IR 05000382/1990024

Insp Rept 50-382/90-24 on 901002-1113.Violations Noted. Major Areas Inspected:Onsite Followup of Events,Monthly Maint & Surveillance Observations,Operational Safety Verification,Ler Followup & Balance of Plant Insp

ML20058K070
Person / Time
Site:	Waterford
Issue date:	11/30/1990
From:	Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20058K063	List: IR 05000382/1990024 ML20058K060 ML20058K064
References
50-382-90-24, NUDOCS 9012070073
Download: ML20058K070 (28)

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8'- . ; ii e APPENDIX B

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U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

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INRC Inspection Report: 50-382/90-24 Operating License: NPF-38

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? Docket: 50-382-

. Licensee: Entergy Operations, In P.O. Box B ! .Killona, Louisiana 70066

~ Facility"Name: _ Waterford Steam Electric Station, Unit 3 (Waterford 3)

Taft, Louisiana

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I 1 Inspection At:

UInspectionConducted: October 2 through November 13, 1990

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'[ Inspectors: ' W. F. Smith, Senior Resident Inspector

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Project Section A, Division of Reactor Projects S. D. Butler, Resident Inspector-

. Project Section A, Division of Reactor Projects x ' W. M. McNeill, Reactor Inspector Materials and Quality Programs Section Division of Reactor Safety i

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Approved: , 2- //- 7 0 - 7 o w T. F. Westerman, Chief Projen Sectio , A , Dat Division Sf Reactor Projects a y .o ,

i ki i alnspection' Summary E'InspehtionconductedOctober2throughNovember i 13, 1990 (Report 50-382/90-24)

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l Areas; Inspected:. Routine . unannounced inspection of onsite followup of events, monthly maintenance observation, monthly surveillance observation, operational

< , safety verification, followup of previously _ identified items, licensee event o report followup, balance of plant (B0P) inspection, and evaluation of licensee

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quality assurance -(QA) program implementation.

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l N I Resillts: During:this inspection period, the following violation was identifie It contained four examples of failureLto comply with approved procedures:  ?

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, The licensee _ failed to' conduct'a: thermal element time delay test of a

- molded case circuit breaker supporting the emergency diesel generator-using_ the technique prescribed by procedure (paragraph 4.4).

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• .The licensee failed to perform a door interlock mechanism test on the !

above molded case circuit breaker upon reinstalling the breaker as required by procedure (paragraph 4.4).

The licensee connected the containment air lock leak testing device to station air. system in lieu of the instrument air system, contrary to specific procedure requirements (paragraph 5.1).

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The licensee. installed a temporary level indicating hose on the spent resin storage tank without authorization, contrary to the licensee's temporary alteration procedure (paragraph 6).

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General Observation

• As a. result:of the lengthy discussions between the staff and licensee,=

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operation with the the existing reactor coolant system (RCS) leakage from

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, . s the 2A reactor coolant pump (RCP) flange and the subsequent issuance of a

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confirmatory action letter (CAL), the licensee was encouraged by the inspector to improve the communications of actions taken and/or.bein considered.' 3 s

Weakness as discussed below:

As indicetadIby the above violation, and as addressed in the cover letter

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for this appendix, the; licensee appeared to be experiencing a recent '

I resurgence of procedure'noncompliances, which had been a previous NRC-issue.in the 1988-1989 timeframe. The programs and training implemented by-the licensee appeared tollack the ability to sustain an acceptable'

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le,el of: procedure compliance and, as such, other long-term actions may be ,

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tecessary.~- ;While this issue is being characterized as a weakness, the ,

licensee's recent self-assessment of maintenance activities came up with a

. similar observation, and actions to correct the problem were already being contemplated when the; inspectors expressed the concern. However, the

' aspect of self-assessment-can be considered a strength (paragraphs 4, 5,.

-and6). .

The followir,q strengths were observed during this inspection period:

• The licens*e's efforts to arrive at the root causes of the fuse. problems, coupled witt, a willingness to notify the industry of the generic implications, o s commendable. The inspectors were informed by the-licensee that this was not a new problem to at least one other utility nor to the affected vendors, but it did not appear to have been communicated to the' industry (paragraph 3.3).

The' operators' response to the blown fuses in CP-29 was timely and appropriate, particularly in view of the unusual circumstances (paragraph 3.3).

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• The licensee's man; gemeni :"arsight, planning, and teanvork was exemplary in the execution r f the Occober out;;e in perfonning the' planned work and

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resaonding to the emerge.it activity broght about by the discovery of

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leakage from RCP 2A (;,aragraph 6).

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• - Monthly System ingineering presentations are provia:ng licensee management with excellent iverviews of the status of individual ;vstems and appear to

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be strengthenin i the engineering group support of operations (paragraph 6).

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'The licensee's operation and maintenance programs applicab;e to the

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BOP appeared comparable to the programs for safety-related equipnent (paragraph 9).

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DETAILS L Principal Licensee Employees _ j

.- J; R. McGaha, General Manager, Plant Operations r

. ~ * V. Prasankumar, Technical Services Manager (D.. F. Packer, Operations and Maintenance Manager

.A. S. Lockhart,. Quality Assurance Manager

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• D. E; Baler, Director, Operations Support and Assessments R. G. Azzarello, Director, Engineering and Construction h t *T.' P. Brennan, Design Engineering Manager D < J. A.- Ridgel . Acting Radiation Protection Superintendent H '*G.' M. Davis, Events Analysis Reporting'& Response Manager k *R. F. Burski, Director, Nuclear Safety L. W. Laughlin, Licensing Manager

~J. G.' Hoffpauir. Maintenance Superintendent

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• R. S. -Starkey, Operations Superintendent

"' A. G. Larsen, Assistant Maintenance Superintendent, Electrical <

D. T. Donnady, Assistant Maintenance Superintendent, Hechanichl

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LK. R. Rauch, Acting Assistant Maintenance Superintendent, Instrumentation and

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. Controls-J. E. Howard, Procurement Progrens Manager

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GUW.! Robin, inservice Inspectior, (ISI) Coordinator

~ *B. R. Loetzerwich, Site Licensing Engineer p' %g ,l*J.:H. Roberts, Procurement Support Quality Assurance-Supervisor

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mh'"' in addition to the above: personnel,-the inspectors held discussions with othe '

L m operations ~, engineering.. technical support,-maintenance;.and" administrative!

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PLANT STATUS- (71707)

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E On' October 2,1990, the plant was operating at 99 percent power with a. reduced

'& ' x .RCS; pressure of 2100 psia'to minimize leakage from the pressurizer codei N

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, safety-relief valves. The licensee,was holding' power at or. below 99 percent. to ','

maintain the desired departure from the nucleate boiling ratio of 0.3.above the N^x s minimum of- I'.26 required by. the Technical Specifications (TS). : Operation at

. 99 percent: power continued until October 5,1990, when a' shutdown.was initiated r-to replace the pressurizer relief' valves. By October 6,1990, the plant was in cold shutdown.(Mode 5). The plant remained in-cold' shutdown'to conduct outage activities until.0ctober 15,-1990, when the plant was placed on the power grid and~ restored to full ' power operation where it remained through the end of this -

.m inspection period, which is November 13, 1990.

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5 ONSITE FOLLOWUP 0F EVENTS (93702)  !

