IR 05000346/1992019

Insp Rept 50-346/92-19 on 921216-930125.Violations Noted,Not Cited.Major Areas Inspected:Action on Previous Insp Findings,Ler Followup,Operational Safety,Surveillances & Maint

ML20128K833
Person / Time
Site:	Davis Besse
Issue date:	02/10/1993
From:	Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20128K822	List: IR 05000346/1992019 ML20128K819
References
50-346-92-19, NUDOCS 9302190021
Download: ML20128K833 (14)

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', U. S. NUCLEAR REGULATORY COMMISSION REGION III'

Report No. 50-346/92019(DRP)

Docket No. 50-346 Operating License No. NPF-3 Licensee: Toledo Edison Company Edison Plaza, 300 Madison Avenue Toledo, OH 43652 Facility Name: Davis-Besse Nuclear Power Station Inspection At: Oak Harbor, Ohio

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inspection Conducted: December 16, 1992, through January 25, 1993 Inspectors: S. Stasek R. K. Walton s

Approved By: 2 _ 2- to n R. D. Lank Mb- -it f ) ate

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Reactor Kajects-Section 3B ,

Inspection Summary

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Inspection on December 16. 1992. throuah January 25. 1993 (Report No. 50-346/92019(DRP))

Areas Inspected: A routine safety inspection by resident inspectors of action on previous inspection _ findings, licensee event report followup, operational safety, surveillances, and maintenanc .

Results : An executive summary follows:

Plant Operations: Overall, performance of the operating crews was good this inspection period. Inspector review of a December 14 inadvertent boron addition to the reactor coolant system determined operators responded appropriately and conservatively. In general, adherence to administrative controls was good. However, a clean waste monitor tank-pump was inadvertently operated for several hours with its suction valve closed due to an-incorrect performance of a valve lineup. Two control room emergency ventilation system valves that were found out of position on December 8 were due to an operator-error while returning the system to standby and were classified as a non-cited violatio '

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Radiolooical Controls: The licensee implemented the revised 10 CFR Part 20-requirements on January 1, 1993. Results of a containment neutron distribution survey conducted in October 1992, were issued this period indicating significant neutron dose savings from those originally use Adherence to radiation protection program requirements was good this period 9302190021 930210 PDR ADOCK 05000346.-

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with no substantive problems noted, liowever, the inspectors noted some inconsistencies with radiological postings on one occasio Maintenance / Surveillance 1 Overall, surveillance and maintenance activ'i'es observed during the inspection period appeared to be conducted in acco t: ace with all applicable requirements. However, on one occasion, the inspector noted a failure of Instrument and Control (!&C) technicians to adhere to personnel red tagging requirements that resulted in a non-cited violatio Additionally, a failure to adequately perform an independent verification on September 16, 1992, that resulted in exceeding a Technical Specification limiting condition for operation was classified as a non-cited violation this inspection perio Engineerino/ Technical Support: Engineering support to plant organizations on '

a day-to-day basis as well as in response to identified problems was gccd. In general, engineering expertise was routinely considered as a substantial resource to be utilized by other parts of the organization onsite, i

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. DETAILS 1. Persons Contacted L Toledo Edison Company D.Shelton, Vice President, Nuclear G. Gibbs, Director, Quality Assurance

• L. Storz, Plant Manager

• J. W. Rogers, Manager, Maintenance

• S. Jain, Director, Engineering

• E. Salowitz, Director, Planning J. K. Wood, Operations Manager

• J. Polyak, Manager, Radiological Protection

• V. Sodd, Manager, independent Safety Engineering l

• D. Timms, Manager, Systems Engineering

• G. Grime, Manager, Security R. Schrauder, Manager, Nuclear Licensing l J. Wissner, General Supervisor, Instrument and Control Maintenance

• C. Hawley, Superintendent, Shift Operations l

• J. P. Hartigan, Supervisor, Design Engineering  ;

• J. E, Blay, Engineer, Independent Safety Engineering l

• Honma, Supervisor, licensing

• K. Peterson, Licensing

• W, Rabe, Supervisor, Quality Assurance  !

