IR 05000237/1990003

Insp Repts 50-237/90-03 & 50-249/90-03 on 900106-0220.No Violations or Deviations Noted.Major Areas Inspected:Lers, Plant Operations,Maint & Surveillances,Safety Assessment/ Quality Verification & fitness-for-duty Training

ML17202G818
Person / Time
Site:	Dresden
Issue date:	02/27/1990
From:	Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML17202G817	List: IR 05000237/1990003 ML17202G816
References
50-237-90-03, 50-249-90-03, NUDOCS 9003120285
Download: ML17202G818 (19)

Text

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U. S.. NUCLEAR.REGULATORY COMMISSION REGION III *

Repor.t Nos. 50-237 /90003( DRP); * 50-249/90003( DRP)

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Docket Nos. 50-237; 50-249. *

• License Nos. DPR-19; DPR-25

• Licensee:

Commonwealth Edison Compan P. o. *sox 767 Chi ca.go, IL 60690 Facility Name:

D*resden Nuclear Power Station, Units 2 and 3 Inspection At: *Dresden* Site, Morris, IL Inspection.Co~ducted: Janua*ry 6 through February 20, 1990

'

Inspectors:

S. G. Du.Pont *

D.. E. Hills

• • D. E. Jones Approved B

. FEB 2 7 1990 Date

Inspection Summary Inspection during the period" of January 6 through February 20, 1990 (Report. Nos. 50-237/90003(DRP); 50~249/90003(DRP)).

Areas Inspected: Routine unannounced reside.nt inspection of licensee event reports, plant operations, maintenance and surveillances, safety assessm~nt/quality ver.ification aod fitness for duty training (TI2515/104).

Results:

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.No violations.or deviations were i~entified during the inspection period..

• Numerous ma i ntenarice af'.ld surveil 1 ance activities* were observed with

.only mi~or problems identified as described i.n paragraph The.planned teh*week refu~lin~ outa~e fo~ Unit 3 was completed only one day behind schedule although several unforeseen major activiti~s had to be added to the sco~e of work: *These incl~ded such items as ins~ect~on ahd/or work on the High P~e~sure Coolant Injection (HPCI)

valves for both ~nits, additional local le~k rate testing volumes and rep 1 acement of the Unit-2 reserve auxiliary transforme The inspect~rs noted that the outage activities were well planned, staged, ahd executed..

t':'. 90022 ~1~028~

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,~~7 9UU5*-A*; 11-11-.~. h~OO~~~.

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The licensee. took conservative actions in regard to safety dealing with identificatio~ and resolution of a recirculation pump inboard seal leak and an inner seal cooler leak as described in paragraph 4~a.(4).

• The inspectors identified conditions allowed in*or~sden procedures for which the single failure analysis for a turbine pressure regulator failure described in the "Final Safety Analysis Report*

{FSAR) would be incorrect. The licensee is performing a 10 CFR 50.59 safety evaluation to determine whether an unreviewed safety question exists as described in paragraph 3.a *

• • DETAILS Persons Contacted Commonwealth Edison Company

~E. Eenigenburg, Station Manager

• L; Gerner, Technical Superintendent E. Mantel, Services Director

• J. Kotowski, Production Superintendent

• D. Van Pelt, Assistant Superintendent - Maintenance J. Achterberg, Assistant Superintenden~ - Work Planning G. Smith, Assistant Superintendent-Operations

• K. Peterman, Regulatory:Assurance Superv~sor C. Allen, Administrative Service Superintendent W. Pietryga, Operating Engineer M. Korchynsky, Operating Engineer B. Zank, Operating Engineer J. Williams, Operating Engineer

• M. Strait, Tech~ical Staff Supervisor L*. Johnson, Quality Control Supervisor J. Mayer, Station Security Administrator

• D. Morey, Chemistry Services Supervisor D. Saccomando, Health Physics Services Supervisor E. Netzel, Quality Assurance Superintendent

• R. Janecek, Quality Assurance/Nuclear Safety Offsite Review

• K. Kociuba, Quality*Assurance Superintendent

• S. Stiles, Training Supervisor

• *

• T. Lewis, Regulatory Assurarice Staff

• G. Bergan, Onsite Nuclear Safety The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shift engineers and foremen, electrical, mechanical and instrument personnel, and contract security personne *Denotes those attending one or mor~ exit interviews conducted informally at various times thrbughout the inspection perio.

Licensee Event R~ports Followup (90712 and 92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine*

that reportability requirements were fulfilled, immediate corrective actions were accomplished, and corrective actions to prevent recurrence had been accomplished in accordance with Technical Specification (Closed). LER 237/89028-01:

Containment Cooling Service Water (CCSW)

Bay Level Reduction Due t6 Intake Structure Flow Blockag This event and initial licensee actions were described in inspection reports 50-237/89b19; 50-249/89018 and 50~237/89022; 50-249/8902 Additional long term corrective actions proposed by the licensee included evaluation

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. of a modification that would automatically measure the water level drop across the. trash bars. A temporary system was proposed to use.a chair and buoy arrangement to indicate low water level in the CCSW suction bay until implementation of the permanent modificatio In addition, the operators' daily round book was to.be revised to include verification of water level in the CCSW suction ba Finally, this information was to be reviewed for inclusion into the operator training p~ogram *.