3.1 Feactor Coolant Pump (RCP) Casing Leakage On October 8, 1990, while the plant was shut down and cooled to about 125*F (Mode 5), the licensee found indications of primary coolant leakage between the casing of RCP 2A and the driver mounting flange at the top of the casin Leakage was normally prevented by two spiral wound, asbestos filled,

"flexitallic" type, stainless steel seals with the annulus between the two seals connected to a pressure switch, which was designed to detect leakage past the first seal. The licensee identified the leakage during their routine inspection arogram for the control of boric acid corrosion of carbon steel _ reactor pressure aoundary components pursuant to NRC Generic Letter 88-05. After the reflective insulation was removed, the licensee detennined that primary coolant was exiting the pump between the driver flange and pump casing, out through three keyway Waterford 3 has Byron-Jackson RCP There were 16 sets of 33-inch long by 4 3/4-inch nominal diameter studs and nuts preloaded hydraulically to attach the driver flange to the pump casin They were SA540, B23, Class 4 steel, _ Recause only'the nuts and the top end of the studs were visible, the licensee could not visually determine if there was

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any wastage of the stud material caused by bric acid corrosion. There were no

' boric acid deposits around the nuts, p In an effort to determine if there was any wastage, the licensee employed two

' - ultrasonic testing (UT) techniques. The first, and primary, method was to scan the area of the studs near the interfacing flanges by usez of a 45-degree UT l

search unit inside the stud elongation hole. The hole did not go deep enough to: allow a 90-degree probe.to obtain UT signals from the area of interest. No wastage was detected on any of the 16 studs. The secondary, or backup, method was to have Southwest Research Institute apply a " Cylindrically Guided Wave"

, technique, which simply _ verified that the geometry of the 16 studs resembled l that of a new stud. The NRC staff expressed concerns over the reliability of L these techniques based on previous experiences at other plant See NRC Information Notice 80-27 and NRC Bulletin 82-0 A conference call was held on October 12,-1990, between NRC Region IV, Headquarters, the resident inspectors,'and the licensee. A_ CAL was subsequently issued on October 12, 1990'(CAL 90-06), documenting the NRC staff's

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understanding that reactor startup (entry into Mode 2) would not comence until the staff had confirmed the licensee's actions for assessing that adequate

? safety existed for continued power operation while the leak at-RCP 2A was not repaired. Three actions were discussed as follows:

(1) NRC inspection at the Waterford 3 site of the UT technique for determining degradation'of RCP studs: On October 13, 1990, in the presence of a regional

. inspector assisted by a nondestructive testing expert contracted by the NRC

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from Sandia National Laboratories, the licensee performed a demonstration of the technique used to ultrasonically examine the RCP studs. For this demonstration, the licensee used the same 45-degree,1/4-inch by 1/2-inch search unit used on RCP 2A, with the sample-examined being a spare stud which

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had been notchcd in the shank area with two circumferential " boat" type cuts of

. 030- and .070-inch maximum radial depth and 2-3 inches long. The notches were noted to have sharp axial characteristics. The demonstration showed that the deep notch could be consistently identified. The shallow notch could be detected with careful manipulation of the search unit. These results appeared to indicate that wastage of the shank area of the studs could be identified and evaluated if it exceeded a depth of- .030 inc The search unit angle and the depth of the stud inspection hole allowed only about 1/2 inch of the threaded area to be ultrasonically scanned. The threaded areas of the stud could not be meaningfully UT inspected beyond the lead thread because of interference by signals from the root of the subsequent thread For-this reason, wastage of the threaded areas beyond the lead thread could not be adequately identified and evaluate It could not be verified that the UT gain setting was exactly the same as that used when RCP 2A studs were inspected because there was no record of the settings used for the "in situ" testing of the RCP studs. The licensee's UT examiner attested that all settings were the same. Subsequent dimens 4nal measurements by the licensee and puinp vendor information indicated tnat the

' interface of the RCP case to driver flange could be as much as to to four threads below the start of the threaded area of the sttJs. The inspectors t-concluded that the UT process may have been of limited value in determining the presence, or absence, of wastage in the immediate flange interface area, which was the area of greatest interes (2) NRC inspection of the licensee's leakage monitoring techniques for detecting RCP flange leakage: The resident inspectors observed the licensee's actions. The licensee connected the annulus between the RCP flange seals directly to .the reactor drain tank (RDT)-so that pressure against the outer seal would be relieved and leakage past the inner seal routed to the drain tank where leakage could be quantified during routine 72-hour RCS inventory surv'eillances ~. The inspector visually inspected the RCP 2A casing flange on October 14, 1990, when the' plant reached normal operating temperature and pressure. No leakage was found at the flange, thus the venting of the annulus

.to the drain tank appeared to be successful. The licensee installed three

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. closed circuit television cameras at the RCP flange with the monitor on th '

platform just inside the containment near the personnel air lock. ' The operation-of the camera was demonstrated to the inspecto LAfter the plant was restored to full power operation and stabilized, the

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i 111censee established base line leakage water to the RDT at 0.215 gallons per

, minute (gpm). Licensee management issued instructions to the operators to Y monitor the trend and, if the leak rate to the RDT increases to 0.415 gpm,

action shall be initiated to determine the extent and source of leakage. If the television cameras failed to function, the licensee indicated an intent to -

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shut down and investigate. During routine tours of the control room, the

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. inspectors monitored the RDT level trend and noted stable conditions through

the end of the inspection period. The licensee's measured leak rate to the RDT J

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increased from 0.215 to 0.290 gpm over a period of about 1 month, indicating a marginally increasing trend. The inspectors will continue to monitor these j parameters'on a daily basis when on sit The licensee also comnitted to enter the containment and visually inspect the RCP any time the plant is shut down to hot standby (Mode 3). In addition, the licer.see found that RCS leakage external to the pump hcd a tendency to cause boric-acid deposits which could foul the RCP motor stator air coolers. The extent _ of fouling had been reflected in stator winding temperatures, which were l monitored in the control room. During the outage, the cooler for RCP 2A was h found severely fouled and was! cleaned. Any significant increases in stator L Ewinding temperatures'would be an indication of leakage past the outer RCP-flange seal.

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.The NRC staff concluded;that the licensee's actions to monitor for increased RCP 2A leakage' were satisfactor .

'(3) 1 Conduct of a conference call between the NRC and the licensee's staff regarding'the results of the licensee's bounding failure analysis for RCP stud degradation': On October 13, 1990, a second conference call was held between the licensee,'thet resident inspector, NRC regional managenent, and headquarters

. staff. The licensee's engineering evaluation of October 12,1990'(Addendumto ConditiontIdentification(C1)L271478), was discussed concurrent with a review of the:11censee's actions taken prior to startup of the plant. The bounding

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i failure' analysis demonstrated-that an observable change in pump casing gasket leakage would occur prior _to catastrophic closure stud failure and thus provide

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an indication of the onset of closure stud wastage so that the plant could be safely-shut.down in plenty of:timeLto make repairs. The preliminary evaluation-i wasu transmittedito the regionalland/ headquarters staff by fax. The NRC staff s y considered 1the . licensee's actions' satisfactory for a safe.startup and resumption L 4 of full power operation while unitoring RCP 2A for leakage. The final

^. . evaluation'was sent to theiNRC on October 17,:199 ,.

