IR 05000346/1988026

Insp Rept 50-346/88-26 on 880816-0930.No Violations or Deviations Noted.Unresolved Items & Weakness Noted.Major Areas Inspected:Operational Safety Verifications,Maint, Surveillance,Lers,Licensee Events & Quality Verification

ML20206K346
Person / Time
Site:	Davis Besse
Issue date:	11/10/1988
From:	Defayette R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206K344	List: IR 05000346/1988026 ML20206K340
References
50-346-88-26, NUDOCS 8811290374
Download: ML20206K346 (16)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-346/83026(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, Unit 1 Inspection At: Oak Harbor, Ohio and Lynchburg, Virginia Inspection Conducted: August 16 through September 30, 1988 Inspectors: P. H. Byron D. C. Kosloff T. D. Reidinger D. L. Shepard T. E. Vandel L. N. Valenti $$

Approved By: RobertW.DeFayette,Cdef // /0 '

Reactor Projects Section 3A Datb Inspection Summary

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Inspection on August 16 through September 30, 1988 (Report No. 50-346/88026(DRP))

l Areas Inspected: Routine, unannounced inspection by resident and region based inspectors of licensee action on previous inspection items; operational safety verifications; maintenance; surveillance; licensee event reports; licensee

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events and quality verificatio Results: No violations or deviations were identified; however, five

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unresolved items were identified (Paragraphs 4, 5, and 7). A weakness also was identified in the licensee's apparent lack of progress in implementing the requalification training program for simulator scenario development and exercise evaluation (Paragraph 3),

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DETAILS

, Persons Contacted  ;

' Toledo Edison Company (TED)

D. Shelton, Vice President, Nuclear

• L. Storz, Plant Manager ,

N. Bonner, Assistant Plant Manager, Maintenance  !

• R. Flood, Assistant Plant Manager, Operations  ;

J. Kasper, Operations Superintendent

• Ricci, Operations Supervisor t

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• E. Salowitz, General Superintendent Outage and Program Management

• L. Ramsett, Quality Assurance Director

• S. Jain, Independent Safety Engineering Director  ! Grime, Industrial Security Director B. Beyer, Davis-Besse Services Director i

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T. Myers, Technical Services Director

• P. Hildebrandt, Engineering General Director i

• D. Timms, Electrical and Control Systems Manager f W. Johnson, Primary Systems Manager

• E. Chimahusky, Test / Project Supervisor i

• K. Ellison, Performance Engineering

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• G. Honma, Compliance Supervisor

• R. Schrauder, Nuclear Licensing Manager ';

L. Harder, Radiological Operations Supervisor T. Haberland, Electrical Superintendent C. Daft, Technical Planning Superintendent

• D. Breese, Facility Hodification Supervisor ('

J. Moyers, Quality Verification Manager S. Zunk, Nuclear Group Ombuds. tan D. Harris, Quality Systems Manager p

• J. Sturdavant, Licensing Principal l

. G. Skeel, Nuclear Security Operations Manager l

! E. Benson, Nuclear Materials Manager  ;

j *R. Simpkins, Nuclear Operations Training Manager r US NRC

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i *P. Byron, Senior Resident Inspector

*D. Kosloff, Resident Inspector [

, L. Valenti, EG&G l

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i * Denotes those personnel attending the October 4, 1988 exit meeting.

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. Licensee Action on Previous Inspection Findings (92701)

(Closed) Open Item (85004-05): Implementation of Facility Change Request (FCR)78-309. FCR 78-309 was a proposed modification intended to eliminate the need to flush radioactive contamination from the radiation monitors, RE 1878A and RE 18788, for liquid radioactive waste. Before implementing FCR 78-309, the licensee discovered that anothee modification, FCR 86-236, eliminated the need fcr flu',hin FCR 86-236 was completed in October 1986 and FCR 78-309 was voide No violations or deviations were identified in this are . Operational Safety Verific.ation (71707) '

