IR 05000382/1997008

Insp Rept 50-382/97-08 on 970406-0517.Violations Noted. Major Areas Inspected:Operations,Maintenance,Engineering & Plant Support

ML20141B603
Person / Time
Site:	Waterford
Issue date:	06/19/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20141B560	List: IR 05000382/1997008 ML20141B556 ML20141B590
References
50-382-97-08, 50-382-97-8, NUDOCS 9706240063
Download: ML20141B603 (27)

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ENCLOSURE 2

1 U.S. NUCLEAR REGULATORY COMMISSION  !

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REGION IV

l l Docket No.: 50-382 License No.: NPF-38 Report No.: 50-382/97-08 Licensee: Entergy Operations, In Facility: Waterford Steam Electric Station, Unit 3 i Location: Hwy.18 Killona, Louisiana Dates: April 6 through May 17,1997 Inspectors: L. A. Keller, Senior Resident inspector T. W. Pruett, Resident inspector l G. A. Pick, Senior Project Engineer G. E. Werner, Project Engineer D. L. Proulx, Resident inspector, River Bend J. F. Melfi, Resident inspector, Arkansas Nuclear One Approved By: P. H. Harrell, Chief, Project Branch D ATTACHMENT: Supplemental information

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9706240063 970619 PDR ADOCK 05000302 G PDR

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EXECUTIVE SUMMARY

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l Waterford Steam Electric Station, Unit 3 l

NRC Inspection Report 50-382/97-08 l I

j This routine, announced inspection included aspects of licensee event response, operations, maintenance, engineering, and plant support. The report covers a 6-week period of resident inspection.

! Operations

• Observed operations activities were generally well coordinated and consistent with  ;

safe operation of the facility, which included a well executed plant shutdown and I draining the reactor coolant system (RCS) for midloop operations (Section 01.1).  !

A new fuel assembly was inadvertently dropped from the spent fuel handling tool during refueling because of personnel error. A noncited violation was identified for an inadequate procedure. The root cause analysis was thorough and represented good self-assessment (Section 01.2).

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Control room supervisors demonstrated conservative decision making during reduced inventory conditions (Section 01.3).

The Shift Support Center was effective in minimizing distractions and providing i additional personnel to aid the control room staff when necessary (Section 01.3).

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Two lapses of control room discipline were observed that involved reading of nonjob-related material in the control room at-the-controls area (Section 04.2). t

• A weakness in operator knowledge was observed when onshift operators could not j explain why shutdown cooling pressure appeared to be high (Section 04.2).

Quality Assurance performed a very good self-assessment by identifying that >

licensed individuals did not have corrective eyewear compatible with emergency ,

breathing apparatus and a plant procedure for proficiency watches inappropriately  !

deviated from 10 CFR 55.53e. These inadequacies were identified as noncited violations (Section 07). -

Maintenance

• Observed maintenance and surveillance activities were generally performed in ,

accordance with procedures and achieved acceptable results (Section M1.1). ,

A noncited violation was identified for the failure to conduct inservice testing of dry cooling tower isolation valves within the required surveillance interval (Section M8.2).

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• Engineering identified and promptly resolved a design deficiency involving l postaccident pH levels in the containment sump (Section E1.1).

! * An example of a violation of design control measures involved a failure to provide ( procedures and instruction to limit the electricalloading for the chemistry laboratory on an emergency diesel generator (EDG) (Section E1.2).

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• A violation was identified regarding the failure to maintain preventive maintenance procedures for safety-related breakers (Section E1.2).

• The failure to appropriately account for instrument uncertainty resulted in operation outside of the Technical Specification (TS) 3.6.2.1-required containment spray riser level and is a second example of a violation of design control (Section E2.2).

• A third example of a violation of design control resulted from adding an extra layer of filter material to a cabinet housing a core protection calculator without an evaluation regarding the effects on cooling air flow (Section E8.1).

• The failure to ensure that all design basis inputs in the groundrules basis document i accurately reflected design basis information is a fourth example of a violation of design control (Section E8.2).  ;

l Plant Supoort

• The use of a remote acquisition display system (RADS) to remotely monitor dose rates in the facility was considered a good health physics practice (Section R1.1).

• The inspectors identified examples of poor health physics work practices involving lack of labeling of radioactive material and inappropriate storage of flammable material (Section R1.2).

• A tour of the protected area determined that accessible areas were adequately illuminated (Section S1.2).

• The reactor coolant pump (RCP) lube oil drain and fill systems were well maintaine The licensee did not have the appropriate administrative controls to verify that the oil transferred into the RCP remote filllines actually reached the reservoir, (Section F2.1).

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R_e. port Details

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Summarv of Plant Status i

The plant operated at 100 percent power until April 11,1997, when the plant began a scheduled shutdown for Refueling Outage {

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l. Operations 01 Conduct of Operations

! 01.1 General Comments (71707)

l The inspectors performed frequent reviews of ongoing plant operations, control i

room board walkdowns, and plant tours. Observed activities were generally l

performed in a manner consistent with safe operation of the facility. Housekeeping

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and material condition were generally good. The inspectors observed good operator performance during the plant shutdown from 100 percent power. Draining of the RCS for midloop operations was well coordinated. The Shift Support Center was effective in minimizing distractions and providing additional personnel to aid the control room staff when necessar .2 Fuel Assembly inadvertentiv Drocoed in Soent Fuel Pool (SFP) insoection Scope (93702. 71707)

During refueling activities, an SFP bridge operator inadvertently dropped a new fuel assembly in the SFP. The inspectors reviewed licensee actions with respect to ,

stabilizing the fuel assembly and determining the root cause ' Observations and Findinas On April 28, at 9:37 p.m. (CDT), a new fuel assernbly (LAR338) was lifted out of the spent fuel rack and moved severalinches when it fell from the spent fuel handling tool. The assembly fell vertically, approximately 3 inches, before hitting ,

the top of the spent fuel storage rack. The assembly then toppled over where it !

came to rest at approximately a 45 angle against the side of the SFP. No other

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fuel assemblies were damaged and there were no increases in radiation levels as a result of the dropped assembly, i

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To stabilize the dropped assembly in this position, the licensee generated ;

Procedure RF-TEM-001, " Dropped Fuel Assembly Stabilization." The licensee was '

successful in using this procedure to put a 40-foot sling and choker hitch around the fuel assembly, which secured it to the side of the SF j The licensee inspected the spent fuel handling tool to determine if the tool had degraded and could have caused the fuel assembly to drop. No problems were identified during this inspection. The inspectors independently inspected the tool and did not identify any damag :

