IR 05000461/1990001

Safety Insp Rept 50-461/90-01 on 900203-0316.Violations Noted.Major Areas Inspected:Licensee Action on Previous Insp Findings,Operational Safety,Event Followup,Radiation Controls,Maint/Surveillance,Engineering & LERs

ML20042E345
Person / Time
Site:	Clinton
Issue date:	04/03/1990
From:	Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20042E339	List: IR 05000461/1990001 ML20042E336 ML20042E340
References
50-461-90-01, 50-461-90-1, NUDOCS 9004200651
Download: ML20042E345 (19)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

e I

i Report No. 50-461/90001(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,1111nnis Inspection Conducted:

February.3 through March 16, 1990 Inspectors:

P. G. Brochman G. F. O'Dwyer l

S. P. Ray bf

$O Approved By: M k

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Reactor Pro ects Section 3B Date

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Inspection Summary Inspection from February 3 through March 16, 1990 (Report _No. 50-461/90001(DRP))

Areas Inspected:

Routine, unannounced safety inspection by the resident inspectors of licensee action on previous inspection findings; operational safety; event follow-up; radiation controls; maintenance / surveillance; engineering and technical support; licensee event reports; regional requests; evaluation of licensee's self-assessment capabilities and quality assurance program; and safety evaluations.

Results: Of the 10 areas inspected, one violation was identifie regarding the review of safety evaluations.

The violation involved an improper safety evaluation in which prior NRC approval of a required change to Technical Specifications was not obtained before a local leak rate test procedure was revised.

This violation was of low safety significance because the licensee was aware that Technical Specifications were required to be changed and had initiated actions to change them.

The procedure was never used in a way that caused a Technical Specification requirement not to be met.

A deviation was identified by the licensee in the area of engineering and technical support involving failure to meet the requirements of Regulatory Guide 1.97 regarding calibration of accident monitoring instrumentation. A Notice of Deviation was not issued due to the minor safety significance.

An unresolved item

.i was identified in the area of review of safety evaluations involving use of alternate decay heat removal methods.

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4During the inspection;. strengths 1were noted in plantLoperations concerning

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control of plant activities, communications, responsiveness to NRC concerns,.

and evaluation of unusuals conditions., Execution ofi"first time activities

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was mixed.. Radiation controls during-outage activities were conside' red a

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= strength.

Maintenancefand surveillance activities were' mixed with planning '

i;i and execution of outage activities Limproved from previous' outages.

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performance in the safety assessment-and quality 1 verification area was mixed.

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' with significant improvements noted in'the corrective action' program but

. weaknesses in one particular corrective _ action activity. < As' noted above, a :

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violation was' identified in:a safety evaluation,-but the violation actually!

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DETAILS l

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p! ' PersonsJContacted

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y Illinois Power Company

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4 *#J. Perry,iVice' President,

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

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• fR. Freeman, Manageri Nuclear Station Engineering

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sg3, Hall Director, Nuclear. Program Assessment

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• 1. J. Miller, Manager, Scheduling and: Outage Management

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'#J. Palchak, Manager, Nuclear Planning and; Support y

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• 1. J. Palmer. Manager, Nuclear Training

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- #F.5Spangenberg, Manager, Licensing;and Safety-

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• 1. R.~iWyatt,; Manager,' Quality Assurance

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• 1. J2 Greenwood, Manager,PowerSupply-u

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The inspector also contacted and interviewed other" licensee and e

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contractor. personnel during the course of this inspection,

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1. Denotes those present'during the management meeting on february 28, 1990.

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• Denotes those present-during the exit: interview on March 16, 1990.

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2.

Astion on Previous Inspection Findings (62703.) _

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(Closed) Open Item (461/89038-02): - Posi-Seal' Butterfly. Valves Insta11ed'

Backwards to Preferred Flow;0rientation Recommended by Vendor;

This item was previou. sly discussed in Inspection Report No. 461/89038, r

Paragraph 5.b.

The issue involved the inspector's observation that.

some Posi-Seal butterfly. valves were not' installed in the; preferred-

= orientation as indicated by the vendor tag on the valves.

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The inspectors were supplied with a record of a telephone conversation

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between the licensee and a representative of Posi International which

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confirmed the licensee's position that the. orientation,was not critical.

Posi International indicated that all Posi-Seal valves, that were

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supplied with a soft seat and elasto'mer back up ring, allowed.

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bidirectional orientation, regardless of the preferred orientation

tflow tag.

The vendor indicated that the preferred flow orientation o;.was critical only on valves which had a Teflon back up ring and/or O

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metal seal ring.

Clinton had no such valves.

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The inspectors noted that: the plant was required to shutdown during this

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-Ph, report period due to seat leakage of Posi-Sea 1Lvalves in theldrywell!

4, < 1 ! 4 < ~ purge. system and that there were several problemt with critical path work t

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> during the outage due to leakage past Posi-Seal valves in the' service-

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water system.

However,'the leakage-through Posi-Seal valves did not seem'to be related to orientation. ' This item is considered closed, t

. No violations or deviatiens were identified.

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Plant Operations

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The unit operated at power levels up to 100% until-February-12,1990, when it was shutdown due.to Technical Specifications. The-licensee declared a Notification-of Unusual Event (NOVE) and terminated it when.

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tne: unit was placed in COLD _ SHUTDOWN.

The plant remained shutdown j

through the remainder of the inspection period.1 The licensee made a.

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transition from a forced shutdown to a preplanned maintenance shutdown i

on January 21, 1990.

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.'1 The inspecto'rs' observations during the report period indicated that_

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' plant operations were conducted in a safe and conservative manner.

!The NOUE and' subsequent plant shutdown were properly conducted.

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discussions with plant operators and superv_isors during the1 maintenance

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outage,'they. demonstrated familiarity with' operations and plant

conditions, positive control of evolutions, and knowledge'of the status i

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of engineering'and licensing questions.

