifie . Fire Protection While conducting a routine tour of the reactor auxiliary building, the NRC inspectors noted that Fire Door D-47, which is an access to the decontamination room on elevation +21 feet, was ajar when it should have been closed. Workers had apparently impaired the door with a scaffold clamp because the door was difficult to open because of the negative

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., ambient pressure'in the room. The NRC inspectors checked the control room for documentation of' the impairment as is required by Fire Protection Procedure FP-1-015, Revision 3, " Fire Protection System' Impairments." No record of the impairment was found. Procedure FP-1-015 also requires an individual discovering an impairment to a fire appliance to immediately report it to the Shift Supervisor / Control Room Supervisor (SS/CRS). It was apparent that several employeer had previously passed through the impaired door without making the notification. Failure to inform the

'SS/CRS and failure to have the impairment documented is contrary to.the requirements of Procedure FP-1-01 During the ESF system walkdown discussed in paragraph 7 of this inspection report, the NRC inspectors also found a cigarette butt on the foundation for EFW Pump "A", and three more on the +46 level of the stairwell to the main steam and main feed isolation valve room. Neither of these areas is designated as a smoking area. Paragraph 6.3.4 of Procedure FP-1-013, Revision 3, " Fire Protection Program Plan," prohibits smoking in areas not '

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designated as smoking areas. Furthermore, the area around EFW Pump "A" is a radiologically controlled area wherein it is specifically forbidden to eat, drink, chew, or smok Five safety cans containing "Fyrquel" hydraulic fluid were found unattended at different locations in the main steam and main feed isolation valve rooms. Procedure FP-1-017, Revision 4, " Transient Combustibles and Hazardous Materials," requires such safety cans to be stored in an approved flammable liquids storage cabinet when not in us On August 19, 1986, the NRC inspectors found six unattended safety cans containing the combustible liquid'in the same area. Violation 382/8616-02 was issued, and in response, the licensee placed an approved flammable liquids cabinet nearby. Such a cabinet was not visible to the NRC inspectors during the present inspection, but the licensee immediately placed one in the are During a subsequent tour, the NRC inspectors noted that Fire Doors 0-136, D-163, and 0-165, also in the reactor auxiliary building, were not latched closed, contrary to a warning sign on the doors informing passers-through that the fire door must be kept closed. Failure of personnel to heed the <

warning signs placed the doors in an impaired condition, which was not identified as required by Procedure FP-1-01 Failure to follow the requirements of the fire protection procedures, as discussed above, is an apparent violation of TS 6.8.1 which requires, in part, that fire protection program procedures be implemented (382/8722-03).

1 Procedural Adherence and Attention to Detail I

The NRC inspectors observed numerous minor problems related to procedure compliance, the details of which are discussed in paragraphs 5, 9, and 10 of this repor The NRC inspectors met with licensee management and ,

expressed concern that the apparent lack of employee vigilance in keeping l

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fire doors closed or formally identifying them as impaired, coupled with the other minor procedural problems discussed in this inspection report, may be indicative of a level of employee discipline less than that expected by the NRC inspectors to assure procedure compliance. Th ,

licensee reacted to these concerns by conducting " procedure awareness j

, meetings" with site personnel during the week of September 28, 198 A I total of 587 employees attended management discussions on procedural compliance, sensitivity to plant environmental conditions, personal discipline, and doing the job right the first tim This represents training of over 90 percent of the site employees for which t'1is training should apply. In addition, the licensee has engaged the sersices of .l Advanced Resources Development Corporation to improve the human factors aspect of plant procedures to better assure compliance. This latter

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action, however, was.in progress before the NRC inspectors raised this concer !

The NRC inspectors will continue to evaluate the licensee's efforts to achieve the level of procedure compliance and the attention to detail necessary to assure safe plant operation consistent with the controls and I documentation required ~by NRC regulations, j 13. Exit Interview The inspection scope and findings were summarized on October 19, 1987, with those persons indicated in paragraph 1 above. The licensee acknowledged the NRC inspectors' findings. The licensee did not identify as proprietary any of the material provided to or reviewed by the NRC inspectors during this inspectio !

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