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r -2-The licensee then generated Procedure RF-TEM-002, Recovery of Fuel i Assembly LAR338," to recover the assembly and put it into the new fuel elevato The inspectors attended the infrequently Performed Test or Evolution (IPTE) brief for the recovery of the fuel assembly and concluded that the brief was thorough. The licensee successfully recovered and returned the fuel assembly to the vendor for inspection. The vendor determined that approximately 30 percent of the fuel pins had damaged fuel pellets. All pins with damaged fuel pellets were replaced and the reconstituted assembly was shipped back to he license Nbh The inspectors reviewed the Ro t Cause Analysis, " Dropped New Fuel Assembly LAR338," completed by . personnel '- vo. . & The licensee attributed the root cause to personnel error. The report noted that the as-found condition of the fuel handling tool was 75 percent open and locked and, therefore, concluded that the fuel handling tool was not positioned correctly. There were a number of causal factors contributing to the personnel error including: inadequate self-checking, lack of independent verification, insufficient operating experience with moving fuel at Waterford 3, poor tool orientation, and inadequate procedur Corrective actions included revising Procedure RF-005-002, " Refueling Equipment Operation," to include additional details on the mechanics of operating the fuel handling tool, requiring a peer check that assemblies are adequately grappled, and retraining of all refueling operators. The inspectors considered the root cause analysis to be thorough and representative of good selfessessmen The inspectors concluded that the f ailure to maintain Procedure RF-005-002 is a violation of TS 6.8.1.a. This licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B1 of the NRC Enforcement Policy. Specifically, the violation was identified by the licensee, was not willful, actions taken as a result of a previous violation should not have corrected this problem, and appropriate corrective actions were completed by the licensee (50-382/9708-01). Conclusions A fuel assembly was dropped due to personnel error in not grappling the fuel assembly properly. The licensee wrote procedures to stabilize and recover the fuel assembly. The IPTE brief attended was thorough. The root cause analysis was thorough and represented good self-assessment. The refueling equipment operating procedure was inadequate, which is considered a noncited violatio .3 Midlooo Operations (71707)

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The inspectors observed draining the RCS for midloop operations between l April 14-18. Draining and midloop operations were well coordinated. Operators l were knowledgeable of plant conditions and indications that identify potentially l adverse conditions. Senior licensee management was present in the control room

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for most of the evolutions observe _ . _ _

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l 3-Control room supervisors demonstrated conservative decision making on several i

occasions, as noted by: (1) redirecting the focus of personnel assigned oversight l responsibility of shutdown cooling and reduced inventory evolutions away from distracting annunciators, (2) securing drain down evolutions on April 16 due to a level deviation of greater than 0.25 feet, and (3) delaying steam generator manway removal and nozzle dam installation until procedure discrepancies were resolve The use of the Shift Support Center to approve work activities significantly reduced the amount of personnel accessing the control room and provided an immediate availability of trained personnel to aid control room operators in monitoring plant condition Operator Knowledge and Performance 04.1 Plant Shutdown From 100 Percent Power (71707)

l The inspectors observed the plant shutdown that was conducted l April 11-12,1997. The inspectors observed that the plant shutdown was well coordinated and conducted in accordance with procedures and that members of plant management were in the control room overseeing the shutdown. Good command and control and communication discipline were exhibited by the control room operators. Plant equipment operated as expected and observed parameters were within established limit .2 Control Room Conduct (71707)

On April 15, the inspectors observed an onshift reactor operator reviewing health plan benefits on a personal computer located in the at-the-controls area of the l

control room. On April 16, during a drain down of the RCS for midloop operations, I the inspectors observed the onshift control room supervisor reviewing electronic mail messages on the same computer. This recently installed personal computer is l principally used for recording operator logs, filling out danger tags, and other administrative functions.

l l The inspectors determined that the review of health plan benefits and electronic mail while on watch represented a lapse in control room discipline. The inspectors were concerned that the licensee's addition of the personal computer to the control room was not implemented with sufficient restrictions that prohibited use during sensitive plant evolutions. Operations management indicated that direction would

, be provided to the operators to obtain permission from the control room supervisor

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or shift superintendent prior to using the computer for nonjob-related activitie On April 14, the inspectors observed that the indicated pressure of Shutdown Cooling Heat Exchangers A and B was in the red band at 520 psig. When questioned by the inspectors, the reactor operator and control room supervisor were unable to explain why the pressure was at this level. The operators questioned the l

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Shift Supervisor, who correctly stated that the pressure indication was due to the RCS pressure (320 psig) combined with the differential pressure of the low pressure safety injection (LPSI) pumps (200 psig). The inspectors determined that the l operators' unfamiliarity with shutdown cooling pressure indications represented a weakness in operator knowledg ' Quality Assurance in Operations (71707)

From February 21 through April 29, Quality Assurance performed an audit of operations departmental activities (SA-97-034.1). There were numerous good findings in this audit. One of the findings involved operators not having corrective eyewear, as required by their license, compatible with emergency breathing gear (i.e., respirator). The concern involved potential inability to carry out duties during a j radiological or toxic gas event due to reduced visibility. This resulted in a control room supervisor and shift technical adviser being relieved from their watch due to  !

not having contact lenses or respirator glasses. Condition Report (CR) 97-0771 I was generated due to this finding. The requirement to have appropriate eyewear l was emphasized with all operators and fire brigade members. Appropriate eyewear I

was issued to all applicable personne To address this issue, the licensee was in the process of revising Procedure 01-024-000, " Maintaining Active SRO/RO Status," Revision 5, to implement requirements for licensed operators to obtain respirator corrective eyewear. The failure to maintain Procedure 01-024-000 is a violation of TS 6.8.1. This licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.81 of the NRC Enforcement Policy. Specifically, the violation was identified by the licensee, was not willful, actions taken as a result of a previous violation should not have corrected this problem, and appropriate corrective actions were completed by the licensee (50-382/9708-02).

Another good Quality Assurance finding during this audit involved procedural guidance inadvertently deviating from the requirements of 10 CFR 55.53e regarding maintaining active operator licenses.10 CFR 55.53e requires that five 12-hour ,

shifts be performed per calendar quarter in order to maintain an active licens !

