ML20210S265

Safety Insp Rept 50-461/87-02 on 861216-870126.Violation Noted:Failure to Follow &/Or Provide Procedures Re Onsite Followup of Events.One Unresolved Item Identified Re Degradation of Secondary Containment Gas Control Boundary

ML20210S265
Person / Time
Site:	Clinton
Issue date:	02/09/1987
From:	Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20210S188	List: ML20210S184 ML20210S203 ML20210S265
References
TASK-2.B.4, TASK-TM 50-461-87-02, 50-461-87-2, NUDOCS 8702170594
Download: ML20210S265 (33)

See also: IR 05000461/1987002

Text

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-461/87002(DRP)

Docket No. 50-461 License No. NPF-55

Licensee: Illinois Power Company

500 South 27th Street-

Decatur, IL 62525

Facility Name: Clinton Power Station

Inspection At: Clinton Site, Clinton, IL

Inspection Conducted: December 16 through January 26, 1987

Inspectors: T. P. Gwynn

, P. L. Hiland

R. N. Gardner

src %<

R. C. Knop, Chief g g

Approved By:

Projects Section IB Date

Inspection Summary

Inspection on December 16 through January 26, 1987 (Report

No. 50-461/87002(DRP))

Areas Inspected: Routine, unannounced safety inspection by the resident

inspectors and a region-based inspector of licensee action on previous

inspection findings; licensee action on 10 CFR 50.55(e) report; applicant

action on Three Mile Island (TMI) action plan requirements; licensee event

report' review and followup; review of allegations; Region III request;

operational safety verification; engineered safety feature system walkdown;

onsite followup of events at operating reactors; and management meeting.

Results: Of the areas inspected, no violations or deviations were identified

in nine areas. One violation was identified in the area of onsite followup

of events (paragraph 10.b. - failure to follow and/or provide procedures).

While the violation was of minor safety significance, licensed operators

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made a number of errors that could have been prevented had they used adminis-

trative controls available.One unresolved item was identified in the area of

operational safety verification involving degradation of the secondary

containment gas control boundary (paragraph 8).

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DETAILS

1. Personnel Contacted

Illinois Power Company (IP)

• 1. R. Campbell, Manager - QA

1. W. Connell, Manager - Nuclear Planning & Support

1. G. Edgar, Attorney

1. R. Freeman, Assistant Plant Manager, Maintenance

1. W. Gerstner, Executive Vice President

1. J. Greene, Manager - Nuclear Station Engineering Department (NSED)

• 1. D. Hall, Vice President, Nuclear

1. H. Lane, Manager, Scheduling and Outage Management

1. J. Miller, Assistant Power Plant Manager, Startup

• 1. J. Perry, Manager - Nuclear Program Coordination

• R.~ Kerestes, Director, NSED Field Engineer

• F. Schwarz, Director, Outage Maintenance Support

• 1. F. Spangenberg, Manager - L&S

• 1. E. Till, Director, Nuclear Training

• J. Wemlinger, Supervisor, Operations Training

• 1. J. Wilson, Manager - CPS

1. R.'Wyatt, Director, Nuclear Program Assessment

Soyland/WIPC0

1. J. Greenwood, Manager Power Supply

Nuclear Regulatory Commission - Region III

1. B. Davis, Deputy Regional Administrator, Region III

• 1. T. Gwynn, Senior Resident Inspector, Clinton

• 1. P. Hiland, Resident Inspector, Clinton

1. R. Knop, Chief, Projects Section IB

1. R. Warnick, Chief, Projects Branch 1

• Denotes those attending the monthly exit meeting on January 26, 1987.

1. Denotes those attending the management meeting on Jarvary 16, 1987.

The inspectors also contacted and interviewed other licensee and

contractor personnel.

2. Licensee Action On Previous Inspection Findings (92701/92702)

a. (Closed) Open Item (461/86028-09): Fire Protection Administrative

Controls. During a previous inspection, the inspector identified

that fire protection administrative controls were not fully

implemented.

During this report period, the licensee stated that their fire

protection program had been fully implemented. In order to verify

implementation, the inspector selected a random sample of five fire

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protection surveillance requirements for review. The inspector's

sample included the following procedures:

CPS No. 9071.01' Diesel Driven Fire Pumps Operability Test

CPS No. 9071.06 Visual Inspection of Spray and Sprinkler System

Piping and Heads

CPS No. 9071.08 Fire Protection C02 System Valve Position Check

CPS No. 9071.19 . Monthly _ Fire Protection Valve Line-Up _

CPS No. 9071.25 Fire Protection C02 Weekly Operability Check '

The inspector reviewed the associated inspection checklists for

the above procedures that had been completed and stored in the

licensee's record storage vault. The review performed was to

ascertain if the administrative controls established were being

implemented. For this review, the inspector verified that required

inspection frequencies (monthly, weekly) were met; that noted.

deficiencies were documented and required maintenance work requests

were initiated; completed inspections were reviewed for acceptable

results; and that when unacceptable results were documented, followup

inspections were performed to verify corrective action taken. For

the sample selected, the inspector _ concluded that the licensee was-

implementing the administrative controls that had been' established.

The inspector reviewed the licensee's action taken in response to

a concern identified by offsite fire department personnel. As

documented in Inspection Report 50-461/86028, offsite fire

department personnel stated that a self contained breathing

apparatus (SCBA)'was found to have an empty cylinder during a

drill. Since the concern expressed was not identified to the

licensee at the time of the drill, the specific SCBA was not

identified. However, the licensee revised its control over SCBAs

intended for use by offsite fire department personnel. Previously,

, offsite fire department personnel . received SCBA equipment from a

licensee storage locker when responding to the Clinton Power

Station. In an " Acquisition Agreement" dated September 30, 1986,

the licensee provided SCBA equipment to three offsite fire

departments for their general use and in particular for their use

when responding to the Clinton Power Station as a secondary fire

protection service.- The inspector concluded the licensee's actions

adequately addressed the expressed concern.

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The inspector noted that construction activities at Clinton Power

Station have been reduced to a level consistent with the startup

phase of operation. Housekeeping requirements have been monitored

i - on a continuous basis by the inspector and minor deficiencies

f identified have been promptly corrected by.the licensee. The

inspector observed the performance of routine fire watches and fire

watches stationed in areas where grinding or hot work was being

,. performed. No deficiencies in fire watch performance have been

identified.

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_ Based onithe inspector's review of administrative records, actions

taken-by the'. licensee regarding concerns with control of SCBA

equipment, and the noted housekeeping and fire watch performance,

the inspector concluded that .the fire protection' program for Clinton

Power Station was being fully implemented. This item is closed,

b. -(Closed)OpenItem(461/86054-05): Deficiencies related to

watertight doors. During a previous . inspection, watertight doors in

the plant were' observed to have numerous minor hardware deficiencies '

indicating inoperable status. Testing and. maintenance programs had ,

not been established for'these doors.

As. documented in Inspection Report 50-461/86060, paragraph 2, this

item remained open pending completion of corrective actions to.

upgrade reliability of the watertight doors-(plant modification

.HC-20).and pending approval of the maintenance procedure for

watertight doors.

The licensee presented this item to the inspector for closure. All

watertight doors in the plant had been modified in accordance with

minor modification HC-20 through supplement 1 and a formal procedure

for maintenance of watertight doors (CPS No. 8250.01) was approved

for use on December 4, 1986. This information provided the basis '

for closure of this item. ,

The inspector-had noted apparent improvement in the reliability

of plant watertight doors through routine tours of the facility.

Discussion with the licensee's licensing staff indicated that

only four_ maintenance work requests (MWRs) had been issued on

-watertight door deficiencies since completion of modification

HC-20 on November 4, 1986. Of those four MWRs, only two involved

inoperability of.the affected door; the other two involved degraded

, performance of the closing mechanism which remained operable. This

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data indicated an improved reliability as compared to previous NRC

, observations.

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! . Finally, the-licensee completed testing of watertight door seals

) in accordance with the manufacturer's specifications. All doors

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had acceptable test results after necessary adjustments by the

maintenance department. This item is closed.

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c. (Closed)OpenItem(461/86074-04): The licensee agreed to review

their maintenance training program to determine if an interim

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program or changes to the existing program were warranted prior

!- to the completion of INP0 accreditation.

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The licensee completed their evaluation of the current maintenance

, training program and presented their results to the resident

inspector for review. Both the IP Maintenance Department and the

i IP Nuclear Training Department participated in the review. Their

( review identified the following:

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'(1) The current INP0 accreditation program will resolve all

training weaknesses observed by the NRC.

(2)l.The current training program is implemented and additional:

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efforts'are being focussed on_ supporting emergent training- ,

requirements that arise from specific problems in the plant.