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The alternatives for long-term corrective action have been under review by ' J

. licensee management. The licensee has indicated that in early December 1990-T(c' ' they,will' perform additional UT demonstrations with a spare stud in an effort ;

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a: totprovide ' adequate-confidence in the process for determining the presence, or W -absence..of stud wastage.: 'The. licensee has been requested to advise the NRC

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Ltracked to completion under Inspector Followup Item (IFI) 382/9024-0 .m ,

l34 3. 2. : Containment Peak Pressure Analysis (

l4 una l0n November 7,1990, the licensee discovered that the value usL for the overall J

heat transfer coefficient -(U)! for' the shutdown cooling (SDC) heat exchangers in l their containment peak pressure design analysis was not conservative. The SDC heat exchangers were designed for use by the containment spray (CS) system to cool containment sump water following an accident. The value used to verify

- adequate containment heat: removal following a design basis accident during the ,

postrecirculation; actuation phase was higher than it should have been and did

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not consider possible fouling of the heat exchangers. The licensee entered

, their "Nonconfomance/ Indeterminate Qualification Process" (N0P-019) procedure

.to confirm operability and to evaluate the affect of the error on containment peak pressure and temperature calculations. A preliminary engineering-evaluationandscopingcalculation(CI 271979), completed on November 8, concluded that the reduced heat removal capability of the CS system based on a U value of 216 versus 244 resulted in a postaccident containment pressure increase of 0.44 psig and a temperature increase of 0.6'F. The peak pressure increase to 41.5 psig was still within the 44 psig design pressure of the containment, and it was considered that the temperoture increase of .6'F would not have an impact on the environmental qualification of safety-related equipment in the containment. The 1fcensee planned to have Ebasco perform containment pressure and temperature sensitivity studies using the lower V value for the SDC heat exchangers to confirm their results. The inspectors will follow up to ensure there are no further problems with the results under IFl 382/9024-0 . L_oss of Power to Process Analog Control (PAC) Cabinet CP-29 On November 7,1990, with the plant at 100 percent power, power was lost to Rack No. 100 in PAC Cabinet CP-29. CP-29, a standard Westinghouse 7300 panel, contained primary control and indication instrumentation. The loss of Rack No. 100 affected main control room Panel CP-4 with its controls and indications for the chemical and volume control system. Letdown from the RCS was isolated when letdown isolation Valve CVC-101 went shut and the charging pump that was operating shifted suction from the volume control tank (VCT) to the refueling water storage pool (RWSP), due to the indicated low level in the VCT. The operator promptly stopped the operating charging pump before a significant amount of borated water from the RWSP was injected into the RCS. Once power was restored to the instrument rack in CP-29, the operator was able to restore charging and letdown and dilute the RCS to prevent a significant decrease in reactor power. Based on control board displays, it appeared that reactor power l was decreased approximately 1.5 percent.

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Inspection of CP-29 revealed that both the primary and backup fuses for the 24 VDC power to Rack No. 100 were blow It appeared that this problem was related to the fuse problem discussed in paragraph 4.5 of this inspection

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report, except that both blown fuses were rated for 20 amperes. The fuse in the primary power supply was a fast blow rather than a slow-blow fuse. The

20-ampere, slow-blow fuse in the backup supply had just been replaced recently, so it was unclear as to why it also blew. No grounds or shorts were found in the rack. The power to rack No.100 was restored af ter both fuse holders were replaced and correct slow-blow, 20 ampere fuses were installed.

l l- On November 9, 1990, the licensee conducted shop tests to help determine why

) the 20-ampere fuses were blowing. They installed a new slow-blow, 20-ampere l fuse in the backup power supply fuse holder that was removed from CP-29, and

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loaded it to 17.2 amperes, 26 VDC. This was identical to the load measured at CP-29. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the fuse holder got extremely hot and the fuse ble Other combinations were tested, and the licensee concluded that although the fuses and fuse holders were rated at 20 amperes, subjecting the devices to l

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greater than 15 amperes can cause them to heat up and fail. As part of the root cause determination, the licensee also sent sample fuses and fuse holders to haterials Evaluation Laboratories, in Baton Rouge, Louisiana, to run independent tests to determine the cause of failure. The results were expected before the end of November 1990. Discussions with the Bussman Fuse Company appeared to support the licensee's findings that, as a minimum, greater capacity fuse holders should be considered for the triple instrument card racks and also, if the circuit designs pennit, greater capacity fuses so that the current loading would not be as close to the limit. The licensee was also pursuing a resolution to this potentially generic problem with the vendor since it appeared that the fuses were overheating and failing under normal operating condition The licensee has indicated an intent to pass this information to the industry via INP0 Nuclear Network, and the inspectors have discussed the issue with regional management for consideration of NRC generic correspondence. To date the licensee has checked all of the triple card frame racks for current values and overheating and has not identified any cases where this fuse problem could cause a significant safety problem that might require reporting under 10 CFR Part 21. The inspectors will continue to monitor licensee actions on this issue to conclusion under IFI 382/9024-0 .4 Broad Range Gas Monitor (BRGM) Wiring Discrepancies On November 7,1990, the licensee discovered discrepancies between the BRGM field wiring and the vendor drawings. The BRGM monitors the air intakes for the control room and isolates the control room ventilation system when toxic

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gas is detected. The licensee initiated a nonconformance condition I identificationworkauthorization(NCI-WA 01067383) to document and resolve the i

differences and verify operability of the monitors. The wiring differences I were analyzed with vendor assistance and compared to revised drawings which the licensee-obtained from the vendor. It was determined that the differences were changes made for improvements to the monitors to reduce electrical noise and improve the ability to calibrate the instrumentation. There were also some jumpers installed because the licensee did not utilize an optional remote control panel for the monitors. The licensee's engineering staff concluded, based on their review of the differences and the successful periodic calibration and functional tests of the monitors, that they were operable. This was completed on November 7. From discussion with the vendor, the licensee also detennined that there were additional enhancements that could be made to further improve the reliability and operation of the BRGM. The licensee has stated that they will evaluate these improvements and update plant drawings as part of the NCI-WA closeout. The inspector will review the final resolution of this NCI-WA af ter it is completed (IFI 382/9024-03).

3.5 Summary of Findings No violations or deviations were identified. The licensee's actions and responses to the events described in this section of the report were appropriate and timely, which was reflective of a positive attitude toward identifying and solving quality problems. Because of the lengthy discussions between the staff and the licensee, which were necessary to resolve the staff's concerns regarding the return-to-power operaticns with the existing RCS leakage from the

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S 2A RCP, the licensee was encouraged by the inspector to improve comunications

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in actions to monitor for precursors to catastrophic failure, as well as an

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e analysis of the probability and nature of failure, was not-fully comunicated ,

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?Y- :4. MONTHLY MAINTENANCE OBSERVATION ~(62703)

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Wi The' station maintenance activities affecting safety-related systems and f h4 - components listed below were observed and documentation reviewed to ascertain c m .that the activities were conducted in accordance with approved work  !

)%? . authorizations (WAs), procedures, TA, and appropriate industry codes' or

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i '4k1 WA 01064986~

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On October 26, 1990, the inspector observed preventive maintenance in progress -

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to inspect and clean the circuit breaker and test the thermal overloads in accordance with the specified procedures. The inspector verifled that the work b was. properly authorized and being performed in accordance with properly controlled' maintenance procedures. Test equipment was properly calibrated and

- the electricians were' familiar:with the work to be performed. The inspector

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verified that the breaker was removed from service in accordance with the

,y~ procedure and later. verified that it was properly restored. No problems were ,

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4.2 1WA 01064653-

• On' October 29, 1990, the inspector observed portions of the removal and Ie installation of the jacking gear for the B emergency diesel generator (EDG).

' This work.was necessary in order to correct a greasing problem with the jacking 1

, ' gear bearings';c The WA>was clearly written and the mechanics followed the- 3 y ' instructions. No problems were identified.- 3 4.3' WA 01063349 On October'29, 1990, the inspector observed portions of the preventive -

maintenance performed on the B EDG air compressor (B2). The inspector noted g that the.WA was being-followed and verified that the correct lubricating oil was:being used..=No problems were identifie .4 lWA 01064951-

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g is 4 On October 29, 1990, the inspector witnessed the thermal element time delay

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. test,' inspection, and cleaning of the 480-volt molded case circuit breaker for the B 'EDG jacket water heater.' This work was to be accomplished in accordance

+ with Testing Procedure'ME-007-002, Revision 7, " Molded Case Circuit . Breakers."

