IR 05000313/1999004

Insp Repts 50-313/99-04 & 50-368/99-04 on 990228-0410. Non-cited Violations Noted.Major Areas Inspected:Operations, Engineering,Maintenance & Plant Support

ML20206S432
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/10/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20206S425	List: IR 05000313/1999004 ML20206S421
References
50-313-99-04, 50-313-99-4, 50-368-99-04, 50-368-99-4, NUDOCS 9905210134
Download: ML20206S432 (21)

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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket Nos.: 50-313 50-368 License Nos.: DPR-51 NPF-6 Report No.: 50-313/99-04 50-368/99-04 Licensee: Entergy Operations, In Facility: Arkansas Nuclear One, Units 1 and 2 Location: Junction of Hwy. 64W and Hwy. 333 South Russellville, Arkansas Dates: February 28 through April 10,1999 Inspectors: K. Kennedy, Senior Resident inspector K. Weaver, Resident inspector J. Hanna, Resident Inspector P. Goldberg, Reactor inspector R. Nease, Reactor inspector Approved by: P. Harrell, Chief, Project Branch D Division of Reactor Projects Attachment: Supplemental Information 9905210134 990510 '

PDR ADOCK 05000313 G PDR ;

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EXECUTIVE SUMMARY Arkansas Nuclear One, Units 1 and 2 NRC Inspection Report 50-313/99-04; 50-368/99-04 This routine announced inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspectio Operations

The Unit 2 chemists failed to inform the operations department to shut the emergency fee.iwater sample valves as required by the sampling procedure. As a result, the valves were out of position for approximately 10 days. This Severity Level IV violation of Technical Specification 6.8.1 is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2-1999-0324 (Section O2.1).  !

On May 26,1997, alternate radioactive gaseous sampling was not established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of losing the normal radioactive gaseous effluent instrumentation for the Unit 2 containment building. This was identified as a noncited violation of Technical )

Specification 3.3.3.9. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2-1997-0288 (Section 08.2).

On June 26,1997, the fuel handling area ventilation system flow rate was less than the minimum required flow rate for transporting a load over the Unit 2 spent fuel storage pool. This Severity Level IV violation of Technical Specification 3.9.11 is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2-1997-0435 (Section 08.4).

Maintenance

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Operations, maintenance, and engineering personnel demonstrated good communications, self-checking, and peer-checking techniques during the observed l surveillance activities (Section M1.2). i

On October 22,1997, the licensee discovered that, due to a deficient test procedure, a Unit 1 auxiliary cooling water valve had not been verified to close on an engineered safety feature actuation signal. This is a violation of Technical Specification 4.5.1. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 1-1996-0086 (Section E8.2).

Plant Support

Health physics technicians provided good coverage of the work activities involved with the Unit 2 Spent Fuel Pool Purification System Filter 2F4A change out. Two health

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2-physics technicians provided continual coverage during this work and thoroughly briefed the maintenance craft involved on dose and contamination rates in the area and the radiation work permit requirements. All personnelinvolved demonstrated very good as i

low as is reasonably achievable practices (Section R1.1 ).

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Report Details Summary of Plant Status Unit 1 began this inspection period at 100 percent power. On March 23,1999, operators commenced a power reduction to 85 percent for turbine and governor valve testing and to replace Feedwater Heater E-1 A Level Control Valve CV-3026. During the downpower, Governor Valve GV-3 started cycling erratically. Operators placed the turbine controls in manual and isolated Governor Valve GV-3. The plant stabilized at 75 percent power. Following subsequent troubleshooting and maintenance activities, operators returned Unit 1 to 100 percent power the same day. Unit 1 remained at or near 100 percent power through the end of this inspection perio Unit 2 began this inspection period at approximately 90 pe'rcent power. On March 2, operators increased power to 100 percent. Unit 2 remained at or near 100 percent power through the remainder of this inspection perio . Operations 01 Conduct of Operations 0 General Comments (71707)

The inspectors observed various aspects of plant operations, including compliance with Technical Specifications (TS), conformance with plant procedures and the Safety Analysis Report, and shift manning. Inspectors also observed the effectiveness of communications, management oversight, proper system configuration and configuration control, housekeeping, and operator performance during routine plant operations and surveillance testin The conduct of operations was professional and effective. Evolutions were generally well controlled and performed according to procedures. Shift turnover briefs were comprehensive. Housekeeping was generally good and discrepancies were promptly corrected. Specific events and noteworthy observations are detailed belo O2 Operational Status of Facilities and Equipment O Emeraency Feedwater (EFW) System Walkdown (Unit 2) Inspection Scope (71707)