(Closed) LER 237/89031:

Additional Volumes Added to Type Band.c Local Leak Rate Testing (LLRT) Program Due to Self-Assessment Initiative. This event ahd corresponding licensee actions were described in inspection report 50-237 /89026;. 50-249 /8902 *

(Closed) LER 237/89032:

LLRT As Found Limit Exceeded Due to Excessive Leakage _From the Unit 2 Drywell Personnel Interlock. This event and corresponding licehsee actioni were described in inspection report 50-237/89026; 50-249/8902 (Closed) LER 237/90002:

Unit 2 Primary Containment Group I isolation and reactor scram due to procedure deficiency during performance of main steam line high flow instrument surveillance on January 5, 199 The details of scram followup and corrective actions are documented in paragraph 4.b.(l) of this report. Additionally, discussions of the effecttveness of the licensee's onsite review and root cause *

determination are also contained within paragraph 4.b.(l).

(tlosed) LER 249/89009:

LLRT As Found Lfmit Exceeded Due to Excessive Leakage From Primary Containment Valve This event and corresponding licensee actions Mere described in inspection report 50-237/89026; 50-249 /89025 *.

(Closed) LER 249/89010:

Imprbper Stationing of Fire Inspections Due to Personnel Error. This event and corresponding licensee actions were described in inspection report 50-237/89022; 50-249/8902 * *

(Closed) LER 249/890li:

Unplanned Primary Containment Group II and Group III Isolati-ons Due to a Labelling Deficiency. This event and corresponding licensee actions were described in inspection reports 50-237/89026; 50-249/8902 No violations or deviations were identified in this are. * Plant Operations (60710 and 7i707)

a.*.

Operational*Activities The inspectbrs -Observed control room operations, reviewed applicable logs and conducted discussions with control room operators during this perio The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper *.

return to service of aff~cted component Tours of Units 2 and 3 reactor buildings and turbine buildings were conducted to tibserve plant equipment conditions, includfo*g potentia 1 fire hazards, fluid

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leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenanc The inspectors, by observation and direct interview, verified that

.'the physical security plan was being implemented in accordance with the stati.on security plan. This in'cluded verification that the

.appropriate.number of security personnel were on site; access control barriers were operational; protected areas were well maintained; and vital area barriers were well maintaine The inspectors verified that the licensees' radiological protection program was implemented in accordance with facil i,ty policies and programs and.was in compliance with regulatory requirement The inspectors *reviewed new procedures and changes to procedures that were implemented during the inspection perio The review consisted of a verification for accuracy, correctness and compliance with regulatory requirement * These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR and administrative procedure On February 6, 1990, the inspectors identified conditions allowed in Dresden procedures for which the single failure analysis for a turbine pressure regulator failure described in the FSAR ~ould be

. incorrect. Section 11.2.3.2 of the FSAR indicated that a regulator failure in the wide open direction would result in a 100 psi vessel pressure drop in the first ten seconds resulting in a Main Steam Isolation Valve (MSIV) closure at less than 850 psi reactor

pressure. A scram would result from the MSIV closure and depressurization would be stopp~d due to the isolation *. However, with reactor*water level initially near the top of the range, allowed

. by the operating procedures, the reactor water level swell due to the single failure would cause a.turbine trip 6n high reactor water level prior to reaching 850 psi reactor p~essu~e. In the condition where reactor power was greater than 40%, the reactor would scram due to the turbine t_ri The MSIV automatic closure on low reactor pressure was bypassed when the mode switch was not in the RUN position. If the control room operator immediately placed the mode switch to the shutdown position following the scram in

• acc.ordance with instructions in the abnormal operating procedures, the MSIV closure would not occur at 850 psi. The FSAR analysis did not account for the possible turbine trip if initial reactor water level were assumed to be near the top of the allowed operating range. *It was not clear how this would affect parameters associated with, and iri c~mparison to.the plant licensing basis. The licensee plans to perform a 10 CFR 50.59 safety evaluation to determine whether an unreviewed safety question exists. Completion of the safety evaluation is considered an open itern (50-237/90003-0l(DRP)).

...

b.: *.Refueling A~tiviti-es

• On January~' 1989, while r~ar*g?niz*ing fue*l assemblies in the Unit 3

• fuel-pooJ, a,minor interference was encountered while lowering fuel

. * assembly X3P089 into storage rack posH ion E2;...M The fuel handler

• attempted to manually rotate the ~ain hoi~t to maneuver the fuel

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• bundle.. This caused.the fuel assemblytodrop approximately six

--inches result'ing in' it being wedged in the storage rack at a slight angl~. The licensee de-thanneled the fuel assembly and inspected for damage with an underwater video camera.. The only tndication of the inciden*t was a shiny spot and some scratches on the. channel

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._iJs.el No d*amage was found*on the fuel assembly internal to th channel.. An i n*spect iOn of the storage rack did not identify any.