Communications b'etween

operations, maintenance, and outage control l personnel seemed to be

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improved over previous outages. -

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Operational Safety (71707)'

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The inspectors observe control room operation, rev'iewed applicable'

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February and March 1990.. During these, discussions and observations, i

the. inspectors. ascertained that the operators were alert, cognizant i

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j of plant conditions, and attentive to-changes in:those conditions,

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verified the' operability of selected emergency _ systems, reviewed _

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tagout records, and verified the proper return to service of

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y affected components.

Tours of tne plant were conducted to observe

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equipment conditions, including potential fire hazards, fluid leaks, t

and excessive vibrations,'and to verify that maintenance requests had been initiated for ' equipment _ in need of_ maintenance, i

The inspectors verified by observation and direct interviews' that-the physical security _ plan is being implemented in accordance with

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the station security plan.

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The ' inspectors observed plant housekeeping / cleanliness conditions j-and verified implementation of radiation protection controls.

The inspectors also witnessed portions of the radioactive. waste system.

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controls associated with radwaste shipments and barreling.

Thel obs'erved facility operations were verified to be in accordance with-the requirements established under Technical Specifications, 10 CFR, and administrative procedures, f

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(f (1)' On February 6,1990, the inspe_ctors _became aware of a finding

.at the Limerick plant in which it.was. determined that actions

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T taken by the operators after'a' reactor scram caused byfa

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electro-hydraulic control-(EHC) systemtfailure could result r

in the plant being outside the design basis as described.in-

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the accident analysis-in the Safety Analysis. Report. LThe

inspectors, determined that the Clinton plant could be subject -

'to the same' problem and informed licensee management.-

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- Off Normal Procedure, CPS No. 4100.01, " Reactor. Scram,"! required

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in-step 3.1 that the: operators place the mode < switch'in SHUTDOWNi j

as the first immediate operator action'after a: reactor scram.

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Clinton Updated Safety = Analysis Report,-Section 15.1.3,l" Pressure" (

Regulator Failure -:Open," assumed that after a failure of the-pressure regulator such that the turbine bypass valves were fully open, the reactor. scrammed on highLvessel. water l level at.

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2.64 seconds-into'the event and theLsteam flow transient was

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stopped at 36 seconds _into the event by.a' main steam isolation

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on low turbine' inlet ~ pressure.. However,-if the operator-performed as expected and placed the mode switch in SHUTDOWN shortly after the' scram, the main steam, isolation would not occur because the main steam line isolation feature was bypassed when the mode switch was-in SHUTDOWN.-

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The inspectors were concerned'that,,1f operators followed the reactor scram procedure as written.it might' result:in an

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' overcooling. transient in the case of an EHC system failure.

This transient could have placed the plant in an unanalyzedi

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condition.

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The inspectors brought this condition to the' attention of plant

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j, management and the licenseeLimmediately wrote Condition Report

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1-90-02-021 to document their investigation and ' corrective.

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actions..They also immediately issued a night order to the j

3 operators alerting'them to the potential'unanalyzed condition-

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'and reminding them of step 4.2.2 of the reactor scram procedure

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which required that the cooldown rate be closely controlled.

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The, night order directed that' main steam isolation valves'be.

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' 11osed.after a reactor trip if bypass' valves failed-open or

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excessive cocidown wasjobserved.

The licensee's actions.

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'demonstratej timely responsiveness to NRC concerns.

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The licensee was still in the process of determining whether'

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permanent; corrective actions would be required based on reviews

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by' General Electric 'and the BWR Owner's Group.

Prelimiilary

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results indicated that the transient was less severe than other b

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similar analyzed transients, such as.feedwater controller

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') failure with maximum demand,'even if the main steam line f

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, -isolat. ion did not automatically occur.

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F (2) On. March 6,1990, the licensee's-trainiiig department identified;

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that when containment spray was initiated using the manual pushbuttons prior to the 610 second' timer completing its

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sequence after a' loss.of coolant accident,: the residual heat

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exchanger bypass valves (IE12-F048A/0) would, cycle open and?'

closed continuously until the~ timer: timed out.

This phenomenon

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'was initially observed on the simulator but a review of<the

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electrical prints for.the actual plant 1 indicated that the

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valves in the plant would-also operate similarly.

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'4 Condition Report 1-90-03-057 was imme'diately; issued to requ' ire

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an investigati_on of the. safety significance,-generic'

. implications, and corrective actions for.the~ condition. The

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licensee placed caution tags on the containment spray manual'

. initiation pushbuttons to: alert,the operators to' the concern.-

'The fact that this condition was observed in the' simulator and'

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. recognized to have potential significance.in the actual' plant-

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was commendable.

' i (3) Execution of "first time" evolutions.were improved with:

specific examples'being the movement ~of:the' reactor. water l

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cleanup (RT) filter septa material'and modifications"of the

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m RT system; An exception was noted in,an incident in which a l

barrel of radioactive waste was, dropped while moving it withi a

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crane using an untried;techriique.

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,0n February 28, 1990, while' attempting to lift a drum containing

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Evalve'1G33-F013B in the radwaste truck bay,'the lifting l

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N mechanism failed and the' drum fell 12. feet to the floor. 'No

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personnel were injured.

No-contamination = occurred.

The drum

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was being used to transport valve.1G33-F013B from the reactor

water cleanup system, where it had just been removed, to

,b radwaste storage. ' Temporary-shielding was hung on_the' drum, i

i duetto the high dose rates from the valva I

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ilAfter7theeventthelicensee-conductedacritiqueL(number

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RW-90-0011)'-and identified the' root causes. The licensee I

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' ' believed;the-root cause to be the use of the wrong type'of Q; 1

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lifting device on the drum,. compounded by the interference of

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' the' temporary shielding.

The inspectors attended the critique

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Based on

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1these observations'the inspectors identified additional

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,contributingscauses to this event.

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First, ~the purpose of the drum was' not clearly' defined., The

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purpose of the drum was not just to act as a shielded container s

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for the: valve, but a shielded container which had to transport

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the valve from the work. site to,the radwaste storage area and-a

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i nbe lifted-by a crane.