• R. Wuokko, Supervisor, Regulatory Affairs i

• C. Zyduck, Manager, Nuclear Engineering i

• L. Tabbert, Supervisor 'n lpendent Safety Engineering i

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• S. Stasek, Senior Resident Inspector

• R. K. Walton, Resident Inspector

• Denotes those personnel attending the January 25, 1993, exit meetin '

l 2. Followup of Previous inspection Findinas (92701) (Closed) Unresolved Item (346/92014-01(DRP1): Improperly performed independent verification (IV) of a Steam and Feedwater Rupture Control System (SFRCS) detector equalizing valve. On September 16, 1992, Instrument and Controls (I&C) technicians completed testing of a differential pressure switch and returned the detector to service with the equalizing valve three turns open in lieu of closed. As documented i Licensee Event Report 346/92-008, this error rendered two SFRCS detectors inoperable and was not detected until September 19, 199 The licensee concluded that the technician directed to " restore the switch" failed to completely close the valve and a second technician failed to properly perform an independent verificatio .

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The licensee's corrective actions for this event were discussed in inspection Reports 50-346/92014(DRP) and 50-346/92017(DRP), ;

The inspectors interviewed the individual, an I&C journeyman level mechanic with about 5 years experience, who incorrectly performed i the IV and concluded that he had been adequately trained to '

perform independent verification activities and had recognized I that an IV was required by the surveillance tes Further, he indicated he could not explain why he failed to perform the independent verification correctl l The initial mispositioning of the valve and the improperly performed independent verification that would have detected the !

error, appeared to be isolated actions by the two technici n The technicians were subsequently counseled and disciplined. This l event was discussed with personnel in the maintenance shop and a memorandum was issued to shop personnel on how to properly perform independent verifications. Maintenance management also subsequently submitted a training action request (TAR) to ensure that all shop personnel received adequate independent verification trainin The inspectors observed the performance of DB-MI-03204, Channel Functional Test and Calibration of SFRCS Actuation Channel 2, on January 6,1993, and noted that the individuals utilized a copy of the procedure in the field and properly performed the independent verifications in accordance with the licensee's guidance memorandu The test procedure, DB-MI-03203, Channel Functional Test and Calibration of SFRCS Actuation Channel 1, revision 1, step 8.4.1.4.c., required that the subject equalizing valve to PDS-2686A be close In fact, the valve was found three turns ope This same step required that the equalizing valve position be independently verified. The step was improperly signed as having been independently verified. Additionally, Technical Specification (TS) Limiting Condition for Operation 3.3.2.2.,

Action statement a., states in part, "With a SFRCS instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3.11, until the channel is restored to OPERABLE status..." With the equalizing valve for PDS-2686A not closed, the detector was rendered inoperable on September 16, 1992, but the channel was not placed in the tripped condition within I hour as required by Action 16 of Table 3.3.1 The failure to adequately implement DB-MI-03203 is considered a violation of TS 3.3.2.2. as well as TS 6.8.1.c (which specifies written procedures be established and implemented for surveillance and test activities relating to safety-related equipment). However, further review determined that SFRCS remained functional with the instrument inoperable, the event was isolated in nature, and that the failure to do the IV was made at the technician leve . . . _

Therefore, the violation will not be cited since the criteria specified in Section Vll.B.2. of the " General Statement of Policy and Procedures for NRC Enforcement Actions," (Enforcement Policy 10 CFR Part 2, Appendix C (1992)) were satisfied, b. (Closed) Open item (346/92017-Ol(DRP.j h On December 8, 1992, during performance of DB-SS-03041, Control Room Emergency Ventilation System (CREVS) Train 1 Monthly Test", an operator found valves HAIS and HA17 open when their required position was closed. Valves HAIS and HA17 are capped isolation valves which are maintained closed to prevent freon from migrating to the air cooled condenser located on the roof of the auxiliary buildin The licensee subsequently determined that DB-0P-06505, " Control Room Emergency Ventilation System Procedure," (which included manipulation of the subject valves), was last used on November 25, 1992, when CREVS was placed into service to provide control room ventilation while the normal ventilation system was unavailable due to maintenance. On that day, to return CREVS to a normal standby lineup, the control room operator secured the CREVS fan by simply operating its control switch from the control room (and documented that the system was secured on the turnover sheet), but valves HAIS and HA17 were not locally repositioned closed as specified by step 5.2.6. of DB-0P-06505. Because the operator apparently felt returning the system to standby was a routine evolution, in-hand usage of the procedure was not required by DB-0P-00000, " Conduct of Operations".

