IR 05000461/1990011

Safety Insp Rept 50-461/90-11 on 900512-0703.Violations Noted.Major Areas Inspected:Licensee Action on Previous Insp Findings,Operational Safety,Event Followup,Radiological Controls,Maint/Surveillance,Lers,Tmi Items & Meetings

ML20055J169
Person / Time
Site:	Clinton
Issue date:	07/25/1990
From:	Shymlock M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20055J168	List: IR 05000461/1990011 ML20055J167
References
TASK-3.D.3.4, TASK-TM 50-461-90-11, NUDOCS 9008010150
Download: ML20055J169 (25)

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U. S. NUCLEAR REGULATORY COMMISSION REGION III ,

Report'No. 50-461/90011(DRP)

Docket No. 50-461- License No. NPF-62 ;

Licensee: Illinois Power Company 500 South 27th Street Decatur, IL 62525 Facility Name: Clinton Power Station Inspection At: Clinton Site, Clinton, Illinois ,

Inspection Conducted: May 12 through July 3, 1990 Inspectors: P. G. Brochman S. P. Ray-P. L. Hiland F. R. Brush D. L. Butler Approved By: M. nt c /jec25, /9po Reactor Projects' ection-3B Jatef Inspection Summary 7 Inspection from May 12 through July 3, 1990 (Report No. 50-461/90011(DRP))

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Areas Inspected: (1) Routine, unannounced safety inspection by the resident I

inspectors and region based inspectors of licensee action on previous

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inspection findings; operational safety; event follow-up; radiological-controls; maintenance / surveillance; engineering and technical support;-

licensee event reports; TMI items; and meetings. (2) SIMS issue status:

(Closed) TMI Item-III.D.3. Results: Of the eight areas inspected, no violations or deviations were-identified in six areas; two . violations were identified in the following areas: (failure to follow procedures - Paragraph 2.c; failure to obtain compensatory samples - Paragraph 3.b(1)). In both cases the violations were of minor safety significance and met the criteria of 10 CFR 2, Appendix C, 1 Section' V.G. and therefore- a Notice of Violation was not issued. One

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unresolved item was identified (failure to wear dosimetry in a radiologically l

controlled area - Paragraph 4). The licensee's performance in the area of

operations improved. Operator response to abnormal events was good and management involvement continued to improve. Several events occurred in the 900s0to150 900725

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i radiological controls area which were of. concer Performance'in the ,

maintenance and surveillance area ' remained constant.E- However, problems were identified in the use:of/ cranes 1(with onelpotentially significant near miss).~

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~ Engineering support-of technical' issues > continued to improve, with problems

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from original construction. continuing to emerge. Safety assessment remained

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i-Persons Contacted

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Illinois Power Company (IH 1 -

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s ,i : : q 7 ' @#J. Perry, Vice President c ' '

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@#*J. Cook, Manager, Clinton Power Station .

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R. Gill, Manager,' Projects and Assessment ,.

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@hJ. Miller,' Manager, Nuclear Station Engineering iU

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' @R.' Morgenstern, Manager, Scheduling and'0utage Management ,

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@J. Palchak,-Manager, Nuclear Planning and Support r,% = i

u, @J. Palmer, Manager, Nuclear Training '

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@*F. Spangenberg, III, Manager, Licensing =and Safety * '

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@*R. Wyatt, Manager, Quality Assurance N j~' ~

1. K. Graf, Director, Operations Monitoring Program D. Holtzscher, Director, Nuclear Safety.:

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1. • R."Phares, Director, Licensing 1 3

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. S. Rasor, Director,. Plant Maintenance

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• P. Yocum, Director, Plant Operations

1. K. Baker, Supervisor, I&E Interface ,

i Soyland Power

@*J. Greenwood, Manager, Power Suppl . .

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,The inspector also contacted and _ interviewed other licensee and contractor personnel during the course of .this inspectio .

@ Denotes those:present during the meeting with Commissioner Curtiss on June 6, 199 . , .

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1. Denotes tho'se present during the management meeting on June 18, 199 !

• Denotes those present during the exit interview 'on July 3,199 . Actica on-Previous Inspection Findings (92702)- (Closed) Open Item (461/89030-01): Questions Regarding Locking -

-Devices on Locked Valves ,

This item was previously discussed in Inspection Report No. 461/89030, Paragraph 3.'1.3. The issue concerned the use of twisted lock wires as locking; devices for valves that were required to be. locked. The-inspectors were= concerned that the locks would not prevent: operation i of the valves'or give an indication that the' valves had been operate '

Other plants that the-inspectors have observed used seals that would indicate that the valves ~.had been tampered wit ,

In their response to Inspection Report No.'461/89030, the-l'icensee provided the b position that the use of twisted wire was acceptable -

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for ::::;ali valve:,. They stated that the use of lock wires, along l

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with their program of attacNing a' " locked valve" label to all locke '

valves and periodic surveillances of locked valves in accordance s with Technical Specifications, were adequate to prevent' inadvertent < 2 operat1,on of the valve ,

The inspectors reviewed Technical Specification Surveillanc a Requirements 4.1.5.b.3, 4.5.1.a.2, 4.6.1.1.b, 4.6.2.1.a, 4.7.3. ,

and'others which required that certain valves be locked.. None_of-the specifications required a certain kind of lock. The license l instituted its" locked valve program in accordance with Administrative ~- =l Procedure CPS 1401.01, " Conduct of Operations,"'which required the

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use of padlocks, chains, lockwires, The; '

procedure, referenced the licenseecommitment

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~to locking device ANS N18.7-1976 as the basis for locked valve control. Section 5.2.6'of- ANS N18.~7-1976- 1 required that procedures shall be provided for control of equipment, y as necessary,' to maintain personnel and reactor safety and to' avoi ,

unauthorized operation of equipment. It also required ~that those i procedures shall require control measures such as locking or tagging-to secure 'and identify equipment .in a controlled status._ The .

-licensee's u'se-of twisted lock wires on some valves along-with the 6 ,

" locked valve" signs and periodic surveillances' appeared to meet,

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the' requirements of Technical Specifications and commitments to ANS N18.7-1976 for the control of locked' valves. This issue is considered close (0 pen) Unresolved Item (461/89030-02): Switchgear Heat Removal Fans

Windmilling Backwards 3 s .This item was previously discussed in Inspection Report No. 461/89050, J Paragraph =3.5.4. The issue involved the inspectors' observation 1

'that the safety-related switchgear' heat removal fans, 1VX03CA/B/C, ,

,, were observed to be windmilling backwards during periods of standby 1

/ ' a operation when the normal nonsafety-related_ fans were running. 3The

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l, q inspectors' concern was that the fans might draw excessive current d whenstartedfromthatconditionand'thatitcouldcausea1 problem.(

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J l' 1 r with the, fan, fan breaker,' or class IE bus in the case of an acciden '

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In their response-to Inspection Report No. 461/8903U,*the'lic'enseel .'

i1< 1 l stated.that the condition had~been evaluated, based ~on,available' fan ( s

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l *l - motor data and operating experience. The evaluation c'oncluded that' "

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backward" rotation caused by leakage through the isolationLdampers ( ,

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did not adversely affect the operations of the fans, motors, or' '

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electrical class 1E busses which supplied the '

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The inspectors reviewed closed Condition Report No. 1-89-09-066 l which contained the licensee's evaluation of the condition. That

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a condition report discussed the effects of'the typical 15 to 20 rpm ,

backward rotation and concluded that, although there was no' specific 3

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manufacture's data or tests to confirm it, the engineering judgement  ;

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of the motor vendor and Sargent and'Lundy Engineers (S&L),was that n

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the effects would be negligible'. However, in a memorandum of a

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telephone conversation between a representative of the motor vendor and'S&L', the vendor representative-stated that "high speed".

backwards rotation could cause problems. Representativess for both the motor and the fan vendors stated that the weak link would 3

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probably be the motor / fan coupling. No specific clarification !o _

what.would be considered "high speed', backward rotation was give '

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li.c normal forward speed of the Division:I and-II fans was 1165 rpm *

and for Division-III.was 1750 rpm.- ,

On June 13,'1990, at a' bout 7:00 a.m.,cthe inspectors-informed the .