Since.this~ purpose was not understood, J

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'the~only engineering analysis made on this drum was for ALARA

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purposes-(the amount of shielding required).

Consequently,-

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c Tinteractions between the shielding and the ability'to lift the

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~ drum using the normal scissors clamp on the radwaste building' -

y crane were not included in the-design of the shielding.

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1shieldingconsistedof,leadblankets,which'werehung,onthe'

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f outside'of the drum;.and were supported by "S"shooksJ which"

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hung'over the top of the drum,$ preventing the drum top:and

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support ring from being installed.-

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Second, since-the use of th'e normal lifting device (scissors i

' clamp) was precluded by the-design of1the shielding. an',.

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p, alternate lifting device was selected (a plate steel clamp)., Af,

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plate steel clamp was designed to' lift-flatssheets:of. metal,7 s

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not to lift" drums - Licensee' training.on lifting and' rigging

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did discuss'the;use'_of plate steel clamps;.but only on sheete

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metal. ~However, common practice duringcconstruction at:Clinton~

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'had been to use! plate steel clamps' totliftsopen drums. EThe usel

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of a plateisteel clamp 1was not. appropriate becauseethe center-

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'of gravity ofethe load was'notfunder the center of the lifting'*

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point, which caused the loadtto' shift as'it was beingimoved;;

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Iandcthe side"of the drum wassnot designed to support'all of the

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4" weight at a' single point, but rather to_ distrib.ute,thel load, 4

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around the circumference. The use of.the_ plate steel-clamp top

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. lift a drum was not an approved rigging method.~

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the crane hook did not know' how to(use the plate steel clanip. _

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The worker who moved the drum to the'radwaste building showed

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initiative in;inspectina the< clamp <tolmake sure11t'was" operable

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and showed the:radwaste aperators and supervisor;how the clamp

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was to.wo'rk.

He then hooked it?to!the drum, after checking l

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g that it was. engaged.

He statedfin the critique that he did not;

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feel that he was responsible 1for the -1ift, but was just helping o ut'.

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P In summary, this wa[a first time evolutid, using inLincorrect:

design, with an improper device, byiworkers whoshad-not been:,

trained in the use of the' device, that.no one wasMresponsible for".

However, there were some positive aspectsJto thisl event.

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The radwaste depai-tment did performra ~ dry run on an empty drum-

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before this event. _ At that time, the' decision to use -the plate.

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clamp was made because'the personnel. felt that"other.fmethods ~

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U such as barrel slings would not. work.. Additiona11y', the area-around the drum was cleared so that if it fell no one would be injured.

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t No violations orJdeviat; ions were' identified.

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Onsite Follow-up of-Events (93702)

y The inspectors performed onsite follow-up activities for events-

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which occurred during February and March 1990.

These: follow-ups included reviews'of. operating logs, procedures., ConditionrReports,

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~ Licensee Event Reports =(where'available), and interviews with.'.

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licensee personnel.

For each' event, the inspectors developed a chronology, reviewed the' functioning of safety systems required

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byplantco'nditions,andreviewedlicenseeaclio$sto: verify-I

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? consistency with procedures, licen,se conditions, and the nature

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.of_ the event. > Additionally,' the inspectors v'erifiedithati the=

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licensee's investigation' had identified the root causes of equipment

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- malfunctions' and/or. personnel errors"and~that the licensee: had

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-taken. appropriate; corrective actions prior to restarting the unit.E

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f; Details of the. events and the licensee's corrective-actions

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A developed through inspector follow-up-are'provided below:

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C On> February 12,1990,.at 2:30 p.m,, the licensee failed a. local h

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' leak rate test (LLRT)Lon containment penetration MC-102-(drywe11=

i-purge).

At'4:30 p.m., when the. leakage through this_ penetration

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was combined with previously< measured: secondary containment

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bypass-leakage, theLtotaleexceeded the-allowed leakage'of. Technical

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Specification 3.6.122.d andLtheslicensee entered'an ACTION statement ~

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trequiring;that primary containmentt integrity; be resto' red within - -

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11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> or be in at' least1 HOT SHUTDOWN'within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s-and

.)

,

in' COLD SHUTDOWN within;theifollowing124Lhours.

.

'

LAtL.9:00'p.m.,!when efforts'to stop the-leakage failedFthe operators.

a started an/ orderly plant: shutdowniand. declared a' Notification of -

Unusual Event'(NOVE).c During.the shutdown,the. licensee experienced -

a problem with controlling'feedwater flow ~using the. condensate.

i booster pumps.- Reactor vessel level exceeded thsilevel-8 trip:

,

n steam lines

, setpoint and increased to within six inches.of theimaidetermined

before being ; brought.under control.' ' An tinvestigation

that the. discharge Lisolation, bypass / valve for' thed'A" turbine driven -

feedwater pump (1FW003A) wasan an intermediate position. <The'

.

'..

valve had beenttagged=out and wastthought"to be closede Drawing

'

M05-1004-1: indicated that the valve: failed closed so it had been

,

-tagged closed by failing the instrument air' supply.. Howevery after

-

'

' the: incident, it was discovered that the. valve actually failed

'

,

,

L"as-is" on loss of air and only failed closed on lossiof electrical i

power.

Pre lssure against the valve disk had apparently cause the

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valve to come partly open.. Condition Report.1-90-02-043 was-

'

'

initiated to investigate the(incident.;

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.

,

The plant reached HOT SHUTDOWN at 5:25 a.m. on Februaryr13 and. COLD E

,

,

SHUTDOWN at 9:34.p.m. cThe NOVE was" terminated at 9:50 p.m. the'same'

day.

The' licensee sub'm'itted a written summary of the event to the" f

.

"

NRC on February 20, 1990, and followed up with LER 90-001 on 1-

.

'Marchi13, 1990.

The licensee attributed the cause of the LLRT:

'

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'

Efailure to leakage through 36 inch Posi-Seal butterfly isolation'

.