With valves HAIS and HA17 open, the licensee indicated that the compressor would still have had starting capability but could subsequently trip on a low suction pressure (caused by the freon migration). Over a short period of time, freon pressure would have returned, the compressor trip would have automatically reset, and the compressor would have started again. The number of times the compressor could have cycled was dependent upon the degree of freon migration to the air cooled condenser, however,-the licensee's position was that the system remained operable with valves HA15 and HA17 open. The inspectors spoke with engineering staff and concluded that the air cooled mode of cooling would be in service well before exceeding any room temperature limit Step 5.2.6 of DB-0P-06505, revision 0, Control Room Emergency Ventilation System Operating Procedure, required that valves HAIS and HA17 be positioned closed. On December 8, 1992, valves HAIS and HA17 were found open. This was a violation of Technical Specification 6.8.1.a. (required written procedures specified in Regclatory Guide 1.33 be established, implemented, and maintained), in that procedure DB-0P-06505 was not properly implemente The violation will not be cited since the criteria specified in Section VII.B.1 of the " General Statement of Policy and Procedures

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for NRC Enforcement Actions," (Enforcement Policy 10 CFR Part 2, Appendix C (199?)) were satisfied.

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The licensee documented this event on a Potential Condition Adverse to Quality Report (PCAQR 92-0462). The individual who improperly secured CREVS was removed from licen ad duties and disciplinary actions were implemented in accordance with company policy. Additionally, during discussions with the Manager-Operations, the inspector was informed that administrative procedure DB-0P-00000 was currently in process of revision to, in part, better address procedural adherence requirements and management expectations in that are c. (Closed) Unresolved item (346/92017-03(DRP)): Inadvertent boric acid addition to the reactor coolant system during a dilution -

operation. On December 14 1992, with the #1 Boric Acid Addition Tank (BAAT) being recirculated to support chemistry sampling, control room operators attempted a series of reactor coolant system (RCS) dilution evolutions to maintain a target of about 96%

withdrawn on Control Rod Group (CRG) These evolutions had been ongoing for some period of time and included using substantial amounts of water due to the core being near the end of the operating cycle. With CRG 7 approaching 98% withdrawn position s operators anticipated adding approximately 2000-2400 gallons of water to reach the target rod positio At 7:39 p.m., operators added 800 gallons of water to the RCS, but noted about 30 minutes later that the water addition had little affect on CRG 7 and the rod group continued to move slowly out of the core. Around 8:20 p.m., CRG 7 was at 100% withdrawn from the core. The shift supervisor thought that the lock of rod response was possibly due to a xenon imbalance in the core from earlier control rod maneuvers and that the water addition was of insufficient volume to overcome this xenon imbalance At i 8:58 p.m., operators added an additional 1611 gallons to the RCS in an attempt to return CRG 7 to the 96% position. Again, after the second water addition, operators noted that the control rods did not appear to respond to the water addition. The operators, suspecung that a boric acid addition to the RCS was made during the second water addition, verified that chemistry had completed sampling of the #1 BaAT and at 9:23 p.m., secured the #1 BAAT pump and checked that valves MU23 and MU356 used to isolate the #1 BAAT recirculation piping from the makeup system were closed. About the same time, the boronometer showed a slight increase in RCS boron concentration and operators commenced decreasing reactor