Shif t Supervisor that they had observed that' the Division II .VX fan ,

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-was rotatingLbackwards much faster than usual. .The Shift Supervisor

,s investigated the problem and at about 11:00 a.m. determined that'

the fan was rotating backward at 320 rpm as indicated by a strobe ,

tachometer. The licensee manually manipulated the outlet damper '

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linkage to cause the damper to close farther which-reduced the speed to aboutt20 rpm. The licensee wrote Maintenance; Work Request 012330, i

'to adjust.the linear l converter on the-damper and the rep' airs'were' s completed by'9:00 p.m. oniJune 13. The' licensee' issued Condition i Report No. 1-90-06-035- to track completion of the' evaluation..of_the V ?

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y y This unresolved item remains,open'pending the!11censee's det'ermination -t of the operability of the safety-related.VX fansfunder high speed

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backward rotation. - The Director - Design and Analysis Engineering ^ v a

stated that the licensee was in the process;of:evalucting the fans  :

at 320 rpm backward rotatio l

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c. ' (Closed) Unresolved Ithm (461/89034-01): Unterminah.edCables'on Division II Nuclear System Protection System Inverter .

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' This item was previously discussed in Inspection Report No. 461/89034,' 4

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Paragraph 3.a.(4). The issue involved an error during construction -

in transcribing design calculations:into field drawings _for thez i

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Division.II nuclear system protection system-(NSPS); inverter such-  :

l' that only two of.four specified conductors suppling power to'the l'

inverter from the DC source were-connected. The cause'of the' error 3 was'apparently a draftsman who overlooked a' note on the master ,  ;

diagram for the; inverter which stated:that all four conductors of the cable were to be' terminated. .The, licensee reported their s

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finding'as LER 89-038 on January 11, 1990,.and stated _that they were )

u conducting a study. to determine if the equipment supp_ lied by the ,

. inverter would,have been able to function under the< reduced voltage conditions-that would have resulted under the DC system accident l loading conditions with only,two of the four conductors connected to-the inverter' suppl g:

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On May 23, 1990, the licensee submitted. Revision'1.to the LER which

- 4 reported the results of their study. The revised LER. stated tha ~

the" Division II NSPS inverter had been capable of. performing its- ,

intended-design. function with only two conductors connected. Thus <

the licensee' concluded that the original finding was not reportable'

as an LER and that LER 89-038 should be considered a voluntary '

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A regional specialist inspector. reviewed the licensee's calculation E Numbr e 19-D-44,- " Division' 2 DC Systems IB _ Inverter Voltage Drop."

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The purpose of_the calculation was to assess the~ voltage drop L '

between Division II DC MCC 1DC14E and NSPS inverter 1C7150018.

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The calculation was based on a minimum inverter input voltage of l 105 VDC and a, normal operating input current of'10 A (measured).,

The Clinton Updated-Safety Analysis Report (USAR) assumed an inverter ll -

input current of 53.11'A. The' licensee _ informed the inspectors'that

l the, USAR value did not~ accurately reflect th Werter current  !

• requirements for Clinton's solid-state prd sthn Ap tea. In . ,

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addition, the licensee i.nformed the inspectors that;the normal inpu l C current would be lthe current, drawn by the inverter during a' design '

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basis accident. This was based on the fact that the inverter' loads -

'f changing " state'l were electronic in nature. verses the' energization I of electro mechanical devices used in older design protection w- systems.-

. The schematic diagrams of the loads supplied ty the inverter were - I reviewed and it appeared to the inspectors that the-. inverter normal /

design basis > accident input' current assumption was reasonabl l The calculation determined th'at the voltage available at the'1DC02E l battery during the first minute of. discharge after an AC power;1oss . '

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concurrent with a design basis accidentswas 108.23 VDC.A The inverter: '

under the above conditions would require a. battery voltage of 108'VDC to remain operable. -Based on the above, the. licensee m sr concluded the Division II NSPS inverter would-be ab1_e to-perform its safety function with only'two conductors connected. ; (

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The inspectors reviewed Surveillance Test Procedure.No. 9382.06, , .

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"125 VDC Battery Service Test," thats was performed on' February 21,- *

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? 89 on battery IDC02E. Thelowestbatteryterminillnitage.was '

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109.8 VDC;throughout the four hour load profile battery. capacity ,

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4' test (TS 4.8.2.1.d.2.b)' The lowest voltage; occurred during the'one' ,

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,%_ minute discharge portion of the test (TS greaterithan 462 amperes -

l , forthefirst60 seconds).;Thisfurthersupportstheilicensee'M *

lT conclusion that there was additional safety margin'available' toy .' , . '.

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ensure operability _of:the' inverter.- ,.

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Theinspectorshaveconcluded'thattheinverterwasoperableandl g l

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that the drawing-translation error was;an isolated occurrenc *'

. l Th'e 1.ER also reported the completion'of addit.ional corrective .

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. l actions'to restore the inverter.to;the design configu' ration, check'

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the;other three NSPS inverters, and review < other similar drawings" for' safety related equipment,to.se'e if there were other instances

of information in the notes tsections that were not incorp'., rated as required. No other'significant errors were~found.-

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t Although'the failure to correctly translate the design basis' for the'

c Division II NSPS inverter into-specifications, drawings, procedures, and instructions.was a violation of 10 CFR 50,, Appendix !

Criterion III, the event was considered a " licensee-identified"

. violation of' low safety significance which met the tests of

'Section V.G. of the Enforcement-Policy. Therefore a' Notice of  :

Violation was not: issued and this item is considered closed (NCV 461/90011-01_(DRP)). LERs89-038 and 89-038, Revision 1 were also closed in Paragraph 6.a of this report; i (Closed) Part 21 Item (461/86003-PP): ; Brown-Boveri K600/K80 ,

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circuit breakers that may have cut harness wires caused by1 contac '

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.of the harness wires with the racking gear inside the breake ,

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Clinton Power Station was using a total of twenty onh;,tyhdiK6003 k >

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breakers in the Divisions I and II; unit substation The licensee ViF <

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physically inspected all twenty-one' breakers and determined that inl- ,

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, seven cases the breakers.showed some sign of harness wear'or t,he H ,

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harness was touching the racking gear.. The following breakers were i, ,

identified:

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(2)--0AP05E-4D= :Scratched Scratched spiral spiral wrap wrap "$'l' ; ^ '+ ,

i-(3) 0AP06E-5C Scratched spiral wrap * N' i #+

(4) 0AP06E-5D Scratched spiral wrap ai ^ r (5) 0AP06E-4D Touching 9 J < '1 y

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(7) 1AP11E-5D Touching ,

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For all of the above, none of the. breakers exhibited any damage,to - .

