ML20206J495

Insp Rept 50-483/86-10 on 860415-0531.Violations Noted: Failure to Maintain Intermediate Head Safety Injection Operable & to Notify NRC within 4 H of Event.Unresolved Items Re Limitorque Valves Also Noted

ML20206J495
Person / Time
Site:	Callaway
Issue date:	06/19/1986
From:	Forney W, Wohld P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206J433	List: ML20206J430 ML20206J495
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-2.F.2, TASK-TM 50-483-86-10, IEIN-86-003, IEIN-86-3, NUDOCS 8606270181
Download: ML20206J495 (18)

See also: IR 05000483/1986010

Text

. - - .-

. .

I

U. S. NUCLEAR REGULATORY COMMISSION

REGION III

L

Report No. 50-483/86010(DRP)

Docket No. 50-483 License No. NPF-30  !

Licensee: Union Electric Company -

l

Post Office Box 149 - Mail Code 400

St. Louis, M0 63166  ;

Facility Name: Callaway Plant, Unit 1

Inspection at: Callaway Site, Steedman, M0 f

Inspection Conducted: April 15 through May 31, 1986

Inspectors: B. H. Little  !

'

C. H. Brown

M.

U

O.L.

R. h

Wohld 4/n/n

Date

LO.L.

Approved by: W. L. orney, lef 4/n/H,

Reactor Projects Section 1A Date

Inspection Summary

Inspection on April 15 through May 31, 1986 (Report No. 50-483/860101',DRP))

Areas Inspected: A routine unannounced safety inspection by the resident

inspectors and one Region III inspector of licensee actions on previous

inspection findings, licensee event reports followup, followup on regional

requests TMI NUREG-0737 items closure, inspection of licensee events, monthly

surveillance, operational safety verification, monthly maintenance, Cycle 2

startup.

Results: Two unresolved items relating to EQ of Limitorque Valves discussed

in Paragraph 4. Two violations were identified in Paragraph 6., failure to

maintain intermediate head safety injection operable, and failure to notify

the NRC within four hours of an event.

$$f2{ggj{ ha g3

0

.

• O

DETAILS

1. Persons Contacted

D. F. Schnell, Vice President, Nuclear

S. E. Miltenberger, General Manager, Nuclear Operations

• G. L. Randolph, Manager, Callaway Plant

C. D. Naslund, Manager, Operations Support

A. P. Neuhalfen, Manager, Quality Assurance

• J. D. Blosser, Assistant Manager, Operations & Maintenance

• J. R. Peevy, Assistant Manager, Technical Services

P. T. Abbleby, Assistant Manager, Support Services

W. F. Powell, Assistant Manager, Materials

M. E. Taylor, Superintendent, Operations

D. E. Young, Superintendent, Maintenance

W. R. Robinson, Superintendent, I&C

R. R. Roselius, Superintendent, Health Physics

V. J. Shanks, Superintendent, Chemistry

J. A. Ridgel, Superintendent, Radwaste

G. J. Czecchin, Superintendent, Planning & Scheduling

W. H. Sheppard, Superintendent, Outages

J. M. Price, Superintendent, Training

G. R. Pendergraff, Superintendent, Security

J. E. Davis, Superintendent, Compliance

D. W. Capone, Manager, Nuclear Engineering

W. R. Campbell, Assistant Manager, Nuclear Engineering

A. C. Passwater, Superintendent, Licensing

T. H. McFarland, Superintendent, Design Control

R. D. Affolter, Superintendent, Systems Engineering

D. C. Poole, Consultant

W. H. Stahl, Supervisor, Engineering

• B. K. Stanfield, Assistant Engineer

• S. Petzel, Engineer

^W. R. Bledsoe, Engineer, Compliance

• Denotes those present at one or more exit interviews.

In addition, a number of equipment operators, reactor operators, senior

reactor operators, and other members of the quality control, operations,

maintenance, health physics and engineering staffs were contacted.