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power to maintain T m constant. At 9:30 p.m., operators noted that the #1 BAAT had decreased bbout 100 gallons and requested t!.at the chemistry department sample the RCS for boron concentration. Reactor power was stablized about 10:45 p.m., with power at 90%. Chemistry tonfirmed that RCS boron concentration

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[ had increased about 10 pp Subsequently, water additions 1to the j, RCS continued during the night and reactor pover was returned to

/- 100% about-5:15-a.m. on December 15, 1992, i The licensee determined that root cause of the boric acid Mditton .

to be seat leakage past MU23. Operators had followed_DB-OP-06031,

"BAAT Operating Procedure", appropriately, but the procedure-itself was deficient. This was because valve MU23 was apparently designed to function as a control valve and was not intended-to provide a positive isolation function. Subsequently, Operations issued a temporary change to DB-0P-06031 to close manual isolation valve MU363 when recirculating a BAAT. This was to provide an-additional isolation to prevent inadvertent addition of boric acid to the makeup tank when the BAAT system is pressurize The inspectors reviewed this event and the licensee's corrective-actions and found operator actions were properly conducted during the event and the corrective actions taken should preclude repetitio This item is close No violations or deviations were identified; however, two non-cited violation was identified in this are . Licensee Event Report followun (92700)

Through direct observation, discussions with licensee personnel, and [

review of records, the following licensee event _ report (LER) was

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reviewed to determine that reportability requirements were fulfilled, immediate corrective actions were accomplished in accordance with Technical Specifications (TS), and correct /e action to prevent recurrence had been establishe (CLOSED) LER 92-006-00. Deficient reactor protective system (RPS)

calibration procedures. This LER documented the licensee's _di:.covery that procedures used for calibration-of "High Flux" and " Flux / Delta-Flux / Flow" and "High Flux / Number of Reacto_r Coolant Pumps 0_n" trip functions did not address associated common circuitry and as-a result, a-non-conservative "High Flux / Number of Reactor Coolant Pumps On" tri function setpoint may not have been promptly dete_cted. The licensee's review determined that no actual- problem existed.. The inspector reviewed the licensee's corrective actions for this event, with the exception of reviewing the modified RPS calibration procedures, to ensure that these procedures did not introduce any advarse variables into the calibration process. To finalize review of this matter, the inspector witnessed the performance of DB-MI-03057, "RPS Channel-1 Calibration of Overpower, Power / Imbalance / Flow, and Power / Pumps Trip Functions" utilizing the subsequent revision and determined that this procedure did not -introduce any adverse variables into the calibration process. Additionally, the . inspectors observed that the output vol.tages for the contact-monitor-(which provides inputs to the "High Flux / Reactor-Coolant Pumps On" bistable) were set more conservatively than previously .

to account for instrument drift. The inspectors found a minor test

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[, deficiency and spoke-to the 180 foreman-and systems engineer about a

, possible procedure change. This item is close No violations or deviations were identified in this are '

4. Operational Safety Verification (71707) (71714)'(40500)

The inspectors observed control. room operations, reviewed applicable logs, and conducted discussions with control room operators during the inspection period. The inspectors verified the operability of selected emergency systems, reviewed tagout records,. 'and verified tracking of-limiting conditions for operation associated with affected component Tours of the auxiliary and turbine buildings were conducted to observe plant equipment conditions including potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests - - -

had been initiated for certain pieces of equipment in need of maintenance. Walkdowns of the accessible portions-of the following systems were conducted to verify operability by comparing system lineups with plant drawings, as-built configuration, or present valve lineup lists; observing equipment conditions that could degrade performance; and verifying that instrumentation was properly valved, function:ag, and calibrate Diesel Generators 1-1 and 1-2 and auxiliaries

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High Pressure Injection System - Divisions 1 and 2

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Hydrogen Dilution System The inspectors, by observation and direct interview, verified that the physicti security plan was being implemented in-accordance with the station security plan, including badging of personnel; access control; security walkdowns; security response (compensatory actions); visitor control; security staff attentiveness; and operation of security equipmen Additionally, the inspectors observed plant housekeeping, _ general plant-cleanliness conditions, and verified implementation of radiation protection control Specific- observations and reviews included the following:

i On May 21, 1991, the NRC issued a revision to 10 CFR Part 20, i

" Standards for Protection Against Radiation". The revision'to the rule was based on changes prepared by the Internacional Council on Radiation. Protection.- This revision changed annual: dose. limits, radiological posting requirements, and required licensees to have

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programs which minimized radiation exposures to individ" 7s to "as low.as reasonably achievable" (ALARA). -The new rule 6. ws effective on January 1,1994, although licensees _had the option of implementing the rule as early as January 1,199 .