During perfonnance of Section 8.4 of ME-007-002, which was the thermal element tine delay-test, the technician adjusted the current generator to 300 percent of the rating of the breaker by running the current through the Phase C thermal

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11 i elenent. Then, without more than a few seconds delay, coninenced timing the

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thermal element time delay to trip the breaker. Since current flow during the adjustnent contributed heat to the thermal element, the time delay was shortened

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by an indeterminate anount. Upon questioning this practice, the inspector

was told by the technician that because there was such a wide tolerance, c experience has shown that the initial heating effect was insignificant. .A note in the procedure after Step 8.4.4.5 required the technician to allow 5 minutes for couldown of the thermal element after adjusting the current. This was not-done on Phase C, contrary to procedure requirements. With or without the 5-minute wait, the breaker would have tripped within tolerance, so the technician may have been technically correct. During restoration, Step 8.1 .

required the technician to test the door interlock nechanism on the breake ~This was.not oone. The technician explained that he did that while the the breaker was on the bench, but the procedure expected the breaker to be mounted

' in the cubicle for this test. The above failuks to follow ME-007-002 were two examples of procedurc- in.ncompliance, which is a violation of HRC regulations (382/9024-04).

The inspector also discussed two possible improvements to the procedure with the licensee. .The practice of. setting the current generator through the thermal elements could be avoided, along with the 5-minute delays, by shorting the test leads and making royh adjustments before connecting to the breake The breaker continuity . test in Section 8.9 of the procedure has an acceptance criterion of "near 0 ohms," with no tolerance specified. . The technicians accepted 0.2 ohms as near enough to zero. At slightly over half the rating, L the above breaker may. have to dissipate about 100 watts of heat energy. This could be a problem for a breaker supplying power to an engineered safety feature if called upon to operate for any length of time. The licensee agreed to consider both items. Actions taken shall be tracked under IFI 382/9024-0 .5 WA 01067402 .!

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On October 31, 1990, the inspector observed work, including troubleshooting and L corrective maint? nance, on the power supply for PAC Cabinet CP-26, which contained the process analog control for protective Channel B. On October 30, power was lost to the No.100 instrunent card rack in the panel causing all

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Channel B-instrumentation in the rack to fail to its deenergized conditio This resulted in Channel B trips in the plant- protection system and engineered I

safety features actuation system and loss of Channel B~ remote shutdown instrumentation and control instrumentation Channel mediate troubleshooting was authorized by the shift supervisor. The i

technicians found both the primary and backup fuses for 24 VDC to the rack blown. . In addition, the fuse holder for the primary fuse fell apart when the fuse was removed. It appeared to be overheated and brittle. Visual. inspection <

. of the panel wiring and all the cards in the rack, and resistance checks with the cards removed and installed one at a time, did not indicate any short circuits or ground . A WA was properly prepared and the work was approved by the shift supervisor '

prior to beginning any corrective maintenance. The WA provided instructions to

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l replace both fuse holders, make some minor repairs to the wiring, and install new fuses. Additional instructions were provided to measure current being carried by the fuses in all four of the protective channel instrument cabinets and verify that the backup power supply fuse in Channel B would continue to supply power to the rack when the primary fuse was pulled. The work was successfully completed and the new fuses did not blow when reinstalle When the primary fuse was pulled, the backup fuse continued to supply power to the rack without blowin Current measurements indicated that the current drawn by Rack No.100 in Channel A was 16.2 amperes and 15.8 amperes in Channel B. The current for similer racks in Channels C and D were considerably less, that is, 9.4 and 9.2 amperes, respectively, since they were two-bay racks versus the three-bay racks in Channels A and B. The fuses found installed in Channel B were rated at 15 amperes. In the absence of information on these vendor-supplied fuses,

1dentical 15-ampere fuses were installed as replacement Discussion with Westinghouse, vendor for the process analog control instrumentation, resulted in the determination that the power supply fuses for the three bay instrument card racks should have been 20-ampere fuses, and 15-ampere fuses should be used for two- and one-bay racks. Discussion with the manufacturer of the fuses and fuse holders indicated that continuous operation of the fuse at 100 percent of its rating would generate enough heat to degrade the fuse holder over a long period. A similar failure occurred in the Channel A process analog cabinet in April 1988. On November 1, after the licensee received written confirmation and technical manual drawings, which were not available onsite, they replaced the 15-ampere fuses in Channel B with the correct 20 ampere fuses on and declared the Channel B instrumentation operabl The licensee converted the original condition identification (Cl) report to a nonconformance CI in order to evaluate the cause of the problem of improper fuse sizes being used in the cabinet. Immediate corrective action included generating another WA to replace the primary fuse holder in the Channel A process analog cabinet and install 20-ampere fuses. This was accomplished on November 2. All other process analog control cabinets were inspected and currents measured to determine how many had three-bay instrument card racks and would require 20-ampere fuses. It was found that CP-29, a cabinet with primary

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L instrumentation and controls, was the only other cabinet with card racks drawing current in excess of 15 amperes. The backup fuses were pulled to

install correct fuses. The inspection and replacement of the fuses and fuse l holders in the remaining cabinets would be scheduled later, when plant conditions permi The inspector questioned the generic implications of fuse control in all electrical equipment at the plant. The licensee indicated that they would evaluate fuse control and 10 CFR Part 21 reportability as part of their root cause evaluation and determine if additional corrective action was necessar Before this determination was completed, the fuses in CP-29 blew on November 7, 1990, resulting in a minor operational event. T % continuation of the fuse

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issue is, therefore, discussed in paragraph 3.3 of this inspection repor ,

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1 4.6 WA 01067577, On November 7,1990, the inspector observed work in progress on chilled water pump recirculation Valve CHW-129A/D. Electricians were disconnecting the valve operator to allow its removal to the shop for inspection and repair. The

- inspector verified that the work instruction was adequate and properly authorized for performance. The lifting of leads was documented as part of the ,

work package. No problems were identifie .7 Sunnary of Findings

, ~In'five of the six maintenance observations described above, the licensee continued to demonstrate the improving trend of overall maintenance compliance with, and quality of, procedures. The technicians appeared to have greater

. confidence that the procedures would get them through their tasks without incident. However, during one observation of molded case circuit breaker testing, the technicians deviated from this positive trend by appearing to use the procedure as a guide rather than as a quality document requiring complianc . 110NTHLY SURVEILLANCE OBSERVATION (61726)

The inspectors observed the surveillance testing of safety-related systems and components listed below to verify that the activities were being cerformed in accordance with the TS. The applicable procedures were reviewed for adequacy, i

test instrumentation was verified to be in calibration, and test data was L reviewed for accuracy and completeness. The inspecturs ascertained that any deficiencies identified were properly reviewed and resolved.