The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the months of August and Septembe At the beginning of the inspaction period, the reactor was shut down with all fuel off loaded to the spent fuel poo ,

On September 4, the plant was placed in Mode 6 (refueling). Refueling ;

was completed and on September 14, the plant was placed in Mode 5 (cold ,

shutdown). The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected component Tours of the auxiliary, turbine, water treatment and service water 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 equipment in need of maintenance. The inspectors by observation and direct interview ;

verified that the physical security plan was being implemented in accordance with the station security pla The inspectors observed plant housekeeping and cleanliness conditions and verified implementation of radiation protection control Curing the months of August and September, the inspectors walked down accessible portions of the Safety Features Actuation, Service Water, Emergency Diesel Generator, Essential 120 Volt AC, Essential 4160 Volt AC, Essential 480 Volt AC, Essential 125 Volt DC and Component Cooling Water Systems to verify operabilit The licensee has identified 425 procedures which must be written or revised prior to restar At the end of the inspection period, 250 procedures nad been approved, which leaves 175 procedures to go through the approval cycle or to be written. In eddition, training must be accomplished en these procedures. Criticality is scheduled for early December 198 On September 14, 1988, Virginia Electric and Power Company shut down Surry Unit 1 af ter determining that both EDGs would be overloaded if a loss af offsite power occurred between five and sixty minutes after a LOC At the request of Region III, the inspectors verified that a similar condition does not exist at Davis-Bess _

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During the inspection period two operator license examiners from Region III accompanied operators from the Davis-Besse site to the B&W simulator. The purpose of the observation was to evaluate the operators'

ability to implement the Davis-Besse abnnrmal procedures and to evaluate !

the licensee's operator requalification program against NUREG-1021, Revision 5 Examirar Standards. While it is recognized that Revision 5 to this NUREG has not been published, copies of the applicable standard, '

ES-601, were provided to the licensee via letter from Geoffrey C. Wright, Chief, Operations Branch, Region III on March 18, 1988, and an updated '

copy was sent on June 23, 198 The impressions of the examiners were that to the extent practical, given the differences between the B&W simulator and Davis-Besse, the operators were capable of effectively using the abnormal procedures. On the other hand, the examiners determined that the licensee's progress towards ,

implementing the new requalification program related to simulator scenario development as defined in ES-601 of NUREG 1021, Revision 5, has been disappointingly littl While it is recognized that some of the aspects of ES-601 are new, many of the observations made by the examiners are applicable to proper conduct of exams regardless of type or vintag The following are discussions of some of the more significant observations made by the examiners, Performance of Evaluators During the conduct of the simulator exercises, the licensee provided only one official "evaluator" (a contractor) and two "observers." The observers had no official rola in evaluating the individual operators' responses to the scenarios. The evaluator was required to be the simulator operator, outside operator, instructor (at times), interpreter of operator actions, and evaluator. The varied tasks required of this individual s'esulted in a dilution of his prime responsibility to evaluate the actions of the operators. The lack of licensee participation during the scenarios would not be acceptable during an NRC administered examination and could have a negative impact on the evaluation of the licensee's progra To further complicate the evaluation process, the licensee had not provided detailed simulator scenario guides to either the contract evaluator or to their own observers. The only information available was a listing of the actions taken by the operators, as recorded by the contract evaluator, as they reacted to the scenario. The lack of specific guidance resulted in the evaluator not being able to objectively and consistently determine operator performance. The lack of evaluation criteria for both the crew and individuals also resulted % the post-scenario evaluation being very genera _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

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The examiners found no evidence that plant operations personnel were I involved in the review and development of the scenarios. Further,