The inspectors performed a detailed walkdown of the accessible portions of the EFW system to verify that it was properly aligned and to inspect the material condition. The inspectors also conducted a walkdown of the ac and de electrical systems and equipment that supports the EFW syste b. Observations and Findinas ]

The inspectors checked the interiors of electrical circuit breaker cabinets and verified i them to be free of debris, loose material, and unauthorized jumpers. Power supplies and breakers were correctly aligned, functional, and available for components that must activate upon receipt of an actuation signal, such as the motor-driven EFW pump. The upper- and lower-piping penetration rooms and the condensate storage tank pits were i found to be free of ignition sources and flammable materials. Cleanliness was )

acceptable in all areas inspected. With only minor exceptions, components were I correctly labeled. The inspectors did identify minor differences between some component descriptions. The deficiencies were between the required valve lineup of Procedure 2106.006, Revision 48, " Emergency Feedwater System Operations," and certain system valves. These discrepancies were referred to the licensee for corrective actio On March 6,1999, the inspectors identified that the EFW sample valves (Valves 2EFW-14B and 2EFW-158) were mispositioned in the open position. Procedure 2106.006 and the system drawing indicated that the valves were normally closed. These valves 3 provide double isolation between the EFW header and the sample system. The  !

inspectors promptly reported this to the Unit 2 shift superintendent. An operator was dispatched to close the valves. The operator checked the sample isolation valves on the opposite train (Valves 2EFW-14A and 2EFW-15A) and found that they were also in the open position. The operator closed the valves. The licensee evaluated the condition and determined that the EFW system remained operable with the sample valves ope The licensee identified the following causes for this configuration discrepancy:

• Procedure 2618.002, Revision 3," Sampling Unit 2 Emergency Feedwater," had specific steps that needed to be performed when the sampling was complet This procedure was not completed after the sampling was secure * Errors in the process to track the configuration of systems (component out-of-position log) prevented the operators from identifying that the valves were not in their proper positio * Operations and chemistry personnel failed to adequately implement corrective actions established as a result of a previous occurrence of this configuration control error (Condition Report (CR) 2-1998-0195).

The licensee identified several potential corrective actions including:

• Changing the procedures to require sign-off steps to aid in tracking procedure usage that spans multiple shift I i

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• Revising Procedure 2102.004, Revision 25, " Power Operation," to have chemistry notified when shifting steam generator feed from EFW to main j feedwater and include instructions to verify that the sample valves are locked close * Operations reviewed this CR with the operations staff, emphasizing the need to implement an effective questioning attitude when performing administrative dutie TS 6.8.1 requires, in part, that written procedures shall be implemented covering the ;

applicable procedures recommended in Appendix A of Regulatory Guide 1.33, j Revision 2, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, Section 3.1, states, in part, that instructions should be prepared for operation of the auxiliary feedwater system. As a result of personnel error, the inspectors determined that the j failure to follow the EFW sample procedure was a Severity Level IV violation. This i Severity Level IV violation is being treated as a noncited violation, consistent with l Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 2-1999-0324 (50-368/9904-01). Conclusions The Unit 2 chemists failed to inform the operations department to shut the emergency feedwater sample valves as required by the sampling procedure. As a result, the valves were out of position for approximately 10 days. This Severity Level IV violation of TS 6.8.1 is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 2-1999-032 Miscellaneous Operations issues (92901,92700) i j

0 (Closed) Licensee Event Report (LER) 50-313/97-003-00: Inadvertent automatic actuation of the EFW system resulting from a lightning induced trip of two EFW initiation and control system power supplies t The inspectors verified the immediate corrective actions described in LER 50-313/97-003-00, dated August 7,1997, and found them to be adequate and complet .2 (Closed) LER 50-368/97-004-00: Alternate radioactive gaseous effluent sampling not established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> as required due to inadequate alarming ccpabilities on radiological dose assessment computer system terminals The inspectors verified the immediate corrective actions described in LER 50-368/97-004-00, dated June 25,1997, and found them to be adequate and complete. This Severity Level IV violation is being treated as a noncited violation

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-4-consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 2-1997-0288 (50-368/9904-02).

08.3 (Closed) LERs 50-368/97-006-00. 50-368/97-006-01. and 50-368/97-006-02:

inadvertent emergency diesel generator (EDG) start, which resulted from inadequate written guidance regarding system component testing The inspectors verified the immediate corrective actions and found them to be adequate and complete as described in LER 50-368/97-006-00, dated July 2,1997; LER 50-368/97-006-01, dated July 30,1997; and LER 50-368/97-006-02, dated September 5,199 .4 (Closed) LER 50-368/97-007-00: Fuel handling area ventilation flow rates less than TS requirement during crane operation with a load over the storage pool due to personnel error regarding verification of equipment condition before evolution authorization On June 26,1997, the licensee determined that the fuel handling area ventilation system flow rate was less than the minimum flow rate required by TS 3.9.11 for transporting a load (filtration / vacuum assembly) over the Unit 2 spent fuel storage poo This is a violation of TS 3.9.11. The inspectors verified the immediate corrective actions described in LER 50-368/97-007-00, dated July 28,1997, and found them to be adequate and complet This Severity Level IV violation of TS 3.9.11 is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 2-1997-0435 (50-368/9904-03).