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the actual breaker wiring. One-breaker was returned to the vendor '

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and the' remaining'six.were removed to spare positions. Tagout

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Number 88-680 was-issued;to prevent the use of these breakers in ,

' safety-related functions.~ ' The. licensee issued Maintenance Work - m*

Request D04715 which was scheduled to be completed by December 31,

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, 1990, to. install wiring harness protection on the remaining six'

spare breakers. Maintenance Procedure No. 8410.02, "480. Volt Power Circuit Breakers," which was used during breaker' disassembly and-

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inspection, provided the. electricians adequate steps,or'how to i inspect the' breaker and' breaker cubicle for damage and loose parts or wires. The licensee has ' adequately addressed the Part 21 notice and the. inspectors.have no further questions t,1 this item. This item'is1 considered close *

One violation was identified for which a Notice of Violation was not' .

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m The unit operated at power levels up to 43% until 4:38 p.m. ,' on May 17, *

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when the reactor was manually scrammed (see Paragraph 3.b.3). The unit .'

was taken critical at 8:00 p.m.'on May 18 and was synchronized to the

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grid at 6:38 a.m. on.May 19. The unit operated-at power levels up;to

! 100% for the rest of the report perio , Operational Safety (71707)

r The inspectors observed control room operation,qreviewed applicable logs'and conducted discussions with control room operators during May and June 1990. During these discussions and observations, the inspectors ascertained that the operators were alert, cognizant of

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plant conditions, and attentive to changes in those conditions, and i that they took prompt action when appropriate. The inspectors verified.the operability of selected emergency systems, reviewed tagout records, and verified the proper return.to service of-affected components. Tours of the containment, auxiliary, fuel-handling, diesel and control, radwaste, 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 equipment- M in need of maintenanc The inspectors verified by observation and direct interviews that-the physical security plan is being implemented in accordance with the station security plan, The inspectors observed plant housekeeping / cleanliness conditions, and verified implementation of radiation protection controls. The .

inspectors also witnessed portions of the radioactive waste system i controls associated with rad-waste shipments.and processin .

The observed facility operations were verified to be in accordance with the requirements established under Technical Specifications, 10 CFR, and administrative procedure '

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(1) Reactor Recirculation Pump Seal Degradation #

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. < During the entire inspection period, the lower seal on the "B"

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reactor recirculation (RR) pump appeared to slowly degrad This was indicated by a slow increase in the' pressure measured

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between the lower and upper seals. .That pressure was normally .

about 500 psig but had increased from about-690 psig;to: ,

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790 psig during the inspection. period indicating-higher than . -

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normal: leakage past the lower seal. In addition,1a high scal'

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l leakof f ' alarm was received on the upper seallof.the ;"B",RR: pump t

. . on May 20, 1990 and continued to be in until the end of the. -

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, , report per bd. This indicated higher than normal tota,lEseal .

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leakage. 'On June 1, 1990, the loose parts monitor alarmed'on s the channel monitoring the "B" RR loop at about the same time ~ ,

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that recirculation flow had been increased to slightly above >

-100%....The alarm cleared when recirculation flow was reduced t below 81 million pounds mass per hour (M1bm/hr). On June 13, 1990, the loose parts monitor again alarmed on the "B" RR loo c channel when recirculation flow was raised to above 84'Mlbm/hr-and the alarm stayed in during the remainder of the. inspection "

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period except for brief periods when recirculation flow.was-reduced below 100% for surveillances. During the last two-

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weeks of the inspection period, the pressure between the upper and lower seals was noted to oscillate by about 15 ps , ,

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4 The licensee has experienced previous problems-with the seals . s on the "B" RR pump. Inspection Report NoJ 461/89021 discussed

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the results of a special safety inspection of the catastrophic- .

failure of the seals on June 1, 1989, including the licensees '-

assessment and corrective actions. That report also discussed . ,

an additional failure of the seal on December 17, 1988,'and an i improper assembly of th~e. seal on January 23, 198 ,o

.During this inspection periodj.he inspectors monitored the .

licensee's actions and decisions regarding the seal. Repair part availability was verified and plans were drawn up'in case

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i the seal had to be replaced. The licensee maintained close

. contacts with the vendor regarding allowable seal parameters and sent them recordings of the vibration signatures for

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analysis. Drywell leakage parameters were closely monitored to. verify that seal leakage was not becoming excessive. The system engineer for the RR system was closely involved;in following the developments and advising the operators of actions to take. Interviews with operators verified that,they L'

were cognizant of the actions to be taken if, seal leakage'. . ,

degraded rapidl Licensee management personnel were,kept closely informed of the seal statu [

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Although licensee management decided not to conduct a special %~

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outage to replace the seals during this inspection period, they took adequate actions to monitor the situation and to prepare 3; for any seal failur '

(2) Waiver of Compliance ,

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At 2:43 a.m. on June 22, 1990, licensee personnel discovered- J

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that the expansion tank for the jacket water cooling system fore the Division II emergency diesel generator's 12-cylinder diesel'

engine was overflowin This had happened previously to b'ot the Division I and=II diesel generators when microbiological 1y ,

influenced corrosion (MIC) had caused cracking of.the'tubesPin *

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the diesel engine's jacket water heat exchanger (HX); thereby' <

allowing the higher pressure service water. system to: fill the:

jacket water system. Previous problems with MIC were discussed- *

in. Inspection Report No. 461/8903 ,

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The licensee had previously performed a 100%' eddy current: 1 y inspection of the tubes on this HX, plugged all tubes with sm . ;

greater than,50% thru wall pitting, and performed a sodium .

.. 4 bromide chemical treatment,-to retard the growth of MIC, during  !

' January 199 The licensee had intended to retube the HXs on the Division II diesel generator during the refueling outag scheduled for this fall; consequently,_the equipment necessary to retube the HXs was not onsit Af ter evaluating the. condition of the HXs the licensee decided to retube the HXs; however, as the necessary equipment wa .not onsite, the licensee estimated that the length of time necessary to complete the retubing would exceed the allowable-outage time ~for the diesel generator technical specification of'

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. ; Consequently, the licensee requested a waiver of

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compliance from.the NRC to extend the allowable outage tim '

This requ2st was contained in letters from J. S. Perry to A. B. Davis dated June 22, 1990 (serial numbers U-601696 and U-601698).