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the evaluation summary sheets used by the evaluators will require significant upgrading to bring it into agreement with the Simulator Examination Summary Sheet endorsed by NUREG-102 Specific Comments on Simulator Scenarios The examiners observed that the same three scenarios were used for all crews during the requalification process. While NUREG-1021 does not explicitly prohibit this practice, it does speak to examination security. It is the examiners' opinion that the use of the same scenarios for all crews is not in keeping with the intent of maintaining exam securit The scenarios were devoid of detailed actions against which the operators' actions could be measured. No crew evaluation standard; were availabl The scenarios were inadequate in regards to the number and quality ofmalfunctions,componentfailures,andmajorplanttransient While the NRC, during this observation, did not make any changes to the scenarios the licensee should note that during the NRC administered exam the NRC may supplement or augment the scenario The licensee's scenarios were about 85 minutes long, whereas the NUREG suggests approximately 50 minute The licensee should work on shortening the scenario Given the differences between the B&W simulator and the Davis-Besse i facility, Lhe scenarios were reasonably realistic and comprehensiv The licensee did not identify Individual Simulator Critical Tasks (ISCTs) for each scenario. ES-601 discusses the identification of ISCTs which can be tied to facility Job Task Analyses (JTAs) or the NRC's knowledge / ability (K/A) catalog and should have an importance factor greater than or equal to 3.5.

The examiners observed that Technical Specifications were consulted when appropriate. Abnormal procedures and the emergency plan were

implemented to the extent practical given the differences between the simulator and the plant.

NUREG-1021 requires that scenarios contain at least one team dependent crew response and at least one time critical response.

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Neither criterion was addressed by the licensee in any of the prepared scenario l

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The licensee did use specific Job and Task Analyses as a basis for the examination development; however, there were no importance factors associated with the JTAs used in the requalification scenarios. The scenarios had in some cases two JTA numbers for the R0, eight JTA numbers for the SRO, and six JTA numbers for the STA, none of which could be cross referenced with the NUREG-1122 K/A catalog to determine that the importance factor of at least was being maintained as required by ES-60 The examiners concluded that the scenarios used during the requalification exercise did r.ot take into account any of the significant modifications currently being made to the plant. In the future the licensee should account for these types of change NUREG-1021 states that the scenario events should involve each crew member and should be composed of related or linked events. Due to the small operating area and close proximity of the bench panels, all crew members were involved in most events. While this arrangement accomplished some of the NUREG-1021 intent, the scenarios lacked individualism. The scenarios presented generally did not itemize the role assignments for crew members as required by the Standard For example: On a T-hot failure, the drill guide by the utility classifies most expected actions as "Expected Student Response" with the evaluator left to determine whose responsibility it is to take action. Not all of the scenarios reviewed contained related events and in one case the scenario consisted of four non-related event malfunctions associated with the Integrated Control Syste c. Emergency Operating Procedure Section 5 of Emergency Operating Procedure (EOP), "Loss of Subcooling Margin," has an immediate action to trip the Reactor Coolant Pumps (RCPs) as the first ste Included with this step is the Specific Rule 1 which, when implemented, per the second step of the procedure, would also trip the RCPs. However, the E0P in this case does not have "response not obtained" guidance in the event that the RCPs are not tripped. The RCP's trip criteria basis is that the RCPs will be tripped within two minutesper Chapter IV.A.pIV.A-2 of the B&W Technical Basis Document to preclude a high void fraction formation inside the RCS for certain size break .

The E0P should have a provision, as dD ineated per Chtpter IV.A, that if the RCPs are not tripped within two minutes of a loss uf subcooling margin (SCM) then at least one RCP in each loop should be operated. If RCPs were tripped beyond the two minute basis it potentially could cause core uncovery and inadequate core cooling.

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While it is recognized that there is an administrative procedure

that guides the operators to trip the RCPs on loss of SCM, there is no direct link between the administrative guideline and the E0P.