Quality Assurance determined that Procedure 01-024-000 allowed deviating from 10 CFR 55.53e without justification or NRC approval. Step 5.1.10.1 allowed active-license status to be maintained without standing five 12-hour shifts per calendar quarter by having up to three proficiency watches performed in the first ,

week of the next quarter count towards the previous quarter. 10 CFR 55.53e has  !

no such provision. CR 97-0812 was generated due to this findin :

Procedure 01-024-000 was revised to comply with 10 CFR 55.53e. The licensee j reviewed the status of all their active licenses and concluded that all personnel had I appropriately stood proficiency watches for the current cycl l

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The inspectors concluded that the failure to maintain Procedure 01-024-000 is a violation of TS 6.8.1.a. This licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B1 of the NRC Enforcement Policy. Specifically, the violation was identified by the licensee, was not willful, actions taken as a result of a previous violation should not have '

corrected this problem, and appropriate corrective actions were completed by the

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08 Miscellaneous Operations issues (92901)

08.1 (Closed) Violation 50-382/9613-01: Failure to follow procedure regarding configuration control (three examples). This item involved the failure of operators to recognize and maintain plant systems in the correct configuration. The licensee's corrective actions included, in part, restoration of the affected systems to the proper configuration, reenforcement of expectations regarding control panel walkdowns, a review of the shift turnover methodology, and a review of normal operating procedures to determine if improvements were necessary. The inspectors ,

determined that the licensee's corrective actions were adequate to resolve this issu .2 (Closed) Licensee Event Report (LER) 95-007: Essential chilled water system degradation. The licensee submitted this voluntary LER on degraded flows that were identified in their essential chilled water system. During a system flow test, the flow rates in some system branches were much lower than expected. Throttle valves set the flow rates in these branches, which had solid deposits on the valve seats. When the valves were cycled, chemistry samples showed a large increase in iron concentration. This event was reviewed at the time by NRC Inspection Reports 50-382/95-04 and 50-382/95-1 Subsequently, the licensee's evaluation concluded that the as-found flow rates would have removed enough heat from the equipment under potential accident conditions. The inspectors reviewed the analysis and deterrrined that this condition did not make the essential chilled water system inoperabl The inspectors questioned the chemical controls on the essential chilled water system to determine why large increases in iron concentration were noted. The licensee informed the inspectors that they changed the corrosior; inhibitor from a nitrate-based to a phosphate-based inhibitor, but the results were not satisfactory and the licensee reinstated the original corrosion control. This change may have caused iron deposits to collect in various portions of the system, but the nitrate-based system has not resulted in any further system degradatio . .

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6-li. Maintenance M1 Conduct of Maintenance (62707,61726)

M1.1 Gerieral Comments l The inspectors observed all or portions of the following maintenance and i

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surveillance activities:

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• WA 01156893 24-Hour EDG B Run (OP-903-116)

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• WA 01157338 Changeout of Hydraulic Actuator for SI-405A ,

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• WA 01159071 Welding of HPSI Header Vent Valve  :

in general, ti'e inspectors found the conduct of these maintenance and surveillance  !

activities to be good. All activities observed were performed with the work 3 authorization package and/or test procedures present and in active use. When applicable, e.ppropriate radiation control measures were implemented. The inspectors observed supervisors monitoring job progress and quality control personnel present whenever required by procedur M8 Miscellaneous Maintenance issues (92902)

M8.1 (Closed) Violation 50-3'82/9612-01: Failure to follow procedures for measuring and test equipment requirements. The licensee's corrective actions included a reemphasis on procedure compliance and self-checking, identification and review of -

a negative trend in human performance issues, and the development of a ,

multidiscipline team to review generic concerns associated with underlying cause The inspectors determined that the licensee's > 3rrective actions should be adequate to resolve this issu M8.2 (Closed) Violation 50-382/9613-04: Failure to perform inservice testing on the dry r cooling tower manual bundle isolation valves. Corrective actions included i successful testing of the affected valves, inclusion of the valves in the inservice testing program, initiation of a review of the design basis tornado event, and initiation of a review of the inservice testing design basis document by Engineering in support of the second 10-year interva The inspectors reviewed CR 97-0843, initiated April 11,1997, to document the l

failure to perform stroke testing of the dry cooling tower isolation valves within 92 days. -Initial testing had been performed on December 7,1996, and the extension date was April 1,1997. On April 11,1997, the licensee declared both trains of the ultimate heat sink inoperable due to the failure to perform required surveillance testing. The licensee invoked the provisions of TS 4.0.3 and completed testing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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-7-During discussions with the licensee, the inspectors determined that operations had initiated the procedure changes to place the affected valves in the testing program; however, final approval had not been obtained. The inspectors determined this

example was an isolated occurrence and that operations had a tracking and trending program in place to ensure most procedure changes were made prior to the next occurrence of the surveillance. In response to this discrepancy, the licensee initiated a review to develop additional enhancements to ensure procedure changes that are requested through alternative mechanisms are placed into the tracking j system. The inspectors concluded that the failure to perform inservice testing of '

the dry cooling tower isolation valves within the required surveillance interval is a ;

violation of TS 4.0.5.b. This licensee-identified and corrected violation is being j treated as a noncited violation, consistent with Section Vll.B1 of the NRC Enforcement Policy. Specifically, the violation was identified by the licensee, was not willful, actions taken as a result of a r,revious violation should not have corrected this problem, and appmoriate <,orrective actions were completed by the licensee (50-382/9708-04).

Ill. En_gineerina E1 Conduct of Engineering E1.1 Containment Sumo Boron Concentrations insoection Scope (37551)

The inspectors performed a review of the licensee's evaluation regarding postaccident containment sump p Observations and Findinas The bases for TS 4.5.2.d.4 require that at least once per 18 months verify that, when a representative sample of 410.01 grams of trisodium phosphate (TSP) from a TSP storage basket is submerged, without agitation, in 4 iO.1 liters of i 120 110 F water borated within refueling water storage pool boron concentration limits, the pH of the mixed solution is raised to 2:7 within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. On March 4, during a licensee evaluation of additional TSP for extended cycles, Engineering

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determined that, while using the new software methodology, sufficient TSP may not exist to neutralize boron solutions near the upper TS limi The licensee performed additional analyses and established administrative limits for the remainder of the operating cycle. The licensee added sufficient TSP to the containment sump during the current refueling outage to allow operation at the maximum allowed TS limits for boron concentrations. The inspectors reviewed Design Change (DC) 3491, which described the inclusion of additional TSP baskets and performed a walkdown of the installation of the TSP baskets in the containment. The licensee installed the TSP baskets per design drawings with l

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-8-some exceptions that were appropriately dispositioned for resolution. During the walkdown, the inspectors noted that the original baskets contained aggregated TSP and the inspectors were concerned that the rate for dissolving the TSP in the containmerd sump water may not be within design basis assumption The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR) and Safety Evaluation Report for descriptions of the TSP baskets. Safety Evaluation Report Section 6.5.2 described TS 4.5.2 as requiring periodic inspection of the TSP baskets for evidence of aggregation and the mechanical dispersion of any aggregation found. Subsequent discussions with chemistry and licensing personnel identified that testing had previously been completed on this concern and TS 4. was changed to remove this requirement as part of License Amendment 8. The inspectors reviewed Amendment 8 and the associated Safety Evaluation and determined that the current TSP aggregation was acceptabl Conclusions Engineering identified and promptly resolved a design deficiency involving postaccident pH levels in the containment sum E1.2 Desian Control Issues Inspection Scope (37551)