related damag Priof_to the inspection, the ~i~ensee had chariged

_the refueling plans due.to the incident such that this.peripheral fuel assembly was 'replaced and would not be used in the upcoming, cycle~

Further review indicated that the hoist cable, which*pro~ided ver.ti-cal motion to the main hoist, had fallen off the main hoi.st

.. drum and was -slung over the smalier diameter motor shaf Thfs was caused by an inherent aesign problem dealing with the calibration of the. load cell~ The function of the loa.d eel l was to stop the main hoist drum when there was less than* a 50 pound load on the hoist cable and to energize the slack ~able light on the r~fueling platform control consol This.would allow the grapple to be raised b_ut not lowere The hoist cable was normally wound onto the main hoist drum i~ a helical. pattetn; Groves.in the main hoist dru~

provided a gujde for the *hoist cabl In"this event, the' load cell apparently allowed enough slack cable by not stopping the main* hoist drum quickly enough which caused* the hoist cable to jump groves and migrate to the end of the main hoi.st' drum.. A maintenan,ce histqry

• • review.did not indicate that this was a recurring proble However, the de~ign of the s~stem*was such that the calibtati~n ~f the load *

ce.11.was on.ly optimal for the particular grappling elevation for which it was calibrated. Therefore, if it was calibrated for use in the reactor core~ it would not be set.exactly right for the fuel poo The li~ens~e.planned to evaluate the ~nstallatidn of a

~estraining mechariism above t~e**hoist cable on. the ~ain hoist d~um to ensure that the hoist cable wo.uld not be. able to jump groves or mi~rate to the end of the. main hoist drum..

The licensee also planned.to.review industry designs for possible replacement or upgrade of the.entire full gra~ple includi~g the cell. This last action was also.in response to.recurring maintenance problems encountered wi~h.the_fuel grapple itsel *

Fuel* loading for Unit 3 waS'.begun on January*10, 1990, but was suspended l at'er -that same day due to concerns *abcfut Source Range Monitor ~SRM) respons The.. control room._operator's and th inspector observ~d.SRM s.hort p~riod alarms and SRM 21 spiking high, Pre-fuel load. SRM surveillances reviewed by the inspector did not indicate ~ny problems.with.the SRM Further revieW,indicated that

~portion of t~e ~able for'SRM 21.was not sh~elded which allowed noise to enter the process *signal:

This_ cable had*been replaced

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- during the out~ge prior to fue1

~oading.. The amount*of noise varied in ~ccordance with activities ar6und -0r over the vessel.. The

.licehse~*~dded additio~al shielding (to the small portio~ of the cable that ha_d* no shielding) a*nd performed a connector inspection'

• to correct the problem.,

Followfog caple repair, SRM 21 failep downscale due to a malfunctidning pulse height discriminate~ car Since no properly qualified replacement part.was available, this

• part was obtained from SRM 24 on Unit 2~

The card.was installed on Unit 3 and SRM 21 calibration/functional test was complete Proper SRM response was *again verified for all four SRMs by moving the appropriate fuel bund*le around-each detector andfuel loading resumed on January 11, 199 *

The inspectors verified impl~mentation of administrative controls prescribed by. Dresden Fuel Procedure (DFP) 800-1,.Master Re*fuel ing Procedure, and other associated ~efueling and operating surveillance

• procedures, during fuel movements *by. review of appropriate completed checklists, logs and surveillances*, direct observation, :personal inter~ie~s and verification that technical specification require-ments for refueling_were me The inspectors verified thit key

~ersonnel possessed an adequate understanding of their individual responsibilities and administrative reQuirements through direct obse~vation and personal intervie~s.. Adequate stafftng for refueling activities, appropriate r~diation pro~ection controls and adequate plant cleanliness conditions we.re also verified by the inspector5~

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No violations or d~viat~pn~'we~e identified in t~is are.

Maintenance and Surveillances (62703, 61726, and 93702) Maint~nance Activities Station maintenance activities of systems and components listed below were observed or reviewed to* ascertain that they were c:onducted in accordance with approved procedures, regulatory guides and in.dustry codes or standards and in conformance with Technical Specification Th,e following items were considered durin~ this review:

the Li~iting. Conditions ftir Op~ration (LCOs) were met while*

components or systems were.removep from service; approvals were obtained prior to initiating the work; activities were ac~o~~lished using a~proyed pro~edures and were inspected as applicable; functional testing.and/or c~librations 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 imple~ent~d. Work req~ests were reviewed to determine status bf outsfa~ding.jobs ~nd to assure that priority is assigned to safety related equipment.maintenance which.may affect system performance.*

Portions of the following maintenance items were observed or reviewed during the in$~ectiori period:

Replacement of the Unit.3 HPCI ~oom Cooler Repair of Damaged Electrical Cables' in Cable Trays 785T and 785.