The licensee's request to extehd the allowable outage time'from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to.7 days was approved by the Region III' Administrator on June 22, 1990; and was documented in a letter from A. B. Davis to.J. S. Perry, dated June 26, 1990. The licensee successfully completed.the retubing of the HXs and. declared the ,

diesel operable by 3:04 p.m. on' June 25, 1990, t

No violations or deviations were identifie b. Og i.te Event Follow-up (93702)

The inspectors performed onsite follow-up activities for events which'

occurred during May-and June 199 These follow-ups~ included reviews-of operating logs, procedures, Condition Reports, Licensee Event Reports (where available), and interviews with licensee personne ,

, For.each event,'the inspectors developed a hronology, reviewed the: .. F

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functioning of safety systems required by plant; conditions, and

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reviewed licensee acti.ons to verify consistency with procedures, I,

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license conditions, and the nature of the event. iAdditionally,,,the

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inspectors verified that the licensee's investigation had~ identified, l' the root causes of equipment malfunctions and/or personnel 'errorsJ * * *

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and that the-licensee had taken appropriate correctiv'e actions ~pri'or i

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to restarting the uni Details of the events and the licensee's'1 . ,

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corrective actions developed through inspectors' follow up,are- ' 1 provided in Paragraphs (1) through (4) below:

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(1) Inoperable Drywell Air Particulate Monitor 3

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Th'is item was briefly d'iscussed in Inspection > Rep [ ort _ g(

No.'461/90006, Paragraph 4.c. The report stated that the

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event.would be reviewed in more detail when LER 90-009 wa , o i

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The licensee issued LER 90-009,' " Sot Screw on Gear Train of <

Moving Filter: Paper Drive Mechanism Results in Paper Not Moving:

and Inoperable Leak Detection Drywell Air Particulate' Sample Panel," on May 21, 1990. In their-LER,!the licensee-stated that the cause of the filter paper not-advancing was apparently a loose set screw on the gear train of the drive mechanis <

The paper-had.apparently not been moving properly between April 4, 1990, and the time of discovery on April 27, 199 The plant had been in Operational Conditions requiring the drywell, leak detection. system to be operable between April 7

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and April-15, and again between April 22 and April 27, 199 j Technical Specification' 3.4.3.1;a ' required;that the drywell i

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atmosphere' particulate radioactivity monitoring system be 0PERABLE-in OPERATIONAL CONDITIONS 1, 2, and The associated

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ACTION statement required that with one of-the required leakage j detection systems. inoperable, operation may continue.for.up to ~

30 days provided grab samples of the drywell atmosphere arel obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. . . otherwise' i be in at least HOT SHUTDOWN within'the'next 12. hours and in

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COLD SHUT 00WN'within'.the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> *

Between April 7.and April 15, 1990, and between April 22 and 's j

. April 27, 1990,-. the plant was operated inL0perational .

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y- Conditions 1 and 2 while the'drywell atmosphere particulate j radioactivity monitoring system was inoperable due to its:

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• particulate filter paper not advancing from the sam'le, p point.to .

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the detector. No drywell grab samples were takeneduring< those' .

time periods. - The failure to obtain and analyze:the! required, , .!

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e' samples ',as a violation of Technical Specification 3 3:4.3.1.a. , .

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i Howeve , the event was . licensee-identified and properly] reported . .

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in 13 90-009. The event.was not considered safety ~signif,1 cant s -

be.ause other portions of the drywell: leak detec' tion l system - -(i; "

, .<ere operable and grab samples taken'upon discovery-of the .

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condition showed that there had been no excesisiv0 0nidentified i . i .

leakage. The' inspectors . reviewed all' violationstand condit$on' .Is i

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reports written on the leak detection-system in the.last tw6 " T

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L d -years and noted~no previously identified-conditions for which l '

corrective actions could have prevented the event. Since the ,; j violation met the criteria of Section V.G. of the Enforcementii '

Policy, a. Notice of Violation was not-issued and.this issue is ' j

,t considered closed (NCV 461/90011-02(DRP)). ,

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The associated LER will remain open pending' completion of the, " 5 J

licensee's investigation of metho.ds to ensure proper operation of,the paper drive mechanism without disassembling'the uni That action was scheduled to be completed by August 31, 199 ,

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> (2) Diesel Generator Failure due to Mispositioned Service Water , m

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CooH ng Valves i . . .

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0n May 15,- 1990, during a routine surveillance of the' '

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. Division-I diesel generator, the engine tripped on'high jes'c+ . 4

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, water temperature shortly.after loading. This1 event >was- ' .

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reviewed in Inspection Report No. 461/90012 and'an enforcement ^ ,

conference was held on this matter'which was do'cumented in: '

. > Inspection Report No. 461/9001 , '* .

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l, Thelicensee'sclassificationoft$testandreporbingof=the1

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event are discussed in Paragraph 6 belo '

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(3) Manual Reactor Scram due to Reactor Recirculation Pump Trip: '

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On May 17,.1990, with the reactor?at about 43%' power, the "B"  ;

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' reactor' recirculation'(RR)^ pump tripped of f and the "A" RR ! '

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pump shifted from fast to slow speed. As a result,-reactor'

power der'. eased to approximately 21% and.the reactor operator inserted a manual scram in accordance with his procedures and instruction Approximately one hour before the event,Pth'e RR pumps had b'en e l shifted to fast speed ~as part of the normal-power. ascension' i procedure. At 4:31 p.m. , reactor feedwater flow channel ."B" >

failed low which caused the sensed total feedwater flow signal to go below the 30% setpoint of the'RR pump's cavitatio interlock. The cavitation interlock feature was designed _ to automatically shift the RR pumps to slow speed whenever total'

feedwater flow goes below 30%. The low flow signal-started timers on both the "A" and'"B" RR pumpo l'gic circuits, which

, were nominally set for a 15 second delay. At 14.44 seconds; the; -

fast speed breaker on the "B" RR p~ ump opened, starting its slow

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speed transfer sequence. At 14.8~ seconds, the "B" RR. pump -

reached 95% speed, where the logic was designed such that the M l ;. slow speed transfer would be bypassed unless,the'other-RR l pump's fast speed breaker was also open. The "A".RR pump fast y speed breaker didn't open until 16.9 seconds, thus-the "B" RR '

pump's auto transfer to slow was bypassed ~and it continued to l

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coast to off. The "A" RR pump completed its transfer to slow

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speed normally, as the "B" fast speed breaker was already ope Off Normal-Procedure CPS 4008.01, " Loss of Coolant Flow,"

contained instructions to manually scram the reactor if both RR pumps have shifted to slow or both have tripped to off. This'

l step was added in response to the flow / power oscillation n .

i problems that were experienced in other, boiling water reactors, d l

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Plant management had determined that they.-did not want to have operators-distracted by.trying to determine whether they wer in'the flow instability region of the power-to-flow map after a loss of coolant flow, but rather specified a manual scram when l

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the potential existed. In this case, the procedure.did not a s ,

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specify actions for one pump off and one in~ slow, but the . *

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operators-took the conservative' action of manually scrammed th +

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reactor. In addition, the' Director - Plant Operations had -

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i recently written a " night. order'! to scram!the reactor if single l loop operation was necessary. This direction was a result of.

l a licensee ~ unresolved question involving the analysis for the; '

Maximum Extended Operating Domain'(ME0D) on the power-to-_ flow ,

i map and single. loop operation i

,- , Plant response to the scram and operator followup actions j l were all correct. Due to the normal reactor level transient i following the scram, a level 3 trip signal-was generated which

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L l caused isolation of containment isolation groups 2, 3, and'20.