The inspectors reviewed several DC packages associated with Problem Evaluation /Information Requests (PEIR) to ensure compliance with regulatory requirement Observations and Findings As part of the review of the PElR process, the inspectors reviewed several DC packages. The DC packages were generally adequate; however, concerns were identified for DC 3055. DC 3055 was initiated to provide power to chemistry lab equipment during a loss of offsite power via safety-related,480-Vac Bus 3AB311-S, which could be powered from either EDG. DC 3055, Revision 0, planned to install one transformer, rated at 7.5 kVA, and one lighting panel (Panel LP-3003);

however, after further reviews of the DC, engineers determined that the one lighting panel was not capable of supplying all the required chemistry loads. Subsequently, DC 3055, Revision 1, added Transformer LVD-EMT-311 AB-6FR2 (rated at 7.5 kVA)

and Panel PDP-305 DC 3055, Revision 0, contained a safety evaluation that increased EDG loading by 8 kW. The UFSAR and the EDG load calculation were updated to include 7.5 kW from Panel LP-3003; however, the inspectors identified that DC 3055, Revision 1, l did not contain a safety evaluation and the UFSAR was not updated to include Panel PDP-3055. Engineering stated that loads on both distribution panels were I

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controlled to be less than 7.5 kW by the Chemistry Department Guide, as indicated in DC 3055, Revision 1. Several days later, Engineering produced the original signed and completed safety evaluation for DC 3055, Revision 1, which had never been placed in the completed DC package. The licensee initiated CR 97-773, which documented the failure to place the safety evaluation in the DC record.

l l The inspectors reviewed the safety evaluation for DC 3055, Revision 1, and found '

l that the safety evaluation relied on the developmenc of tables in the Chemistry l Guide, which were to list permissible loads to be placed on the two electrical j panels. These listed loads were intended to be developed to ensure that the l additional loading attributed to all the chemistry loads was limited to 7.5 kW. The l inspectors reviewed Chemistry Department Standing instruction 32, "Use of Safety Bus Power in the Primary Chemistry Lab and Count Room," and identified that the

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l instruction did not list permissible loads, it only listed electrical circuits. The

licensee was unable to supply a list of loads for either of the electrical panels.

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The inspectors performed a walkdown of the chemistry lab and noted that Panels LP-3003 and PDP-3055 included electrical outlet plug connections or power l strips with numerous electrical outlet plug connections. The arrangement would l allow any portable 120-Vac electrical equipment to be plugged into the outlets. The l inspectors identified that the electrical plug connections were not labeled to identify l that circuits were restricted to only certain specified loads.

Maintenance personnel measured the electricalload on both circuits and determined present loading was less than the maximum assumed load of 7.5 kW. However, l the panels had the combined capability to supply approximately 15 kW of loads, which would have exceeded the reviewed loading by a factor of 2. The failure to provide procedures and instructions to limit the electricalloading of an EDG as i specified in DC 3055 is an example of a violation of 10 CFR Part 50, Appendix B, ( Criterion lll (50-382/9708-05).

Further review by the inspectors identified that safety-related Breakers LTN-EBKR-311 AB-6FL and -6FR, which isolated the safety-related portion of the 480-Vac electrical system from the nonsafety-related chemistry loads, did not have any type of preventive maintenance or periodic testing to ensure that they would operate correctly when required to isolate the circuit. Technical Manual 457000956, "lTE Gould Molded Case Circuit Breakers Testing and Maintenance Procedures," specified that the breakers should be exercised and tested at times to ensure proper functioning. The Technical Manual listed various l testing to demonstrate proper breaker operation. The licensee initiated CR 97-769 i l to identify that the preventive maintenance task for the two breakers had been l improperly deleted. The failure to maintain procedures and/or instructions to

periodically exercise and test the breakers is a violation of TS 6.8.1

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(50-382/9708-06). )

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- 10- Conclusions DC packages reviewed were generally adequate. However, DC 3055 was inadequate in that: (1) the safety evaluation for Revision 1 to the package was inadvertently left out of the package, (2) Revision 1 added additional electrical outlets to a bus that would be loaded onto an EDG without adequate controls to ensure the loads would remain below the analyzed value, and (3) required preventive maintenance for safety-related breakers was inadvertently delete E2 Engineering Support of Facilities and Equipment E2.1 Review of Containment Atmosphere Release dvstem (CARS) (37551)

UFSAR Section 6.2.5.2.3 described the CARS as a backup to the hydrogen recombiner system, which following a loss-of-coolant accident (LOCA) transfers combustible gases from inside the containment to the reactor building annulu Section 6.2.5.2.3 of the original FSAR submitted in 1978 specified that the CARS would maintain hydrogen concentration below 4 volume percent following a LOC Safety Evaluation Report, July 1981, Section 6.2.5, specified that the CARS was provided in addition to the hydrogen recombiner system in accordance with 10 CFR 50.44(e), Regulatory Guide 1.7, and Branch Technical Position CSB 6- Additionally, the CARS was a redundant system that, if required, purges containment atmospher The maximum operating pressure of the CARS supply fans, which dilute the containment atmosphere, is 2.0 inches of water gauge (iwg). The maximum design pressure of the CARS system is 12.0 iwg. TS 3.6.1.4 allowable containment pressure is between 14.375 psia and 27 iwg. The licensee conservatively estimated that the 30-day post-LOCA containment pressure would be approximately 19.0 psia. Because the TS 3.6.1.4 allowable pressure and the post-LOCA accident pressure exceeded the operating pressure of the CARS, the inspectors questioned crediting the CARS for compliance with 10 CFR 50.44(e). The licensee maintained that the CARS was not essential for safe shut down, would only be operated in the unlikely event that both recombiners fail (the licensee noted that the ability to withstand two failures was not a design consideration), and that the system provided the capability for controlled purging of the containment to aid in postaccident cleanu NRC Region IV submitted Task Interface Agreement 97TIAOO2 to the Office of Nuclear Reactor Regulation (NRR) on January 17,1997, to determine if CARS met regulatory requirements. NRR concluded that Waterford 3 was in compliance with all governing criteria. NRR determined that the fully redundant recombiners are all that are required to control the hydrogen concentrations below the lower flammability limit following a LOCA and that the CARS was not related to the control of hydrogen. The criteria of Regulatory Guide 1.7 is to provide a system for post-LOCA cleanup. Therefore, there are no rate requirements, redundancy, nor