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Intermediate Ran~e* MohitOr Cable Adjustment The inspectors also ~itnessed or reviewed various aspect5 of the following o~currenc~~=

(1) *On January 16, 1990, with Unit 2 at.100% rated thermal power, a

~eactor.scram occurred in conjunction with various equipment malfunction A faulted motor caused a trip of condensate pump 20 whic~ left only two condensat~ pumps runnin The standby tondensate pump either failed to automatically start 6r did not

. start in sufficient time to prevent feedwater.Pumps 2A and 2C from tripping on low suction pressure~ Although feedwater pump 28 automatically started, the reactor scrammed on low leve The operators a]so manually scrammed the reactor.at about the same time as the automatic s~cra Groups II and III primary containment *i so 1 at ions. a 1 so occurred due to 1 ow reactor water leve The operators restarted feedwater pump A and a main turbine trip resulted from high reactor water leve Following an unexpected trip of reserve aux-11 iary transformer TR 22, busses 22 arid 24 automatically transferred to unit auxiliary transformer TR 21 as designe During normal operation Unit 2 loads are usually shared between the onsite power source, TR 21, and the offsite power source, TR 2 Once the main turbine had coasted down, the *operators opened the *ma in generator output breakers to prevent generator motorin The operators manually started the Unft 2 diesel generator and the swing diesel generator automatically started on undervoltage and soon after the 2C outboard MSIV drifted close The operators manually isolated the,MSIVs to conserve inventory since reactor water level was again decreasing and manually controlled pressure at various times with the is.elation *

condenser, elettromatic relief*val~es and the HPC An Unusual Event was d~clared. Reactor pressur~ was reduced sufficiently t6 initiat~ shutdown cooling and cold shutdown was reached on January 17, 199 Unit 3 power was cross-tied to Unit 2 and the Unusual Event was terminated on January 17, 199 *

Various equipment malfu~ctions were identified including the following:. condensate,pump 20 motor burnup, condensate pump 28 possible failure to automatically start, reserve auxiliary transformer unexpected trip, premature closure of main steam isolation valve 2C,

~lectromatic relief valve 28 control room switch burnup, possible failure of main g~nerator output breakers to automatically open, and.failure of 28 shutdown cooling pump discharge valve to open remotely.* An NR Augmented Inspection Team responded to the event and issued inspection report 50-237/90004 which describes the event in

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{2)

(3)

(4)

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. detail including th~ root cause of the equipment failure On January 18, 1990, while observing licensee investigation and repair of the MSIV 2C DC 'pilot solenoid in accordance with work request D89930, the inspector noted a labeling discrepancy for terminal blocks on control room panel 902-While attempting to take a voltage check across th~ DC pilot solenoid coil on terminal points FF60 and FF61, the electrical technician instead tried to t~ke readings across the corresponding numbered points on terminal block E When the expected voltage was.not observed, the instrument technician realized which was the correct terminal block and continued in*

accordance with instructions. The electrical technician had located the incorrect terminal block by labels mounted on the control panel fram The EE and FF terminal block labels on the frame had been reverse L~bels internal to the pahel above each terminal block were correct. However, these other labels were much more-difficult to identify due to various visual obstructions in t~e -pane The licensee correctly

• re-labeled the terminal blocks and reviewed the other similar terminal block labels.in the control room panels for discrepancies. A total of two additional discrepancies were identified and corrected during.this revie On January 18, 1990, while observing activities in the Unit 2 X-area {steam tunnel) the inspector noted air leaking from a valve fitting on an instrument air line to the ~utboard MSI accumulator; A work request identification card in that location indicated that the leak had been identified in

. November 1988. A licensee search of records could not find a work request dealing with this air leak. *It appeared that a work request identification card had been written with no *

corresponding initiation of a work reque~t. However, the air leak was on a non-safety related instrument air line and thus the.long lead time in correcting the leak was not a safety conce~n. The licensee repaired the leak. after being notified by the inspector*.

A* previous occurrence of a possible work request identification card without a corresponding work request was described in unresolved item {50-237/89022-01) *

. As a result, the inspectors have increased observations in thi~ area of concer *

During the Unit 3 startup from the refueling outage on

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February 11, 1990, recirculation pump 38 inboard and outboard seal cavity pressures were noted to both be reading 940 psig indicating a possible inboard seal leak. - Drywell equipment drain sump leakage was approximately 2 gpm, well below technical specification limits. The seal had been rebuilt and successfully hydro tested for norma 1 preventative maintenance.

during the refueling outage. A subsequent drywell entry revealing excessive ~ater in the normal controlled leakoff sight glass and the pump inboard seal cavity high flow alarm

• relay contacts being closed further indicated a seal failur '-' '

As a result, Unit 3 was shutdown later that same,day for repair of the seal.* The seal assembly was removed and tested again revealing excessiye flow and equalized pressures in the seal cavities. *The seal was disassembled and a liquid penetrant inspection of ~he hardface surfaces and shaft sleeve were

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performe.No discrepancies to the internal components were note The seal assembly was rebuilt and successfully hydro teste *

While working on the recirculation pump 38 seal problem, reactor building closed cooling water {R8CCW) was isolated from the drywelL. This resulted in a water level increase in the RBCCW head.tank, indicating.a possible recirculation pump seal cooler leak. Additional testing located the leak in the recirculation pump 38 inner coole Due to the-substantial time inv.olved in replacing the.inner cooler cover, a temporary al:teration was instead performed which involved rendering the inner cooler inoperable*.