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Their review concluded that any attempt to develop an interim

training program would take nearly as long as developing and

-implementing the INP0 required program and that development of

an: interim training program would result in costly delays in the

accreditation schedule.

In view of. Policy Statement on Training and-Qualification of Nuclear

Power Plant Personnel (50 FR 11147 dated March 20,1985),the

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licensee's schedule for achieving INP0 accreditation of their

maintenance training program, and the lack of any substantive

evidence that maintenance personnel are not adequately trained. -

this item is closed.

d. (0 pen) Open Item-(461/85005-32): Verify that procedures to ensure

independent verification of system lineups are complete before fuel

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loading (TMI Item II.K.1.10).

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This item was previously reviewed as documented in Inspection Report

-50-461/86064, paragraph 2.a. Since that inspecticn, the licensee

revised procedure CPS No. 1401.01, Conduct of Operations, to include

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clarified criteria for independent verification of system lineups

and to include a listing of plant systems that required independent

verification. In addition, the licensee reviewed operating. "

procedures containing valve and/or electrical lineups to determine

if the clarified criteria were met and initiated action to make

necessary revisions.

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The inspector reviewed the actions taken by the licensee and

verified that necessary reviews and revisions were either completed

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or scheduled to be completed in a meaningful time frame. In

!- particular, the licensee had reviewed all system operating

! procedures for systems to be declared operable to support the

initial criticality milestone and had scheduled reviews / revisions

for other operating procedures to be completed prior to required

milestones. (Some exceptions were taken to this general statement

where the licensee had a high. level of confidence in the currently

h approved. procedure being conservative). The inspector verified that

L the following procedures had been revised to include independent

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verification of important valve and electrical lineup:

E CPS No. 3315.01, Containment Monitoring (CM)

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CPS No. 3101.01, Main Steam (MS, IS, & ADS)

I CPS No. 3310.01, Reactor Core Isolation Cooling (RI)

l CPS No. 3302.01, Reactor Recirculation (RR)

CPS No. 3402.01, Control Room HVAC (VC)  ;

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CPS No. 3306.01, Source / Intermediate Range Monitors (SRM/IRM)

CPS No. 3308.01, Local / Average Power Range Monitors-(L/APRMS)

The licensee intends to complete review and revision of all

operating procedures requiring independent verification by June 30,

1987. That will include review and revision of some procedures

that currently (conservatively) require independent verification of

omponents that exceed the criteria established in CPS No. 1401.01.

The inspector identified some minor discrepancies in CPS No. 1401.01,

Appendix C (for example, the reactor recirculation [RR] system was

not listed but did require and was provided with independent

verification) which were pending correction by the licensee.

e. (0 pen) Open Item (461/85015-07): " Confirm necessary revisions to

EPGs made, E0Ps upgraded, and operators trained before fuel load

(SSER4-13.6.3.1)." Paragraph 13.6.3.1 of Supplement 4 to the SER

required verification that revisions were made to the CPS emergency

procedure guidelines (EPGs), that emergency off-normal procedures

(E0Ps) were upgraded, and that the operators were trained prior to

fuel load. In Inspection Report 50-461/86059, the inspector

determined that the requirements were fulfilled with the exception

of the combustible gas control EPG and E0P which was scheduled for

completion after fuel load.

The inspector reviewed the status of this item with the licensee

and with the NRC Licensing Project Manager (LPM). The licensee

indicated that a generic combustible gas control EPG had been

developed by the Hydrogen Control Owners Group (HC0G) and submitted

, to the NRC Office of Nuclear Reactor Regulation (NRR) for review on

l December 1, 1986. The licensee plans to endorse the HC0G submittal

once NRC review has been completed. The licensee estimated that

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six months would be required to complete the NRC review and that

additional time would be required to complete plant specific work

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necessary to achieve an E0P for use at CPS.

Discussion with the LPM indicated that the licensee's schedule for

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this item was consistent with the rest of the industry and that

operation above 5% of full power using interim combustible gas

control procedures was acceptable. The inspector will review this

, matter further when the licensee has prepared the applicable E0P.

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f. (0 pen) Open Item (461/86011-01): The licensee committed to having

seven radiation chemistry technicians (RCTs) complete all (36)

qualification cards by 5% power.

! The licensee provided information to the inspector for closure of

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this item. That information indicated that nine RCTs had completed

all qualification requirements necessary to act as the on-shift

(ANSI /ANS 3.1 qualified) RCT. Only five of those RCTs were

l qualified to operate the Post Accident Sampling System Panel (PASS),

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A. sixth qualified PASS operator assigned to the Nuclear Training

Department as an instructor was available to respond to emergencies.

The licensee stated that another individual in the Chemistry

Department was being trained to operate PASS.

The Supervisor-Chemistry had a high level of confidence in the

ability of the chemistry group to augment the normal shift

complement with PASS qualified personnel to respond to any emergency

in the required time. The inspector noted that three of the six

qualified individuals (a team leader, a PASS operator, and a third

individual preparing the chemistry laboratory for PASS analysis)

were needed to perform post-accident sampling; that the licensee

had not specifically demonstrated the ability to augment the normal

shift to meet PASS requirements; and that the ability to augment the

shift with a sufficient number of qualified personnel was related

to the number of qualified personnel available. The inspector, in

consultation with Region III management, agreed that the. licensee

had met their commitment concerning the number of qualified

personnel necessary to man the shift and thus their commitment to

5% power was met. This item will remain open pending review and

verification of RCT qualification records by a Region III based

specialist inspector.

The quidelines of CPS No. 1890.30, Post Accident Sampling Program,

indicated that a minimum of six PASS qualified individuals was

desired to ensure the availability of qualified personnel. The

licensee stated that a plan was being formulated to enhance the

PASS program to provide three staff professional (technical)

individuals to act as PASS team leaders. When implemented, that

plan will provide additional PASS qualified individuals to respond

to emergencies, increase the depth of the organization (i.e., more

than the minimum number of qualified personnel available), and

improve leadership provided for PASS teams. The licensee stated

that this plan will be finalized and the appropriate individuals

qualified by April 1, 1987. This is.an open item pending NRC

review of the licensee's actions (461/87002-01).

g. (0 pen) Open Item (461/86054-14): Deferred Testing Activities.

The Clinton Power Station Operating License paragraph 2.D.

granted a number of schedular exemptions to the performance of

test activities. These exemptions deferred testing to a specific

milestone. The status of these deferred test activities was

reviewed by the inspector during this report period and is

tabulated below:

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Deferred Tests Deferred Tests

System Milestone Completed Remaining

Turbine Electrohydraulic Reactor Heatup ATP-EH-01 NONE

Control (EH)

Traversin Incore 5% Power PTP-TP-01

Probe (TP

Off Gas (0G) Reactor Heatup PTP-00-01 NONE

PTP-0G-02

PTP-V0-01

XTP-00-12

Containment Initial PTP-CM-01 NONE

Monitoring (CM) Criticality

Leakage Initial PTP-LD-01 NONE

Detection (LD) Criticality

Fuel Pool Cooling 5% Power * PTP-FC/SM-01

and Cleanup (FC)

Fuel Handling (FH) 5% Power * PTP-FH-01

In-place. Filter on Initial XTP-00-12(VC) NONE

Control Room HVAC (VC) Criticality

HVAC Testing For: Reactor Heatup* PTP-VA-01 NONE

PTP-VQ-01

Aux. Building (VA) PTP-00-01(VA)

Dry Well Purge (VQ) PTP-00-01(VQ)

Dry Well Cooling (VP) PTP-00-01(VP)

Containment XTP-00-12(VQ)

Building (VR) PTP-00-01(VR)

Turbine Building (VT) PTP-00-02(VW)

Radwaste Building (VW) PTP-00-01(VT)

Fuel Building (VF) PTP-00-02(VT)

PTP-00-01(VW)

PTP-00-02(VF)

PTP-00-02(VR)

XTP-00-12(VW)

PTP-00-02(VA)

• This milestone or before removal of the reactor pressure vessel

head after initial criticality.

During this report period, the inspector verified the licensee

had evaluated the results of the above completed deferred test

activities. The inspector reviewed each of the above completed

test summaries and verified the test results were reviewed and

approved in accordance with the licensee's program.

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-This item will remain open pending the completion'of the remaining

deferred tests.

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h. (0 pen) Open Item (461/86074-02): _ Procedure comment control forms

(CCFS) were being used to-identify suggested procedure improvements.

This.use was not controlled by plant administrative procedures.

The NRC. inspector.was-concerned that these CCFS had not been .,

. reviewed to determine their technical . impact and the need for an '

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immediate-procedure revision.