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In summary, the licensee's program for conducting the annual operating evaluation required by 10 CFR 55.59(a)(2)(ii) is inconsistent with the guidance provided in draft ES-601, Revision 5. The material used by the licensee in conducting its most recent examination would be unacceptable for use during an NRC administered exam. While NUREG 1021, Revision 5, which includes ES-601, is yet to be published the information contained in ES-601 was provided to the licensee training organization as an enclosure to a letter on March 18, 1988. An updated version of ES-601 was forwarded to the licensee on June 23, 19PA and further information was made available during a joint INP0, NUMARC, NRC meeting with all Region III licensees in July 1988. It was the NRC's expectation that each licensee would begin to incorporate the new guidance provided. It is apparent that in the area of simulator scenario development and exercise evaluation, even though both INP0 and NUMARC endorse the methodology, the licensee has made very little progress in implementing the proces These reviews and observathns were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedure No violations or deviations were identified in this are . Monthly Maintenance Observation (62703)

Station maintenance activities of safety related systems and components listed below were observed or reviewed to ascertain that they were conducted in accordance with approved procedures, regulatorv guides and

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industry codes or standards and in conformance with technic'al specification The following items were considered during this review: the limiting conditions for operation 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 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; radiolugical controls were implemented; and, fire prevention controls were implemente Work requests were reviewed to determine status of outstanding jobs

and to assure that pricrity is assigned to safety related equipment maintenance which may affect system performanc The following maintenance activities were observed or reviewed

• Installation and wiring of LVDT positten indicators for Atmospheric Vent Valves and eight Main Steam Safety Valves (FCR 86-201).

• Installation of venturis FE 6456 and 6457 (FCR 86-330).

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• ~ Low voltage circuit checks of Steam and Feedwater Line Rupture Control System actuation channel circuits for feedwater control valve * Reinstallation of a Hydramotor actuator for a Component Cooling Water (CCW) Room ventilation dampe * Preventive maintenance on a battery dischargo alar * Rer air of CCW pump bearin Following completion of maintenance on the CCW room ve W lation system, the inspectors verified +. hat the system had been retvinrj to service properl The inspectors review all Potential Conditions Adverse to quality Reports (PCAQRs). Recently, they have noted an increasing number of PCAQRs which have been written to document either missed quality control (QC) hold points or performing work outside the scope of or without a maintenance work order (HWO). Discussions with the licensee have revealed that most of the PCAQRs relating to missed QC hold points are attributable to the Faciiity Podification Department (FMD) and most of those relating to MWCs can be attributable to the Maintenance Department. The inspectors consider that those problems were caused by inattention to detail. It was determined that most of the incidents were discovered during HWO package closecut review rather than at the time of occurrence. The inspectors are concerned that these problems were discovered as a result of an historical review and not at the time of the incident. The inspectors have discussed their concerns with the license This is an cpan item (346/8A026-01(DRP)).

The licensee discovered that work had been performed to install pressure transducers in min steam safcty valves (SP 1787 and $P 1788) without a work order. PCAQRs 88-0779 and 88-078G were issued to document thi Investigation revealed that the vendor installed the equipment under the direction of two design engineers. In addition to installing the transducers without an MWn or a temporary mechanical modification ('IMM),

wires were jumpered without using jumper / lifted wire tags. This is the second incident involving work being directed or performed by engineering department personnel without benefit of an MWO (Inspection Report No. 50-346/87008). This is an unresolved item (50-346/88026-02).

During this inspection period, a Maintenance Team inspecticn was also performed and the results of that inspection will be documented in Inspection Report No. 50-346/8802 No violations or deviations were identified in this are _ _ _ _ _ _ _ _ _ _ _ -

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5. , Monthly Surveillance Observation (61726)

The inspectors observed technical specifications required surveillance testing on the Fire Protection System, ST 5016.05 (DB-FP-030 1 , "Three Year Fire Hose Station Hydrostatic Test" and verified that testing was performed in accordance witn adequate procedures, test instrumentation was calibrated, limiting conditions for operation were met, removal and restoration of the affected components were accomplished, test results conformed witt technical specifications and procedure reo.cirements and were reviewed / personnel other than the individual directing the test, and any defi acies identified during the testing were properly rev:ewed and resolved s sppropriate management personne The inspectors also witnessed portions of the following test activities:

• DB-MI-03201, "Channel Functional Test and Calibration of Steam and Feedwater Line Rupture Control System (SFRCS) ACH 1 Pressure Inputs"

• 08-PF-10002, "Emergency Diesel Generator Lube Oil System Test"

• DB-PF-10037, "Makeup Feed rd Bleed Mode 5 Test" e DB-PF-10049, "Startup Feedwater Pump Acceptance Test"

• DB-PF-10050, "Motor Driven Feed Pump System Flush and Acceptance Test." The inspector's review of the chronological log for this test revealed several documentation errors. These errors were discussed with licensee management; the importance of high quality documentation of test activities e s stressed. Chronological logs for other test activii.ies were of higher qualit The inspectors also observed high piping vibratian during testing apparently

aused by venturis installed as part of FCR 86-330. The inspectors initiae review of documentation for FCR 86-330 did not indicate that this vibration had been evaluated by the licensee. The presence of this vibraticn is an Unresolved Item (346/88026-03) pending further review by the inspectar * DB-PF-10058, "Auxiliary Feedwater (AFW) Level Control System Response Test"

• 08-PF-10059, "AFW Sy; tem Auxiliary Steam Test"

• DB-PF-10061, "SFRCS Acceptance Test"

• DB-PF-10076, "Makeup Feed and Bleed Hydrostatic Test"

• DB-SC-10075, "CR0 Rod Stop Circuit Test"

• ST 5016.02 (CB-FP-03002), "7 Day Electric Driven Fire Pump Test"

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• ST 5016.11 (DB-FP-03032), "Fire Barrier Penetration Seal Surveillance Test"

• ST 5051.04 (DB SP-03932), "ECCS Sub Systen Refueling Test" No violations or deviations were identified in this ar . Licensee Event Reports Followy (92700) Through direct observations, discussions with licensee personnel, and review of records, the following event reports weni reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specification LER 87-009, Liquid Radwaste Discharge Interlock Inoperable Due to Procedural Error. Radiation monitors RE 1878A and RE 1870B monitor

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the radioactive liquid release path for the Miscellaneous Liquid Radwaste Syste The monitors provide continuous radiation level indi*.ation, alarm on high radiation levels and provide automatic valve closure signals on high radiation levels. On March 20, 1987, Instrument end Control (I&C) technicians started calibrating RE 1878A in accordance with procedure IC 2005.04.12 "Process Rad Monitor Calibration," dated December 12, 1986. RE 1878A was recognized as being inoperable at this time and the technical specifications require only one operable radiation monitor in the effluent release path. RE 1878B was considered to be the operable radiation monitor. IC 2005.04 requires that electrical leads be lifted for the detector being calibrated; however, at the time of the event, 10 2005.04 did not. specify which of three possible sets of leads should be lifted. The I&C technicians lifted the wrong leads (the leads that prevented both RE's from sending a close signal to the isolation valves). The monitoring and alarm functions of RE 1878B were not affected by the lifted lead On April 5, 1987, surveillance test ST 5032.01 was being performed on RE 1878B to verify its operability. Since the leads were lifted, ST 5032.01 revealed that a high radiation input to RE 1878B would not cause the isclation valves to (. lose. The shift supervisor declared RE 18788 inoperable, verified compliance with Action 18 for technical specificaticn (TS) limiting condition for operation (LCO) 3.3.3.9 and directed I&C technicians to investigate. The incorrectly lifted leads were discovered, restored to the correct condition and RE 1878B was returned to operatio Further investigation by the licensee revealed that niae miscellaneous radwaste releases had occurred between March 20 and April 5, 1987, when both REs were inoperable. Se licensee's review of the event indicated that the concentratir of radioactive material in the nine liquid effluent "ses did not

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exceed regulatory limit During the event, the alann and monitoring functions of RE 18788 were operable and no abnormalities were note The inspectors verified that Procedure Development C-526, dated July 17, 1987, modified IC 2005.04.12 so that it provides adequate