08.5 (Closed) Inspection Followup Item (IFI) 50-368/9803-02: Corrective actions addressing reactor coolant system (RCS) level indication anomalies During a draindown of the RCS to reduced water inventory conditions on February 25 and March 17,1998, operators experienced problems with RCS levelindication due to moisture and air accumulation in the instrumentation lines. The licensee initiated a number of corrective actions to address these problems, including changing Procedure 2103.011, " Draining the Reactor Coolant System," to flush the instrument lines at specific levels while draining the RCS and modifying the RCS level instrumentation lines during Refueling Outage 2R13 to increase the size of the l instrument tubing and improve the slope. Although the modification was installed after the reduced inventory periods during Refueling Outage 2R13, the actions to flush the instrumentation lines appeared to alleviate the level indication anomalies experienced during previous outage .6 (Closed) LER 50-313/98-002-00: Automatic actuation of the control room emergency ventilation system due to higher than normal radiation at the detector when a filtration / vacuum assembly was moved in an adjacent area

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5-The inspectors verified the immediate corrective actions described in LER 50-313/98-002-00, dated June 1,1998, and found them to be adequate and complet II. Maintenance M1 Conduct of Maintenance M1.1 General Comments Inspection Scope (62707)

The inspectors observed all or portions of the following maintenance activities:

. mal 5033," Dry Fuel Storage Preparation for Loading of MSB-004," performed on April 2 (Unit 1) ,

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. MAI 964707, " Inspect, Clean, Lubricate, Megger Motor and Check for Leak of j CV-3813, Service Water Isolation Valve to the Reactor Building Coolers VCC-2C and VCC-2D," performed on April 9 (Unit 1) Observations and Findinas On April 2,1999, the inspectors observed the prejob briefing conducted in preparation for the Unit 1 dry fuel storage loading for Cask 9. The briefing was thorough, comprehensive, and covered all compensatory measures, including measures to be taken due to inclement weather condition On April 8, the inspectors observed craftspersons perform maintenance on Valve CV 3813, which is the service water isolation valve to Reactor Building Coolers VCC-2C and VCC-2D. The inspectors observed that the maintenance craft could not remove the valve motor housing cover to inspect the limit switch compartmen The maintenance craft rebolted the housing cover and stopped work to evaluate the situation. A 1-inch line that went to the standby hydrogen purge system, which had been abandoned in place, interfered when removing the Limitorque cover. The maintenance craft learned that personnel had experienced similar difficulties during previous maintenance activities. Subsequently, the maintenance craft were able to ,

l remove the housing cover after several manipulations. No deficiencies were identified i

with the internal components during the inspection. Due to difficulties in reinstalling the housing cover, the maintenance craft were careful to minimize the stress applied to the j valves'intemal wiring contacts. The inspectors observed that one technician pulled !

back the interfering 1-inch line while two other technicians maneuvered the housing cover back into place. The maintenance technicians and the inspectors were concerned that the difficulties experienced in removing and reinstalling the housing cover caused ,

severe stress to the internal components and presented the potential to damage or disconnect the internal wirin :

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I-6- l The maintenance craft supervisor at the job site initiated an engineering request for engineering personnel to evaluate a possible plant configuration change to resolve the proble Conclusions During maintenance activities for Valve CV-3813, the service water isolation valve to the Reactor Building Coolers VCC-2C and VCC-2D, concerns were identified that removing and reinstalling the Limitorque housing cover in the present plant configuration caused severe stress to the internal components and presented the potential to damage or disconnect the internal wiring. The maintenance craft supervisor at the job site initiated an engineering request for engineering personnel to evaluate a possible configuration chang !