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The licensee revised CPS ~No. 1005.01, " Preparation, Review,

i- Approval,.and Implementation of and Adherence To Station Procedures

and Documents" on January 8, 1987, to include requirements concerning

control of CCFS initiated against issued station procedures. The

procedure changes were responsive to the.NRC concern. In addition,

all CPS departments reviewed outstanding CCFS to determine if any

'were of sufficient significance to warrant revision of the affected-

procedure prior to the normal biennial review. A small number of

CCFS were identified which resulted in the initiation of procedure

revisions. Those revisions were scheduled for completion by

! required. plant milestones.

The licensee's QA organization performed a surveillance'of the

Operations Department procedure files (Surveillance Q-09456 dated

December 15-16,1986) to determine how CCFS generated against

issued procedures were handled. Their surveillance verified the

information discussed.above and also determined.that the procedure ,

files were not up to date (i.e., the files contained CCFS which had

already been resolved, contained CCFS against procedures that had

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been cancelled,.etc). The licensee's QA department scheduled an

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additional surveillance to verify action taken to correct the 1

identified concern. '

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This item will remain open pending completion of tne' licensee's

actions and verification that the plant staff is adhering to CPS

{ No. 1005.01 for control of CCFS.

1. (Closed) Unresolved Item (461/86059-01): The basis for closure .

r of CR 1-86-07-009 concerning performance of safety related work

without approved procedures required additional justification.

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The licensee presented this item to the inspector for closure.

CR 1-86-07-009 was revised to provide assurance that work performed

, prior to issuance of approved work procedures for core drilling and

concrete expansion anchor installation was adequately controlled,

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, documented, and inspected. No violations were identified. Work

p control procedures were approved, as follows:

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CPS No. 8901.16, Core Drilling, revision 0 dated September 13,

1986.

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CPS No. 8199.01, Concrete Expansion Anchor Work, revision 0

dated August 25, 1986.

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In addition, the licensee scheduled training for maintenance

supervisors and planners to assure that all cognizant personnel

understood the need to have approved procedures to control safety

related work. That training was scheduled for completion on

February 2,1987. Completion was being tracked by centralized

consnitment tracking item (CCT) No. 044013. This item is closed.

j. (Closed) Violation (461/86037-02): Procedure CPS No. 9052.02,

Low Pressure Core Spray Valve Operability Checks, did not provide

sufficient detailed instructions and/or appropriate acceptance

criteria for determining that important activities had been

satisfactorily performed.

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This item was previously reviewed, as documented in Inspection

Report 50-461/86060. At that time, this item remained open pending

completion of revisions to certain surveillance test procedures

identified in attachment B of the licensee's letter U-600689.

Those revisions were required to be completed prior to initial

reactor criticality. In addition, CPS No. 1011.05, CPS Surveillance

Procedure Guidelines, was scheduled for revision by October 20,

1986, to address the reporting of all failures to meet surveillance

test acceptance criteria to the shift supervisor. The licensee had

provided interim guidance to all plant personnel in plant manager's

standing order (PMS0) No. 30 regarding the reporting of test

failures.

The inspector verified that the licensee had completed revision

to surveillance test procedures required to be completed prior to

initial criticality. Several minor editorial /non-technical

discrepancies identified during this inspection were corrected by

the licensee.

The inspector noted that CPS No. 1011.05 had not been revised as

scheduled by the licensee. Discussion with cognizant licensee

personnel indicated that PMS0 No. 30 remained in effect and that the

revision was scheduled and expected to be completed by January 30,

1987. This information provided a sufficient basis for closure of

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k. (0 pen) Violation (461/86060-02): Corrective actions in response to

IPQA Audit Q38-86-10 and IPQA Surveillance Finding M-86-005 were

not effective to prevent recurrence. The licensee had identified

l deficiencies in the processing of Maintenance Work Requests (MWRs)

l for evaluation of post maintenance testing. The corrective action

l performed was not effective as evidenced by additional deficiencies

l identified by an NRC inspection conducted subsequent to the

licensee's corrective action.

During this report period, the licensee formally responded to the

subject violation. The licensee was unable to respond to the

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violation in the thirty days required by the Notice of Violation

! dated October 17, 1987. The licensee verbally communicated to NRC

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Region III their inability to meet the thirty day requirement and

the written response dated December 19, 1986, was considered

acceptable.

The inspector selected a sample of 47 MWRs that had been closed

between August and December 1986, to verify the specific corrective

action taken by the licensee.

The review performed was to ascertain if the closed MWRs were

being evaluated for post maintenance testing (PMT) requirements

in accordance with the licensee's controlling procedure CPS No.

1401.01, " Conduct of Operation", revision 11, dated December 31,

1986. For each of the MWRs selected, the associated PMT evaluation

was performed in accordance with CPS No. 1401.01. The inspector

was able to locate each PMT evaluation form in the licensee's record

storage vault; in the system status files maintained in the main

control room; or in the Plant Staff Technical Department. The

inspector concluded through this review that the licensee's specific

corrective action was adequate.

The corrective action taken to prevent further violation included

revising the implementation procedure to require a copy of the

completed MWR be received by the PMT evaluator prior to closing out

the MWR in the computer file. In addition, the PMT evaluators had

been relocated with maintenance planners. The inspector verified

the above actions were in place; however, the formalized change to

the MWR Preparation and Routing Procedure, CPS No. 1029.01 was not

completed at the end of.this inspection period. The licensee stated

that the revised procedure would be issued January 30, 1987. This

item will remain open pending the issuance and the inspectors review

of this revised procedure.

1. (0 pen) Violation (461/86065-03): Procedure CPS No. 1016.01, CPS

Condition Reports, was not followed in that corrective action plans

were not approved prior to implementation; block 2 of the condition

report form was not always filled out; and reviews of condition

reports (CRs) by various departments did not identify and correct

the violations that existed.

The licensee responded to this violation in letter U-600806 dated

January 6, 1987. This letter was late in meeting the 30 day

response requested by the notice of violation. The licensee's

response to the violation appeared adequate to address the substance

of the violations.

The inspector reviewed CPS No. 1016.01, revision 15 dated

November 24, 1986 and verified that the changes reflected in the

licensee's letter, Attachment A, paragraph I.a., had been

incorporated. The inspector also reviewed several recent CRs and

verified that they had been processed in accordance with the revised

administrative controls.

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The-inspector reviewed records'of training provided to personnel

responsible for the review of condition reports and verified that

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the cognizant records coordinator had been included in the required

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training.

Discussion with plant staff personnel indicated that the additional

procedure revision was scheduled to be completed on March 31, 1987,

and that the revision was expected to be completed on schedule.

Thisl violation will remain open pending completion of the actions

discussed in Attachment A, paragraph II.a.

m. (0 pen) Violation (461/86065-04): Three examples of inadequate

surveillance procedures.

The licensee responded to this violation in letter U-600806 dated

January 6, 1987. Review of the licensee's response indicated that

the response adequately addressed two of the three examples in the

NOV (examples B & C). However, that response limited the scope of

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the licensee's corrective actions to first time performance mode 1,

2, & 3 surveillance procedures. The inspector noted that the first

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example of the violation involving the Standby Liquid Control Pump

Operability Test procedure was not a first time performance

3 surveillance procedure-and that the problem encountered did not

involve installation of jumpers or lifting of leads. The licensee

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agreed to review this matter further to determine if additional

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corrective' action was needed and to provide a supplementary response

L to this NOV.

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The inspector reviewed PMS0-30, revision 3 and verified its

implementation. The PMS0 provided the controls identified in

the licensee's response and appeared to have been effective in

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reducing the number of events resulting from first time performance

of surveillance procedures.

This violation will be reviewed further after receipt of the

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licensee's supplemental response.

n. (0 pen) Violation (461/86065-05): Eight examples of failure to

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follow procedures during the conduct of initial fuel load

operations.

The licensee initially responded to this violation in letter

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- U-600806 dated January 6,1987. At the request of Region III,

the licensee provided additional information concerning the

corrective actions taken for each of the eight examples cited

in letter U-600823 dated January 21, 1987. The license's

supplemented response to the violation appeared adequate to

address the substance of the violations.

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_m. , ,._,c.---.,_. _,_...,.m, , , . , . , , , , , . . - , . - , , , - . . . _ , , _ , , _ _ _ , . , _ _ _ _ . . _ _ . _ , _ . , , , . , _ _ . . . _ _ . _ . , _ , , . , _ . _ _ . -

.

The inspector reviewed the specific corrective actions taken by the

licensee in response to each violation cited and verified, based on

a sample ~ of the actions taken, that their corrective actions had

been implemented as stated.

Concerning the generic corrective actions addressed in letter

U-600806, Attachment B, the inspector verified a sample of_ the

corrective actions taken by direct observation of the corrective

actions in progress and through interviews of various plant and

plant management personnel. The actions taken by the licensee

appear to have been effective in reducing the nuinber of personnel

errors and the frequency of reportable events. Additional NRC

concerns regarding the conduct of plant operations were identified,

as documented in paragraph 10.b. of this report. The licensee's

additional corrective actions will be reviewed with their response

to that violation.