,

-

valve IVQ004B.' The leakage was believed to be caused by a buildup.

of dirt and dust'on.the! seating: surface due to failure to remove the:Cosmolene preservative'from -the valve seat prior.to the valvefs'

.

'

'

initial 3 installation during construction.

Both the inboard and-

+

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outboard penetration valves were removed. Thecinboard valve was

,

refurbished and the outboard valve was replaced.' l Subsequent' testing

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of the penetration ~ demonstrated-that leakage was well within, specifications.

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f-No violations or deviations were identified.

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Radiation Cont'rols (717.07)

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Severaleventsinttiis-inspeci,ionperiod.involvedradiologlcal' cont'rols.

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Most of the events occurred during the outage which begari on February 13, i

,'

1990. The inspectors ^ observed aggressive-involvement ~of > radiation

~

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protection personnel at'all: levels-in attempting to. minimize exposure and contamination; Communications seemed to.be improved <over, previous-

' outages and. worker radiological awareness appeared to be~ heightened.

The

.

specific radiological events in this inspection' period were followed up, r-

,

-

,

by regional inspectors-in Inspe.ction Report No.r461/90004;

.

,

,

a.

While the inspectors were observing a local'le'ak rateLtestiof the5

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inclined fuel transfer tube blindt flange, the"Radiat. ion Protection ( "

.-

Shift Supervisor-(RPSS) was;notifiedithat'a pipefitter hat

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disconnected a= hose used to~ drain part of the fuel pool > cleanup.

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. system piping without a radiation protectioni(RPh techn'ician; f

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present.

This resulted in contamination-of about two, square feet'

'

.

on the floor below the hose connection.

The RPSS went to the:jobt.

,,

-site and determined that the pipefitter and the test director had'

violated the applicable radiation work permit (R_WP) in'that the.

• a RWP required that 'an.RP technician be'present when the system was

-

breached.

The RPSS immediately ordered that the. system be placed in a safe condition and escorted the.pipefitter and test-director out of the Radiologically Controlled Area to' his' office.

The RPSS's investigation determined thatuthe_ root cause of the incident was a misunderstanding of the RWP requirement and that the pipefitter was; inexperienced.with working with potentially contaminated systems.

a The RPSS independently took the following corrective actions:-

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explained the requirement to haveia: technician,present:when

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. breaching a system,

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counseled the pipefitter on worker responsibility-:and RWP.

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compliance,

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' counseled the test director on his responsibility'for work'er

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safety,

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changed /the RWP to clarify the requirements;for system breaching,

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required that both individuals review the RWP. proced'ure,

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hafthe area around the tes't_ decontaminate'd, [

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documented the event on a Radiological? Improvement? Report, and

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informed the Outage Coordinator and Director - Radiation

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Protection of his actions.

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~RPSS,. good immediate corrective actions,-and decisions beir.g madefat (

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thetappropriate level.

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SurveysMf; reactor recirculation piping completed during February '

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i, indicated that; significant; accumulations of Cobalt-60 were building;

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a e ie up on'the piping.

Radiation fields.near Clinton's recirculation I

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. piping were the highest. in the industry after the first fuel cycle:"

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s and were increased-even mere during the;second~ fuel cycle.

These'

,

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dose rates' represented a significantichallenge to the ALARAl program Land the licensee was considering actions'such as chemical-

'

-decontamination and: treatment systems in an attempt to. reduce the-

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dose rates.

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No violations or deviations were identified.

,

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-

5.

Maintenance / Surveillance (61720, 61726 & 62703)

,

Station maintenance.and surveillance activities;of the safety-related.

,

. systems and components listed below were. observed ortreviewed to

'

p ascertain that they were conducted in'accordance'with. approved'

'

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procedures,. regulatory guides,-and industry codes' or standards, a,fidl L

'

~

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in conformance:with Technical Specifications.

The following items?were considered'during this'reviewi the limiting-

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conditions for operation.were met while affected components or systems were removed from-and restored to. serv. ice; approvals were obtained prior to in_itiating work or testing; quality; control records were, maintained; parts and materials used were properly certified; radiological"and fire-J

,

prevention controls were~ accomplished in accordance with approved'

procedures; maintenance and testing were1 accomplished by qualified.

personnel;. test instrumentation was within its calibration interval;'

functional testing and/or calibrations were performed prior'to returning.

,,

components or systems to service; test results. conformed with. Technical

'

Specifications and procedural requirementsLand.were' reviewed'by personnel.

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'

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 weretreviewed to

,

P determine the status of outstanding jobs and.to assure;that: priority was-

-

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. assigned to safety-related equipment ~ maintenance which may affect system

,

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performance.

The followin7 maintenance and surveillance-activities were-I

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' observed:

-

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Activity Title

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" CPS 9861.02-Local Leak Rate Testing G

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'N MWR D12948 Remove and Repair.1VQ004A yu m

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MWR D06457, Fuel Trolley Repairs

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MWR,C51443'

Containment Area Radiation Monitor Repair

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MWR'004858 Containment Electrical Penetration r

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~ Modification'.

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'MWR1006457 Tack Welding of Fuel Handling Bridge

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- Activit'y.

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I'spection'of.DivisionIIIVX-

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MWR'008328 n

CPS 2800,10 Service Water System Flow Balancing

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MWR D08284 &1

. Diesel Generator Heat Exchanger. Retubing:

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008385

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'Sev'eral other maintenance and surveillance act nitie5 were ~ observed ~

during.a Balance of. Plant. Maintenance Team Inspectiom conducted-during'~

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this inspection period.' The results of that. inspection were documented?

'

-

,

in Inspection Report 461/90002.

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For CPS 9861.01, " Local Leak Rate Testing Requirements, thei linspectors noted that there were'no procedural; requirements to.

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correct'the indicated' leakage rate as measured on the air rotometers for the temperature, pressure,;and specific gravity of the test..

.

- fl u'i d.

The calibration: data sheet for the' air rotometers stated:

,

that corrections must be' applied if standard conditions lof.70

.F.,

9.0 psig.and specific-gravity of 1.0 (air) did not exist.