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The licensee reported the event to the NRC.via the ENS on .

May 17, 1990, and submitted LER 90-012-as a followup report'. j

i The cause of the failure of thel"B" feedwater flow channel was  ;

I determinedtobeduetoafailure'ofthepowersugply'spower converter. The same power supply also fed the "C reactor 3 level and upset range level channels. About 14! hours before ,

the event, operators had noted an intermittent spike on-

!- the upset range level channel but efforts to determine the:

cause had not been successfu The power supply had been continuously energized for a period of about nine years'before q the event and the vendor considered the failure to be a normal J end-of-life failure. The power suppl 9:was Ja sealed unit and i l was returned to the manufacturer for additional' analysis. The y licensee replaced a similar power converter in the.other

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channel of feedwater flow,-and instituted a five year _-

preventative maintenance _ task to replace both converter . t The cause of the "B" RR pump not completing its automatic s

transfer to slow speed was that the difference between the actual settings of the two 15'second timers was too larg ~ q Thus the "B" RR pump coasted to below 95% power before-the "A" . i RR pump transfer sequence started; thereby causing.the "B" RR-pump's transfer logic to lockout, with the pump coasting to ,

a sto The system engineer for' the RR' system has held

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discussions with General Electric to try'to determine the .

reason for the design of the automatic fast to slow sequence ,

a l bypass if the other pump's fast speed b~reaker'isn't open._ It- 1 l has been determined that-some plants hwe modified the circuit

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and the licensee was considering that_ option. The 1icensee's *j investigation was scheduled to be completed by December'30,

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The licensee's Facility Review Group. met on May'18,~1990',sto i conduct a post trip review of the event and to review the corrective actions. The failed power supply was replaced and '

the similar power supply of the "A" feedwater flow channel was  ;

checked. As a precaution, that power. supply was also replace *

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Failure history of the' power supplies was reviewediith no' .. 1

5 previous similar failures being noted. ! Longer term corrective ~* ,

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actions-included contacting the vendor concerning the, expected 'C i

I service life of the power supplies, checking the Nuclear Plant *t

Reliability Data System for similar failures, and checking thez

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calibration procedures and allowed tolerancesJof the two 15:, 3 , !

second timers. The reactor was restarted on May' 18, 1990' and

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synchronized to the grid on May 19, 199 s

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,y a (4) Possible-Waterhammer on Division III Shutdown Service Waters Piping , -'

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o At 12:46 p.m. on May 24, 1990, the Division III shutdown 3 ,,3

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L service water (SX) pump was being started for a surveillance'. *

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when it_ tripped on-thermal overload. Simultaneously, a gasket ,  ;

blew out at the 781' elevation.of the control building in a SX _  :

line to the-division.III switchgear chille '

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The SX loads'are-

; normally supplied by plant service water (WS) system; however, '

. a " when a start signal is received by the SX pump, the WS valves go closed and the SX' valves go open to' shift the water sourc .

In this case sinc ~e the SX pump'did not start'the valves'

repositioned themselves to their _ origin ~al position. The licensee believes that-a portion of the SX piping'may have

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drained as the valves were stroking (with the SX pump off);. J l

, consequently, when the pump restarted,;the surge of water.back'

into the system cause an. apparent-water hammer. -This transient also cause a locked throttle valve in. the SX system to change

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its position. The licensee initiated condition. report 1 _

l 1-90-05-096 to resolve this' event. The: licensee's analysis had -

not been completed by.the end of the report and the inspectors willfollowthis_asanopenitem(461/90011-03(DRP))..

One licensee-identified violation was' identified for;which a Notice of

. Violation was not issue One open item was identifie . Radiation Protection (71707 & 93702) 1

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On June 19, 1990, an individual entered the Radiologically Contro'11ed Area (RCA) at Clinton without dosimetry. The individual was employed

'by Sargent & Lundy Engineers'(S&L), the architect / engineer for Clinton,

-and had recently been assigned _to the site. The individual had take radiation worker training in February; however, he had not completed training for the new radiation work permit and dose tracking system (SR-31 modification which was not yet operational). The individualL and a second S&L engineer entered the protected area at approximately!  !

10:00 a.m. on June 19 and proceeded to;the~ dosimetry window where the

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individual was told he could not be issued dosimetry because he had not yet completed SR-31 training. The two individuals then toured portions ,

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of the service building before entering the turbine building sTh1 * ,

, r turbine building is inside the RCA. The two individuals then, entered- -;

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y the radwaste building, where they were seen by a-maintenance.wo'/ker,. ',

[ who noted that one of the individuals did not have. dosimetry. {he " ,,

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maintenance worker told the two individuals that they needed dosimetry ,

and that they should leave the RCA'and report to radiation protection

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(RP). As they were leaving the RCA, they were found'by a RP technician, '

, who escorted them out of,the RCA and-to dosimetry where an exposure  ;

investigation was' initiated; The two individuals then notified their 4

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supervisor of this event, who notified the S&L site manager and, S&L The S&L site manager notified. licensee management

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corporate managemen ,

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in the nuclear station engineering department (NSED). The S&L site- it manager directed'all of his supervisors-to brief all of their personnel ,

on this' event and its seriousnes The S&L site manager also directed:

the;two individuals not to enter the plant again until this event was resolve ,

The'RP shift supervisor (RPSS)~was informed of this' event and, notified

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the Director - Plant Radiation Protection. The following information

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was communicated: the entry did'not appear to be willful, the individual 1 was not in the RCA' for very long,-and an accurate dose assessment could ..

be made from the dosimetry,of the second individual. The Director instructed the RPSS to have dosimetry initiate'a dose assessment and a J condition report. The Director notified the acting Plant Manager of this -

event. On June 22, 1990, the Director discussed this event and the need -

to be vigilant with RP operations personne ,

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Based upon initial review by the inspectors, the actions taken above by RP, NSED, and S&L management' appeared to 'be appropriat However, the inspectors had some concerns with other actions which did not appear to .

have been taken in a timely manner. The RPSS did not log this event in-the RPSS log, but did enter it on his turnover sheet. The condition report was not initiated for more than-24 hours by the dosimetry personnel, due to questions.over whether it should-be initiated, even though RP .

management had given direction to do so. A critique was not held in a timely manner, such that management had difficulty in answering questions raised by- the inspectors. Access to the RCA was not suspended for-either

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i individual, pending completion of the investigation, as was required by f o ,

licensee policy, until July 3, 1990, after question's were raised by the inspector a

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Subsequently, a critique was held on July 5,1990. Some'oftheffacts *

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which emerged were that the individuals did not believe that they needed: ,

fl dosimetry or needed to sign in on a radiation work permit (RWP) inside g ,' .

the RCA.if they ' restricted themselves to certain pathways. Training i

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given to both individuals specified that dosimetry was required; however,'_ ']

t r i signing in on an'RWP was not- required for certain very . limited cases.; -

'l '(Note: This policy will end with implementation of.the SR-31 modification

'in August 1990.)- Neither individual saw the posting at.thenentrance to 'N

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> theRCAwhichstatedaTLDandpocketdosimetrywerereqbiredjfo'rentry.'; .