This violation remains open pending licensee verification that

all corrective measures indicated in the response to the notice

of violation, attachment B, have been completed.

o. (0 pen) Violation (461/86065-06): Two examples of performance of

plant operations without approved procedures.

'

The licensee responded to this violation in letter U-600806 dated

January 6, 1987. That letter was late in meeting the 30 day

response requested by the Notice of Violation. Review of the

licensee's response indicated that the response did not address the 4

apparent violation of CPS No. 1011.01, Test Programs and Control.

The licensee stated that CPS No. 1011.01 would be revised to provide

controls over the type of activity described in the Notice of

Violation. The licensee is planning to revise their response to

this Notice of Violation to reflect the corrective actions to be

.

taken. This violation will be reviewed further after receipt of

l the licensee's revised response.

t

p. (0 pen) Violation (461/86065-07): Four examples of failure to meet

plant technical specifications.

The licensee responded to this violation in letter U-600806 dated

January 6, 1987. The inspector performed a preliminary review of

the response to this violation during this report period. In

conjunction with the response provided, the inspector performed

a detailed review of Licensee Event Report (LER) 86-009-01

associated with this violation. Results of the inspector's review

l of LER 86-009-01 are contained in paragraph 5.a. below. At the

, conclusion of this report period, the inspector's review of the

!

licensee's response to this violation was still in progress. The

results of this review will be reported in a future inspection

report. This item remains open pending completion of that review.

!

l

l

13

l

-- . .- .

- .

-

.

p

,

an <c""

. <. .

~.

q -(0 pen) Violation (461/86074-05): Failure't'o follow approved

procedures for control of Temporary Modifications. .This violaf. ion

.

identified a number of deficiencies in the 1Nplementadon of

administrative controls for temporary modifications.

The licensee responded to this violation in letter U-600819 dated

January 20, 1987, in a timely manner. The inspector noted that

,

the licensee expected to be in full compliance on January 31, 1987.

This item will remain open, pending the inspector's review of

corrective actions taken by the licensee.

No violations or deviations were identified. +.,

y .

y.

,

3. Licensee Action on 10 CFR 50.SS(e) Report (92700)

a. (Closed) 10 CFR 50.55(e) Item (461/86006-EE): Watertight Seals s

and Openings in Vital Area Boundaries.

This item was previously inspected as documented in Inspection

Report 50-461/86060.

During this inspection, the inspector reviewed the licensee's final

report submitted by letter U-600765 dated November 24, 1986; a '

supplemental final report submitted by letter U-600825 dated

January 26, 1986; and portions of additional quality records

related_to corrective actions taken by the licensee. Those

documents included the following:

CPS No. 1029.01, Maintenance Work Requests, revision 10 and 14

CR 1-86-11-171

CR 1-86-08-020

CR 1-86-12-014

CR 1-86-12-029

' Plant Modification Packages A-67, A-71, and A-73

Plant Modification Package A-47 (10 CFR 2.790 information)

Chairman's Final Report on 55-86-06, letter Y-82470 dated

October 31, 1986.

!

Review of the above documents indicated that the licensee's

, corrective actions had been completed; that additional findings

concerning the floodproofing of the CPS Screenhouse had been

'

i submitted to the NRC in a supplemental report; and that the

I

licensee's corrective actions had addressed both the specific

and generic implications of the identified deficiency.

The inspector noted that a violation related to this matter

(461/86048-03) was pending enforcement action by the NRC.

Additional reviews related to this matter will be tracked by

the violation. This item is closed.

b. (0 pen) 10 CFR 50.55(e) Report (461/86007-EE): Broken Tack Welds

on Anchor Darling Globe Valves.

14

_ , . _ . _ - . _ _ . . _. _ - . _ - . _.

y } '

~ =L+s:

yQ v -

Q -*

[?

, .

5f This matter was previously reviewed as documented in Inspection

rM _

'

Report 50-461/86072. That report determined that the licensee's

m planned corrective actions were deferred to the first refueling

outage but the licensee had not provided sufficient justification

'

~for operation of 32 potentially affected valves during the first

-

,

operating cycle.

C The licensee provided letter Y-83108 dated January 13, 1987 to

supplement the final report on this deficiency. That letter

-

provided the engineering justification for operation of the affected

valves through the first operating cycle. However, the licensee's

review did not account for system operation to provide long term

decay heat removal after a postulateo accident involving damage to

the plant. Although the likelihood of such an accident is small,

the plant systems are designed to operate under those conditions

. and should not be adversely affected by this identified deficiency.

'

' The3fcensee conducted additional reviews and determined that two

of the affected valves (1E12-F003A/B) may be operated in a throttled

mode during long term decay heat removal after a postulated accident.

The licensee's engineering justification provided a sufficient

i basis to justify removal of administrative controls from all valves

except the two valves documented above. The licensee stated that

administrative controls would remain in place for those two

potentially affected valves pending completion of additional

. engineering reviews.

This matter will be reviewed further during a subsequent inspection.

No violations or deviations were identified.

4. - Applicant Action on Three Mile Island (TMI) Action Plan Requirements

(25401)

The NRC Office of Inspection and Enforcement issued Temporary Instruction

(TI) 2514/01, Revision 2, dated December 15, 1980, to supplement the

Inspection and Enforcement Manual. The TI provides TMI-related

inspection requirements for operating license applicants during the phase

between pre-licensing and licensing for full power operation. It is

divided into two parts. Part I lists requirements that were closed prior

to fuel load. Part 2 lists requirements that must be closed prior to

full power operation. Part 2 of the TI was used as the basis for

inspection of the following TMI item found in NUREG-0737, " Clarification

of TMI Action Plan Requirements".

(0 pen) Item II.B.4.2: Training for Mitigating Core Damage. The licensee

was to complete training prior to full power operation.

During a previous inspection (50-461/86023), part 1 (II.B.4.1) of this

TMI action item was closed based on the licensee's established Mitigating

Reactor Core Damage (MRCD) training program. During this report period,

the inspector verified through review of training records that the Power

Plant Manager had successfully completed the MRCD training. In addition,

15

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- _ _ _ - _ _ _ _ _ _ - _ _ .

.

.

the inspector verified that nonlicensed technicians had been provided

training in accordance with the licensee's commitment contained in

section 13.2 of their Final Safety Analysis Report (FSAR). However, the

inspector noted that several technicians had not received the required

training and the licensee was unable to provide the inspector evidence

that those technicians would be trained as committed in the FSAR. This

item remains open pending the inspectors review of actions taken by the

' licensee to complete the training of nonlicensed technicians.

No violations or deviations were identified.

5. Licensee Event Report'(LER) Review and Followup (90712 & 92700)

a. In-Office Review Of Written Reports Of Nonroutine Events At Power

Reactor Facilities (90712)

For the LERs listed below, the inspector performed an in-office

review of each LER to determine that reporting requirements had

been met; that the corrective action discussed appeared appropriate;

that the information provided satisfied the applicable reporting i=

requirements; to determine if appropriate actions had been taken on L_

ay generic issues present; and to determine if any additional NRC

inspection,' notification, or other response was appropriate. Where =

determined appropriate, the LER was scheduled for onsite followup

inspection or other necessary action by cognizant NRC personnel.

(1) (Closed)LERNo. 86-006-00 (461/86006-LL) [ ENS No. 06499 and

06569]: Automatic Initiation Of Essential Service Water Due

To Transient Pressure Drop In Nonessential Service Water.

(2) (Closed)LERNo. 86-008-00 and 86-008-01 (461/86008-LL) [ ENS

No. 06552]: Containment Isolation Of The Instrument Air System

Due To Procedural Inadequacy.

LER 86-008-01 indicated that LER 86-008-00 had been superseded

in its entirety by LER 86-009-01. As discussed in (3) below,

the information previously contained in LER 86-008-00 was

included in LER 86-009-01. This LER is closed.

(3) (Closed) LER No. 86-009-00 and 86-009-01 (461/86009-LL) [ ENS

No. 06568]: Automatic Actuation Of An Engineered Safety

, Feature Due To Procedural Inadequacy and Technical

Specification Violation Due To Operator Error.

As documented in Inspection Reports 50-451/86072 (paragraph

6.b.) and 50-461/86073 (paragraph 3.b.), LER 86-009-00 did

not accurately describe all the facts surrounding the subject

event. The inspectors onsite followup of this event was

documented in Inspection Report 50-461/86065 which resulted

in the issuance of several violations (461/86065-04C, 06B,

07A,B,C). Follcwup of the licensee's corrective actions

will be tracked by the open violations.