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guidance on which wires should be lifted when calibrating RE 1878A, i RE 1878B, RE 1412. RE 1413, RE 1770A and RE 17708. This LER is close LER 88-016, Electrical Cir:.uit Bridging in Safety Related System I This condition was inspected by a Region III inspector and documented in Inspection Report No. 50-346/88031(ORS), Paragraph 4. This LER is a closed based on the results of that inspectio LER 88-019, Improper Crimps on Safety Related Electriccl Cable Lug l l

The licensee identified loose cable termination legs (size 4/0) in DC low voltage switchgear ('..estinghouse Electric Corporation Type W MCC). The switchgear was installed in the plant in 1972 and the lugs had been installed by the manufacturer prior to that. Forty I lugs had been installed and thirty-two wve toose. There was no evidence that any cable or equipment damage was caused by the loose  ;

lugs. Westinghouse, upon completing analysis of samples of the '

cables and lugs, concluded that che defective lugs would perform their intended function. However, the licensee's evaluation  !

conc!uded that, during a seismic event, the loose lues had the

, potential to prevent the starting of the Emergency Ulesel Generato t The licensee determined that no other lugs of this size had been installed at Davis-Besse by Westinghouse Electric Corporatio The licensee reported the loose lugs in accordance with 10 CFR 2 The inspectors verified that all forty lugs were replaced. This LER

, is close The following LER's were reviewed during the inspection period but  !

could not be closed.

l LER 88-017, Roving Fire Watch Tour Times Exceede This event was '

reviewed by Region III fire protection inspectors and will be i i

documented in Inspection Report No. 50-346/88028(DRS).  ;

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] LER 88-018, Inoperable Station Vent Radiation Monitor This event

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appears to be a violation uf Technical Specifications and the 1 inspectors' review of the event is continuing to ascertain the '

exact nature of the violation. This event had previously been identified as an Unresolved Item (346/88021-03).

LER 88-020, Monthly Test of Anticipatory Reactor Trip System and l

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Safety reatures Actuation System Used Some Inactive Logic Gate ,

i This event appears to be a violation of Technical Specifications i and the inspectors review of the event is continuing to ascertain the exact nature of the violation. This event had oreviously been identified as an Unresolved Item (346/88021-05).

No other violations or deviations were identified in this area

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7. Onsite Followup of Events (62702), (82201), (82206) and (9370_2,2 During the inspection period, the licensee experienced several events, some of which required p.ompt notification of the NRC pursuar.t to 10 CFR 50.72. The inspectors pursued the events onsite with licensee personnel. In each case, the inspectors verified that the notification was correct and timely, if appropriate, that the licensee was taking prompt and appropriato actions, that activities were conducted within regulatory requirements and that corrective actions would prevent future recurrence. The spesific events are as follows:

• August 12, 1988 During the triennial fire protection program audit, the licensee discovered that there are three manholes that have redundaltt trato safe shutdown equipment electrical cables that are not separated as required by 10 CFR 50, Appendix R, Section III.G.2. The m6nholes are normally bolted closed. The Region III fire protection specialist's review of this event will be documented in Inspection Report No. 50,346/88028. The licensee documented this event on Pctential Condition Adverse to Quality Report (PCAQR)88-063 On September 1, 1988, the inspectors photographed the interior of one of the three manholes and informed the contract electricians who were working on the manhole sump pump that the cables in the manhole appeared to be safety-related. Since it appeared that the cables were inadequately supnorted, the inspectors showed the photographs to the Shift Supervisor and the Electrical Superintendent and ,