Efforts had been ongoing since May 1991, under th~e guidance of Radiation Protection, to identify and prepare the necessar '

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changes to procedures, training, and equipment necessitated by the rule revision. As of January 1, 1993, the licensee implemented the mandatory portions of the 10 CFR Part 20 revision, b. During the inspection period, a review was conducted to verify that the plant had been adequately readied for cold weather operations. The inspector reviewed DB-0P-06913, " Plant Winterization Checklist," DB-0P-06331, " Freeze Protection and Electrical Heat Trace," and DB-0P-06222, " Condensate, Domineralized, and Primary Water Transfer and Storage system" as well as several preventative maintenance activities that had been completed in assure proper implementation of the licensee's program. In addition, walkdowns of selected portions of associated piping were conducted to verify proper operation of heat trace circuits. No substantive concerns were identifie The licensee's winterization program was found to be acceptable ind edequately implemented, c. During a tour of the auxiliary building on December 18, 1992, the inspectors noted th;t the entrance way to the #< lean Waste Receiver Tank (CWRT) room was posted with a high radiation area sign due to draining reactor coolant water on the evening of December 14, 199 The inspectors noted that the radiological survey map posted tiear the entrance was dated December 7, 1992, and as such, did not show the high radiation area. The inspectors then reviewed the radiological survey logs maintained by the radiological protection foreman at the entrance to the auxiliary building. These maps were dated December 14, 1992, and since the survey was taken before discharging reactor coolant, the log likewise did not reflect the change in radiological status of the

1. 2 CWRT room. The inspectors reviewed the auxiliary building display boards in the hallway outside the entrance to the radiological control area and found that they too did not reflect the radiological conditions in the #2 CWRT roo The inspectors questioned the foreman and found him to be knowledgeable of the change of radiological conditions in.the room. The insoectors reviewed the licensee's administrative procedure DB-HP-00003, Radiological Surveillance Program, and found that there was no requirement to keep the survey maps or the display board updated. The General Supervisor, Radiological Controls, indicated that each radiological work permit wat annotated to check with the radiological controls foreman prior to entering any radiation are Since the #2 CWRT room was properly posted for the existing radiological conditions, no violation of licensee procedures or regulatory requirements resulted. _ However, the licensee agreed that all postings should be consistent and that actions would-be taken in the future to keep postings updated in an appropriate timeframe.

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d. In mid-October 1992, the licensee and a contractor performed a spectral analysis of neutron energy distributions in containment for use in both dose assessment and shielding evaluations. The study, issued this inspection period, concluded thct the neutron energies encountered inside the containment building with the reactor operating at 100% power were typical of those seen in other presurized water reactor (PWR) containments. In addition, the study indicated that the spectrum of neutron energies in containment at full power were lower than previously anticipate The licensee intended to apply this data to modify the exposure received by individuals who made containment entries at power this past year and reduce the mean quality factor as allowed by 10 CFR 20.1004. This matter was being tracked by the Region 111 Division of Radiation Safety and Safeguards via open item (346/92005,-01 (DRSS)) .

e. On December 17, 1992, at 6:42 p.m., a zone operator noticed that t the #1 Clean Waste Monitor Tank (CWMT) pump was operating at

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elevated temperatures. The operator found the tank outlet control valve (located on the suction of the pump) closed and contaded the control room. Subsequently, the #1 CWMT pump was stoppc1 The licensee subsequently documented this condition in a Patential Condition Adverse to Quality Report (PCAQR 92-0474).