M1.2 General Comments on Surveillance Activities Inspection Scope (61726)

The inspectors observed all or portions of the following surveillance activities:

. Procedure 1104.002, Revision 51, " Makeup & Purification System Operation,"

Supplement 6,"HPI Pump P-36C Test," performed on March 9 (Unit 1)

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- Procedure 2304.037, Revision 28," Unit 2 Plant Protection System Channel A Test," performed on March 10 (Unit 2) ,

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. Procedure 1104.005, Revision 38, " Reactor Building Spray System Operation," )

Supplement 5," Reactor Building Spray Pump P-35B Ouarterly Test," performed {

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on March 11 (Unit 1)

. Procedure 1104.036, Revision 38, " Emergency Diesel Generator Operation,"

Supplement 2,"DG2 Monthly Test," performed on March 15 (Unit 1)

. Procedure 2104.036, Revision 42, " Emergency Diesel Generator Operations,"

Supplement 1 A, "2DG1 Monthly Test (Slow Start)," performed on March 31 (Unit 2) Observations and Findinas The inspectors found that these surveillance activities were performed in accordance with approved procedures by knowledgeable personnel using calibrated test equipmen Operations, maintenance, and engineering personnel involved in these activities i demonstrated good communications, self-checking, and peer-checking technique The inspectors noted that system engineering personnel observed system performance during the Unit 2 EDG monthly surveillance testing and demonstrated good attention to detail when monitoring for abnormal conditions. During interviews with the Unit 2 EDG system engineer, the inspectors found the individual knowledgeable of deficiencies

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-7-associated with the EDG and that he had verified that the deficiencies would not impact system operability. Prior to conducting the surveillance, personnel were painting in the EDG A room. As a conservative measure, the painting was suspended during the EDG B monthly test. The inspectors learned that the system engineer monitored the EDG painting activities on a daily basis to ensure no impact on system operability would oecer Conclusions Operations, maintenance, and engineering personnel demonstrated good communications, self-checking, and peer-checking techniques during the observed surveillance activitie M8 Miscellaneous Maintenance issues (92700, ' -2702)

M8.1 (Closed) LER 50-313/98-004-00: Inadvertent actuation of the EFW system This LER was prev.ously reviewed and discussed in NRC Inspection Repor150-313/98-08; 50-368/98-0 E8 Miscellaneous Engineering issues (92902,92700)

E (Closed) LERs 50-368/97-003-00 and 50-368/97-003-01: TS, allowing operations with one channel of refueling water tank level or steam generator differential pressure tripped, established the potential for premature initiation of recirculation actuation signal during a loss-of-coolant accident or inability to automatically isolate a faulted steam generator for certain steam line break scenarios placing the plant outside its design basis The inspectors verified the immediate corrective actions described in LER 50-368/97-003-00, dated June 13,1997, and LER 50 368/97-003-01, dated October 14,1997, and found them to be adequate and complete. The NRC issued Amendment 195 on December 29,1998, to modify the Unit 2 TS to restrict the length of time that a steam generator differential pressure channel or refueling water tank level-low channel can be placed in a tripped conditio E8.2 (Closed) LER 50-313/97-005-00: Surveillance testing did not verify Valve CV-3643 i would close on an engineered safeguards signal

During a quality assurance audit of surveillance testing, the licensee discovered that l testing of the circuitry for Valve CV-3643, the valve that isolates the auxiliary cooling water system from the service water system, did not verify that the valve would close upon receipt of an engineered safeguards signalif the Train B service water pump was running. The licensee conducted special testing and verified that the valve was operable. The licensee determined that the root cause was attributed to a deficient test procedure that was developed in 1973 and remained unchange _ _ . _ .

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8-The inspectors reviewed CR 1-1996-0086, dated March 21,1996, which documented a review of the reactor protection system and the emergency safeguards actuation system in accordance with NRC Generic Letter 96-01," Testing of Safety-Related Logic Circuits." In addition, the inspectors reviewed Procedure 1305.006, Revision 16,

" Integrated ES System Test," which was revised to incorporate changes due to a lack of testing Valve CV-3643. The inspectors noted that the licensee completed all of the corrective actions described in the LE TS 4.5.1.1.2 for the low pressure injection system requires that once every 18 months a system test shall be conducted to demonstrate that the system is operable. The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal. The failure to test Valve CV-3643 is a violation of TS 4.5.1.1.2. This Severity Level IV violation is being treated as a noncited violation consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR 1-1996-0086 (50-313/9904-04).