L~ Procedure OP-903-111, Revision 1, " Containment Air Lock Door Seal Leakage Test"

- On October 13, 1990, the inspector observed the containment personnel air lock

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-inner and outer door seal test. This was-in preparation for heatup and startup

! of the plant following a cold shutdown outage. The test was properly performed L

and the results were well within the acceptance criteria; however, the shif t i technical advisor (STA) performing the test initially connected the leak rate monitor to the station air system when Step 2 of the procedure required the leak rate monitor to be connected to the instrument air system (or a regulated nitrogen source). Af ter the inspector identified the error, and before any air-was used, the STA disconnected the leak rate monitor from the stati on ai r system and reconnected it to the instrument air system. The inspectors followed

<s up on.0ctober 14,-1990, to determine if the li e nsee had entered the procedure violation into the corrective action program e accordance with Nuclear Operations Procedure NOP-005, " Corrective Act.on." In this case, a quality notice (QN) should have been initiated in accordance with Quality Assurance Procedure QAP-012, but as of October 14, 1990, a QN had not been initiate It l

was not until October 16, 1990, after the inspector discussed the issue with licensee management, that a QN (QA-90-227) was issued. Failure to follow Procedure OP-903-111 is a third example of procedure noncon.pliance, which is a violation of NRC regulations (382/9024-04).

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5.2 Procedure OP-903-008 Revision 6, " Reactor Coolant Systens isolation Leakaoe Test" On October 14, 1990, the inspector observed portions of the RCS isolation leakage test of the high pressure safety injection (HPSI) to RCS Loop 1A Check Valves.SI-241, SI-242, and safety injection header Loop 2A Check Valve SI-336 The operator conducted the test satisfactorily in accordance with the procedur The leak rates varied from 0-0.05 gpm with an acceptance criterion of 1.0 gpm maximum. The inspector did not identify any problems; however, the operators noted that the remote position indicator for SI-226A did not indicate correctl It indicated 80 percent when the valve was actually at 75 percent, which did not neet the acceptance criterion of 76 plus or minus 2 percent. A condition identification report was written which initiated maintenance actions, 5.3 Procedure NE-02-002 Revision 4, " Variable T-average Test" On October 18, 1990, the inspector observed the performance of NE-02-002, which was performed to satisfy the surveillance requirements of TS 4.1.1.3. for the moderator temperature coefficient (MTC). The test involved varying RCS

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temperature using tu bine load and observing the affect on reactor power to

ensure that the MTC vas as predicted and met the TS requirement The inspector

' attended the test br? efing.for operations personnel performing the test, reviewed the procedure, and observed several turbine load cycles between

.approximately 92 pert ent and 98 percent power. The inspector verified that n

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initial test conditions and prerequisites were satisfied and that test

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.' ' requirements were met during conduct of the test. The completed test package

) was reviewed to verffy that acceptance criteria were met. The measured MTC of L .-l.6272 E-4 was acceptably close to the predicted value for MTC ano met TS

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5.4 Procedure OP-903-068, Revision 6, " Emergency Diesel Generator Operability

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Verification" On October 29, 1990, the inspector observed the performance of-Procedure OP-903-068 for the B EDG. The surveillance test was performed to restore the machine to an operable status after it was taken out of service for corrective and preventive maintenance. The inspector verified that the surveillance was performed in accordance with properly controlled procedures by qualified individuals and.that the EDG ran fully loaded for the required tim Operation of the machine was observed locally while the load run was in progress, and logs taken by an auxiliary operator were reviewed to ensure that important engine parameters remained within their specified range during the tes .5 ~Sumary of Findings The licensee's perfonnance of surveillance activities during this inspection period was excellent in all respects, with the exception of one error made by the STA during the containment personnel air lock seal leakage test discussed in paragraph 5.1 above. The licensee has sustained a high level of performance in the control, tracking, and execution of TS surveillances over the past three

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<SAlp cycles. The violation above became more significant when the STA and his  !

supervisor both failed to implement the corrective action program by initiating a QN, so that -licensee managenent would be placed in a position to properly evaluate the implications of the incident as a potential precursor to procedure ccmpliance problem .

' ,0_perational Safety Verification (71707)

The objectives of this inspection were to ensure that this facility was being_

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' operated safely and in conformance with regulatory requirements, to ensure that the licensee's management controls were effectively discharging the licensee's E

' responsibilities for continued safe operation, to assure that selected ictivities of the licensee's radiological protection programs were implemented in conformance with plant. policies and procedures and in compliance with 1 regulatory requirements'. and to inspect the licensee's compliance with the -

!= approved physical security plan, ,

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, The inspectors conducted control room observations and plant inspection tours and reviewed logs and licensee documentation of equipment problems. Through

, in eplant observations and attendance of the licensee's plan-of-the-day meetings, f

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the inspectors maintained cognizance 'over plant status and TS action statements:

in effec '

i .On October 3,1990,: the inspectors attended a system engineering presentation l

ton the emergency diesel generators. This was the second session of this type _ <

J Jattended by the inspectors. On September 5,1990, the inspectors attended a ^

L , similar session'en the component cooling water system. The system engineers L explained the design and operation of their system and discussed problems that a h existed and what was being done about,it. They also discussed maintenance L, trends and. proposed stagion modifications to the' system. The inspectors L obsenved that these sessions were candid and informative and, as such, enabled E the inspectors to evaluate the professionalism, . system knowledge, and personal ;

involvement demonstrated by:the engineers. During these last two sessions, the 2 inspectors'noted a professional pride in ownership of the systems for which the 4 Lengineers were responsible. These sessions not only brought licensee management up to speed on system status'and current problems, but they also served to help

- evaluate progress in'the development of a strong engineering group in support

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of operation '

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On October 6 and 15,1990, .the inspectors observed shutdown and startup '

l activities (respectively) related to the. outage to replace leaking pressurizer l code safety relief Valves RC-317A and RC-3178. Throughout the portions of the ,

evolutions observed, the operators conducted plant operations in an orderly and L'

' professional manner and in accordance with the applicable procedure During the above outage, the inspectors noted excellent comunications between the various-licensee organizations, teamwork, and continuous management R I

E involvement. At the daily meetings, emphasis was placed on safety and minimizing' radiation exposure. Based on self-reading dosimetry, the licensee estimated:about=14.8 manRem expended. The following significant tasks were accomplished:

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Replacement of both leaking pressurizer code safety reliefs. (Inaddition, temperature sensors were-installed temporarily to record the ambient temperature of the reliefs to ensure that they are set at the proper pressure). j

Resetting of low temperature overpressure protection (LTOP) relief l Valve SI-406A, which was previously found incorrectly set (see LER382/90-006).

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Replacement of pressurizer quench tank rupture disk to preclude premature failure caused by pressure cycles from leaking pressurizer safety relief

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Added packing to No. 1 main steam isolation valve (MSIV).

.Recalibrated Channel e and A steam generator flow instruments to prevent spurious trip signal '

Emergent work in support of RCP 2A leakage problem described in

' paragraph 3.1 of this inspection report, including cleaning of boric acid-fouling from the motor stator cooling heat exchange On October 24, 1990, the licensee held their annual emergency preparedness exercise, which was evaluated by an NRC Region IV inspection team. The resident inspectors participated in the exercise as " players," 1.e., responded as th would during an actual event. The senior resident ~ inspector manned the technical support center (TSC), and the resident. inspector reported to the'

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simulator control room where the scenario was simulated. The results of the exercise were published in NRC Inspection Report 50-382/90-20. The notable observations made by the resident inspectors were transmitted to the regional-

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inspection team and were discussed in.that report. The resident inspector -

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noted that the control room operators performed in a well-trained, professional 4 : manner. _ Good comunications were maintained between the operators' and the

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resident inspector such that he could maintain adequate knowledge of plant- '

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status and the challenges that were presented to the operators and that they

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desponded in a manner consistent with safety and their emergency operating . .

= procedures. The senior resident' inspector-noted good teamwork and excellent communications in the TSC. The emergency coordinator' appeared to keep.the

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supporting activities in perspective in terms of.the continuous. prioritization -

A ,u , of goals leadin;; to plant stabilization and temination of radiological releases.