There:

,,-

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'were no provisions for those corrections in the-local; leak rate-

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test (LLRT) procedure.

Discussions:with technical staff aersonnel a r

indicated that this-issuelhad been previously considered aut-that" the correction factors: tended to be'small and' tended to.. cancel each

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other out because the air'used-in the-tests was usuallyislightly.

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' J above 70 F. but also.was slightly above 9.0 psig;since the test

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- ~ acceptance criteria required that test pressure be 9.1. to-9.9 psig.

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The inspectors 'were provided.with calculations that showed that thej-

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errors-introduced by those conditions would be* expected to be le s

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.The inspectors, pointed out that certain LLRTs~were accomplishedtcfp

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using a high pressure nitrogen' bottle ~as a pressure source and in Y.

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those cases all three.of tne correction-factors.would probably be'

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non-conservative and could result-'in an error of more than~ 5L -The

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Mtechnical. staff personnel agreed-that-the LLRT proce' dure-should be

'

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' changed to' apply the correction factors for' nitrogen' tests and b e '-

,

'

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comrpjtted to submit a revision which was being tracke'd by Clinton ' ' <'

Commitment Tracking No. 52827

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b.

For the repair uf the containment area rad'ation monitor i

.(1RIX-AR022) the-inspectors.noted that the System Impact-Matrix in the. Maintenance Work Request (MWR) was written and approved about ~a-

-year ~ and one half earlier., The-matrix had' been reviewed bylthe

.l

'

Assistant Shift Supervisor ~ a'little over sixLmonths'earliero As'

, documented in Inspection: Report -461/89038,. Paragraph 3.b.(2),-~as a: _

.-

~

result.of'an event on January 4,1990, which;resulted in tripping of,

the "A" reactor recirculation pump, one of the' corrective actiots

,

was to be that any System Impact Matrix more than six months old be reevaluated before use.

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Theinspectorsfquestionedthe:05I: technicians;onth$jobtksee

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' whether they, were; aware of' the requirement.

They-stated that they

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p-realized that the matrix'was more than.six months old and had-s pointed that out to the Staff Assistsnt" Shift;Supervis'orl(SASS)

+

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'when checking in on the_ job.~.. The SASS noted thatsthe' work was~on r.

nonsafety-related equipment that had no impact on the plant'and/

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allowed the work to proceed. _However, the SASS's reviewfofithe'

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impact matrix was not' documented'so there was no objective evidence t

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i-being carried out.

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The inspectors notedlthat -the-C&I -technicians' working the, job were

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wearing safety belts when. working.inside the railing;arou'nd the -

.

,

Je reactor pool. ' The failurei,to wear safety: belts had been aLproblem; a

',

.in_the f,irst refueling outage.

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While observing preparations for' tack. welding on the fuel handling h

bridge the inspectors.noted that the welder was about'to weld using

,

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-covered electrode filler material,..However,<the. Weld-Requirement 1

, Data: Form,~ CPS.No.. 1509.01F001,.specified the(use of bare wire;.

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filler, _ The welder stated that he, had not noticed' the requirement l

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for' bare wire and was not qualified for. bare wire welding.'!The1jobi

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supervisbr stopped the job and: brought the Weld Requirement Data:

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Form to engineering for review'.'

Engi'neering personnel-later -

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' determined that; covered electrode welding was2 acceptable for that l

_ application.

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The inspectors discussed this incident'with management asian examp e

p of poor work control.

The licensee took' prompt corrective" actions

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to prevent recurrence and inform' appropriate personnel of thes f

.

lessons learned.

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The, inspectors observed'severalfoffthe activities' associated with

.

cleaning, inspecting and flow balancing of the service water system'

in response.t_o NRC Generic 1 Letter'89-13.

Thiscissue'is~ documented

-

i further in' Inspection Rep ^ ort No.;461/90005.

j In additiosto the above maintenance and stirveillance activitiese the inspectors observed the overall. control of ' activities du'ing the major r

f

,

maintenance outage during the inspection period.

The' inspectors attended

-

pre-outage planning meetings,' pre-outage.-information seminars, daily

'

.

outage control meetings, and several jpecialJoutage: activity meetings.

.

In general,.the outage activities appeared to be well planned and y

coordinated.

0bservations by the-inspectors as well as comments by workers in-the field indicated that communi' cations, decision making',

control of contractors, and response-to events were improved over

'

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previous outages.

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No violations or deviations were identified.

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6 h Engineering and Technical Support (71707)

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On February 21, 1990' the inspectors noted a jam nut missing from, l

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the seismic support for.the drywell purge inlet; valve (1VR006A).

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N The condition was brought to the attention of a. licensee director'

% Eho% as with the inspectors.

The. licensee issued Condition Report 3 M.

f,(

.f 1-90-02-050,to document the condition and assign an investigation of

'/ G v +, m the safety significance. 'Immediate corrective action consisted of

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issuin'g;MWR D06650 to install the jam. nut and take as found data for [ @

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j*ha determination of seismic operability.The licensee's evaluation t

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of4the CR indicated that the breakaway torque on the nuts was:

f greater th'an the torque required for the hanger to meet its seicmic ~ [\\'7

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7 ndesign-requirements.

The licensee retorqued the nuts.and installed

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-r jam nuts. ? Based on this evaluation the inspectors have no further-

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concerns and this issue is considered closed.

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10n(March 17,1990, the licensee identified that certain accident-y cfd

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mon'itoring instrumentation-required by Regulatory Guide 1.97 had not I

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been evaluated for calibration. frequency.

Several; instruments had

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not been calibrated since initial construction and were not-in the

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identifiedbyengineeringpersonnelwhowereconductinga.

.

preventative maintenance program.

This condition was initially i

.

reverification of vendor s-environmental ' qualification (EQ)

requirements.

This' action was being taken in responseLto commitments made in response to violations identif.ied'in the licensee's EQ program.