E8.3 (Closed) IFl 50-313/9718-02: 50-368/9718-02: Review of licensee's corrective actions for main steam safety valve (MSSV) setpoint deficiencies The licensee documented in CR 2-1997-0152, dated May 10,1997, that the as-found setpoints of eight of ten Unit 2 MSSVs exceeded the maximum 1 percent tolerance permitted by the Unit 2 TS. The licensee sent nine of the valves to a vendor for inspection, repair, and testing. The valves were subsequently reinstalled and declared operable. Further NRC followup of Refueling Outage 2R13 test results and confirmation of licensee corrective actions for the MSSV setpoint deficiencies was identified as an IF The inspectors reviewed CR 2-1999-0011, dated January 9,1999, which discussed all of the corrective actions taken to resolve the Unit 2 MSSV as-found high setpoint During Refueling Outage 2R13, the licensee replaced the original valve discs with flex-discs in an effort to eliminate seat leakage problems associated with the flat seat design disc. In addition, the licensee, with approval from the Office of Nuclear Reactor Regulation, revised their TS tolerance for setpoint acceptability from +/-1 percent to +/-3 percent. The inspectors reviewed the as-found setpoint results from Unit 2 Refueling Outage 2R13 and found that the setpoints for only one of the ten valves fell outside of the +/-3 percent tolerance. The inspectors concluded that the licensee's corrective actions were effectiv E (Closed) Violation (VIO) 50-313/9721-06 (EA 50-313/97574-01): Failure to initiate temporary alteration for the borated water storage tank (BWST) work Procedure 1000.028, Revision 21, Attachment 1, " Control of Temporary Alterations,"

provided instructions that a temporary alteration package was required if a change was made to a safety-related component remaining in service during the period of the change. Changes were made to the BWST without a temporary alteration package being initiated. Specifically, on December 4,1996, the pressure vacuum relief valve was removed from the Unit 1 BWST, and the tank's valve flange was covered with a plastic bag for foreign material exclusion. On December 5,1996, the plastic bag l

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-9-covering the Unit 1 BWST was removed and replaced with a foreign material exclusion cover that had a 1-inch gap between the cover and the valve flange on the tank. On December 12,1996, the foreign material exclusion cover on the Unit 1 BWST pressure and vacuum relief valve flange was adjusted to establish a 3-inch gap between the cover and the valve flang The inspectors reviewed CRs 1-1997- 0019,1-1998-0068, and 1-1998-0143, which addressed the above violation. The inspectors found that a temporary alteration package to document the 3-inch flanged gap foreign material exclusion cover was completed in February 1997. The licensee walked down the accessible areas of Units 1 and 2 to identify undocumented temporary alterations, and only minor deviations were found. The licensee performed an industry benchmarking study, which identified the attributes that would best enhance the temporary alteration program. The screening criteria for the temporary alterations was incorporated into the job order planning process to be used for planning jobs. The engineering request process was revised to include links to the temporary alteration process, and risk management guidelines were updated to reference temporary alterations. Engineering, planning, and operations personnel received training on the use of these revised procedures. The inspectors determined that the licensee's corrective actions were appropriat E8.5 { Closed) VIO 50-313/9721-07 (EA 50-313/97574-02): Inadequate procedure for BWST work Procedure 1306.034, Revision 2, paragsph 4.3.1, " Testing of Unit 1 Pressure Vacuum Relief Valves PSV-1617, PSV-2423, and PSV-1412," provided steps for performing testing of the Unit 1 BWST vacuum relief valves. However, on December 4,1996, Procedure 1306.034 was not appropriate to the circumstances. The procedure failed to establish appropriate guidance to ensure that the pressure and vacuum relief valve was tested only while the plant was shut dow The inspectors reviewed Procedure 1306.034, Revision 2, and found the licensee revised it to state that the procedure was only to be performed during a refueling outage. The inspectors reviewed Action item 3 of CR 1-1997-0031 which requested a justification or change in design such that the BWST would not be susceptible to a single failure of an active component. The licensee stated that the pressure vacuum relief valve had not been installed on the tank since December 1996 and that a new vent cover was being designed. Currently, a temporary alteration was installed on the tank vacuum valve flange. The inspectors reviewed CRs 1-1997-0019 and C-1995-0098, which also addressed the above violation. The inspectors found that the licensee's corrective actions were appropriat E8.6 (Closed) IFl 50-313/97201-09: Use of commercial-grade cable ties in containment and use of commercial-grade polyvinyl chloride (PVC) flexible conduit This IFl was opened to address the discrepancy between Drawing E-59, which required that cable grips be used to support cables on vertical risers of more than 20 feet, and Procedure 6030.109, which described the use of cable ties for support but did not

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require the use of cable grips for support. In addition, the IFl would review the resolution of CR C-1997-0105, which was written to address the use of PVC conduits !

inside the reactor buildin The inspectors reviewed Procedure 6030.109, Revision 3, " Installation of Electrical ,

Cable & Wire," which was revised to require the use of cable grips on cables supported !

in vertical tray risers of 20 feet or more and to agree with Drawing E-59, " Conduit and Cable Tray Notes and Details." During discussions with the licensee, the inspectors determined that a CR had not been written to document this discrepancy. The inspectors learned that the cable had originally been installed in accordance with Drawing E 59, which required the use of cable grips for a vertical riser of 20 feet. Once Procedure 6030.109 was prepared, cables were installed in accordance with this