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. 10 CFR 20.202 required that individuals who enter restricted areas ;I +t* *

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where they are likely to receive 25% of the quarterly doseilimit -of- A ,j 10 CFR 20.101.a, be supplied with and be required ~to wear dosimetry., f This event will be followed-up on by regional specialists and will be, ' ,

I tracked as an unresolved item (461/90011-04(ORP)). 2 1 Yl ')

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Other events which happened during:this report period included: an  !

individual entered the RCIC pump room on June-20,1990,:after leaving his low range dosimetry in the change area (he was wearing high range

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dosimetry). High radiation door number 245M was-found-unsecured on May 25 and again on June 2, 199 Additional problems (lifting ~of resin liners with cranes) which had radiological1implications are discussed in; Paragraph ~ The. inspectors"viewtheidentificationofkheindividualwithout

' dosimetry by'a maintenance worker as a very positive aspect of . ,

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the licensee s radiation protection progra ,

,t One unresolved item was identifie t

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' ' Maintenance / Surveillance: (61726 & 62703)

Stationmaintenanceandsurveillanceactivitiesofthesafety-relAted 4 7 systems and' components were observed or reviewed to ascertain that they were conducted'in accordance with: approved procedures, regulatory guides,' <

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and industry codes' or standards, and in conformance with Technical Specification ,

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The following' items were considered during this review: . the'limiling conditions for operation were met while affected components or systems >

were removed from and restored to service; approvals were obtained prior to1 initiating work or testing; quality control records were maintained; parts and materials used'were properly certified; radiological and fire- ..'.

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prevention, controls were accomplished in accordance with approved

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procedures; maintenance and testing were accomplished-by qualified- *

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personnel; test instrumentation was within its; calibration interval; t functional testing and/or calibrations were performed prior to returning

. , components or systems to service; test results conformed with Technica . .'

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Specifications and procedural requirements-'and were' reviewed by personnel other than the individual directing the test; any deficiencies identified

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during=the testing were properly documented, reviewed, and resolved by appropriate. management personnel; work requests were reviewed to determine the status of> outstanding jobs and'to assure that priority was assigned ^

to safety-related equipment maintenance which may affect system performanc Concerns with Safe Crane and Hoist Operation On May 8, 1990, the inspectors observed a new fuel receipt worker climb to the top of:the new fuel uprighting stand. As the worker ,

. reached the top of-the stand he kneeled over to secure his safety 4 line. At the same time, the crane operator was lowering the hookE which would be employed to-lift the new fuel bundles-out of Ltheir-

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shipping containers and' move them to the inspection stand. However, the hook was lowered too much, so that the worker on the stand

'3- struck-his hard hat on the~' hook when he straightened up from the

. kneeling position'. The worker was startled but not injured by the "

blow because'his hard hat provided. protectio .

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', e At the time the worker struck his head, he had-not'yet secured his

, safety line. The consequences 'of the incident' could have' been very

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serious had the workar not been wearing a hard hat or had the worker; lost:his balance and fallen from the' platform approximately.20; feet ~.

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Prior to that observation, all crane maneuvers were closely controlled s >

by the crane signalman through hand signals. The' signalman also had ,

, radio communications with the crane operator. It appeared that the signalman was not directing the crane operator when he lowered th '

hook above the worker on the platform although the inspectors could ,

not tell;if the radio was being used. .The-inspectors? discussed thei

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event with the fuel receipt supervisor, shortly'after the occurrence;

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. and noted that he was not' aware of the' incident. The supervisor -

stated that he would ensure that crane operations.were more. tightly'

controlled. Plant management and'the safety: department were not

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i informed of this near miss till some time afterythe event; during  ; ,

discussions with the senior resident inspecto

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"Also on May 8, 1990, a near miss-incident involving improper crane  : .

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operations was identified by the licensee. - In that' case,_inadequates t

communications between the signalman and the crane'operatorf and improper direction given to the crane operator by the, super. visor,.

resulted in a truck being in motion while the crane operator was trying to lower a liner containing highly radioactive resin' . ~ s (approximately 40 Rem /hr dose) .into,a shipping container on the truck bed. The liner started to tip' over. A radiation -technician ,

next to the truck saw the problem and stopped the. truck andz the' load j was stabilized. That event was very serious ~ and-could have led- to i the spill of radioactive-resin onto the. floor of the turbine 3 '

building, if the drum had fallen and, broken ope The licensee had conducted extensive briefings and dry 1 runs before moving the resin liner. These briefings included discussions on command and~ control, lines of_ communication, and independent observer However, when the actual lift took place the radiation fields were higher than had been anticipated: consequently, radiation protection personnel reo': ired that pei:9nnel involved in J the lift stand farther away te reduce their exposnr This changed the lines of communication and where command and control would be ,

exercised from. The supervisor in charge;of the: evolution decided not to postpone it and have another dry run. 'The actual instructions to lower.the;1oad were given by.the supervisor'(even though he was?not supposed to) and were accepted by the crane?

operator (even though he knew only the: load. handler could give him instructions to lower the load). The inspectors _believe only the '

fortuitous action of several' individuals prevented the spill of s highly radioactive resi '

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On May 16, 1990, while attempting-to attach a chain-fall hook'

to a new fuel channel, the worker dropped the hook and it hitione i cof the'new fuel bundles-in the. inspection stand. No damage, l- was,done to the fue , ' "

. . 1 As a result ~ of those.three incidents and the. dropped radioacti'& 3, material discussed in Inspection Report No. 461/90001, .

F  : Paragraph 3.a.(3), the inspectors expressed their concern to p ,' - management over the control of crane and hoist operations. Thi y Manager .-~ Quality Assurance had independently raised the same c* '

concern at about the same-time.: ,

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, On,May 21, 1990, the Vice President discussed the recent er e and ae s-[

L* F . hoist problems with all supervisory level personnel at-his periodic

, Project Status Meeting. In addition, on June 7,1990,'; the(Manager: : A

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Clinton Power Station approved enhancements to'the; crane operator's

,, 7 - tra'ning program to emphasize the lessons learned from the events.; '

? /g Those enhancements were scheduled to be completed.by'J'ly u 31, 1990i

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3" All crane operators in the Maintenance Department were briefed ont' ,

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their responsibilities and authority relative to safety and, t *

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appropriate' personnel in Radwaste=and Plant Support Services 1' . Departments were-to be trained using the enhanced program beforee ',

operating any lifting devices. The. Director' : Industrial Safety, cp '

+ required that safety belts be attached to the uprighting' stand ' s *

l l- before the workers entered the top platform. The actions described t above resolved the inspectors concerns on this issu '

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l Review of 125 VDC Battery < Service Tests o

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1 As a result of a concern raised by a regional inspector, the

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c inspectors reviewed the last performance of Surveillance ProcedureL CPS 9382.06,'"125 VDC Battery Service 1 Test," fer the four class .