16

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.

..

During this report period, the licensee issued LER 86-009-01.

This LER incorporated all the information that had been

contained in LER-008-00 as noted in (2) above. The inspectors

review of LER 86-009-01 indicated that the licensee had

provided a complete description of the subject event. The

inspector confirmed by review of training records and licensee

correspondence that corrective action stated in LER 86-009-01

had been or was being implemented. Since several of the

corrective actions identified in this LER are also applicable

to the violations issued in Inspection Report 50-461/86065,

completion of all corrective actions will be reviewed and

documented during the inspector's followup to those violations.

This LER is closed.

(4) (Closed) LER No. 86-020-00 (461/86020-LL) [ ENS No. 06857]:

Tripping of Level Transmitter Results in Automatic Switching

of High Pressure Core Spray Pump Suction Valve Alignment.

The inspector noted that a similar event occurred on January 7,

1987, (see paragraph 10.b.(3) of this inspection report) which

indicated that the root cause of this event may not have been

accurately identified. Further review will be performed when

the licensee completes.their investigation of that event. This

LER is closed.

(5) (0 pen) LER No. 86-019-00 (461/86019-LL) [ ENS No. 06856 and

07000]: Engineered Safety Feature Actuation Due To A Spurious

High Output Alarm on the Main Control Room Air Intake Process

Radiation Monitor.

This matter will be reviewed further on rect et of the

licensee's supplemental report, scheduled u. January 30,

2

1987.

(6) (0 pen) LER No.'86-017-00, 86-017-01, and 86-017-02

(461/86017-LL) [ ENS No. 06670]: Engineered Safety Feature

Actuation Due To Spiking On Intermediate Range Monitor A.

'

This LER remains open pending receipt and review of the

licensee's supplemental report. The licensee's supplemental

report was scheduled for submittal on January 31, 1987.

(7) (Closed) LER No. 86-023-00 (461/86023-LL) [ ENS No. 07123]:

Automatic Actuation of the Reactor Protection System (RPS)

Due To Utility Personnel Error.

No violations or deviations were identified.

b. Onsite Followup Of Written Reports Of Nonroutine Events At Power

Reactor Facilities (92700)

For the LERs listed below, the inspector performed an onsite

followup inspection of each LER to determine whether responses to

17

.

.

the events were adequate and met regulatory requirements, license

conditions, and commitments and to determine whether the licensee

had taken corrective actions as stated in the LER.

'

(1) (0 pen)LERNo. 86-004-00(461/86004-LL)[ENSNo.06413]:

Unplanned Automatic Initiation Of Standby Gas Treatment System

Due To Inadequate Procedures.

This LER was previously reviewed as documented in Inspection

Report 50-461/86072. At the conclusion of that inspection,

there was an open question concerning this LER.

The licensee stated that their engineering review of trip logic

seal-in circuitry indicated that there were no additional uses

of logic similar to that which caused this event. For that

reason, no additional corrective action was required and a

supplement to the LER was not necessary. After receipt of this

information, another event occurred (see paragraph 10.b.(9) of.

this report; ENS No. 07565) which may involve similar logic

functions. The licensee's review of that event was considering

the potential similarity in trip logic but was not complete at

the conclusion of this inspection. This LER remains open

pending review of the licensee's results and verification that

the use of trip seal-in logic which caused this event was

isolated to the five radiation monitors discussed in the LER.

(2) (0 pen)LERNo. 86-021-00 (461/86021-LL) [ ENS No.06913]:

Reactor Water Cleanup Pump Room High Temperature Trip Due To

Personnel Error.

This event was previously reviewed as documented in Inspection

Report 50-461/86073, paragraph 3.e.

During this inspection, the inspector reviewed the LER and

verified implementation of selected corrective actions being

taken by the licensee. No significant discrepancies were

identified but actions were not complete. In particular,

LER 86-020-00 corrective actions 7, 8, and 9 were not complete

at the time of this inspection. One minor item concerning

inclusion of specific information in the LER related to a

personnel error was discussed with the IP licensing department.

The inspector interviewed the Manager - Nuclear Station

Engineering Department concerning corrective actions regarding

lifted leads and jumpers. The recommendations of the

licensee's jumpers and lifted leads task force had been

forwarded to NSED for evaluation. The licensee was scheduled

to have a general plan for addressing jumpers and lifted

leads (in terms of the CPS design and the need to interrupt

electrical continuity for maintenance or surveillance work)

by the middle of February 1987. The licensee was planning to

complete the identification and scoping of necessary plant

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_ _ _

.

.

modifications by mid-1987 with the modification themselves to

be completed based on an individual schedule.

This LER remains open pending completion of the licensee's

corrective actions and review of the licensee's plan for

addressing the Lifted Leads Task Force recommendations.

No violations or deviations were identified.

6. Review of Allegations (99014)

a. (Closed) Allegation (RIII-86-A-0161): On October 27,.1986,

Region III submitted the following concerns to IP for their review

and followup. On December 11, 1986, IP notified Region III by

letter U-600779 that their review and followup was completed. The

inspector reviewed IP's response to the concerns as documented

below.

Concern No. 1

Transco did not have the insulation top to place over valves, so

they continued to work without the valve tops. As a result, dirt

from the insulation process was trapped between the pipes and the

insulation. The individual thought the feedwater piping was the

system involved.

Review

The inspector reviewed the licensee's letter; reviewed the Transco

stainless steel cleanin

surveillance (CQ-01693)g of procedure SC-0394;

stainless steel piping reviewed

washdown an IPQA

techniques; and performed an inspection of a portion of the

insulated carbon steel feedwater piping. The inspector determined

the following:

(1) The allegation was assumed to be substantiated due to the ,

! lack of documentation which clearly reflected the date of

installation of the insulation tops for feedwater valves.

(2) The feedwater piping was carbon steel covered with a fairly

tight fitting insulation. Carbon steel would not be adversely

i

affected by dirt.

. (3) Due to the tight fit of the insulation surrounding the

l feedwater piping, only a limited amount of dirt and debris

.

could have been accessible to that piping. Any such foreign

niatter trapped between the piping and the insulation would

have negligible affect on heat transfer across the insulation.

.

l

(4) A specific Transco stainless steel cleaning procedure was used

at Clinton. Furthermore, the IPQA surveillance (CQ-01693)

conducted in November 1985 concluded that stainless steel

i

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cleaning operations were being performed in accordance with the

Transco procedure. Thus, it appears that sufficient controls

existed to prevent foreign matter from being trapped between

stainless steel piping and its insulation.

Conclusion

The allegation was assumed to be substantiated based on the lack

of documentation which identifies the date of installation of the

feedwater valve insulation tops. However, there is no safety

concern associated with this matter. This allegation is closed.

Concern No. 2

The individual heard that several voids existed in the concrete wall

between the Diesel Generator and Fuel Building. The individual

thought that the approximate locations of the voids were between

Column Lines AD and AC on either the 762' or 770' elevations. The

individual also heard that the voids were probably repaired.

Review

The inspector reviewed the licensee's letter; reviewed the concrete

pour traveler for the concrete wall immediately above the area

identified in the concern; and performed a visual inspection of the

subject wall.

The inspector identified the following:

(1) As documented in concrete pour traveler F-W-8-4, voids were

previously identified in the concrete wall in the general area

identified in the concern. The voids were subsequently

repaired by the licensee as documented on Quality Control

Inspection Reports C-83-879 and C-83-1350 and Nonconformance

Report (NCR) 3451.

l (2) A visual inspection of the subject concrete wall did not reveal

l the existence of additional voids.

Conclusion

t

l This allegation was substantiated in that voids had been identified

'

by the licensee in the general area of concern. However, the

j licensee had properly identified and documented these voids and .had

completed action to repair these voids. The actions were determined

to have been effective based on a satisfactory visual inspection of

the subject concrete wall by the inspector. This allegation is

closed.

I

I

20

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Concern No. 3

The inner bioshield wall was covered with "20 gauge reflective

insulation" and the insulation was improperly installed. Gaps and

dents existed in the insulation and dirt was trapped behind the

insulation. Also, sheet metal screws were missing and holes existed

in the insulation.

Review

The inspector reviewed the licensee's letter; reviewed the April 4,

1986, Bioshield Insulation Completion / Acceptance punchlist; and

performed a visual inspection of the inner bioshield wall

insulation.

The inspector determined the following:

(1) On April 4, 1986, a walkdown of the inner bioshield wall

insulation by Illinois Power (IP) identified approximately 33

items requiring rework. A number of these items were similar

to the types of concerns identified in the allegation. The '

insulation was subsequently rejected by IP on May 5,1986. On

July 3, 1986, the insulation was accepted based on the rework

of the 33 items and a subsequent satisfactory walkdown by IP.