, procedure which only required cable ties. The licensee stated that commercial-grade nylon cable ties were used. The inspectors reviewed Drawing E-59 and Plant Change 1 to Procedure 6030.109, Revision 3, and noted that the drawing and procedure were revised to state that cables in vertical riser trays shall be properly supported per the National Electrical Code Article 300-19 and the values in Table 300-19a. The inspectors noted that Table 300-19a listed support distances based on the size and material of the cable. In addition, the table listed five methods by which the cable could be supporte The licensee stated that they reviewed the plant drawings for the Unit 1 reactor building to determine if any vertical tray runs existed that exceeded the 20-foot distance. The licensee found only three safety-related cable trays that were installed in excess of 20 feet. In addition, the three vertical trays fell well within the National Electrical Code guideline The licensee stated that the use of tie wraps was to keep a cable tray orderly. In addition, the licensee stated that the vertical cable trays should not see normal exposure doses that would exceed the threshold limit of the tie wraps. Concerning the tie wraps inside containment for Unit 1, the licensee does not have to postulate a seismic event; therefore, no seismic loads would be placed upon the cable tie wrap The licensee initiated CR C-1997-0105 in response to being questioned about the PVC flexible conduit used in the containment. The inspectors reviewed this CR and determined that the PVC conduit had been installed in the plant during constructio The licensee discovered that flexible conduit used in both the Units 1 and 2 containments had not been formally evaluated for use in high temperature or radiation exposure conditions. In addition, the PVC jacketed conduit had not been evaluated in the fire loading calculations. The licensee determined that the PVC conduit impacted containment systems in three categories: (1) impact on the sump screen delta pressure, (2) impact on emergency core cooling system components due to the limited quantities of the PVC conduit inside the sump, and (3) impact on fire loading assessments. The inspectors found the impact on the sump screen was minimal, the impact on the emergency core cooling system components was bounded by previously evaluated debris, and the impact on the fire loading would be an insignificant contributor in the event of a fire. The inspectors reviewed the licensee's corrective actions and found them acceptabl i

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-11-E (Closed) IFl 50-313/97201-13: EDG exhaust stack tornado missile evaluation This IFl was opened to address the concern that a tornado-driven missile could directly ,

strike the EDG exhaust stack causing it to deform inwardly and appreciably reduce the !

EDG exhaust flow during operation. Since the original calculation could not be produced during the inspection, the licensee agreed to reconstitute the analysi The inspectors reviewed CR C-1997-0275, which the licensee prepared to document this IFl. The licensee's corrective actions included removing the nonseismic rain hoods and adding stiffener rings to the tops of the Units 1 and 2 EDG exhaust stacks. The inspectors verified that the modifications were complete for both units. Furthermore, the inspectors noted that IFl 50-313/9216-06 was opened to address a similar issue concerning whether the EDG exhaust pipes and associated exhaust hoods could become crimped from a tornado-driven missile and affect the capability of the EDGs to perform their safety functions. This item was closed satisfactorily in NRC Inspection Report 50-313/93-29; 50-368/93-29 based on revised abnormal operating procedures, which required inspection and clearing of any damage to the exhaust piping after a tornad j E (Closed) VIO 50-313/9718-04: 50-368/9718-04: Violation of 10 CFR 50.59 for making changes to Section 9D.1.3 of the Units 1 and 2 Final Safety Analysis Reports and to Procedure 1000.120, Revision 8, "ANO Fire Watch Program," without a written safety evaluation in accordance with Appendix C of the Enforcement policy, this violation is being closed because the licensee had entered this issue in the corrective action program as CR C-1997-003 E _

( Closed) VIO 50-313/9721-01: Violation of 10 CFR Part 50, Appendix B, Criterion V, for Procedure 1010.002, " Transient History / Transient Cycle Logging," inappropriate to the circumstances because a vendor-recommended maximum EFW flow limit of 1500 gpm to a single steam generator was not incorporated into the procedure in accordance with Appendix C of the Enforcement policy, this violation is being closed because the licensee had entered this issue in the corrective action program as CR 1-1997-008 l E8.10 (Closed) VIO 50-368/9801-01: Violation of 10 CFR 50.65(b) for failure to include the turbine building sump system in the scope of the maintenance rule in accordance with Appendix C of the Enforcement policy, this violation is being closed l because the licensee had entered this issue in the corrective action program as ;

CRs C-1998-0017 and C-1998-008 .

-12-E8.11 (Closed) VIO 50-313/9801-03: Violation of 10 CFR 50.65(a)(3) for failure to establish appropriate performance measures for availability of risk-significant systems, structures, and components in accordance with Appendix C of the Enforcement policy, this violation is being closed because the licensee had entered this issue in the corrective action program as CR C-1998-008 E8.12 (Closed ) VIO 50-313/9801-04: 50-368/9801-04: Four examples of a violation of 10 CFR 50.65(a)(1) for failure to monitor the performance or condition of certain systems against established goals .