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l safety class criteria associated with the CARS. The staff wanted the ability to vent l containment when the need is established during the cleanup process following a LOCA. NRR also determined that since the CARS is only needed for post-LOCA recovery, it is assumed that there will be no containment pressure when the system is called upon. Therefore, there is no containment pressure at which the system is required to be functiona E.2.2 Containment Sorav Riser Level and LPSI Flow Loon inaccuracies Scope (92903)

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The inspectors evaluated Engineering's resolution of inaccuracies related to the i containment spray riser level instrument and the LPSI flow loop uncertainty, as j documented in CRs 97-0682 and 97-0649, respectivel .

l Observations and Findinas Containment Sorav Riser Level ,

On March 21,1997, engineers determined that operation of the controlled ventilation area system decreased the containment spray riser levelinstrument indication by 5 inches. This effort was being performed as a result of level indication problems identified with the refueling water storage pool, as documented in NRC Inspection Report 50 382/97-12. TS 4.6.2.1 requires that operators verify that the containment spray riser level is 2: 149.5 feet mean sea level (181 feet)

once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The containment spray riser level ensures, in the event of a LOCA or main steam line break, that full containment spray is achieved within the time limits assumed in analysis, which ensures peak containment pressure does not i exceed the design value of 44 psi However, Engineering determined that TS 4.6.2.1 failed to account for instrument uncertainty, Calculation EC-191-027, " Containment Spray Riser Level A & B instrumentation Loop Uncertainty Calculation," Revision 0, demonstrated that the instrument uncertainty for the levelinstruments could be as much as 9 fee Similarly, the containment pressure and temperature analysis failed to account for instrument uncertainty. The licensee iaitiated CR 97-0682 to document the potendai ici faiiing to mee't the upesab!!itt/ requirements for the TS limiting condition

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for operation. As immediate corrective action, operators raised the containment spray riser level for Train B to greater than 190 feet and verified the Train A level to be greater than 190 feet to ensure compliance with TS 3.6. The failure of the instrumentation circuit to account for the appropriate design j uncertainty resulted in a f ailure to comply with TS 3.6.2.1 and is the second l example of a violation of 10 CFR, Appendix B, Criterion 111(50-382/9608-05).

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-12-The !!censee performed an analysis to evaluate whether the containment building peak design pressure would have been exceeded, assuming worst case instrument uncertainty. The licensee initially determined that a maximum pressure of 43.97 psig would result during the worst case accident. From discussions with licensing personnel, the inspectors noted that on numerous occasions in the past the containment spray riser level was maintained between 181 and 190 feet for significant period The inspectors reviewed Calculation EC-S93-002, "75% Power Main Steam Line Break With MSIV Failure and With 1.5 Second Delay," Revision 0, assessed the impact on the peak design pressure for a 1.5 second delay in spray flow. The licensee selected the 1.5 second delay because it corresponded to a 10.68 foot decrease in the containment spray riser level assumed in the safety analysis and bounded the 9-foot uncertainty. The calculation indicated that the peak containment pressure would be 43.97 psi Because of recent concerns related to lew containment fan cooler flows and the resultant increase in containment pressure (refer to NRC Inspection Report 50-382/97-03), the inspectors questioned licensee personnel as to whether the analysis for this deficiency considered the degraded containment fan cooler conditions, since both deficiencies existed at the same time. The subsequent CONTEMPT evaluation with only two containment fan coolers in operation resulted in peak containment pressure slightly exceeding the design pressure by 0.05 psi (44.05 psig).

A design engineer who reviewed the revised CONTEMPT design inputs for accuracy identified that the vendor failed to use 1.3 seconds in the initial conditions as a delay period in spray flow initiation and subsequently contacted the vendor regarding the error. The vendor performed another CONTEMPT computer code run with a 1.3 second delay and the resulting peak pressure was 43.93 psig. The 1.3 second delay corresponded to a decrease of 9.5 feet in the containment spray riser below the TC minimum, which bounded the instrument uncertaint LPSI Flow Loon On March 19,1997, while revising Calculation EC-191-052, "LPSI Flow Loop Uncertainty Calculation," to incorporate data from Calculation EC-196-002, " Orifice Calculation," a design engineer noticed that the instrument uncertainty for this application was 725 gpm. These instruments ensure compliance with TS 4.5.2.h, which requires during a flow test that the single pump flow be 24810 gpm. The design engineer determined that the analyti::al limit for TS 4.5.2.h was 4275 gp The engineer determined that a minimum flow of 5000 gpm would be required to ensure the analytical limit was enveloped, which exceeded the allowable TS minimum flow value and provided an opportunity to be outside the design basi . -. - - - __ .- -

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Subsequently, the design engineer initiated CR 97-0649 to identify this deficiency and ensure corrective actions were implemented. The licensee reviewed Procedure OP-903-108, "LPSI Flow Balance," conducted on October 17,1995,and determined that the LPSI Trains A and B flows were 5300 gpm and 5400 gpm, j respectively. The licensee requested from the nuclear steam supply system (NSSS)

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vendor a confirmation of the LPSI flow rate used in the accident analyses. The vendor indicated that, until the most recent core reload analysis, a value of 4084 gpm instead of 4275 gpm had been used. The results indicated no impact l

since the lower flow rate affected the amount of water that flowed from the brea The NSSS vendor had determined m December 1983 that the capacity of a single l LPSI pump at 15 psia, with one diesel generator out of service, was 2137.5 gpm; 4 l therefore, the total LPSI flow is 4275 gpm at 15 psia. The NSSS vendor indicated

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that the 4084 gpm flow rate was based bpon a similar facility and that they had l continued to use 4084 gpm since this value was more limitin !

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l The inspectors reviewed test data for the LPSI Trains A and B flow tests for l Cycles 3-6 and confirmed that the flows exceeded 5000 gpm. The inspectors '

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noted that the licensee could have had flows less than 5000 gpm, since the

! acceptance criteria was 2 4810 gpm. The inspectors noted that the vendor j information demonstrated that the system was always capable of performing its ,

l safety function and that sufficient margin existed. The 4084 gpm flow used in !

l previous cycles was selected to ensure that the design remained conservative. The

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inspectors noted that 725 gpm added to the original 4084 gpm approximated the TS 4.5.2.h flow rate. This deficiency did not result in exceeding a TS limit nor did it l result in the plant operating outside of its design basis. However, the inspectors

! concluded that the deficiency was the same as the containment spray riser level l deficiency identified above in that the licensee did not have a minimum specified flow that accounted for the instrument uncertaint The licensee initiated a long-term corrective action to use vendor guidelines to review, evaluate, and document Waterford 3 TS instrument uncertainty calculation This long-term program was scheduled to begin by October 1996 and is presently scheduled for completion in November 1997. The licensee determined that 130 individual instrument uncertainty determinations would be needed for the TS limiting conditions for operation. The inspectors initiated an inspection followup item to ensure review of the program and results related in determination of uncertainties for TS limiting conditions for operation (50-382/9708-07).