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(On October 26,'1990, the inspector observed preparations for the transfer of-spent resin from the fuel pool ion exchanger (FPIX) to the spent resin- l

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tank (SRT).= The inspector attended the prejob briefing required by Procedure HP-002-224, Revision.1, " Transfer of Spent Resin." Licensee personnel in

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l attendance included operators, health physics technicians, and radwaste personnel involved-in the resin transfer. The briefing included radiation work-permit -(RWP) requirements, radiation survey requirements, and precautions

. delineated'in HP-002-224 The operational requirements included in l

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~'y ' Procedure OP-007-005, Revision 6. " Resin Waste Management," were also discussed, iThe inspector considered the briefing and RWP requirements adequate for the work

' ' to be perfonne o : The' inspector then observed equipment preparation and system conditions in the

. reactor auxiliary building, Elevations -35 -4, and +21. The inspector noted t

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• nthat the normal level indication for the SRT-did not appear to be operating

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A properl L 'This was discussed with the lead operator who indicated that the

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indication for both the water and resin level in the tank had been y , malfunctioning for several years. CI reports had been submitted for both j < . instruments, but troubleshooting.of the problem was unsuccessful. To provide l Mevel indication of the water in the tank, a temporary-level hose was Installed l . Lfrom the suction of the dewatering pump to the upper level tap for the normal 4 ". ' * Lwater level, instrument. The resin transfer was not started when it was

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determined that not enough-water had been drained from the SRT the previous day

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I h Oto' accept the resin from the FPIX as well as the water required to flush the  ;

flines after the transfer.' . The transfer was successfully completed later in-the

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r .1 > lThe-inspector later. questioned the operations superintendent about the lr i1 rinstallation of the temporary level hose on the SRT. A: temporary alteratio ,

[ l request (TAR) was not posted on the hose or logged as required by '

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f Procedure UNT-005-004,- Revision 7. " Temporary Alteration Control ." nor was the ' "

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1 . ltemporary level indication referenced in: Procedure OP-007-005. The inspector ,

Em ' . ' S'was later informed that the hose installation was accomplished under WA 01030073t The inspector

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A ,treviewed WA 01030073and, whichtherefore, was issueda TAR was not required to troubleshoot per the and repair Procedure UNT-005-004

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w "s7 Jmalfunctioning level indication on the SRT. The WA documentation indicated < '

l - & athat;the hose was installed and-then removed on August 3, 1989..-The WA was'

r iclosed"and, therefore, the hose was installed ~without an approved WA or TAR'asr , a 6 Prequired by Procedure UNT-005-004. - FailuretofollowProcedure.UNT-005-004 sis (

o % the fourth example of procedure noncompliance, which is a violation of NRC ~a p + W ! requirements-(382/9024-04).- l '.

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1 ifWhilefconductingaroutinetouronNovember6,1990,theinspector, verified n? ' W ithat all: required notices to workers were appropriately and conspicuously I

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posted in accordance with'10 CFR Part 19.11. The inspector also verified-that' r ;

i y " 4 4the licensee had been meeting the posting requirements of 10 CFR Part4 21.6L

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1TheLinspector noted that the September 1984 version of NRC Form 3 was posted, "

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"12 andithen made copies of the October 1989 version-available to the licensee for-I '

Jupdating their. postings, if. desired. Since 10 CFR Part 19.11(b) requires; the

1. June 1982 or later versions.of NRC Fonn 3 to be posted, this update'would be at

$the: licensee'soption.. There was virtually no difference between the two e 4 version ;7; Followup of Previously Identified items (92701,92702)

7.'lV (Closed) Violation 382/8819-07, Enforcement Action-(EA) No.89-212, Part '

, This part of the. violation involved a deficiency in Maintenance

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? Procedure ME-13-100, Revision 6, " Fire Barrier Installation and Rework." A '

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" ,ection hold point was-mislocated in the procedure. Licensee personnel were e of.the problem but worked around it instead of identifying and correcting

. deficiency before proceeding. This practice continued from June 1985 to

..pril 25, 1988, when the procedure was changed to place the hold point in the proper sequence. As a result of an enforcement conference conducted on December 1,1989, the violation was classified as a Severity Level IV with no civil penalty. The enforcement conference was also held to resolve Part 8 of '

tMs viciation where a quality control inspector misled an NRC inspector over the manner in which this hold point was handled prior to the April 25, 1988, chan9e. Part B.was closed in NRC Inspection Report 50-382/90-15. Actions verified satisfactorily completed for Part A included: (1)replacementof ME-13-100 with two restructured procedures in January 1989; (2) Nuclear Training Course Description NTC-129 and tne associated lesson plan were revised in the third quarter of ~1988; (3) a QN (QA-89-278) and a potentially reportable event report (PRE-89-138) were generated on December 19, 1989, to enter the problem into the licensee's corrective action program; and (4) the procedures upgrade program was implemented and completed for operations procedures and was nearing completion for the more numerous maintenance procedures. This violation is

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7.2 (Closed) Violation 382/8931-01 (EA 89-192)-

This violation involved failure to comply with TS 3.7.1.1.a which required, in short, a reduction in reactor power because of an inoperable main steam code safety relief valve. As a result of an enforcement conference conducted on October 24, 1989, the violation was c hssified as a Severity Level IV with no civil: penalty. The licensee issued LER 382/89-018 to report the incident. The

, LER was_ satisfactorily' closed in NRC Inspection Report 50-382/90-01. .In addition, the licensee took' appropriate administrative actions to ensure that '

plant staff supervisors and senior reactor operators were aware of this incident l and were instructed on how to handle similar circumstances in-the future. The l- inspectors reviewed the documentation which supported the above actions and '

L found no' problems. This violation is close .3.-(0 pen) Violation 382/8931-02 (EA 89-192) '

This violation addressed procedural inadequacies that led to.the failure to comply with TS 3.7.1.1.a. which was discussed 'above.. The licensee comitted to '

revise' Procedure MM-007-015, Revision 2. "Trevitest of Main Steam Safety

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Valves." The procedure was revised on May 29, 1990, which was before using

- the procedure again, as committed. The inspector conducted a detailed review g

' . of the new revision and had many coments. Some were human factors F .' deficiencies, others were technical. The inspector's comentsLwere discussed '

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'with the licensee, and a comitment was obtained to revise the procedur ,

satisfactorily prior to the next relief valve test. This violation shall 7 t remain open until a satisfactory revision to the above: procedure is published.