In accordance with Appendix C of the Enforcement policy, this violation is being closed because the licensee had entered this issue in the corrective action program as CR C-1998-008 E8.13 (Closed) VIO 50-368/9801-07: Violation of 10 CFR 50.65(a)(1) for failure to ensure that the 125-Vdc system was capable of performing its intended function in accordance with Appendix C of the Enforcement policy, this violation is being closed because the licensee had entered this issue in the corrective action program as CR C-1998-008 E8.14 (Closed) VIO 50-368/9812-02: Violation of 10 CFR Part 50, Appendix B, Criterion IX, for failure to ensure that Engineering Sheet 4 contained instructions for using the appropriate phase rotation as specified in Electric Power Research Institute "PWR Steam Generator Examination Guidelines."

In accordance with Appendix C of the Enforcement policy, this violation is being closed because the licensee had entered this issue in the corrective action program as CR 2-1998-009 E8.15 (Closed) VIO 50-368/9812-03: Violation of 10 CFR Part 50, Appendix B, Criterion V, for failure to follow procedure of Engineering Standard HES-72,"ANO Eddy Current Data Acquisition," in making a change in the calibration practice described in the "ANO-2 Calibration Standard Dimensions" sheet in accordance with Appendix C of the Enforcement policy, this violation is being closed because the licensee had entered this issue in the corrective action program as CR 2-1998-008 l

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-13-IV. Plant SuDDort R1 Radiological Protection and Chemistry Controls R General Comments During routine tours of the plant and observations of plant activities, the inspectors found that access doors to high radiation areas were properly locked, areas were properly posted, and count rate meters installed at specific locations were properly calibrate On April 2,1999, the inspectors observed that health physics coverage of the work activities involved with the Unit 2 Spent Fuel Pool Purification System Filter 2F4A change out was very good. Two health physics personnel provided continual coverage during this work. The two health physics technicians measured the dose rates and contamination levels for the area prior to the start of the maintenance activities and .

I documented these readings on the job coverage survey map. The technicians continually monitored the dose and contamination rates until the work was complete The inspectors observed that the calibration of the instrumentation used was curren The maintenance personnelinvolved with the filter change out were throughly briefed by the two health physic technicians, were knowledgeable of the radiation work permit requirements, and demonstrated very good ALARA practice V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of the licensee's staff at the conclusion of the inspection on April 13,1999. The licensee acknowledged the findings presente l The inspectors asked the licensee whether any material examined during the inspection should be considered proprietary. No proprietary information was identifie ;

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ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee C. Anderson, General Manager, Plant Operations T. Armstrong, Jr., Chemistry G. Ashley, Licensing Supervisor W. Barnes, Unit 1 Electrical Supervisor D. Bauman, Unit 2 Outage Management T. Bennett, Unit 2 System Engineer i V. Bhardwaj, Unit 1 Electrical Superintendent J. Bradford, Unit 2 Instrumentation and Control Superintendent M. Chisum, Manager, Unit 2 System Engineering E. Christian, Unit 1 Instrumentation and Control Superintendent M. Cooper, Licensing Specialist J. Cotton, Senior Engineer, Nuclear Safety Analysis E. Dietrich, Senior Engineer, Mechanical / Civil / Structural Design P. Dietrich, Unit 1 Maintenance Manager  !

J. Ekis, Unit 1 System Engineer R. Harris, Senior Lead Technical Specialist, Nuclear Safety Analysis D. Harrison, Supervisor, Engineering Programs ,

A Hartman, Maintenance Coordinator )

A. Hatley, Unit 2 Acting Superintendent, Mechanical Maintenance !

K. Head, Manager, Nuclear Design Engineering i

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G. Hettel, Supervisor, Unit 2 System Engineering R. Hutchinson, Vice President Operations D. James, Manager, Nuclear Safety K. Jeffery, Security R. Lane, Director, Design Engineering B. McBride, Unit 2 Operations D. McKenney, Unit 1 System Engineering Supervisor T. Morrison, Modification Superintendent H. Northrop, Turbine Coordinator S. Pyle, Licensing Specialist A. Remer, Unit 1 Mechanical Maintenance Coordinator J. Smith, Jr., Radiation Protection Manager C. Zimmerman, Unit 1 Plant Manager