c. Conclusions The inspectors concluded that the licensee appropriately addressed the instances related to instrument uncertainty affecting or potentially affecting operaHUty of safety-related equipment. An inspection followup item to review the limsee's actions to address this concern was initiated. A second example of a design control

violation was identified for the failure to account for instrumentation uncertaint _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ . _ _ . . _ -. _ _ . ~ _ _ _ _ . - _

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- 14-E8 Miscellaneous Engineering issues E (Closed) Unresolved item 50 482/9702-06: Review open PElRs to determine if ,

issues were adequately addressed. NRC Inspection Report 50-382/97-02 identified l concerns with the management of the PElR process, which included: (1) no j requirement existed for followup if the due date elapsed, (2) licensee could not j determino the status of numerous PElRs, (3) personnel routinely bypassed the log coordinators who provided tracking, and (4) licensee management was unaware of the status of the program. Although PElRs by definitiori did not involve operability l

concerns, the failure to appropriately manage PElRs or other processes for resolving technical concems that do not initially meet the threshold for a CR could result in the failure to resolve an issue before it becomes an operability issu As of March 31,1997, the licensee located all open PElRs. Most of the open PEIRs had due dates long since expired with no apparent answer to the documented technical problem. The licensee rescreened each of the open PElRs, assigned new l due dates for resolution, and reviewed the PEIRs for safety and operability concerns. No safety or operability concerns were identified by the licensee as a result of this review. As of the end of this inspection, all of the open PElRs were l

being tracked through the Waterford Action Tracking System and a monthly meeting to discuss the status of open PElRs was being held. In addition, the licensee was in transition to the Engineering Review process, which was an j Entergy-wide initiative for improving the management of engineering work. The

[ inspectors concluded that the concerns regarding the management of the PElR process had been adequately resolve Subsequent to the licensee's review of open PElRs, the inspectors reviewed the i PElRs and noted that PEIR OM-90, initiated in July 1995, questioned a j long-standing practice ("several years") of adding additional filter material to the original metal filter screens in the cooling air supply to safety-related Cabinet CP-22, which housed a safety-related core protection calculator. The additional filter l

material was added to remove more dust prior to the air stream reaching the core protection calculator. The inspectors considered the addition of l unapproved /unanalyzed filter material to the original configuration a de facto field change and were concerned that the cooling flow may have been reduced to the l

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core protection calculator, potentially affecting its performance. The inspectors noted that this action had been taken without design documentation or engineering l calculations to verify the configuration's acceptability. The inspectors informed the

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licensee and CR 97-1045 was written. The failure to apply design control measures when adding additional filter material to Cabinet CP-22 is the third example of a violation of 10 CFR, Appendix B, Criterion 111(50-382/9708-05).

3 E8.2 (Closed) Unresolved item 50-382/9605-07: Review of identified groundrules

! document discrepancies. The inspectors initiated this unresolved item to ensure j review of discrepancies in the Cycle 8 groundrules document after a vendor l identified a significant error in the groundrules document for maximum emergency i

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feedwater flow. The identified error was related to a maximum assumed emergency [

feedwater flow of 1050 gpm, whereas the maximum possible flow was 2300 gp ~

CR 96-0704 was issued to document this problem and identify the root causes as:

(1) failure to use the correct steam generator pressure for a main steam line break, which resulted in the low emergency feedwater flow, (2) inadequate references for groundrules parameters, and (3) inadequate review of the groundrules document by plant personnel. Corrective actions included: (1) evaluating the impact of higher maximum emergency feedwater flow on the applicable accident analyses, (2) reviewing the Reload Groundrules Bases document, which was developed by the vendor to assist plant personnel in reviewing and understanding the groundrules parameters, (3) resolving the groundrules document parameter discrepancies and identifying specific references, and (4) completing a self-assessment of the ,

groundrule l The inspectors evaluated the corrective actions for CR 96-0704 and the associated root cause analysis. The vendor demonstrated that the positive reactivity addition caused by emergency feedwater addition remained conservative. The inspectors reviewed the vendor evaluation of the effects of maximum emergency teedwater flow (at 400 psig) to the intact steam generator during a main steam lins break and

, concluded that the increased emergency feedwater flow had negligible effect on the

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reactivit The inspectors evaluated the significance and the affect on the safety analysis for the other identified discrepancies. The inspectors discussed each individual discrepancy, including the vendor assessment of each discrepancy, with the engineer responsible for maintaining the groundrules document. The inspectors verified that the licensee had established for each parameter in the Cycle 9 groundrules document a valid reference, such as a vendor or licensee calculation, l TS value, and vendor specification. The inspectors noted that the licensee l determined that the following groundrules document parameters were incorrect (list not inclusive):

• maximum zero power inlet temperature

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• minimum pressurizer heat loss through walls l

• maximum and minimum letdown flow values  ;

l * steam generator inventory versus power table '

• steam generator pressure at full power

• maximum feedwater flow rat The failure to ensure that all design inputs in the groundrules basis document accurately reflected design basis information is the fourth example of a violation of 10 CFR, Appendix B, Criterion lli (50-382/9708-05).

i j As corrective action to prevent recurrence, the licensee ensured that each I groundrules document design parameter had a valid reference. Following

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l discussions with the inspectors, the licensee initiated actions to establish formal methods to ensure that calculations that change values listed in the groundrules document result in changes to the groundrules parameters. The licensee I established a calculation checklist item to require comparison to the current version i

! of the groundrules document. Similarly, the licensee established procedure requirements for engineers implementing design changes to review for effects on i the groundrules document.

l E8.3 (Closed) Inspection Followuo item 50-382/96202-16: Adequacy of operability confirmation procedure. Specifically, the inspectors wanted to ensure that, during l evaluation of a CR related to nonconforming conditions, personnel would not take ;

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an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to determine that TS 4.0.3 needed to be entered for a missed surveillanc I The inspectors reviewed Procedure W4.101, " Operability / Qualification Confirmation Process," Revision 1, and determined that the procedure does allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the time limit of the TS limiting condition for operation for performing operability evaluations related to nonconforming equipment conditions. The inspectors noted that this procedure makes no mention of entry into TS 4.0.3. Further, the inspectors noted that operators have had opportunities over the last year to enter TS 4.0.3 and the applicable limiting condition for operation. The inspectors verified that Procedure OP-100-014, " Technical Specification and Technical Requirements Compliance," Revision 5, required that, when invoking TS 4.0.3 to complete a surveillance requirement, determine the applicable limiting condition for operaimo to be entered and log that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are allowed for completion of the surveillanc I E8.4 (Closed) Violation 50-382/9605-05: 10 CFR 50.59 safety and environmental impact screening. This violation resulted from a failure of engineers to perform a !