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l Earlier, day shift operators had aligned the CWMT system in accordance with DB-0P-06101, Clean Radwaste System operating l procedure, section 3.7., " Recirculation of the CWMTs". A valve l lineup was performed using Attachment 30 of the procedure which allowed the operator to choose a position for pump suction valve, WC1704. The operator erroneously circled the " closed" position on the valve lineup sheet and checked the valve to be in the closed positien. Attachment 30 was independently verified by a second operato At 1:10 p.m., operators started the #1 CWMT pump with the tank in a recirculation condition to allow sampling of the tan The pump continued to operate with its suction valve closed for approximately 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> until the misalignment was identified the following shif Subsequently, it was determined that the pump had not been damaged by the even Although there was no effect on the safe operation of the plant, the inspectors were concerned that the improper position of a valve on a lineup shut could also occur on a safety-related system. The inspectors spoke to the operations superintendent and sampled valve lineup sheets of safety system operating procedures and found that two of the procedures provided a choice for the position of valves. However, the licensee indicated that those valve positions were selected by the shift supervisor before being given to the operato ...

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The inspectors also noted that the valve lineup sheet had a signoff requiring a review by a senior operator, and that this review signoff was not made prior to starting the pum Operations management specified that the review signoff was performed at the completion of the procedure and not after the completion of the valve lineup. However, a shift supervisor stated that the lineup review signoff could be perfermed after the lineup was completed during infrequent system operations but may not be performed during routine evolution The inspectors could not locate a written policy stating when the valve lineup review signoff should be signed. The inspectors spoke with operations management about the corrective action to PCAQ 88-0934, (which documented an event involving starting decay heat pump #1 with its suction valve closed) which stated "a review of major plant component prestarts is performed to ensure that i

equipment can be placed in service in a safe and logical order."

l Operations management indicated that Management Corrective Action Report (MCAR) 89-0001, which implemented the corrective actions to PCAQ 88-0934, made no reference to the review signoff and could nct determine what initiated the signoff requiremen The inspectors were concerned that the system was placed into l service prior to supervisory review of the system's status and

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that a signoff for such a review was included in the valve lineup.

l It appeared that step 3.7.2.a., which required performance of Attachment 30, was also improperly initialed by the operator as being completed, when in effect, the attachment was not completed l since the review signoff was not made.

Operations management, subsequently, issued a memorandum to all

, operators specifying the policy for completion of procedural steps that referenced attachments in the procedur This appeared to be an example where management expectations were not adequately communicated to the operators. In addition, discussions with the Manager-Operations, revealed that

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administrative procedure, DB-0P-00000, was under revision to l incorporate certain Institute of Nuclear Power Operations (INPO)

, recommendations, some of which addressed procedure adherence. The

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revision was also to clarify procedure usage to provide better guidelines to address the type of operator errors as discussed in paragraph 2.b of this report. It appeared that the revision to DB-0P-00000 may address the inspectors' concerns with providing more specific guidelines on procedural adherence to the operator Therefore, this matter is considered an open item (346-92019-01(DRP)) pending incorporatior. of the INPO recommendations into DB-0P-00000 and subsequent review of the procedure by the inspector No violations or deviations were identified in this are . __

, 5. Surveilj ance (61726)

The inspectors observed safety-related surveillance testing and verified i that the testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions of operation (LCO)s were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specification and procedure requirem1nts and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne The following test activities were observed and/or reviewed:

DB-MI-03057 Reactor Protective System, Channel 1 Calibration DB Mi-03204 Channel functional Test and Calibration of Steam and Feedwater Rupture Control System Channel 2 DB-SL-03070 Emergency Diesel Generator #1 Monthly Test DB-SC-03113 Safety Features Actuation System Channel 4 Functional Test DB-SP-03321 Hydrogen Dilution System Train 2 Quarterly DB-SS-03091 Motor Driven feedwater Pump Quarterly Test ST 5091.01.10 Source Range functional Test No violations or deviations were identified in this area, 6. Maintenance (62703)