R8CCW flow to the inner cooler was isolated and.reactor water leakage into the cooler was hard piped to the drywell equipment drain sump such that it would constitute identified leakage. Dresden technical specifica-tions allowed 25 gpm drywell leakage of which up to five gpm could be unidentified. Redundant pressure switches were

• installed on the new leakage line to provide an a la rm in the control room at 75 psig (equivalent leakage flow of 15 to 17 gpm).

This was to ~nsure that the cooler internal (R8CC side) design pressure tif 150 psig would not be exceede (The cooler external pressure during normal operation is * *

approximately 1000 psig.) A *temperature sensor was also ins ta 1 led to prov'ide redundant means of detecting leakag Additional vibration monitoring capabilities were also added in 1 ight of genera 1 industry concerns on redrcul~tion pump shaft cracking. The licensee had not yet decided exactly how the vibration monitoring capability would be used. Appropriate procedure changes were.made to account.for operation with the temporary alteration and training was given to each operating shift. These changes provided specific administrative controls regarding actions to be taken *upon indicatiori of excessive..

leakage into and/or pressurization of the inner seal cooler. A safety evaluation was performed in accordance with 10 CFR 50.59, which indicated operation in this condition did not constitute an unreviewed safety question. Unit.3 startup was begun again *

on February 18, 1990. A drywell entry made when operatin pressure was rea*ched indicated no leakage ;n the cooler leakoff sight glass. However, a seal cavity high flow alarm was.

received at about 700 psig ~eactor pressure indicating that some seal leakage still" existed. Seal cavity pressures were as expected *for normal operation.*.

The inspectors regarded licensee actlons *with respect to the sea 1 leak to be conservative with respect to safety. Although indications were that the outer seal was still functioning properly and conditions were still far from approaching

.. technical.specification.limits, the unit was shutdown to repair

. the sea 1 assem~ ly'.

(5)

Upon rolling the main turbine during the Unit 3 startup on February 20, 1990, condenser vacuum was noted to be decreasin The licensee began inserting rods and reduced power and reactor pressur~ to preserve vacu~ Aft~~ considerable investigation, the licensee found that charcoal absorber bypass valve A0-3-5418 closed indicator light was burned outr It appeared from the indication that the valve was fully open~ but the valve was in a throttled position and almost completely close It was not easily identifiable that the valve position was

. blocking system flow. After rectifying this problem, addi-tional condenser vacuum problems became apparent. A blown *

rupture disk and leaking relief valve were discovered on the recombiner train.Bpreheater inlet. The licensee was still investigating ~ondenser vacuum probl~ms at the end of the inspection perio Surveillance Activities The inspectors observed surveillance testing, includi~g required Technical Specification surveillance testing, and verified for actual activities observed that testing was performed in accordance with adequate procedures. The inspectors also verified that test instrumentation was calibrated, that Limiting Conditions for Operation.were met, that removal and restoration of the affected components were accomplished and that test results conformed with Technical Specification and procedure requirement Addition~lly, the inspectors ensured that the test results 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 persor.ine The inspectors witnessed or reviewed portions of the following test activities:

Main Generator Reverse Power Special Test Control Rod Drive Friction Testing

.Quarterly Low Pressure Coolant Injection (LPC.I) System Pump Operability' with Torus Available for Inse.rvke Testing (IST)

Prog~m*

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HPCI Motor Operated Valve Operability Verification

• HPCI System Operabi 1 ity Verification

• Bus Undervo1tage and Emergency Core Cooling System Integrated Functional.Tests for Unit 3 and Unit 2/3 Diesel Generators Quarterly Valve Timin~

Drywell Closeout Inspection The* inspectors also witnessed or reviewed various aspects of the*

following occurrences:

(1)

On January 5, 1990, while Unit 2 was operating at 100 percent power, a primary containment isolation and subsequent reactor scram occurred during performance of the monthly main steam line high flow isolation switch calibration and surveillance

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test.: The ~esultin~ group 1 isolation signal initiated the automatic closure of the*main steam isolation and drain valves, and an automatic reactor scra Subsequent groups 2 and 3 isolation signals arid" automatic start of the standby' gas treatment system* occurred as designed on a low reactor vessel

water level following the scra The Nucle6r Station Operator (NSO) took all of the required immediate actions and subsequently manually initiated pressure control with the

• i~olation condense~ (IC). :The IC was operated using clean demineralized*water for make up' of the shell side of the I *The IC was secured once the vessel pressure *stabilized and.

the reactor water cle(lnup system adequately pr~vided decay heat remova *

The inspector independently.-vertfied that all required actions were taken, that all required system responses occurred and that site contamination associated with isolation condenser-0peration*wtth unclean ~emineralized water sources did not occu The licensee.. conducted a. compr~hehsive onsite review and root cause determination. the inspector independently observed these. meetings and determined that. they were effective. The onsite review evaluated the eveht and found that the surveill-*

ance procedure.(DIS 250-1) provided inadequate precautionary controls to prevent a potential pressure transient in the common instrument sensing li.ne header. **. * The procedure. was also inade.quate in providing instructions for vent and filling the instrument sensing line during installatio~ of test gauges *. These resulted in trapped air in the instrument line causing a pres~ure transient when the instrument isolation valve was opened. Since the common instrument sensing line contained four flow switches,.'two per channels A and B, the

• pressure transient resulted in the completion of.the isolation logic for bot~ chanpels a.n.d. th_e subsequent closure of the main steam i~olati6n v6lves (group 1 isblation).*.