safety evaluation for the installation of 50 x 40-ft curtains in front of the Train A I wet cooling tower basin. The inspectors reviewed the corrective actions for this violation along with the corrective actions for Licensee Event Report 96-005. The licenseo performed a root cause analysis that determined: (1) poor administrative controls for use of engineering inputs, required level of review, and need for independent review; (2) lack of standards and expectations for the use of engineering inputs; (3) less than complete work instructions; and (4) ineffective interdepartmental communication i immediate corrective actions included issuing a letter to all Engineering, Operation, I and Construction personnel to provide guidance on the limitations of Engineering i inputs, required interfaces between Operations and Engineering, and new l requirements for an independent reviewer of Engineering Inputs. In addition, ( operations issued Stand!:7 Instruction 96-07 that required operators to verify that

, all Engineering inputs had two signatures, compare the engineering input against

TS, and ensure that for Enginaering inputs that support an operability determination

a 10 CFR 50.59 safety evaluation is completed. As long term corrective actions, the licensee revised Procedure UNT-007-053, " Engineering Work Authorization," to i  ;

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specify that Engineering inputs will not be used to make operability determinations

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l or configuration changes, all nonconforming conditions will have a CR initiated, -

Engineering inputs require a second technical reviewer signature, and complex engineering questions will be asked and answered with a PEl ;

The inspectors confirmed that the licensee had modified Procedure UNT-007-053 to limit the use of Engineering inputs. The licensee incorporated the requirements for !

operator review of engineering inputs into Procedure OP-100-014, " Technical Specification Compliance." Also, the licensee recently implemented new processes l for requesting and receiving engineering assistance. Procedure W4.104,

" Engineering Request," will be used to request and receive engineering suppor The inspectors noted that over the past year the licensee successfully prevented i inappropriate use of Engineering inputs by the administrative controls that were established and by increasing management expectations. The inspectors determined that the licensee had provided continuing training to Engineering personne IV. Plant Support R1 Radiological Protection and Chemistry Controls l

l R1.1 Use of a RADS to Remotelv Monitor Dose Rates (71750)

t During tours of the facility, the inspectors noted the use of a RADS to monitor the dose rates in various areas of the plant. The inspectors reviewed the use of a RADS to determine how the licensee was utilizing the system. A RADS provides the capability to remotely monitor accumulated dose and active dose rates. The licensee utilized a RADS in the safeguards pump rooms to monitor dose rates during shutdown cooling evolutions and a RADS was used in containment to monitor workers performing activities in high radiation areas. Inputs from a RADS were added to the craft supervisor work stations, Shift Support Center, and outage l l

control work stations to provide workers and supervisors the capability to obtain l real time dose rate information updates. The licensee was evaluating additional uses of a RADS to determine if additional enhancements could be made to the as l low as reasonably achievable program. The use of a RADS to remotely monitor dose rates in the facility was considered a good health physics practic R1.2 Control of Radioactive Material a. Insoection Scope (71750)

The inspectors performed tours of the radiologically controlled area (RCA), on April 24 and May 6, to determine if the licensee had implemented effective controls i for radioactive materials.

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l-18-l b. Observations and Findinas l

On April 24, during tours of the reactor auxiliary building, the inspectors observed bags of radioactive material that were not sealed or labeled and bags of labeled radioactive material with improper contamination controls. Specific observations included: (1) an unlabeled bag of used anticontamination clothing outside the containment equipment hatch that was not located in a contaminated area and (2) an open and unlabeled bag of valve parts, gloves, and rags in Boric Acid Tank Room A (contamination surveys on the contents indicated less than

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1000 disintegrations per minute / smear). The licensee secured and labeled the bags

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following the inspectors' observation On May 6, the inspectors performed additional tours of the RCA. Specific observations included: An untabeled bag of used anticontamination clothing inside the waste ,

solidification building that was not located in a contaminated area. Surveys i

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indicated that the external dose rate was less than 2 mR/hr. The licensee l secured and labeled the bag following the inspectors' observatio I l An unattended radioactive materials vacuum cleaner hose in the Boric Acid Makeup Tank A room with the attachment end unsealed. The vacuum cleaner was not located in a contaminated area and the radioactive materials tag indicated that the internal contamination levels were 10,000 disintegrations per minute. The licensee sealed the vacuum cleaner hose following the inspectors' observatio . A can of spray paint labeled as extremely flammable in the hot tool issue room. Additionally, a flammable storage locker was not available in the hot tool room for the attendants to place flammable materials. The licensee initiated a CR, placed a flammable materials storage locker in the tool room, properly stored the flammable materials located in the tool room, and provided training to the attendant c. Conclusions The inspectors concluded that the identified examples involving lack of labeling for bags containing contaminated material and unapproved storage of flammable materials in a radioactive materials storage location represented poor health physics work practice . - - - - - . . - . . . . . - - - - - . - - - - - - - . - . - . _ . _ . - ~

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-S1 Conduct of Security and Safeguards Activities S Protected Area illumination (71750)

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On April 24 at 10 p.m., the inspectors performed a tour of the protected area and ;

determined that accessible areas were adequately illuminate )

l S8 Miscellaneous Security and Safeguards issues (92904)

S (Closed) Violation 50-382/9614-04: Failure to follow security lighting reporting i procedures. This item involved the failure of a security officer to report a deficient lighting condition in the protected area. Corrective actions included counseling of l the individual and supervisor, reenforcement to security officers on reporting deficient conditions, and a review of security procedures. The inspectors determined that the licensee's corrective actions should be adequate to resolve this

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issu F2 Status of Fire Protection Facilities and Equipment '

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F insoection of Reactor Coolant Pomo (RCP) Lube Oil collection System , Inspection Scoce (64100, 37551) i

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Based on an insulation fire from RCP lube oilleakage at anotner nuclear plant, the i

inspectors reviewed the licensee's RCP lube oil drain system and remote filllines to

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determine compliance with 10 CFR Part 50, Appendix R. The inspectors reviewed !

the lube oil collection system integrity and layout and selected licensee's procedures, Observations and Findinos The inspectors walked down the lube oil collection system approximately 1 week after the plant shut down. The inspectors found that the lube oil collection system was in some disarray due to ong:oing work, but substantially intact. The inspectors found the lube oil collection piping well supported and little evidence of oil on outside surfaces. The high pressure lobe oillines were shrouded and the high pressure lube oil pumps were in an enclosure to preclude potential spray. The inspectors identified several minor drawing discrepancies, which were promptly correcte The licensee used remote filllines to replace tube oilinto the RCP reservoirs to

, reduce potential radiation doses to operators. The inspectors walked down the i remote filllines and found the stainless steel pipes had a continuous slope and were l well supported, with braided lines connecting the remote filllines to the RCP. The j braided section is to allow for thermal growth and to prevent RCP vibration from i

affecting the remote filllines. There was no separate oil collection system under

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-20-the remote filllines, which does not meet Appendix R. The licensee previously recognized that their system did not meet Appendix R requirements and submitted an exemption request (Letter W3F1-97-0021, February 19,1997).