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7.4 (Closed) Violation 382/8941-02 This violation involved the failure of maintenance technicians.to perform independent verifications that were required by a work instruction. This

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raised questions as to whether or not the licensee was clearly defining what was required of independent verifiers, and whether or not all affected employees understood. The licensee assessed the problem as conmitted and concluded that an upper tier procedure which would clearly delineate what licensee management required, followed by proper training would solve the problem. On September 24, 1990, the licensee issued Administrative Procedure UNT-005-010 Revision 1

" Independent Verification Program." The inspectors reviewed the revised procedure and found it to be clear and concise. However, the inspectors noted one section that listed examples which were "not typical" of the independent verification program. One example was " critical steps where errors could easily be made." The inspectors questioned the licensee as to why such a situation could not be a candidate for independent verification. The licensee's response was that, as a result of questions raised during employee training sessions, the entire section was going to be deleted. The inspectors attended one of the independent verification program training sessions on October 5, 1990. The material was well presented, and the employees appeared attentiv The inspectors considered licensee actions on this issue to have been responsive and adequate. During future routine inspections, the inspectors will observe the implenntation of independent verifications as they occur. This violation is close ;5 (Closed) Inspector Followup Item 382/8941-03 The purpose of this item was to follow up on a licensee commitment to provide better guidance in the alarm response procedures for loss of nitrogen pressure in the nain steam isolation valve (MSIV) operators. On April 10, 1990, the i licensee changed the two applicable annunciator respo m e procedures to indicate if MSIV nitrogen pressure is less than 2100 psig, then the MSIV shall be declared inoperable. With an alann trip setpoint of 2300 psig, this allowed time for the operators to add nitrogen before having to declare the valves inoperable. This item is close .6 (Closed) Inspector Followup Item 382/9004-01 During an event on February 7,1990, when a feedwater regulating valve operator flexible air tubing section failed, the licensee made immediate nonconformance repairs in accordance with procedures which were confusing and fragmented. As a result, a 10 CFR 50.59 review or safety evaluation was not implemented when it should have been. On March 1,1990, the licensee comitted to review the process and make appropriate procedure improvements within the next 2 month On May 11, 1990, Administrative Procedure UNT-005-002, Revision 9, " Condition Identification," was changed to more clearly specify the engineering interface This item is close .7 (Closed) Inspector Followup Item 382/9019-02 This item was opened to review additional guidance for maintenance personnel when issue The inspector reviewed a memo dated September 26, 1990, from the maintenance superintendent to all maintenance personnel. The memo provided guidance for procedure use and completion when jobs required work at more than one location in the plant. The inspector considered the guidance clear and adequate. This item is closed.

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7.8 -(Closed) 10 CFR Part 21 Followup Item 90-008 In NRC Inspection Report 50482/90-22, the inspector described a followup inspection conducted on a 10 CFR Part 21 report from the Anchor / Darling Valve Company, relative to a problem with backup seal rings for the feedwater isolation valve (FWlV) actuators. The actions taker, by the licensee were

' adequate, except that there was not sufficient documentation from the vendor to support the licensee's conclusions, based on a telecon report, that the correct material was installed in the FWlVs, and thus no action was required on the installed equipment. The licensee was requested to furnish a more concrete >

basis for this conclusion. The licensee has since provided the inspector with a copy of an Anchor / Darling letter dated October 5,1990, which provided the necessary docurrentation. This issue is closed for Waterford . ONSITE LICENSEE EVENT REPORT (LER) FOLLOWUP (92700)

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The following LER was selected for onsite followup inspection to determine l whether the licensee has taken the corrective actions as stated in the LER.

j In addition, the inspector determined whether responses to the events were

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adequate and met regulatory requirerrents, license conditions, and commitment The NRC tracking status is indicated belo .1 (Closed) LER 382/89-002, " Inadequate Qualificati(n of Instrument Air Tubing During Initial Construction The inspector reviewed a random sampling of 5 design drawings among the total of 63 reported by the licensee as having been revised to show the proper safety class. The drawings were appropriately changed. The inspector also verified that the appropriate design changes, nonconformance rework, or engineering disposition was accomplished and closed by noting record closure on each item and sampling work document records. No problems were found and all corrective actions addressed by this LER appeared to have been completed satisfactoril 'This LER is close ~ BALANCE OF PLANT (BOP) INSPECTION (71500)

The purpon cf this inspection was to complete the objectives of NRC Inspection Procedure 71500 for the 80P. This inspection was 60 percent completed during the inspection conducted July 1-31, 1989. See NRC Inspection Report 50-382/89-22. The areas examined during this inspection period included a determination of the adequacy of BOP operating procedures and a review of the licensee's BOP erosion / corrosion. monitoring progra .The inspector reviewed selected operating procedurcs for the B0P to determine their availability and adequacy. It was found that PORC approved procedures were available for essentially all of the 80P systems. The procedures varied from detailed operating instructions to basic system alignment instructions depending on the system involved. Although the adequacy of all the B0P procedures could not be determined, the below listed weaknesses were noted in light of recent events involving the B0P:

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Operating Procedure OP-003-003, Revision 8, " Condensate and Feedwater  ;

System," did not include instructions for removal of strings of feedwater heaters from service with the plant at power. This evolution was required

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twice recently to repair a tube leak in the No. 2B feedwater heater (NRC l Inspection Report 50-382/90-09) and to gag a leaking relief valve on the i No. 4A feedwater heater (NRC Inspection Report 50-382/90-19). The

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inspector'was informed that informal written instructions were prepared prior to.these evolutions, but as yet they have not been incorporated into l the approved procedure ;

• _Off-normal procedures for transients in the condensate and feedwater system that do not cause a reactor trip were 'not available. A recent transient involving recovery from the loss of extraction steam to the N feedwater heaters- (NRC Inspection Report 50-382/90-22) was not covered !

by grocedure. This transient resulted in reactor power exceeding 100 percent for a short period. The licensee initiated training to correct this weaknes > : 0perating Procedure OP-003-031, Revision 7 " Condensate Polisher and

' Backwash Treatment System," appeared adequate to operate the system but component labeling on the control panel for the system did not agree with the procedure. The procedure utilized the licensee's unique identification number (UNID)' designation for the valves to be operated, while the labels m , on the control panel still indicated the original architect / engineer's L

valve designation. . Operators indicated that the valve description and

! additional. letter designation on the' valve 1sbel plates made the procedure usable. The licensee indicated that a program was being proposed to get

. UNID tags'in agreement with the licensee's databas It was not approved L, yet as tof the end of this inspection perio ,

tThe: inspectors routinely conducted inspection tours of 80P systems and equipment underJInspection Procedure 71707. The systems and equipment generally. appeared r!well maintained, with the exception of a few minor steam and oil leak .

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i- LHousekeeping practices were commensurate with safety-related areas, which have L been exellent with'few exception l c The inspector reviewed the liceMee's erosion / corrosion monitoring program for

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' .high energy systems in the 80P. The program wasLdocumented in Procedure .

EN0ECP-254, Revision 1, " Control of Crosion/ Corrosion." The program involved

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lperiodie. ultrasonic testing of selecteo piping and components in B0P high L; energy systems with criteria for acceptable wall ~ thickness' of measured L' components, ca',culation of prcjected wear rates, and: requirements for L

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reexamihation based on wear rate and:engineerino evaluations to determine if-E compoNnt replacement was necessary. The program.was set up to monitor l> 110 d~signated systems, but systems could be added or deleted based on results-J 'of testingfor other indications of problems. At the request of maintenance 4'

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l engineering,. inspection of the A-turbine cooling water pump discharge piping

!< wasiperfonned during the third refueling outage. The responsible engineer s

oindicated that the safety-related auxiliary component cooling water system i . would be> examined during the next; refueling outage to determine if the system

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was experiencing any microbiologically-induced corrosio s 1

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The inspector reviewed the sumary report of the erosion / corrosion inspections perfonred during the licensee's third refueling outage. The licensee examined approximately 183 components with additional examinations perforned inside the steam generator blowdown tank and on the turbine crossunder piping. These inspections resulted in 33 problem evaluation information requests (PElRS)

submitted to engineering for disposition. Eleven of the evaluations resulted in corrective action such as component replacement, weld overlay repair, and replacement or addition of wear plate .1 Sumary of Findings No violations or deviations were identified during the 80P inspection. The inspectors concluded, based on the specific B0P inspection activities, past and present routine tours and observations, and the performance of BOP equipnent in general, that the only significant differentiation the licensee makes between the B0P and safety-related programs and equipment is the level of QA and certificatio The preventive and corrective maintenance practices have been essentially the same. Housekeeping and equipment conditions have been the same, with the exception that B0P equipment leaks appeared to go uncorrected for a longer period. In many cases, nothing could be done, within reason, on BOP steam or feedwater leaks until the plant was shut down. The inspectors considered the BOP programs at Waterford 3 to be a strengt . EVALUATION OF LICENSEE QA PROGRAM IMPLEMENTATION (35502)

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On August 27, 1990, regional management performed an evaluation of the effectiveness of the licensee's-QA program implementation by conducting an in-office evaluation of the following:

NRC inspection reports for the past 12 months; Systematic assessment of licensee performance (SALP) reports for the past 2 years;

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Region IV outstanding open items list; I

Licensee corrective actions for NRC inspection findings; and Licensee event reports for the past 12 months.