The onsite r~view a*lso determined that a similar group 1

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. isolation occurred on Unit 3 in 1987. 'In this previous case, the ro6t cause was.attributed to an apparent valving erro Since this root cause was corrected, the onsite review concluded that correcting the valving.error alone would not prevent possible recurrence. Licensee corrective actions to prevent recurrence-included procedure revision, training, installation of pressure gauges interna 1 to the sensing line.

to verify the* lack of trapped air, and revision of the venting method to ensure positive, venting during surveH lance testing. *

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The inspector's review of these corrective actions found them to be conserva~ive in manner and demonstrated a.commitment to

. :resolution of technical problems:

C2)

On January*24, 1990, an une.xpected start of the Unit 3 Diesel Generator occurred during fnstallation of test equipment in accordanc~ with Special Test (SP) 90-1-39, Bus Undervoltage and ECCS Integrated Functional Test For Unit 3 Diesel Generator. -

• As a prerequisite delineat~d in.the test procedure~ a mobile*

electrical.technician.was installirig an eight channel strip chart* recorder that was to assist in determinatio.n of proper load sequenc~s durin~ testing. While connecting electrical leads to the. appropriate termina 1-posts on the re lay specified in the procedure the techni.cian noticed a spar To ensure that a fuse had not been. blown, the technician used a voltage*

tester {wiggy) to ensure power was still available to the relay by connecting it across various combinations of terminal posts

-_on the rela The technician was not aware that the portion of circuitry dealfng with the particular terminal posts to which the recorder had 'been connected would not have been energized under the current conditions. Thus, the trouble-shooting performed by the technician_ could not have realistically

provided any useful informatio!'). A relay was picked up when the technician connected the wiggy to positive and negative terminal posts to compl~te the circuit~ The current draw in the.circuit was enough to actuate the auto-start circuitry for the diesel generator but not enough to actuate the larger amperage trip ~oils and breakers. Thus, the diesel generator automatically started but the trip breakers did not open between Bus 34 and* ESF Bus 34-L The licensee inspected the break~r cub.icles and performed mechanical testing to ensure.

co~rect functioning of the trip breakers. Further review

-- indicated that wiggys are commonly used at Dresden to ensure

. circuits are dead prio~ to-working On them.* However, using a wiggy on a circu.tt beHeved to be live is an inadvisable practice due to the low DC resfstance.of this instrument compared to other available instruments such as a Fluk The licensee_presented this disti~ction to the electrical techni~ians durin~ a subsequent _tailgate session.*.Further review indicated that instructions on how to use a voltage tester such a*s a wiggy including proper applieations were *

included iri the initial trainirig program for both plant electrici~ns and mobile electricians at the Production Training Center. _This training did not, however, cover possible imprope~ uses: The Jice~see planned to add this training bn

• improper uses to-the continuing training program for plant electricians *. The licensee also planned.to request the Production Training Center to*include specific training on the proper use of this type of voltage.tester* into training for electricians. *In this way, the issue would be addressed with

. respect' to mobile ~lectricians. Finally, the mairitenance staff was to review the possible purchase of high resistance leads for the voltag*e. testers.*

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N6 ~iolatioris or deviatirins were identified in this are,,

5..

Security (TI2515/104)

Fitness for Duty Inspection of Training Program The.inspector observed the Fitness for Duty (FFD) training session presented.du~ing Nuclear General Employee Training (NGET) on January 31, 199 Thfs consisted of combined training for both general employee policy a~areness and escorts lasting about one and a half hour The training consisted of three video. tapes and a short question and answer s~~sion~ The inspector noted that actual drug equipment displays

~ere not utilized in the presentatio The inspector also verified that questions dn FFD were included in the NGET examination administered to the training ~e~sion attendee The inspector verified that the following items were generally addressed in the presentation:

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0 Licensee policy and procedures, including methods that were used to implement the pdlic The personal and public health and safety hazards associated with abuse of drugs and misuse of alcohcil. *

Th~ effect of prescripti6n drugs and over-the-counter. drugs and

• dietary conditions on job performance and chemical test results and the role of.the Medical ~eview Office *

Employee assistance programs provided by the licensee. *(Thi was brie*fly mentione A more detailed description of* this

. program ~as provided in new employ~e.orientation train{ng.)

What is expected of employees and what consequences may result from lack of ~dherence to.the policy..