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The inspectors identified an inconsistency from the drawing requirements and the l remote fill system. The drawing required that the remote filllines be continuously sloped into the RCP lube oil reservoir. The inspectors noted two examples where the braided connections had a slight bend or swag where lobe oil could potentially I collect in the braided secdon. A section of the remote fillline from the wall to the RCP reservoir had been pushed down, which resulted in a continuous slope not being maintained. The licensee initiated CR 97-0985 and corrected the condition The inspectors reviewed a pending licensee exemption request to the Office of Nuclear Reactor Regulation to understand the licensee's controls when the remote fill lines were used. The inspectors reviewed the licensee's administrative controls that would be used if they had to use the RCP remote filllines, which were documented in Letter W3F1-97-0021. The licensee would use a repetitive task and monitor level changes in the reservoir (s) with direct communication from the field to the control room. The licensee would also remove any excess oil out of the lube oil i collection tan !

l l The inspectors concluded that the controls were generally adequate, except for monitoring the level in the reservoirs. The licensee's letter states, in part, that during communications between the technician in the field, it is verified that a level l increase change occurs in the reservoir. During questioning,'the inspectors !

determined that the licensee did not have a formal reservoir volume versus indicated I level curve for either the upper or lower reservoir. The licensee did have informal information on the volume versus indicated level for the upper reservoir. The licensee informed the inspectors that they would generate a formal volume versus level curve for both the upper and lower reservoir before startup. The licensee also resubmitted their exemption request to clarify the administrative controls discussed in the letter. Review of these licensee actions to verify appropriate implementation will be tracked as an inspection followup item (50-382/9708-08).

c. Conclusions The RCP lube oil drain and fill systems were wellinstalled and maintained. Several minor drawing discrepancies were identified, which were corrected. The licensee di'i not have a formal level curve to verify that the oil transferred into the RCP l remote filllines actually reached the reservoir.

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V. Manaaement Meetinas i

X1 Exit Meeting Summary i The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on May 20,1997. The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the ,

inspection should be considered proprietary. No proprietary information was '

identifie ,

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• 6 ie i ATTACHMENT

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SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee R. G. Azzarello, Manager, Maintenance C. M. Dugger, Vice-President, Operations E. C. Ewing, Director Nuclear Safety & Regulatory Affairs "

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T. J. Gaudet, Manager, Licensing T. R. Leonard, General Manager, Plant Operations  :

D. C. Matheny, Manager, Operations D. W. Vinci, Superintendent, System Engineering A. J. Wrape, Director, Design Engineering i

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INSPECTION PROCEDURES USED 37551 Onsite Engineering 61726 Surveillance Observations 62707 Maintenance Observations 64100 Postfire Safe Shutdown, Emergency Lighting And Oil Collection Capability At Operating And Near-Term Operating Reactor Facilities 71707 Plant Operations 71750 Plant Support Activities 92901 Followup - Plant Operations 92902 Followup - Maintenance 92903 Followup - Engineering 93702 Event Response i

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ITEMS OPENED. CLOSED. AND DISCUSSED Opened 50-382/9708-01 NCV inadequate fuel handling procedure (Section 01.2)

50-382/9708-02 NCV inadequate corrective lens procedure (Section 07)

50-382/9708-03 NCV Inadequate proficiency watch procedure (Section 07)

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{ 50-382/9708-04 NCV Failure to perform inservice testing of dry cooling tower ,'

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isolation valves (Section M8.2)

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l 50-382/9708-05 VIO Failure to maintain design control with four examples l (Sectionn E1.2, E2.2, E8.1, E8.2)

50-382/9708-06 VIO Failure to maintain preventive maintenance procedures for j safety-reiated breakers (Section E1.2)

50-382/9708-07 IFl Review results of licensee evaluation of adequacy of I

instrument uncertainties applied to TS parameters l (Section E2.2)

50-382/9708-08 IFl RCP oil fill administrative controls (Section F2)

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50-382/9708-01 NCV inaasauate fuel handling procedure (Section 01.2)

50-382/9708-02 NCV Inadequate corrective lens procedure (Section 07)

50-382/9708-03 NCV Inadequate proficiency watch procedure (Section 07)

50-382/9708-04 NCV Failure to perform inservice testing of dry cooling tower l isolation valves (Section M8.2)

50-382/9612-01 VIO Use of measuring and test equipment that did not meet the full scale range requirements for the procedure (Section M8.1),

j 50-382/9613-01 VIO Failure of operators to recognize and maintain plant systems in correct configuration (Section 08.1).

50-382/9613-04 VIO Failure to perform inservice testing on dry cooling tower manual bundle isolation valves (Section M8.2). )

50-382/9702-06 URI Review open PEIR to determine if issues have been adequately addressed (Section E8.1).

50-382/9605-07 URI NRC review of identified groundrules document discrepancies (Section E8.2)

50-382/96202-16 IFl Adequacy of operability confirmation procedure (Section E8.3)

50-382/95-007 LER Essential chilled water flow degradation (Section 08.2)

50-382/9605-05 VIO 10 CFR 50.59 safety and environmental impact screening (Section E8.4)

! 50-382/9614-04 VIO Security lighting problems (Section S8.1)

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l LIST OF ACRONYMS USED i

CR condition report  !

l l CARS containment atmosphere release system

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CFR Code of Federal Regulations DC design change ,

EDG emergency diesel generator gpm gallons per minute iwg inches water gauge kVA kilovolt-ampere kW kilowatt LOCA loss-of-coolant accident LPSI low pressure safety injection mR/hr milliroentgen per hour NRR Office of Nuclear Reactor Regulation NRC Nuclear Regulatory Commission NSSS nuclear steam supply system pH logarithm of the reciprocal of hydrogen ion concentration in gram atoms per liter psia pounds per square inch absolute l psig pounds per square inch gauge  !

PElR problem evaluation /information request RCA radiologically controlled area RCP reactor coolant pump l

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RCS reactor coolant system RADS remote acquisition display system i SFP spent fuel pool TS Technical Specifications TSP trisodium phosphate UFSAR Updated Final Safety Analysis Report f Vac volts-alternating current

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