Although the failure.to-include the instruments in a' calibration

,

program is a deviation from-the licensee's commitments to Regulatory Guide 1.97, the condition was self-identified as a result. of licensee's actions in response'to previous violations and was of minor safety. significance considering the relatively short time the plant has operated since'the. instruments were. calibrated _during construction.

Therefore, a Notice of Deviation.was not(issued.

No violations were identified.

One deviation was identified.for-which a i

Notice of Deviation was not issued..

7.

Safety Assessment / Quality Verification'

,

a.

Licensee Event Report (LER) Follow-up-(90712 & 92700)

'

.

Through direct' observation, discussions 1with licensee personnel, and review of-records, the following LER was reviewed to determine that the reportability requirements were fulfilled, immediate corrective "

action was accomplished, and corrective action-to preventirecurrence had been accomplished-in accordance with Technical Specifications.

.

Based on the inspectors' review, the following LER is closed:

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Violation of the Plant's Technical n s

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, J '. ' ' . Specifications Du'=to Utility _ Personnel + e A , Error Resulting from a Deficient Procedure

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. , - Nol violations or deviat' ions--wereLidentified.

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' Follow-uptof Region'al Requests'(71707) ' . + .. . , . ., ~

'In response to'a Part 21 notification from Automatic-Valve Company

E (AVC), the Region III' office asked that the: resident: staff-check , with licensees to. determine if.'they hadl received -information ~

• -

_regarding the subject and wer2 takingcappropriate actions.- 'The t , . ,' .' licensee informed the inspectors that-a checklof-.Clinton's? ' ,

" configuration documentationiindicated.that they had no AVClequipment.

' ons ite.- . _

. ' ,. . ; . ._ No' violations' or deviationVwereL identif.ied.i . , -

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Evaluation' of Licensee = Self-Assessment Capability and Quality , c.

. Assurance Program (40500 & 35502).

> . a , ,, o (1) As a response to NRC_conderns',~ peer-evaluations;<and self ,, L ' -identified findings, theflicenseeirecently. initiated several . ' actionstte improve their' corrective action program,- Results of thoseiiniti'atives began to becometapparent during this ~ , inspection period.: y , , y _ . ~ , < - , The ' licensee created a' taskzfo'rce to-review the corrective " actions. for tall' Condition Reports, LERs, violations,;and!QA^ . . findings which were documented during the.first 11) months-of!1989.

With most'of the reviews completed,cthe licensee' ' < discovered that approximately'18% of the1 corrective actions.

> were either not adequate to: prevent; recurrence of an adverse

condition or'were not properly' completed and documented; New ' , ' Condition Reports were. issued to, tracki correction of-these , ' problems ~in most cases. The. lessons learned from:these reviews l' . werelbeing.used:to strengthen the ! corrective? actions for more I J ' recent findings.

. . , P .. .. L . ,

The' inspectors also noted' improvements in"theiquality of recent.

- Condition Reports.. Most of_the reports contained immediate 'b L i corrective actions to correct'the condition orLprevent recurrence even;before a detailed evaluation was completed.' - , ' In addition, more:of the Condition Reports were-evaluated for n s ' generic implications,and the: scope of the'co rective actions , was_cxpanded where appropriate.

Departments demonstrated an ', improved sense 1of," ownership" ofx.the Condition Reports and were ,

more informed.about the status ofJ corrective actions.

In t n ' o c t addition, personnel were observed ~ to be doing a more complete ' . f job of verifying that' corrective' actions were complete.before

• -

closing the Condition Reports.

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c.y; .g ,- One: exception totthe above" observations was.the>1icensee'sc k , nc X ' ' , actions in response tolloose covers'.found on Rosemount-S l ' ransmitters. =The;conditiontwasfinitially discovered shortly? t t , before al peactor startup in November 1989, Land. action taken tok r , . j~. check other Rosemount1 transmitter covers * was conducted dsingian' + "i K - ' ' -inappropriate techniquemas poorly docume'ntedt and appeared 'toJ p^4 _ 'T be~ incompletely planned and scoped..The) licensee's tesponseito_ ^ . ; > q

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, s _ V this, condition appeared tocbe:very similar to'theiraactions7 p ; (1. % . ! e taken when unqualified buttnsplices were discovered ust_ prior? ! V ji[. t'o a reactor startup inJ1988.J Corrective 1 actions ta enifor.

'~ ' p? a _ithat event were'determitied to be1 inadequate,and resultedin i ', j'

m . j givilfpenaltye.This issue was;b'eing; followed4up byta regional i l, ybL f - -+ s 4.1_ q g Henvironmenta) qualification sp9cialistc > r, f a ,L,s -

s , a < . . - p - - ,, b, z u f i $ (2) JhYlicenjbe identifiedithree^ examples oflinadeqdt'e or .. , ' W W aA ? i !4 C* untimelf actionsErelating1to NRCLissues dUring(the; inspection /9

V( @E i ' ' ' , ' per'iodl 6A' commitment,s in responseito -NRC Bulletin"88-087 ?_ V

' f ' ,r y y 4 Q t[ f to"perfo'rm thermal < stress analysis on reactoh coolant system, T ~ "' m .' >d p,iping?was not completed as scheduled'and was"not,being

' y(( ~ < - i < l tracked. Actions' committed in: responsesto NRCvGeneric Letter e .- >3.

Y / 88-14 to verify the quality of instrume'ntr irNereg not given? M

'" ' a " the appropriate priority. to ensure complet. ion'asisched' led.I q t ' u + Review'of NRC Information. Notice"88-24< received'in May '1988 '

, y, - regarding the failure of air-operated = valves under certain 1 < ' ' - e C conditions.was not completed until March 1990. 'In'that case l?it: ' . _ was determined thatiseveral safety-relatediair4 operated valves: C - , at'Clinton could be' subject'to-failure. Ths'licenseets-actions" - to analyze and correct the dondition:were'beingtfollowed by a[' _ J~ regional environmentalLqualification specialist'/ i, ', * ,, . , , , . l ., ,.