I On the basis o' the evaluation, the NRC staff found no negative performance trends in any of the seven SALP functional areas. The staff noted centinued strengths in operations, radiological controls, and securit Emergency planning had not been fully evaluated due to the annual exercise being scheduled for October. 24, 1990. See NRC Inspection Report 50-382/90-20 for the results of that exercise. There are also some brief conments in paragraph 6 abov Maintenance and surveillance has continued to improve with the completion of maintenance procedure upgrades approachin The NRC Staff determined that no adjustnents to regional inspection plans will be required as a result of the above evaluatio ,

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4 On October 23, 1900, the Region IV Chief. Project Section A, and the senior resident. inspector met at the Waterford 3 facility with licensee management to a discuss the results of the evaluation. Subjects discussed are outlined in

. Appendix C of this repor .1 ' EXIT INTERVIEW i

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The inspection scope and findings were suninarized on November 10,.1990, with those porsons indicated in paragraph 1 above. The licensee acknowledged the inspectors' findings. The licensee did not identify as proprietary any of the

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material. provided to, or reviewed by, the inspectors during this inspectio * *

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, APPENDIX C

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OUTLINE

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WATERFORD 3 MIDCYCLE MEETING

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OCTOBER 23, 1990 (10VEMBER 1, 1989 THROUGH APRIL 30,1990(SALPPERIOD)

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ATTENDEES R. P. BARKHURST, VICE PRESIDENT, OPERATIONS J. R. MCGAHA, CENERAL MANAGER, PLANT OPERATIONS R. G. AZZARELLO, DIREC10R, ENGINEERING AND CONSTRUCTION R. F. BURSKI, DIRECTOR NUCLEAR SAFETY

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, D. E. BAKER, DIRECTOR, OPERATIONS SUPPORT AND ASSESSMENT

P. V. PRASANKUMAR, TECHNICAL SERVICES MANAGER

]

D. F. PACKER MANAGER, OPERATIONS AND MAINTENANCE '

JT. P. BRENNAN, DESIGN ENGINEERING MANAGER

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.NRC T. F. WESTERMAN, CHIEF, PROJECT SECTION A I W. F. SMITH, SENIOR RESIDENT INSPECTOR

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') PERFORMANCE INDICATORS (LAST 4 QUARTERS)

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- 3 SCRAMS 3SAFETYSYSTLMACTUATIONS(ASSOCIATEDWITHSC.',AMS)

ESCALATED ENFORCEMENT

}

3 ENFORCEMENT CONFERENCES 1 LEVEL 111(N0CP)-

2 LEVEL IV (N0 CP) '

EA 90-117 - HP TECHNICIAN FALSIFICATION OF SURVEY REPORT PROPOSED NONCITED DUE TO GOOD LICENSEE ACTIONS NORMAL ENFORCEMENT 12 -' LEVEL IV VIOLATIONS  :

2 - NONCITED VIOLATIONS (NCVs)

1 - LEVEL V VIOLATION ,

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ENFORCEMENT INSIGHTS TREND FAVORABLE LESS INSPECTIONS /SALP RATING

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SELF IDENTIFICATION LERs 13 LERs, ALL SELF- OR LICENSEE-IDENTIFIED, EXCEPT ONE

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PLANT OPERATIONS I OPERATION - SAFE AND PROFESSIONAL i

GRID FAULT REACTOR TRIP l HOUSEKEEPING AND EQUIPMENT CONDITIONS ISLOCA INSPECTION, BUT ALL CONTAINED PRIMARY SYSTEM LEAKS

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OPERATING PROCEDURE UPGRADE PROGRAM SLIPPED JULY TO SEPTEMBER 1990 2500F260(ASOFSEPTEMBERI)

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TRAINING AND QUALIFILATION PROGRAM

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, APRIL 1990 INSPECTION

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ENTERGY OPERATIONS INC. TRANSITION REQUALIFICATION EXAMINATIONS i L RCP 2A FLANGE LEAKAGE

. RADIOLOGICAL CONTROLS RESIDENT INSPECTOR OBSERVATION CONTROLLED SURFACE CONTAMINATION 3.4 PERCENT TO 2.7 PERCENT GOAL UNDER 5 PERCENT WORK PRACTICES ALAPA RADI0 ACTIVE MATERIAL TRANSPORTATION PROGRAM SELF IDENTIFICATION HP TECHNICAL' FALSIFICATION OF SURVEY REPORT PORC APPROVAL OF RAD CONTROL PROCEDURES

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MAINT; A E AND SURVEILLANCE MAINTENANCE PERFORMANCE TEAMWORK PROCEDURE ADHERENCE PROCEDURES HECHANICAL AND ELECTRICAL - ECD OCT 1990

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1 & C - ECD FIRST QUARTE 8' 1991 SURVEILLANCE NO MISSED SURVEILLANCES IN PAST QUARTER SURVEILLANCE PROCEDURES IMPROVING

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PROBLEMS CONDENSERVACUUMPUMPGASMONITORSETP0 INT (APRIL 199d)

LTOPLIFTSETPOINT(MAY1990) ,

CONTAINMENTLOCALLEAKRATETESTING(JUNE 1990 INSPECTION)

TWOFWCHECKVALVESANDEDGFUELOILPUMPOMITTEDFROM1ST(JUNE

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1990) ,

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OVERALL PERCENTAGE OF DEFICIENCIES SMALL EMERGENCY PREPAREDNESS

'[ EXERCISE IN OCTOBER 1990

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JUNE 1990 INSPECTION ,

SECURITY NEW CAMERA SECURITY SYSTLM ONE ISSUE OF UNESCORTED ACCESS ONE DRSS INSPECTION

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ENGINEERING AND TECHNICAL SUPPORT PROGRAMS INITIATED ,

TRANSITION TO A FULL-DESIGN ORGANIZATION NEW ENGINEERihG BUILDING ON SITE PLAN OF DAY ATTENDANCE CAD SYSTEM DBD SYSTEM

<. 12 PACKAGES OF 50 COMPLETED SYSTEM ENGINEERS 22 IMPROVED EVALUATIONS WEEKLY PRESENTATIONS

'0BSERVATIONS RETEST HVAC SYSTEM DAMPERS POORLY ENGINEERED

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SLOW RESPONSE TO TECHNICAL SAFETY CONCERNS HVAC DAMPER RELAYS

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SG BLOWDOWN VALVE CLOSURE TIMES

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REG GUIDE 1.97 INSPECTION WEAKNESSES WEAKNESSESINDOCUMENTFILINGANDRETRIEVALPROGRAM(NRCINSPECTION REPORT 50-382/89-39)

ISLOCA INSPECTION INDICATED NO OPERABILITY PROBLEMS FIRE PROTECTION PROGRAM OPEN ISSUES

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QA ORGANIZATION CHANGES  ;

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SELF-IDENTIFICATION AND CORRECTI0tl l

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