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IE batteries. The purpose of the! surveillance ~was to meet. Technical

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Specification' 4.8.2.1.d.2 ~which required that' attleast each 18 months, the battery capacity be verified to be adequate to supply. a: dummy d load of the profile specified, while maintaining the battery terminal voltage greater than or equal to 105 volts.- The inspectors reviewed f

the following surveillance 9382.06 records-

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Division l' January 11, 1989 Division II February 21, 1989 ,

' Division III January 5, 1989 >

Division IV January 17, 1989 '

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.The surveillances were conducted by programming an Alber M'odel l O BCT-30 Battery' Capacity Test. System with the load profile specified ,

in Technical Specifications' and conducting the test:while using an -

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Electronic Control / Data Logger Unit to record the.various battery; t

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' parameters each 15 minutese The specifiSd load profil'staIl U'

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i e required a high current for.60 seconds, a medium cQrrentifor the , , , ,

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next 59 minutes, and a lower current for the.next.180 minutes, .

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except for Division III which required a high currentifor<60 seconds-7 o and a lower current for the next 239 minutes. The.15, minute , 3 1 ,

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printouts.from the data logger were required tc be filed with thet

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surveillance record 'y 1a -

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, g 's acceptance criteria of the surveillances were met'due to the: data" .

,' logger printing out the. load only every 15 minutes. Specificalyy: y ]

, * For Division I, there was no printout giving objective evidence I

j that the load had actually been the required 561 amps for the

, first minute and the- printout at~ 60 minutes showed e 126 amps- 't rather than'the required 159 amps. 'here , was an, extra printout;  %.- {

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at 61.5 minutesLthat showed the proper 159 amps, lAll later 1 printouts also showed the proper 159 amp * -For Division II', there was;a printout for 31 seconds which showed 460 amps. load rather than'the' required 462 amps'and the *

printout for 60 minutes showe'd 87 amps rather than the required-108 amps. The printouts at 75 minutes and-later showed the proper 108 amp * For Division III, there was~no printout =giving objective "

evidence that the load had actually been the required 112 amps l for the first minute,

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• For Divisi:. IV,.there was no printout'giving objective evidence tiit the load had actually been'the. required i 127 amps fc? the first minute and the printout for 60 minutes ,

showed 20 amps rather-than the required 44 amps.- The' printouts at 75 minutes and later showed the proper 44 amps.

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The inspectors verifie'd that'.the printouts'of the BCT-30 program showed that the'BCT-30 had been programmed to deliver the correct load profile for each surveillanc That provided'some evidenc j that the surveillances met their acceptance criteria butidid not

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provide sufficient records cf actual load current; The' discrepancy i in the: Division II test at .L seconds was within.the specified- l accuracy of the BCT-30 load current but was slightly less that the 'J Technical Specification . load profile. . The discrepancies at  !

60 minutes appeared to be due to-a temporary load swing'because the BCT-30 was programmed to change from medium to low load at exactl , ,

the same time that a printout was demanded.. The vendor's manual for- "

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the BCT-30 stated that the unit's computer would readjust the .

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resistance value if the actual current flowing was-not equal toL+1%' , e C *

+1 amp -0 amps of the programmed current. Thus the_ low cur. rents

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All of the final battery terminal voltages at the end of the four hour tet.ts were well above the specified 105 volt The battery surveillances demonstrated that the batteries could have met the specified load profiles. However, with the way that CPS 9382.06 ,

was written, it did not provide objective evidence that could be used by the Shift Supervisor or other independent individuals to verify that the dummy load profile actually met the acceptance criteria. One way in which the surveillance could have been '

improved te resolve that concern would have been to require that, in ,

addition to the 15 minute printouts, a data logger printout be manually demanded during the first minute of the test and also shortly after 60 minutes into the test, Improper Calibration Currents Specified for Local Power Ranae Tonitor (LPRM) Adjustments

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On May 24, 1990, the Heensee discovered that calibration currents supplied by the nuc1,ar engineer for use in Surveillance Procedure i CPS 9831.01, "LPRM Calibration " were incorrect. A full set of '

traveling incore probe data had been obtained on May 23 which were used in a computer program containing what an engineer thought was the last set of LPRM calibration currents to calculate a new set ofLPRMgainadjustmentfactora. On May 24, when a technician was

, adjusting the Division I LPRMs to the newly calculated currents, he noted some unexpected responses in the instruments. Some of the .

LPRM outputs changed excessively with the new currents. The l technician stopped after completing the calibration of eight LPRMs y on Division I and informed the Shift Supervisor of his observation .

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The Shift Supervisor called the nuclear engineer. to reviewed the calculation , The nuclear engineer discovered that the previous caHbration data

that was used by the computer to calculate the new gain' adjustment s ;

factors was not from the most recent calibration. An, investigation determined that the nuclear engineer had failed to transfer:the data from the most recent calibration to the hard disk on the .'

engineer's computer. This may have been due to the fact that the engineer's office and computer were being moved frt,m the service building to the engineering building at about the time that the; previous calibration was done. When the date was taken from the ,

hard disk to perform the calculations on May 23, no check was l accomplished to verify that it was actually data from the most'

recent calib stion. The latest data did exist in paper form on i the engineer % desk,

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l- The event was not safety significant because the currents that were

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set into the eight LPRMs resulted in the instruments being off by less than the allowed acceptance criteria of 10%.-- In addition, th *

error was discovered while LPRMs from the first of the four instruinent divisions were being calibrated. Procedures _ required "

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that the new gains be verified by conducting a heat balance between Sach division 50 it is likely that even if the technician had not stopped to investigate, the error would have been: discovered before the gains on the next division were se The technician who noted the unexpected instrument indications and ,

stopped to inform the Shift Supervisor demonstrated an excellent understanding of nuclear instrumentation and displayed a,high degree, of concern for reactor safety. The licensee's investigation-of the event disclosed several weaknesses in the control of safety-related

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computer data filos which they have taken action to correc '

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No violations or deviations were identifie '

, Engineering and Technical Support -

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Inspection Report No. 461/90012 discussed a failure of the Division I emergency diesel generator on May 15, 1990, shortly after loading, due to. . i mispositioned service water cooling valves. As a result of the failure, 1 the licensee inspected the Division 11 diesel and discovered that its 4 service water cooling valves were also mispositioned in'such a way that ,

it would have failed had it been ru ,

A The licensee initially categorized the failure of Division I and the .

- l condition discovered on Division Il as valid failures in accordance with

- Regulatory Guide 1.108, Revision.1, August 1977, Regulatory Positions C.2.e.(5), which stated that successful starts followed by an

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attempt should be considered valid tests and unsuccessful loading (8) which stated that cranking and. venting procedures failures, and C. that lead to the discovery of conditions that would have resulted in the failure of the diesel generator unit during test or during response to a i bona fide signal should be considered valid tests,and failures. The-event was the second valid failure of the Division II diesel generator in'

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the last 20 valid tests and thus the licensee was required to. increase the monthly test frequency to at least once per seven days in'accordance , j with Technical Specification Table 4.8.1.1.2-1. The Division I diesel generator was already on the increased test frequency because of previous failure '

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On May 22, 1990, the licensee reevaluated their decision and redesignated the events as invalid tests in accordance with Regulatory Position .