discussed the cables with the On September 9, 1988, the manholes were opened again for inspection and the licensee made videotapes and took photographs of the interior of the manholes. About this time the inspectors discussed the electrical cables in the manholes with Operations Department management personne The condition was documented on September 19, 1988, when PCAQR 88-0732 was written by a member of the Technical Planning Department. It appears that the writing of PCAQR 88-0732 was untimely. However, this will remain an Unresolved Item (346/88026-04) pending the inspectors' review of the licensee's evaluation of the PCAQ * August 30, 1988 At 9:50 p.m. EDT, the motor operator (Limitorque) for auxiliary feedwater system valve AF 3869 failed during testing. This was the third Limitorque failure for this valve since 1985. Although the Limitorque failed, the cause of the failure appears to be the valv The licensee docum nted the event with PCAQR 88-067 The licensee removed the valv and sent it to Limitorque for analysi A new valve was installed and tested. The determination of the root cause of this failure is an Open Item (346/88026-05).

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• September 2, 1988 At 7:30 p.m. EDT, a small reactor coolant system (RCS) leak was found on the seal assembly for Reactor Coolant Pump 1-1. The RCS was being filled and was at ambient temperature and atmospheric pressure at this time. The leak was found by licensee personnel who were inside containment looking for any abnormal conditions during the RCS fill. Because of the location of the leak, all the leakage (estinated to have been 110 gallons) was directed to the containment normal sup. The leak occurred because a thermocouple had not been installed, leaving a small opening in the seal assembly.

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The reactor coolant fill was stopped, a pipe plug was installed in the opening and the reactor coolant fill was completed. Tne licensee

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documented this event with Potential Condition Adverse to Quality Report (PCAQR)88-068 This is an Open Item (346/88026-06) pending the inspectors' review of the licensee's corrective actio * September 27, 1988 During a QA audit of Technical Specifications, the licensee determined that an overcurrent relay (51V-2/DG) is not tested nor bypassed during a loss of voltage on the essential bus or on a safety features actuation system (SFAS) signal. Technical Specification 4.8.1.1.2.d.3.c requires that every 18 months, the licensee must verify that ali diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon the loss of voltage on the essential bus and/or on an SFAS test signal. Potential Condition Adverse to Quality Report (PCAQR)

No. 88-0768 was written to document the finding. The licensee declared both EDGs inoperat,le as a result of the PCA The licensee's engineering department evaluated the PCAQ and determined that Technical Specification 4.8.1.1.2.d did not apply because even though voltage controlled overcurrent relay 51 V-2/DG trips the EDG output circuit breaker AC 101 (AD101), it does not short circuit the generator field nor stop the diesel. Since neither the generator field is shorted nor the diesel stopped, the licensee considers that the functioning of this relay does not cause the EDG to tri The EDG's were declared operable after the engineering evaluation. The inspectors do not agree with this conclusion and believe that the opening of the generator output breaker is the same as tripping the EDG, because power is not supplied to the essential power bus (C1 or 01) and, therefore, Technical Specification 4.8.1.1.2.d is applicabl Regulatory Guide 1.108, "Periodic Testing of Diesel Generator Units used as Unsite Electric Power Systems at Nuclear Power Plants,"

Revision 1, dated August 1977, included the diesel generator

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. Revision 10 of the Final Safety Analysis Report (FSAR) dated December 1974, states that each EDG is equipped with mechanical

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prevent or limit equipuient damag The FSAR further states that loss of essential bus voltage or a LOCA will cause a bypass of all l EDG trips except engine overspeed and/or generator differential !

l relay action. It also states that each EDG is protected by a time i

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overcurrent relay (51V) which trips at higher overload or fault i currents. Revision 7 of the Updated Safety Analysis Report (USAR)

dated July 1988, Section 8.3.1.1.4 added overcurrent relay 51V-1

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(breaker trip only) as a device which is not bypassed. In addition, t the USAR now lists 51V, 51V-1, and 51V-2 as time overcurrent relay ;

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Prior to plant licensing in the early 1970's, the licensee was requested to provide additional information relating to its FSAR l submittal for the EDG's. The following is listed in the question and answer section of the FSAR: -

l Question 8. " Provide a discussion of the installed protective type devices that are incorporated in the design to protect the diesel generators from exceeding operating limits or otherwise prevent i them from performing their intended function during a DB '

What measures will be taken to minimize the possibility of

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the above devices from needlessly preventing the diesel from operating when required?"