(2) The visual inspection of the inner bioshield wall conducted by

the inspector on January 23, 1987, identified no deficiencies

in the installed insulation.

Conclusion

The allegation was partially substantiated in that a number of the

identified concerns were confirmed to have existed in May 1986.

However, the licensee had properly identified and documented these

concerns and had taken action to resolve these concerns. These

actions were determined to have been effective based on a

satisfactory visual inspection conducted by the inspector. This

allegation is closed.

! No violations or deviations were identified.

L 7. Region III Request (92701)

l

During this inspection period, the licensee performed a review of their

'

l computerized maintenance work request (MWR) milestone listing. The

'

purpose of this review was to identify, from the list of MWRs associated

l with the initial criticality milestone, those MWRs which more correctly

[ were required for either plant heatup or 5% power. As a result of their

j review, 74 MWRs had their associated milestone revised from initial

criticality to heatup, 5% power, or were given a "can be worked anytime"

l status.

l-

l

l

21

e

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.,-

The' inspector selected 13 MWRs from the 74 to determine if their removal

from the initial criticality milestone was compatible with established.

requirements. A subsequent review included a review of associated

technical specifications (TS) and a discussion of each MWR with plant

. personnel.

During the review, the inspector determined that three of the MWRs had

recently been closed. This was indicative of the licensee's stated

objective of completing all MWRs in an expeditious manner. The review'of~

all. but three of the remaining MWRs resulted in no identified concerns.

The-inspector did. identify initial concerns with MWRs C29044, B34737 and

C15708. Each of these MWRs had their. associated milestone changed from'

initial criticality to a "can be worked anytime" status. MWR C29044

pertained to the installation of the Reactor Core Isolation Cooling

(RCIC) headspray line to the reactor pressure vessel. A review of the

Clinton TS for.the RCIC system identified specific operability

. requirements for RCIC depending on vessel head status and plant pressure.

Therefore, the inspector was concerned with the removal of this MWR from ,

. the initial criticality milestone. MWR B34737, which pertained to-

electrical . cables for RCIC, was the subject of similar concern. Finally,

MWR C15708 dealt with an investigation of Intermediate Range Monitor

(IRM) spiking. The concern with this MWR pertained to the fact that the

, subject IRM continues to evidence spiking characteristics while being

classified as operable and shorting links removed. Continued operation

,

in this manner would increase the probability of future reactor trips.

During discussions with the licensee, the-inspector determined that MWRs

.

.C29044.and B34737 were identified on the IP Open Vessel. Testing Schedule

-as items required for completion prior to initial criticality. In

>

addition, IRM Special Test CPS No. 2830.11, which addresses the IRM

'

spiking problem identified in MWR C15708, was similarly identified on

the Testing Schedule. The inspector reviewed the Testing Schedule and

,

concluded that the work activities associated with MWRs C29044, B343737

and C15708 were adequately identified and controlled.

4

I

Based on the review of the 13 selected MWRs and the discussions with the

licensee, it appears that the milestone revisions for the 74 MWRs were

acceptable.

f No violations or deviations were identified.

.

8. Operational Safety Verification (71707)

The inspectors observed control room shift turnovers and operations,

attended selected pre-shift briefings, reviewed applicable logs, and

conducted discussions with control room operators during the inspection

period. The inspectors verified the operability of selected emergency

4

systems and verified tracking of LCOs. Routine tours of the auxiliary,

3'

fuel, containment, control, diesel generator, and turbine buildings and

the screenhouse were conducted to observe plant equipment conditions

including potential for fire hazards, fluid leaks, and operating

conditions (i.e., vibration, process parameters, operating temperatures,

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C

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'

etc). The inspectors verified that maintenance requests had been

initiated for discrepant conditions observed. The inspectors verified

by direct observation and discussion with plant personnel that security

procedures and radiation protection (RP) controls were being properly

implemented.

During a plant tour on January 7,1987, at about 2:30 p.m., the inspector

noted that sealing material (" Bisco Sealant") had been degraded on a

penetration through the secondary gas control boundary. The subject

penetration was located on the 781' elevation of the Auxiliary building

inside the East airlock. The penetration was used for a non-safety

related 1" conduit passing through the airlock into the annular space

within the secondary gas control boundary. The seal provided was

approximately 10" x 20" and the sealant had been cut into several pieces.

The inspector informed the Shift Supervisor of the condition noted above

and the licensee initiated Condition Reports 1-87-01-033 and 1-87-01-034

to document the deficiency and investigate the cause. At the conclusion

of this report period, the licensee was still performing their

investigation. This item will remain unresolved pending the inspectors

review of the licensee's investigation (461/87002-02).

The following routine surveillances were observed by the inspector during

the report period:

-

CPS No. 9080.01, revision 22 , " Division II Diesel Generator

Operability"

-

CPS No. 9031.12, revision 20, "APRM Channel Functional"

The inspector's observations of the above surveillances were limited

in scope. However, the inspector noted that the surveillances being

perfomed were current revisions; the personnel performing the

surveillances informed the control room operator when required by the

procedure; and the personnel performing the surveillances exhibited a

good working knowledge of the surveillance in response to the inspector's

questions.

The inspectors observed plant housekeeping / cleanliness conditions. No

i discrepancies were noted.

The above reviews and observations were accomplished to verify that

l facility operations were conducted in conformance with the CPS technical

l specifications and the conditions of the operating license.

One unresolved item was identified.

9. Engineered Safety Feature System Walkdown (71710)

The inspectors performed a complete walkdown of the High Pressure Core

Spray (HPCS) system during the report period to verify the system status.

l

At the time the walkdown was performed, the licensee had declared the

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HPCS system operable and meeting all requirements of the plant's

-Technical Specifications.

For the purpose of this walkdown, the inspector utilized the following

system drawings and the checklists contained in the system operating

procedure:

CPS No.;3309.01V001,= revision 1, HPCS Valve Lineup

CPS No.-3309.01V002, revision 0, HPCS Instrument Valve Lineup

CPS No. 3309.01E001, revision 1, HPCS Electrical Lineup

CPS No. 3309.01E002, revision 0, HPCS 120V AC Electrical Lineup

-P&ID M05-1074. sheet 1, revision Y~

C&ID M10-9074, sheets 1 through 4, revision A q

For the inspection performed, the following attributes were observed:

-

System lineup procedures matched the plant drawings.

-

Valve and electrical switch / breaker positioning agreed with

the lineup checklists.

-

Valves were locked when "equired.

-

Equipment conditions appeared correct with no evidence of

damage. (

-

. Equipment and components were properly identified.

-

Interiors of electrical and instrumentation cabinets'were free

of debris, loose material, uncontrolled jumpers, with no

evidence of rodents.

--

Instrumentation was properly installed and functioning.

-

Lubrication was provided, where observable.

-

Housekeeping was adequate and appropriate levels of cleanliness

were being maintained.

-

Support systems essential to system actuation (Division III

Shutdown Service Water and Division III Emergency Diesel) were

operational.

In conjunction with the above, the inspector reviewed the results of

current surveillances perfo ned on the HPCS system to verify Technical

Specification requirements were met. The following surveillance test

results were reviewed:

Surveillance No. Title Frequency Test Date

CPS No. 9051.01 HPCS System Pump Quarterly 10/29/86

Operability

= CPS No. 9051.02 HPCS Valve Operability Quarterly 11/01/86

Test

CPS No. 9051.03 HPCS System Functional 18 months 08/23/86

Test

CPS No. 9051.04 HPCS Automatic Suction 18 months 07/30/86

-0perability

CPS No. 9051.05 HPCS Discharge Header Monthly 01/06/87

Filled and Flow Path

Verification

24

.

.

The inspector concluded that the HPCS system was operable based on direct

field observations of the above lineups and inspection attributes. In

addition, the inspector's review of current surveillance tests for the

HPCS system indicated that the plant's Technical Specifications were

being met.

No violations or deviations were identified.

10. Onsite Followup of Events at Operating Reactors (93702)

a. General

The inspector performed onsite followup activities for events which

occurred during the inspection period. Followup inspection included

one or more of the following: reviews of operating logs;

procedures; condition reports; direct observation of licensee

actions; and interviews of licensee personnel. For each event, the

inspector reviewed one or more of the following: the sequence of

actions; the functioning of safety systems required by plant

conditions; licensee actions to verify consistency with plant

procedures and license conditions; and attempted to verify the

nature of the event. Additionally, in some cases, the inspector

verified that licensee investigation had identified root causes of

equipment malfunctions and/or personnel errors and were taking or

11ad taken appropriate corrective actions. Details of the events and

licensee corrective actions noted during the inspector's followup

are provided in paragraph b. below.

b. Details

(1) Engineered Safety Feature (ESF) Actuation - Partial Division II

(Inboard) Containment Isolation (ENS No. 07312)

On December 26, 1986, at about 1:06 p.m. CST, a partial

actuation of the division II containment isolation logic

occurred during performance of post modification testing by

the licensee's C&I maintenance technicians. The actuation

-

resulted in start of division II standby gas treatment and

shutdown service water systems and closure of several inboard

containment isolation valves. The actual cause of the event

was not known but the licensee suspected an error made by the

maintenance technician. The licensee returned plant equipment

to its normal operational status and notified the NRC

Operations Center of the event via ENS at about 3:55 p.m. CST.