C.2.e.(2) which stated that unsuccessful start and load attempts that can definitely be attributed to operating error. . . should not.be considered ,

valid tests or failures. The mispositioned service water. valves were attributed to operator error in that the operators did not use the proper technique to set the throttle valves. The licensee discussed their* ~~ '

decision with the NRR Project-Manager at that tim f

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On May 30, 1990, the inspectors expressed their concern to the licensee over the classification of the events as invalid tests. The inspectors .

informed the licensee of their opinion, that by reading Regulatory Position C.2.e.(2) in context, the operating errors that could be ,

considered invalid were only those associated with testing, and were such'

that they would not have prevented the diesel generator from operating in the emergency mode. The mispositioned service weter valves would have caused the diesel generator to trip.on high temperature shortly af ter loading on an undervoltage event and would have resulted in failure of the diesel on a loss of coolant accident event. The NRC Regional and Headquarters Staffs agreed with the inspectors' interpretatio The licensee then stated that the events could still be considered invalid tests because of another portion of Regulatory Position C.2.e.(2)

which stated that malfunctions of equipment. . . that are not part of the defined diesel generator unit design should not be considered valid tests or failurn. The " diesel generator unit" referred to in the Regulatory .

Guide was defined as consisting of the engine, generator, combustion air n l system, cooling water system up to the supply, fuel supply system, lubricating oil system, starting energy sources, autostart controls', ' -

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manual controls, and diesel generator breakers. The licensee's ,

7 interpretation was that the " cooling water system up to the supply".did not include the service water throttle valves. Although that expression ,

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was not further defined in the Regulatory Guide, the licensee referred 5

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, the inspectors to Institute of Electrical and Electronics Engineers,-In (IEEE) Standard 387-1977, " Standard Criteria for Diesel-Ge'nerator Units' ,

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Applied as Standby Power Supplies for Nuclear Power Generating Stations," ,,

which defined the " diesel' generator unit" to include the cooling system ,

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starting at the point where the cooling medium is introduced to the t -

diesel generator unit. However, the inspectors noted'that'the IEEE ' !,, i, Standard definition of the " diesel generator unit" was not/ fully ,

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o compatible with the Regulatory Guide definition becausehfor. example, the '

Regulatory Guide clearly included the diesel. generator output breaker"as 4 part of the unit and the IEEE standard excluded the breaker ' ' -

The NRC's interpretation remained that Regulatory Positions C.2.e.(5) and- [

C.2.e.(8) applied and the events should be counted as valid' failures. In addition, it was the NRC's position that the " cooling water system up to -

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the' supply" should include the service water supply and return valves to the heat exchanger. The licensee relogged the events as valid failures,u i and implemented the once per 7 days test program for Division 11 required by' Technical Specifications. The licensee reported the events to the NRC ' '

as LER 90-011. The LER also contained the information required pursuant-to Technical Specification 6'9.2, "Special Reports," concerning the valid

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test failures of the diesel generators. The corrective actions discussed in the LER will be followed up in a later repor <

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e No violations or deviations were identifie . Safety Assessment / Quality Verification '

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- Licensee Event Report (LEf') follow-up (90712 & 92700) into Field Drawings Results - .noperable _

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Nuclear System Protection Syc.em Inverter - .

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89-038-01 Error in Transcribing a Design Calculation "

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into Field Drawings Results in Incorrect ,

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Connecting of Power Supply to Nuclear System .

Protection System Inverter ,

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b.; TMI Action Plan Requirement follow-up- (25565) '

C (Closed) TMI Item III.D.3.4.3: Control Room Habitability This item was previously discussed and closed by the Staff in

'Clinton's Safety Evaluation Report, NUREG-0853, Supplement No. 1, 'l Section 6.4. This item remains closed, and is listed for documentation purposes in this repor . No violations or deviations were~ identifie . Management Changes

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On May 14, 1990, the licensee announced that Richard Gill had been named to the newly established position as Manager - Projects and Assessmen In his capacity, Mr. Gill will report to Mr. J. Perry, the Vice President. Mr; Gill had previously worked at.the Clinton station as an engineering consultant and most recently as a member of Mr. Perry's executive staf On Mey 21', 1990, the licensee announced that Patrick Yocum had been named to be the Director - Plant Operations reporting to Mr._ J.' Cook, the

, Manager - Clinton Power Station. Mr. Yocum had previously served as the Supervisor - Plant Operations and most recently as the Director -

Maintenance and Technical Training,

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. Unresolved Items

Unresolved items are matters about which more information is required in >

order to ascertain whether they are acceptable items, violations, o i deviations. An unresolved item disclosed during the inspection is '

discussed in Paragraph . Open Items i Open items are matters which have been discussed with the licensee, which i will be reviewed further by the inspector, and which involve some action i on the part of the NRC or licensee or both. An open item disclosed  ;

during the inspection is discussed in Paragraph 3.b(4).  ;

1 Violations for which A " Notice of Violation" Will Not Be Issued i The NRC uses the Notice of Violation as a standard method for formalizin the existence of a violation of a legally binding requiremen However,'

because the NRC wants to encourage and support licensee's initiatives for i self-identification and correction of problems, the NRC will not generally issue a Notice of Vio*eation for a violation that' meets the criteria of-10 CFR 2, Appendix C,Section V.G. These tests are: (1) the violation was identified by the licensee; (2) the violation would be  ;

categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including  :

measures to prevent recurrence, within a reasonable time period; and (5) it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violatio Violations of regulatory requirements identified during the inspection for which a Notice of Violation will not be issued are discussed in Paragraphs 2.c and 3.b(1).

12., Meetinos .

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NRC Commissioner Visit (30702) -

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, On June 6,1990, NRC Commissioner James R. Curtiss visited'th ,

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plant. The Commissioner was accompanied by his.Technica1' Assistant,  ; ,

the Deputy Regional Administrator, and the Director ofEthe Division  !

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of Reactor = Safet The visit included briefings .by' the inspectors,; *

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a plant tour led by the licensee, observations of a; portion of NRC-administered operator requalification exams, andia meeting wit i4

licensee managers denoted in Paragraph 1 of this report.< At.thei ,' ,

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i meeting, the licensee discussed their management team;.the, .

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y reliability engineering program, and their self assessment program, i  ;

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u- Management Meetinos (30702) ' ' '

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On June 18, 1990, Messrs. J. G. Partlow.-Associate Director for

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Projects; D. M. Crutchfield, Director, Division'of Reactor _ Projects

- III/IV/V and SP; J. A. Zwolinski, Assistant Director for ,

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Region III Reactors; J. N. Hannon, Director, Project Director III-3; J. B. Hickman, Project Manager - Clinton, all of NRR, and E. G. Greenman,' Director, Division of Reactor Projects and R. D. Lanksbury, Chief, Reactor Projects Section 3B, bbth o . Region III, and the NRC senior resident inspector met with licensee managers and supervisors denoted in paragraph 1 of this report at

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NRC headquarters in Rockville, Maryland. This meeting was held t t discuss licensee recent operating performance, progress on 1990s .

initiatives and_ plans to improve performance during the upcoming RF- m c. . Exit Interview- (30703)

The inspectors met with-the licensee representatives denoted in Paragraph I at the conclusion. of the inspection on July '3 '

Theinspectorssummarizedthepurposeandscopeoftheins.1990.Lpection and the findings. The inspectors also discussed the likely' rs

informational content of the inspection report, with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any_such documents or; procesr,es as proprietar ,

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