Techniques for recognizing aberrant behavio The procedures for r.eport i ng problems to. supervisory or security personne However, the inspector noted that various other items were not included in the training or were not as detailed as would.be expected in regard to FFD training for escor.ts.. *These items were forwarded to regional safeguards specialists for future* followu The inspector also observed the conducted on February 15, 199 of l~ctures but also included a equipment displ~~ was utilize questions and answet FFD training.session for supervisors This one day session consisted mainly videotape and role playing. A drug

• Sub~tahtial time was also allowed for The inspector verified that the following items were genera~ly addressed in the presentation:*

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Supervisor role and responsibilities in implementing the progra The roles and responsibilities of others, such as the personnel, medical and employee assistance program staff~.

Techniques f~r-~ecognizing drugs and indications of the use, sale or possession of drug *

Behavioral observation techniques for detecting degredation*

in pe~formance, impairment or changes in employee behavio Procedures fdr initiating appropriate corrective action, including referral to the Employee Assistance Progra No violations dr deviations were identified in this area.*

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Safety Assessme~t/Quality Verification (40500)

The inspectors performed an-evaluation of the licensee's self-assessment capabilit T~~ e*::~~.:.:.~:::-: i:-:*:cl*:d review of the li.censee's Offsite Quality Assurance (QA) audits, team assessments and both the onsite review committee and the Onsite Nuclear Safety Group (ONSG) functions; Based upon this review, the inspectors concluded that the self-assessment function was.being adequately implemented by the license The offsite QA audits and assessments*were scheduled based upon audit and assessment experience in the previous year, to effectively target problem*

a~eas. Results, including corrective actions, were tracked utilizing the Nuclear Tracking System (NTS).

Regulatory assurance issued a daily

• activities list which listed overdue responses as.well as cu~rent and

• future action item A department head. meeting is held once a week,

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segment of* which is devoted to the review* of the s_tati.Js of regulatory..

activitie *

The inspectors reviewed the 'following: Dresden off-site audit report number 12-90~1v, which reviewed activities and docurilentatfon associated with station Quality Control (QC), Engineering and Construction (ENC) and substation construction.

receipt inspections and the qualifications of receipt inspection

personnel, dated January 19, 199 * Dresden Dperations Assessment, June 19-23, 1989, and response dated

. September 14, 1989. * Dresden Station On~the-Job Training (OJT) Process A~sessment, November 6-8,. 198 Dresden Primary Containment L~ak Testing Program Assessment, May 2, 1989 and response dated June 9; 19~ Dresden Station Regulatory Activities List for.Tuesday, February 7, 199 *

The Onsite N~clear Safety Group functioned as a NUREG 0737 Independent Safety Engineering Group (ISE~). It should b~ noted that Dresderi was

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ired by their technical specifiptions to have an ISE The ONSG activities included the following:

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Exa~ination of plant operating charaeteristic Exa~ination of NRC issuances Examination of industry advisories Examination of Licensee Event Reports Surveillances of plant activities The ONSG maintains a data base for tracking their recommendations and the station's dispositio The ONSG issues a monthly report, a quarterly report, and an annual repor The inspector reviewed the ONSG's January 1990 monthly r~port, The inspectors also observed the onsite revtew committee meeting conducted on February 8, 1990, for the Unit 3 startu The inspectors noted that the meeting was conducted in accordance with OAP 10-1) On-site Review and Investigative Function and OAP 10-9, Selection of On-site Revi~W ~articipants. The inspectors regarded this meeting as thorough in that managem~nt adequately addressed the relevant issues and indicated good knowledge of outage activities and.a safety oriented, *aggressive attitude toward startup.* Items discussed as relevant to the issues, completion of the Unit 3 outage and subsequent Unit 3 startup include * the fo 11 owing:

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0 0 Outstanding.Equipment Outage Status Degraded Equipment Status Temporary Alteration Stat.us Qu~lity Assurance Audit Surveillan~e of Equipment Lineups Tech~ical Specificati-0n Surv~illance Status Control Ro6m Work Request and Lit Annunciate~ Status Post Modification and Post Maintenance Testing Status Status of Safety Related and Non-safety Related Work Request Completion

Preventative Maintenance O~tage Item Status Quality Control and Inservice Inspection Outage Discrepancy Report Status

Assurance that no Unit 3 parts were presently being used on Unit 2

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Reactor Vessel Inspection * *

Control Rod Drive System Work Main Condenser Inspection/Cleaning Torus* and Orywell Coating* Inspections Local Leak Rat~ Testing Results Snubber Testing and Inspection Modification Status

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Battery Performance Testing Heat Exchanger and Feedwater Heater Eddy Current Testing*

Effectiveness of Outage ALARA efforts

• Inserv{~e Inspectfons Erosion/Corrosion Inspection '

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0 0 Housekeepin~ and Material Condition Status Fire Protection Surveillance Status Environmental Qualification Circuit Walkdowns High Pressure Coolant Injection Valve Irispections Emergency Core Cooling System Undervoltage Testing Reactor Bottom Head Drairi Flow Verification Test Secondary Containment Leak Rate Testing Cominitment Status