, The licensee has-initiated a" review ofiitsccommitment tracking y J - l_ osystem (CCT) tojensure that regulatory commitment _s are being9 ~ ~ ~ met.

The licensee.was also reviewing the priority +a'ssigned.to; .i CCTs and the level--:of review necessary to change;CCT-due--dates.

~ > - ( .om' y a-(3) In response to th'e Operational Safety' Team Inspection- ~ ' n 1 ' (Inspection Report Nof461/89030)iand complaints'by, licensee n'". ~ g" personnel -about the timelinessjofTincorporating suggested;. m

• '

procedure changes into -p. ocedures,ithe licensee;i_nvestigatedT . "# L r .O L the procedure review program requiredjbyl Technical, ' ' c , - Specification 6.8.2.1 It was ~ noted.thati Administrative'., - k ' - Procedure CPS No. 1005.08, " Periodic Review of.. Station A ^ ~ v a , I ' Procedures and Documents," re' quired a biennialereview'of; all.. M ' L station procedures and doc'uments but?did ~notfrequire that:all i - " outstanding procedure Comment Control; Forms'be considered 1for ,, _. * _- incorporation during_thoseJreviews. -CPS!1005.08 further stated W that revisions of procedures' and documents 1shall fulfill _the;.. a} requirement of the periodic reviewJ - - s._ s ~ O' , g / f -. < .- . , z d,* ' '

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iThat practice?resultedlin suggested) procedure improvements ] - submitted:by plant' personnel <on Comment Control Forms'not being1 D ' ' . . considered forfincorporation_ for. periods of several years andc i . contributed to frustration:and: noncompliance with. cumbersome '7 a - .. , W or incorrect procedures.

Although taking credit for thex x biennialireview in the case'of a procedure' revision was-not a' slo 1ation of the Technica1'SpecificationsJ plant procedures," and was allowed by the licensee's commitment to'American?

- . = National > Standards Institute (ANSI) N18.7.-J1976., the practice.

" idid'not meet the. intent-of. ANSI N18.7', Paragraph 5.2.15 for' ' < . ; feedback of information' based; on 'use.- tThe _ licensee,was., ' .. ' ',. ' > ' . ' .considering atrevision to CPS 1005.08.to require:that all' ~

' - s outstanding ~ Comment Control Forms be considered during? thel 7

. ' " ' & j s .. . biennial review.

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' No violations"or deviations we~reEidentified.

q j , , .t= a q .( d.

Review of' Safety Evaluations (37700).

, f A - , ,

' (1) As the licensee made p' reparations to enter,the planned;

, . . ' maintenance ~ outage 'in' February s1990, the jinspectors became j c &

aware.that;the licensee' planned to'use a nonsafety grade y system to provide the alternate. decay heat removal,(DHR)' h capability required by Technical Specifications?3i4.9.1,.. ~H 3.4.9.2, 3.9.11.1,:and 3.9.11.2.

Neither.the Technical,l ' ' ' ... Specifications, Updated' Safety! Analysis. Report,= nor any1NRC1 , - generic correspondence-provided guidance on the design; . + requirements of alternate lDHR systems.

In addition, no' time restraints existed'for dependance on alternate ~DHR.' AcreviewT of logs for the firstarefueling outage i.ndicated that alternate.

' '! '~ , DHR systems depending on nonsafety grade equipment had been ,- used on several occasionss In on.e' case, two alternate DHR " , ' systems were used at:the same' time'for a 7 day-period due'to- , both loops'of residual heat removal 7(RHR),being dnoperable.

The inspectors were concerned that DHR mightjnot be available< ,' , following a loss of offsite power' accident',iseismic event, or other failure.orl accident.

The inspectors;als'o hadbother~ ,~ q s i

questions' regarding the requirement to demonstrateithe' *

. , alternate DHR: systems operable and the requirement to provide ~ . reactorecoolant circulation when. normar RHR or. recirculation. ~ ~ ' " g loops sereLnot in operation.

, - - - - .

Discussions with the licensee,-istaff, and regionalfmanagement

'did'not fully resolve the. issues and the inspectors drafted;a?

' request for technical assistance to-the NRR staff?' For-the e maintenance outage'in' progress, NRC management made the- -r ' decision.to allow,the planned use of alternate DHR to. proceed - E based on the licensee's 10 CF.R 50.59 review of the, procedure and their intention that at'.least one normal loop'of'RHR wouldi L , ' remain available..TheJadequacy of"10 CFR 50.59. reviews of'

s . previous uses of. alterriate DHR was considered an Unresolved, ' Item (461/90001-01(DRP)) pending a response byythe'NRR staff to ' ' ' the request for technica,1 assistance.

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. ! (2) lDuring a review of the shift supervisors log, the' inspectors' ,

• noted that a temporary procedure: change had been r.equired to-

. . add valve 1E22-F022 to procedure-9861,02V014;;a' type C local-leak-rate testing (LLRT) procedure for containment penetration-1MC-35.

Further discussions revealed that the1 valve had been - previously deleted from the procedure; however,:~it was stillL listed as a containment' isolation valve in Tochnical, ' - ' i Specification. 3;6.4,: Table' 3.6.4-1, Item 3(24);_ andl ' ' , consequently should not.have been-deleted.

' ' . . . , . . Further investigation revealed that valve'1E22-F022 had been ^

3c.

deleted by Revision 24 on January 23, 1989. 'This change?

, , ' was precipitated by a fletter from' J. A. Stevens, NRR, to - l g'it!t ,

ti D. L. Holtzscher, IP, dated 0ctober 31,L1988, which stated' ' the'NRCistaff's position regarding;the. boundaries and- '

• methodology-for testing certain containment penetrations',

' ' ' > , 'one of which was 1MC-35.- In-respohsetolthis.Lletter,.the-. ,. ,' '< ' , , ,* ' gicensee prepared a license amendment: request to add certain.

Mi, '

, ,. 4 .,Y,, -valves to Technical Specification Table. 3.6,4-1 and to delete .