This matter will be reviewed further during review of the

licensee's LER.

(2) ECCS Auto Initiation and Injection Into Reactor Vessel (ENS No. 07359)

At about 1:15 p.m. CST on January 2, 1987, division II

Emergency Core Cooling Systems (ECCS) automatically started

25

_ _ _

- . --_ _ - - _ ._ _ _

.

b

~

j

-and injected water into the Reactor Vessel in response.to a

spurious low reactor water level signal. At the time of

occurrence the plant was in mode;5 maintaining reactor water

level at about +170 inches. The division II RHR pumps (B

and C).were secured by plant-operators at a reactor vessel

' level of +200 inches after determination that an actual low

reactor vessel level condition _did not exist. All division

II= equipment responded as-expected and were returned to the

required: standby conditions for mode 5 operation. The

licensee initiated an investigation to determine the cause

of the spurious signal. This matter will'be reviewed further

-during review'of the licensee's LER.

(3) ESF Actuation - Shift of High Pressure Core Spray Suction

(ENS No. 07424)-

~

At about 10:00 a.m. CST on-January 9,1987,' the licensee

-discovered that the High Pressure Core Spray-suction path

had shifted from its preferred source (RCIC storage tank)

to the suppression pool. The licensee's initial investigation

indicated that the realignment occurred the previous day when

a RCIC storage tank level transmitter failed. At the time

of occurrence, the plant was in mode 4 and the High Pressure

Core' Spray system was not an operable ECCS. The licensee is

continuing-to investigate the cause for the RCIC level

transmitter failure. This event was similar to LER

50-461/86020-LL. This matter will. be reviewed further during

-

review of the licensee's LER.

(4) Inadvertent ESF Actuation During Alternate Rod Insertion

Surveillance Test (ENS No. 07468)

At about 8:58 p.m. CST on January 13, 1987, the scram discharge

volume vent and drain valves unexpectedly closed during

performance of surveillance testing on the alternate' rod

insertion system 1 (an ATWS protection feature). Investigation

by the licensee indicated that this ESF' actuation occurred due

to a missing step in the surveillance procedure being used. A

temporary procedure change was initiated and the surveillance

test was subsequently completed without further incident. The

licensee notified the NRC -Emergency Operations Center of this

event at 11:24 p.m. CST via ENS. This matter will be reviewed

further.during review of the licensee's LER.

.

I

(5) Loss of Emergency Response Facility (ENS No. 07472)

On January 14, 1987, at about 9:00 a.m. CST, the licensee

experienced a loss of power to the Emergency Offsite Facility

(EOF). The loss of power occurred when the offsite 138KV

transmission system feeding the EOF had a power failure

apparently due to a tree contacting the transmission lines.

26

r

4

.

4

The licensee restored power to the 138KV transmission system

and the EOF at about 11:00 a.m. CST. The licensee notified the

NRC. Emergency Operations center of this event at about 10:00

a.m. CST on January 14, 1987. This matter will be reviewed

further during review of the licensee's LER.

(6) Degraded Emergency Response Capability Due to Snow (ENS No. 07520)

-At 11:00 a.m. CST on January-19, 1987, the licensee determined

that their emergency response capability was degraded due to

the degraded condition of area roads that resulted from a

winter snow storm. Approximately 8 inches of snow had fallen

in the preceding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;: blowing and drifting snow caused

decreased visibility and made driving in the area hazardous.

The licensee notified the NRC Operations Center of this event

-

at 11:35 a.m. CST. By 1:20 p.m. CST, area road conditions had.

improved to the point that the licensee determined their

emergency response capability was no longer degraded. This

information was comunicated to the NRC Operations Center.

This matter will be reviewed further during review of the

licensee's LER.

(7) Inadvertent Actuation of Division I ECCS Equipment (ENS No. 07545)

>

At about 2:35 p.m. CST on January 21, 1987,. during performance

- of an operational pressure test of the reactor coolant pressure

boundary, the licensee experienced an inadvertent division I

,

!

ESF actuation. The actuation was caused by a hydraulic

transient in an-instrument sensing line that occurred while

'

an operator was restoring pressure transmitter C34-N005 to

service.

l The hydraulic transient caused a transient reactor vessel water

!

level low - level 2 signal which initiated the Low Pressure

Core. Spray system (LPCS), started the division I emergency

l diesel generator. and shut two containment isolation valves in

l- the instrument air system. The low pressure coolant injection

L mode of the residual heat removal system did not actuate since

r

that system was lined up in the shutdown cooling mode at the

time of the event. In addition, the LPCS did not inject into

the reactor vessel since the reactor vessel pressure was

elevated for the operational pressure test in progress.

i

Upon initiation of the above division I ECCS equipment, the

control room operators verified that an actual low level

condition did not exist in the reactor vessel and restored the

i division I ECCS equipment to the standby mode. The licensee

l

rotified the NRC operations center of this event at 5:10 p.m.

( CST. This matter will be reviewed further during review of the

i

l'

.

27

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_ licensee's,LER; however, as discussed below, the inspector _

performed'a review ~of thelsequence of plant operations that:

lead-up to thi.s event. 1

-

_

Discussion- >

sThe--inspector interviewed licensed operators on shift at the-

time of this. event to evaluate plant operations that:resulted

in'the need to isolate pressure transmitter C34-N005. :As

discussed above, the event described was initiated when

-

pressure transmitter C34-N005 was being returned to service.

The inspector's review was primarily limited to interviews of.

-on shift' operators since the sequence of" events documented

below was not documented in the main control room operator's

-log.

-

The licensee was performing an operational pressure test

maintaining'an elevated pressure in.the reactor-vessel via the-

Control Rod Drive system. With the plant in mode.4, Reactor-

Recirculation Pump-A operating in slow speed,.. reactor; vessel-

-and reactor coolant temperature at

_about

pressure at.about

160 degrees 700 psig,t,-the control. room operators were

fahrenhei

requested to-start Reactor-Recirculation Pump-B to allow

continuous recirculation while. testing-the Reactor Recirculation.

-

Pump-A Flow Control Valve. ' Operating the Reactor Recirculation

~

Pump was preventing thermal stratification _in the lower _ portion

-of the-reactor vessel.

Sequence of Events

Sometime before 2:00 p.m., the control room operators attempted

tostartReactorRecirculationPump-B(RRPumpB). After this

first attempt failed, the control room operators noted that?an

annunciator light on control room panel 1H13-P680 (same panel

as the RR pump controls) for "RECIRC PMP B TEMP-INTLK ACTUATED"

was lit..~The applicable annunciator ~ response procedure, CPS

No. 5003.21 identifies the possible cause for this annunciator

as follows:

~

1. Delta-T >50 degrees F between recirgulation loops.

2. Delta T >50 degrees F between vessel dome and either~

recirculation loop.

3. Delta.T >100 degrees F between vessel bottom drain

and vessel dome.

These thermal interlocks prevent undue stress to the reactor

vessel, reactor vessel nozzles, bottom head region, recircula-

i tion pumps, and recirculation nozzels. CPS No. 5003.21 also

provides the operator with actions to take in order to

determine which temperature interlock is causing the alarm.

.

28

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In addition, identification of the interlock relay (K687).and

~

the specific. electrical drawing reference (E02-1RR99, sheet

~10) is provided.

The control room operators then discussed the possible cause

for the annunciated interlock and reviewed control- room-

electrical drawings. The operators decided to place the " Steam

Line Delta T Interlock" bypass switch-located on the Reactor

Recirculation system Low Frequenc

Auxilary Relay Panel (1833-P0018)y Motor Generator

in bypass. This action(LFMG)

was

carried out and a second attempt was made sometime after

2:00 p.m. to start RR Pump B. This second attempt also failed.

The inspector assumed that the annunciated interlock was still

-lit on 1H13-P680 during the second attempt, since the " Steam

Line. Delta T Interlock"~ switch that was placed in bypass

appeared to be the 8 degree F pump cavitation interlock

described in section 5.4.1.3 of-the Final Safety Analysis

Report. 'That bypass switch would have no effect on the

interlock that was preventing the RR pump start.