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-2-INSPECTION PROCEDURES USED 37551 Onsite Engineering 61726 Surveillance Observations 62707 Maintenance Observations 71707 Plant Operations 71750 Plant Support Activities 92700 Onsite LER Review 92901 Followup-Plant Operations 92903 Followup-Engineering issues ITEMS OPENED AND CLOSED 50-368/9904-01 NCV Failure to shut EFW sample valves in accordance with sample procedure (Section O2.1)

l 50-368/9904-02 NCV Failure to establish alternate radioactive gaseous sampling as required by TS 3.3.3.9 (Section O8.2)

50-368/9904-03 NCV Fuel handling area ventilation flow rates less than TS requirement during crane operation with a load over the storage pool due to personnel error '

(Section 08.4)

50-313/9904-04 NCV Violation of TS 4.5.1.1.2 for failure to test Valve CV-3643 (Section E8.2)

Closed 50-313/97-003-00 LER Inadvertent automatic actuation of EFW system resulting from lightning induced trip of two EFW initiation and control system power supplies (Section 08.1)

50-368/97-003-00,-01 LER TS allowing operations with one channel of refueling water tank level or steam generator differential pressure tripped established the potential for premature initiation of a recirculation actuation signal during a loss-of-coolant accident (Section E8.1)

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-3-l 50-368/97-004-00 LER Alternate radioactive gaseous effluent sampling not established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> as required due to inadequate alarming capabilities on radiological dose assessment computer system terminals l j

(Section 08.2)  !

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50-313/97-005-00 LER Surveillance testing did not verify Valve CV 3643 l would close on an engineered safeguards signal l (Section E8.2)

50-368/97-006-00,-01,-02 LER Inadvertent EDG start which resulted from l inadequate written guidance regarding system

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component testing (Section O8.3)

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50-368/97-007-00 LER Fuel handling area ventilation flow rates less than TS requirement during crane operation with a load over the storage pool due to personnel error ,

(Section 08.4)

50-313;50-368/9718-02 IFl Review of licensee's corrective actions for MSSV setpoint deficiencies (Section E8.3)

50 313;50-368/9718-04 VIO Failure to perform a safety evaluation in making changes to the Units 1 and 2 FSARs and to Procedure 1000.120, "ANO Fire Watch Program" (Section E8.8)

50-313/9721-01 VIO Procedure 1010.002, " Transient History / Transient Cycle Logging," inappropriate to the circumstances (Section E8.9)

50-313/9721-06 VIO Failure to initiate temporary alteration for the BWST work (Section E8.4)

50-313/9721-07 VIO Inadequate procedure for BWST work (Section E8.5)

50-313/97201-09 IFl Use of commercial grade cable ties in containment and use of commercial grade PVC flexible conduit (Section E8.6)

50-313/97201-13 IFl EDG exhaust stack tornado-driven missile evaluation (Section E8.7)

50-368/9801-01 VIO Failure to include the turbine building sump in the scope of the maintenance rule (Section E8.10)

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-4-s 50-313/9801-03 VIO Failure to establish appropriate performance measures for availability of risk-significant systems, structures, and components (Section E8.11)

50-313;50-368/9801-04 VIO Four examples of failure to monitor the performance or condition of certain systems against established goals (Section E8.12)

50-368/9801-07 VIO Failure to ensure that the 125 Vdc system was capable of performing its intended function (Section E8.13)

50-313/98-002-00 LER Automatic actuation of the control room emergency ventilation system due to higher than normal .

I radiation at the detector when a filtration / vacuum assembly was moved in an adjacent area (Section 08.6)

50-368/9803-02 IFl Corrective actions addressing RCS level indication j anomalies (Section 08.5)  !

50-313/98-004-00 LER inadvertent actuation of the EFW system j (Section M8.1)

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50-368/9812-02 VIO Failure to ensure that Engineering TS Sheet 4 provided the appropriate phase rotation as specified in Electric Power Research Institute guidelines (Section E8.14)

50-368/9812-03 VIO Failure to follow procedure in making a change to the Unit 2 calibration standard dimensions sheet (Section E8.15)

50-368/9904-01 NCV Failure to shut EFW sample valves in accordance with sample procedure (Section O2.1) .

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50-368/9904-02 NCV Failure to establish alternate radioactive gaseous sampling as required by TS 3.3.3.9 (Section 08.2) i 50-368/9904-03 NCV Fuel handling area ventilation flow rates less than TS requirement during crane operation with a load I over the storage pool due to personnel error l (Section 08.4)

50-313/9904-04 NCV Violation of TS 4.5.1.1.2 for failure to test Valve CV-3643 (Section E8.2)

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l LIST OF ACRONYMS USED BWST borated water storage tank CR condition report j EDG emergency diesel generator i EFW emergency feedwater HPl high pressure injection IFl inspection followup item LER licensee event report MSSV main steam safety valve PVC polyvinyl chloride NCV noncited violation RCS reactor coolant system TS Technical Specification VIO violation

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