4 + s .. v1'. xj

p be superfluous.

This amendment r'equest was submitted-on!.. &Mg 4 77 certainivalves already in Table.3.6.4-1 asithey would now .. . , /M ? MJp, ,3,' y June 30, 1989.

In parallel.with the' amendment 'equest the ( ^1 ' i r " ' ' '

-

Alicensee's staff was preparing a-revision to'the LLRTs y,. < ' '" ' Procedure'9861.02.

V014-is.the specific portion which f , , addresses containment. penetration 1MC-35.

_, , ' > + y ~ ' - e , '

• '

n10 CFR 50.59 (a)(1) sllowed licensees to make changes to.the' , -%'E , j? facility.as-described in the-Safety Analysis Report without i -

pri,or NRC

approval,, provided that the change did not require

~ -

, 'a changeLto the Technical Specifications.

Technical ' ,u. l , Specification 3.6.4,LTable 3.6.4-1, Item 3(24),: listed: valve f ' <1E22-F022 as a manual containment'isolatina valve'for' 1f ' k containment penetration 35 (high pressure core spray pump. " discharge).

Technical Specification 4.6.1.'2.d required that ~ type C leakage tests.be-performed on all containmentzisolation valves.... per Table 3.6.4-1 which penetrate primary-containment.

. Procedure No. 9861.02V014 was revised on January 123,11989"by; - ' the issuance of Revision 24, which deleted valve 1E22-F022,= ~ before the' Technical Specification amendment'to~ delete valve L 1E22-F022 had been approved by.the NRC.

The-failure to obtain prior NRC approval'before: making'a' change to'the. facility is a violation of 10'CFR 50.59 (461/90001-02(DRP)). In subsequent discussions with'! licensee management,Lthey. stated, ~ that they believed the root cause.to this violation to be a misunderstanding of the ' phrase "phior NP,C: approval"; and that: ' the licensee staff personnel.who prepared and? reviewed the . safety evaluation for' Revision 24 to;V014. thought'that upon receipt'of the NRC letter they'did not need'a prior approved j Technical. Specification amendment.. _In response-to this, the ' licensee has revised the. lesson plan for; training on performing - .

, - ,.

. d. m - . < . . . . , g ' - . . , ,g %*, . ,

, ' .

.f ~ ^ . .- .- ,c - , 5* ,j - n safety. evaluations and.'was~ revising;the-safety evaluation ' , ., , manual to make clearer what constitutes, prior NRC approval, l

1. ,

s ,- -AJdiscussion;of:this problem with all; staff?who prepare "and review safety: evaluations will~be accomplished.

,r

._

The inspect' ors belisveathst contributing.to the root cause was

! ~ " , ' the inadequate des'cription of the,8 cope..of the' revision, on the; " document. change' form.

The stated purpose:of the revision-was,- Added valve physical location and.new containment isoletion " " j , valves.".No mention was made of the deletion-of valve 4- ,.. ., !.- y , ': 1E22-F022 nor,were change. bars used in the revision.' 'V" = ' '

" , .3 . t o'1 ' > ' ' ' . . ,. s, ^ ' ' The-inspecto_s:also Necommend-that the Facility. Review Group; r 'j;, w (onsitereviewcommittee)examineitsaiiproachto1 reviewing t > %procedurechangestoensurethattherequiredTechnicals ..

i <- Specification amendments have'been obtained and'that no 'M ' b' - o%, '" gnreviewed safety' questions exist.~ t - j .p+ . .. o . ,- , , , r ,, Onei violaINo.n and one unresolved; item were identified.

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'4 g g , + a , - ,. p'. 1,18.,; Management; Changes A j a, , ~ ,c +'

, & ,3 - > .~ _

, _ , Duri,ng this inspection. period the Manage'ri-Quality Assurance resigned- ' ' ' s ,.

. Jand was replaced by. Mr. R. Wyatt the former Manager Nuclear. Training.

. ' - > , H l'Mr. JL Palm'er, former Director - Maintenance'and Technical Training y::. J 2 a e, > i named as the, new Manager - Nuclear Training,and Mr; P. Yokum wat Mmed' T ' ,

< f5 J to.Mr.j Palmer's' former position. : Also, Mr.- F.- Spangenberg resamed his . J position:as' Manager - Licensing and Safety after completing SR0 training ( 's c ,. .

In addition, the Director - Licensing _ resigned-and was(replaced by ~

' - - Mr."R. Pha,res.

c

.y - . '9. ~ Unresolved Items , Unresolved items are matters about which more-information is required in order to ascertain ~whether they are acceptable itemsF violations,-

, or deviations. An unresolved item disclosed during the inspection is.

' discussed in Paragraph'7.d.(1).

' 10.

Meetings ' -l a.

-Management Meetings (30702) ' o . . On February 28, 1990, Mr. C. E. Norelius, Director, Division of . Radiation Safety and Safeguards:and other members of the NRC1 staff , ' met at the site with-the licensee as-' denoted in Paragraph 1 of this'

report..The meeting was held to discuss current items of concern

• j

. included recent management changes, plant performance, degraded and; ' , out-of-service equipment, control of outage = activities, and radiation protection issues.

The licensee also presented an update on , performance improvement initiatives for 1990.

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. ...g , 9-f . , ~ .37 ."- ' 7 Paragraph 1 at:the conclusion of,the 11nspection on; March?16,3

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.; ig The inspectors * summarizedithe' purpose' and; scope of/the' inspection:, 9-J pfg. u. ' w-

- - Q"itinformational;contentofLthe'inspectionreport,with_ regard 1tof34 4 f/ 4'? and the<Lfindings'.' ;The1 inspectors: also-discussed:the likely.. u;. ; ; "' M YN' :' ' f ;lf

f* % e *,doctments or processes' reviewed byf the1 inspectors duringl thel, J.4 - 4%

7 T Jinspectiori. gTheLlicensee:didinotlidentify:any such documentsfo@ff,L ' ' " fprocesses as' proprietary.

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