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"(c) Voltage controlled, overcurrent relays (51V-1/DG) provide '

generator fault back-up protection as well as system fault [

back-up protection." l

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"The measure that is taken to minimize the possibility of '

the above devices needlessly preventing the diesel from operating when required as during a DBA is to reduce the j number of these devices bypassing selected ones." -

"During a non-emergency diesel generator operation (for example, on-line testing) each of the above mechanical and electrical protective devices is capable of initiating a diesel generator trip. However, during an emergency operation, namely on a loss of essential bus voltage or a 7 LOCA, controls limit the diesel generator trip to generator ,

differential relay action (87/DG) and engine overspeed (OTS)." i i

The inspectors consider that the activation of overcurrent relay !

51V-2/DG, which opens the generator output breaker, prevents the l EDG from performing its intended functio !

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The licensee still contends that the actuation of the overcurrent relays (51V-2/DG) does not trip the EDG and the relays do not have to be bypasse This is an unresolved item (346/88026-07(DRP)),

pending review of the licensee's interpretatio * September 29, 1988 The licensee determined that small break loss of coolant accident (LOCA) analyses performed by Babcock and Wilcox (B&W) assumed an initial power level of 100% instead of 102% as required by item I.A of Appendix K of 10 CFR 50. The small break LOCA analyses were included in B&W topical report BAW-10075A which was used to license the plant. The licensee documented this event with PCAQR 88-078 This is an Unresolved Item (346/88026-08) pending the inspectors'

review of the licensee's evaluation and corrective actio No violations or deviations were identified in this are . Review of Changes to QA Program Description (35001)

A Region III inspector met with the Director of Quality Assurance and members of his staff on August 31 and September 1, 1988, to discuss the changes included in the Revision 7 of Chapter 17.2 (Quality Assurance Program Description) of the Updated Safety Analysis Report. Seven items were discussed and the licensee agreed to change five items to reflect results of the discussion. Three reductions of previous commitments were involved in two of the five discussion items. The five items which will be re/ised and included in change submittal (Revision 8) planned for early 0;tober 1988 are listed below: Page 17.2-3, Fifth paragraph (Organization) Page 17.2-15, Section 17.2.2.5 (Program review) Page 17.2-37, Section 17.2.12.3 (Calibration accuracy) Page 17.2-42, Section 17.2.15.1 (Nonconformances) Page 17.2-53, Table 17.2-1 Item 4 (ANS 3.2 commitment)

These changes will be reviewed by the NRC during the review of the Revision 8, USAR submittal scheduled for October 1988. Based on review, discussions and agreements regarding the above five items, the Quality Assurance Program meets the requirements of 10 CFR 50, Appendix B, and therefore is acceptabl?.

No violations or deviations were identified in this are ,

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. Enforcement Conference j On September 26, 1988, the Vice President, Nuclear and members of his staff met with the Regional Administrator in the Region III offices for an enforcement conference. Members of the Region III and headquarters staff also attended. The enforcement conference was related to an employment discrimination finding which is documented in Inspection Report No. 50-346/8803 . Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or

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deviations. Unresolved items disclosed during the inspection are

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discussed in Paragraphs 4, 5, and ,

11. 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. Open items disclosed during the inspection are discussed in Paragraphs 4 and . Exit Interview (30703)  ;

The inspectors met with licensee representatives (denoted in Paragraph 1) f throughout the month and at the conclusion of the inspection and '

summarized the scope and findings of the inspection activities. The  ;

j findings related to the requalification training at the B&W simulator i

! (Paragraph 3) were discussed telephonically with the licensee's training manager on October 28, 198 The licensee acknowledged the finding t

After discussions with the licensee, the inspectors have determined  :

i there is no proprietary data contained in this inspection repor i

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