Station maintenance activities of safety-related systems and components were observed and/or reviewed during the inspection period to ascertain that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with Technical Specification The following items were considered during this review: the limiting conditions for operation (LCO) were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and fire prevention controls were implemente Maintenance work orders (hW0s) were reviewed to determine status of cutstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which may affect system performance,

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The following maintenance activities were observed and/or reviewed:

f MWO 3-93-2841-01 Calibration of'#2 Boric Acid Addition Tank pump '

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discharge gag MWO 3-93-4801-01 . Clean and Inspect #1 Emergency Core Cooling Room Coole *

- MWO 7-92-0275-04 Replacement of Service Water Piping =to-Emergency Core Cooling System Room Coolers. ' ._

. MWO 7-92-0497-01 Troubleshoot Reactor Protective System Channel

1. 1 Bypass Switc Regarding MWO 3-93-2841-01, on December. 15, 1992, the inspectors-witnessed I&C technicians reinstall a calibrated gage for the #2 BAAT pump discharge pressure gage. The inspectors observed that there were no red tags hung.to isolate the maintenance area from the BAAT system, but instead the technicians had closed the gage isolation valve and placed a pressure tight plug at the gage fitting. The technician i indicated that red tags were not required for the performance of the job-and an administrative procedure allowed the technicians to set their own isolation.

Administrative procedure DB-0P-00015, Safety Tagging, section 6.1.21.b, stated "that a Personal Red Tag Clearance is not required when the necessary isolation can be provided by closing valves normally operated-by the shop performing the work and those personnel involved in the work remain in the vicinity of the instrument and the isolation' valve (s) at--

all times during the activity." The inspectors found the pressure gage removed and no personal red tags hung for the maintenance activit There were no I&C technicians in the vicinit The ins;;ectors spoke to the I&C General Supervisor about the . ,

administrative requirement and the observed work practice. The supervisor stated that other I&C technicians in the shop were aware of the administrative requirements for hanging personal red' tags for maintenance activities and that this was practiced by the technicians.~

The work activity witnessed by the inspectors was classified by the -

supervisor as an isolated event. The individual was reinstructed on.the

- administrative procedural requirement The inspectors noted that there was no safety significance to the observed equipment configuration but were concerned'that_the:

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administrative requirements were not being adhered to by the individuals involved. The inspectors will continue to monitor the licensee's: adherence to.its administrative procedure The failure to implement DB-0P-00015, revision 1, section 6.1.21.b,- was a violation of Technical Specifications 6.8.1.a, Safety Guide 1.33,.

Appendix A, A.3, Administrative Procedure for Equipment Contro However, since the criteria specified in Section VII.B.1 of the " General-Statement of Policy and Procedures for NRC Enforcement Actions,"

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, (Enforcement Policy 10 CFR Part 2, Appendix C (1992)) were satisfied,

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this violation will not be cited.

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No violations or deviations were identified; however, one non-cited violation was identified in this are ]

7. Liplations for Whigit A "Noijce of Violation" Will Not_3g_ listed The NRC uses the Notice of Violation to formally document failure to meet a legally binding requirement. However, because the NRC wants to ,

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encourage and support licensee's initiatives for self-identification and I correction of problems, the NRC will not issue a Notice of Violation if I the requirements set forth in 10 CFR Part 2, Appendix C, Section Vll.B.1 !

i or Vll.B.2 are met. Violations of regulatory requirements identified I during the inspection for which a Notice of Violation will not be issued are discussed in paragraphs 2.a, 2.b, and . Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspectors, and which involve some action on the part of NRC or licensee or both. An open item disclosed during the inspection is discussed in paragraph . Exit Interview The inspectors met with licensee representatives (deneted in f paragraph 1) throughout the inspection period and at the conclusion of the inspection on Janus.y 25, 1993, and summarized the scope and l

findings of the inspectior activities. The licensee acknowledged the findings. After discussionr with the licensee, the inspectors have determined there is no proprietary data contained in this inspection repor _