Safety Related AC/DC Bus Alignment Inspections Isolation Condenser Inspection

State Pressure Vessel Inspections Detailed Design Control Room Work Status Refueling Activities Review Main Transformer 3 Repairs Main Steamline Temperature.Switch Replacement Feedwater Heater Level Instrumentation

Drywell Electrical Penetration Repair Integrated leak Rate Test Grease on Reactor Building to Torus Vacuum Breaker Check Valves MSIV IB Pilot High Temperatures

. Bypass Valve Osei l lations Non-traceable Safety-re lated Molded Case Circuit Breaker Replacement Status Large Bore Piping Configuration Verification Program Reactor Water Cleanup Pipe Settling/Displacement HPCI Oil System Problems Electrohydraulic Control Oil in Low Pressure Heater Bay Cable Trays Testing of Single Circulating Water Pump Operation Inservice Testing Surveillances Core Spray Suction Valve LLRT Analysis Reactor Vessel Head 0-ring* Repair Standby Liquid Control Testing Problems

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The inspectors observed a portion of the onsite review committee meeting conducted on February 17, 1990 regarding the Unit 3 recirculation pump cooling system leakage alteration as described in paragraph 4.a.(4) of this repor The inspectors also r~garded this meeting as thorough in that management adequ(ltely addressed the relevant safety issues in a conservative fashion *. The inspectors' favorable evaluation of the onsite review committee meeting regarding the January 5, 1990 Unit 2 scram is contained in paragraph 4.b.(1).

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The inspectors observed the Nuclear Safety Quarterly Meeting conducted on February 20, 199 Issues discussed focluded various Nuclea*r Manageme and Resources Council (NUMARC) initiatives, American Society of Mechanical Engineers (ASME) N-Stamp certification status and a planned rewrite of the quality assurance manua In addition, an identified negative trend in-deviation report processing timeliness was noted during the meeting. This was most probably attributable to outage activities competing with technical staff engineer availability. It was, however, also noted that Dresden was most likely understaffed in regards to technical staff engineers with respect to.the increased work load for 17 this* grou Management indicated that* steps were being taken regarding the timeliness issu The recurring events report dated February 5, 1990, was also reviewe The report categoriz~d deviation reports from 1984 through 1989, that had. been through offsite review, into identified re~urring eveni type categorfes:

Each was assigned a specific cause code and a brief descri~ti6n of the corrective actions.. In this way, recurring ~vent categories were identified and t~e effectiveness of th corrective actions could be evaluated.* The inspectors regarded this as an excellent tool toward accomplishing this goa It was.also note that Dresd~n's.threshold for issuing,and processin~ deviation reports wa~ much lower than oth~r CECo nuclear plant The inspecto~s have, jn the.past, also noted this *low threshold and have considered it to be extremely beneficial to the process* of identifying and determining the root cause of ~roblem Various'othe~ issues discussed included Nuclear Safety Group r.~views of temporary.alteratio.ns to 10.CFR 50.59 safety

~valuations f6r procedures, *techni~~~ specification charige submittals anci Dresden administrative technical requirement The Nuclear Safety Offsite Review group,indicated that their post-implementation revie required by* technical spectficat~ons of procedure change ~afety *

~v~luations had, in fact, not been accomplished for a long period of tim This was apparently due to their receiving the revised procedures without bein~ on automatic distribution for the c6rresponding 10 CFR 50*.59 safety evaluation As this was a.required post-implementation review, technical specifications had_ no specific time limits regarding this review, the problem was.identified by the licensee and indications were that the problem was being remedied, the inspectprs have no further concerns regarciing thi.s issu Finally-, a discussion of recent NRC inspe~tion items ~as ~onducted. 'The inspector regarded the meeting as ben~ficial towird ensuring adequate communications between Onsite and

• Offsite Nuclear Safet~ ahd plant miriagemen The in~pectors also reviewed the monthly status. report for the m6nth of Decembe The inspect6rs fbund it to be an excell~nt management*tool for remaining c'ognizant and identifying trends in various departmental indicator,

No violations' or deviati6ns* were identifie.

_Report Review Duri.ng the inspection period, the inspectors reviewed the licensee' Monthly Operating Report f~r December, 198 The inspectors confirm~d

• that the information provided met the requirements of Technical Specification 6.6.A.3 and Regulatory Guicie L1 Open Items Open items are matters':~hich have been disc~ssed with ihe license~ which will be reviewed further by the inspector arid which involve some actions on the part of the NRC or licensee or bot An open item disclosed during the inspection is. disc~ssed in*paragraph..

• . Exit* Interview (30703)

The inspectors met ~ith licensee representatives (denot~d in Paragraph 1)

on February 20, 1990,.and informally throughout..the inspection period, and summarjz~d.the scope* and findings of the inspection *acti~ities:

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The*:tnspectors also discussed:the l.ikely i.nformat.i.onal content.of 'the inspection report with regard.to documents or processes reviewed by the inspector during the inspect~on. The lic~nsee did not identify any such documents/processes. as proprietar The *1 i censee ai:knowl edged the findings of.the inspection:

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