The control room supervisor then reviewed the electrical

drawings again with the assistance of an individual more

intimately familiar with the Reactor Recirculation system

interlocks and determined that the annunciated interlock could

be bypassed by isolating pressure transmitter C34-N005 and

venting the instrument. This pressure transmitter was sensing

the elevated pressure (700#) present in the reactor vessel

due to the ongoing operational pressure test. However, since

reactor coolant temperature was only 160 degrees F, the

relationship for the interlock Saturation Temperature =

Saturation Pressure (Tsat=Psat)[] was Thenot valid.

control '

room supervisor then directed that pressure transmitter

C34-N005 be' isolated and vented. This action was carried out

and the RR Pump B was successfully' started on the third attempt

at 2:30 p.m. The only documentation of the events described

above was'then made in the Control Room Operators log book:

"1430 - Started RR pump B".

Technical Specification Surveillance 4.4.1.4 (applicable in

mode 1, 2, 3.& 4) requires that temperature differentials and

flow rate shall be determined to be within specified limits

within 15 minutes prior-to startup of an idle recirculation

loop. The operating procedure in use, CPS No. 3302.01,

revision 2 " Reactor Recirculation (RR)". required in step

8.2.3.1 that "within 15 minutes prior to starting each Recirc

Pump] verify

flow ... are steps

at the . . . [di/ferential

required temperatures

flow and temperature and and

range loop

log the data in the CR0 [ Control Room Operator] log". As noted ,

above, contrary to the procedural requirements, the operators

attempted to start RR Pump B twice and successfully started the

pump on the third attempt without documenting in the CR0 log

the differential temperatures and loop flow. Subsequent review

29

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.:

.

by the licensee identified 6 of 8 additional RR Pump starts

while in mode 4 where the required CR0 log entries were not

made. The failure to follow approved procedures is a violation

of 10CFR50, Appendix B, Criterion V.(461/87002-03A(DRP)).

The procedure in use, CPS No. 3302.01, " Reactor Recirculation

(RR)", revision 2 dated February 27, 1986, did not address the

plant conditions under which plant operators attempted to start

'RR Pump interlock

an invalid B on January)21,1987 -(i.e.,

. The control room elevated bypassed

supervisor pressure causing

the differential temperature interlock by directing pressure

transmitter C34-N005 be isolated and_ vented. .No attempt was

made by control room operators to revise the procedure or use

other administrative controls immediately available. Those

administrative controls would have provided independent review

of the safety significance and approval of the method used

to bypass this interlock. The failure to provide adequate

instructions in procedure CPS No. 3302.01 or other documented

instructions _is a violation of 10 CFR 50, Appendix B, Criterion

V(461/87002-03B(DRP)).

The inspector noted that the sequence of events described

above may indicate a general unfamiliarity with Technical

Specification requirements. The inspector's basis for.this

observation is the multiple failures to record information

required for RR Pump starts-in mode 1, 2, 3, & 4 coupled with

the short time that the plant has been in mode 4 (about one

month). In addition, the difficulties encountered by the on

shift crew in starting a RR pump under the conditions present

may indicate a weakness in the level of system knowledge by

the operators. These observations were discussed with licensee

management at the conclusion of the report period.

(8) ESF-Actuation of Two Containment Isolation Valves (ENS No. 07564)

At about 7:30 p.m. CST on January 22, 1987, during performance

of a channel functional test (CFT) for reactor water level 1,

the licensee experienced an inadvertent actuation of two

containment isolation valves in the instrument air (IA) system.

. The isolation resulted from the failure of a C&I technician

to properly insulate leads lifted in accordance with the CFT

,

procedure to preclude actuation of the affected containment

isolation valves. The technician had successfully completed 2

of 4 channels to be tested prior to the event by individually

taping back the lifted leads. During performance of the 3rd

channel, the technician taped the lifted leads together thereby

maintaining electrical continuity which resulted in the ESF

j

actuation as reactor vessel water level (simulated for the CFT)

i reached the actuation setpoint. The logic involved is a 1 out

of 4 actuation logic. The licensee restored the IA system to

,

normal and successfully completed the remaining portions of the

j' 30

1

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v.,.,

,.- . , - - .- _ . - - . - . ...-- _

'

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'

'CFT. 'The licensee notified the NRC Operations Center of this

event at 11:15 p.m. CST:on January 22, 1987. This matter will

be reviewed further during review of the licensee's LER.

(9) ESF Actuation - Closure of Shutdown' Cooling Suction Valve-

-

(ENS No. 07565) ,

,

f

At.about 8:00 p.m. CST on January 22,.1987,-whil'e restoring

from'an operational pressure test of the reactor coolant

system, the licensee experienced an unexpected closure of the

shutdown cooling inboard suction isolation valve (IE12-F009).

During performance of. the operational' pressure test, valve

1E12-F009lwas open and the valve motor controller was -

deenergized to prevent automatic closure when reactor vessel

pressure was increased. During restoration from the test, the

. procedure in use did not direct the reset of the reactor high

'

pressure seal-in logic (actual reactor vessel pressure was

,

'

below the actuation setpoint at the time of the event). The

plant operators immediately reset the seal-in logic and opened

'

valve 1E12-F009. At the time of occurrence, the plant was in

j . mode 4 and depressurized. The licensee notified the NRC

'

Operations Center of this event at 11:15 p.m. CST on

. January 22, 1987. This matter will be reviewed further during

j review of the licensee's LER.

v

i One violation was identified.

I 11. Management Meeting (30702)

J

, - On January 16,1987, NRC management met with IP management at the Clinton

'

Power Station Visitor Center to discuss the status of the facility, the

licensee's Monthly Performance Monitoring Management Report and actions

, being taken to enhance the licensee's performance in several areas, and

to discuss the readiness of Clinton Power Station to perform initial

critical reactor operation and to generate electricity. Personnel

attending the meeting are identified by (#) in paragraph 1. of this ,

! report. "

p

Mr. A. B. Davis, the Deputy Regional Administrator, opened the meeting >

. with a brief discussion of Region III procedures for NT0Ls at the-full

l' power license stage. Mr. Warnick then identified the scope of the

meeting; discussed current Region III areas of concern and an overview

4

of Region III plans for additional inspections at Clinton prior to

making a decision regarding the recommendation for issuance of the full

power license; and requested that periodic management meetings continue

monthly for the near future.

.The licensee then provided the status of testing deferred beyond fuel

load; the status of surveillance testing required to be completed to

support plant operation in modes 1 and'2; the status of maintenance and

,

'

modification work required to be completed prior to initial criticality

and subsequent milestones; and the status of deficiency documents

i

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(condition reports, nonconforming material reports, and licensee event

reports) applicable to plant milestones.

The licensee discussed the ::tatus of actions being taken to address

recent NRC concerns related to their maintenance and modification

programs; recent accomplishments and current problem areas. The

licensee stated that management attention was being directed to the

radiation protection (RP) areas to assure the readiness of the RP

program to support critical reactor operation.

The licensee projected completion of all work necessary to achieve

initial reactor criticality shortly after January 25, 1987. The

licensee stated that the Region III Administrator would be contacted

by the IP Vice President - Nuclear prior to his authorizing the plant

operators to make the reactor critical. The licensee further stated

that IP expects to proceed directly in their power ascension program

through test condition 1 at which time they intend to shut down the

reactor for a short, scheduled maintenance outage. The licensee

expected to be ready to need a full power license by the end of February.

NRC (Region III) management acknowledged the licensee's status and plans,

and noted that the Commissioner's agenda currently planned a full power

license briefing for February 24, 1987.

The meeting concluded with a tentative agreement to meet again on

February 13, 1987 at the Clinton site with a similar agenda.

12. Open Items

Open items are matters which have been discussed with the licensee, which

will be reviewed further by the inspector, and which will involve some

action on the part of the NRC or licensee or both. One open item

disclosed during the inspection was discussed in paragraph 2.f.

13. Unresolved Items

Unresolved items are matters about which more information is required in

order to ascertain whether they are acceptable items, violations, or

deviations. One unresolved item disclosed during this inspection was

discussed in paragraph 8.

14. Exit Meetings (30703) '

The inspectors met with licensee representatives (denoted in paragraph 1)

throughout the inspection and at the conclusion of the inspection on

January 26, 1987. The inspectors summarized the scope and findings of

the inspection activities. The licensee acknowledged the inspection

findings. The inspectors highlighted the need for management attention

to internal commitments and the CPS emergency off-normal procedures.

The inspectors also discussed the likely informational content of the

inspection report with regard to documents or processes reviewed by the

32

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! inspectors during the inspection. The licensee did not identify any such

documents / processes as proprietary.

.

The resident inspectors attended exit meetings held between Region III

based inspectors and the licensee as follows:

Inspector (s) Date

.Wohld .1/15/87

Hasse 1/15/87

l Foster- 1/15/87

.

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