2. Licensee Actions on Previous Inspection Findings (92701)

(Closed) Open Item (483/84-48-01(DRP)): Licensee plans for modification

of the pressurizer power operated relief valves (PORV). Based on problems

identified during testing of the PORVs at Wolf Creek Generating Station,

and the similarity of the Callaway and Wolf Creek Units, Westinghouse

issued a Field Change Notice (FCN) No. SCPM 10712 for Callaway to inspect

and record measurements of annular orifice gaps formed by the valve body

2

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

. . _ -

. ._ _ _ .

.

. .

1

and valve cage.

Valvemodification(machiningofthecagerib)includedissubject

to inspection findings of inadequate clearances. The licensee

this FCN in Callaway Modification Package CMP-84-0651A, and accomplished

the work during the Cycle 1 refueling outage.

The licensee's inspection determined that machining of the valve's cage

rib was re The work was accomplished under Work Request Nos. 54996

and 54997. The quired.

inspector reviewed the applicable WRs including the

quality control inspection records. The inspector determined that the

specified work was completed and the valves were successfully tested.

(Closed Open Item During inspection 50-483/84-16,

,

thelice)nseeagreed(483/85006-01(DRSS)):to calibrate their gaseous and l

with appro)riate gaseous and liguid sources during the first refueling

outage. T1ese calibrations would be in addition to the solid source

calibrations currently performed. The monitors involved include the

. radwaste building monitor GH-RE-10B, the unit vent

and the liquid radwaste effluent monitor HB-RE-18. gas monitor GT-RE-21B,

Inspection in this matter included the review of Callaway vendor procedures

HTP-ZZ-04154,iological

issued by Rad and Chemical Technology, Inc.HTP-ZZ-04155 and

The licensee's calibration of the gaseous and liquid effluent monitors

has been completed.

I (0 pen)UnresolvedItem(483/85007-01(DRS)): Closure surveillance testing

, of normally closed check valves that perform a safety function in the

closed position. Per a )revious agreement

50-483/86012()RS))thelicenseep(asreportedinInspectionrovided

'

Report No.

a related check valves at the Callaway Plant for inspector review. A

preliminary, onsite review of the check valve list was done in an attempt

, to categorize each valve as tested or not and to determine at least one

! obvious closure requirement for each valve. The results of this review

indicate that there are a number of check valves not being closure tested

j that should be.

'

In the listing total of 225 safety related check valves, the review

results were as follows

) There are 143 check valves not tested for closure

l 8 - Prevent reverse flow through an idle pump in parallel pump

1 combinations.

4

22-Isolateseismic/non-seismicpipeboundariesordifferentpipeclasses

j (suchasthosewhichdefinetheLOCA/non-LOCApipeboundaries).

8 - Direct auxiliary feedwater flow to the proper steam generator during

,

postulated accident conditions.

i

3

!

,

_ - - - . - . -.

. .

10 - Prevent overpressurization of auxiliary feedwater pump suction

piping.

4 - Prevent blowdown from one interconnected steam generator to another

during steam fault conditions.

23 - Had no apparent safety related closure requirement.

,

l 25 - Did not appear to be properly listed as safety related (apparent

listing error) or the v11ve is not used (internals removed).

43 - Were not identified either as having or not having a closure safety

function during the inspection because their function was not

apparent from a cursory drawing review.

There are 65 valves tested for closure

'

10 - Are normally open check valves which are tested per the ASME Code,

Section XI, closure stroke test requirements.

31 - Are tested to provide a high/ low piping interface isolation

< (includes WASH 1400, Event V valves).

14 - Are leak tested for containment isolation.

10 - Are tested per the concerns raised by the inspector under this

unresolved item.

} Others (17)

I 2 - Are visually inspected for " free swing" on a periodic basis.

13 - Are fire protection system check valves which are covered by

separate fire protection requirements and inspections.

2 - Are essential service water check valves that are not tested but the

i failure of which would be readily identified by the affect on normal

j system operation.  ;

The inspector emphasized that the onsite check valve review was only a

" quick look" and that an in-depth evaluation is needed by qualified

system engineers. The licensee agreed to perform and document a review

i of the check valves on the list for their required closure functions and  !

closure test requirements. The staff indicated that this would be

i available for inspector review in approximately 90 days.

No violations or deviations were identified.

l

3. Licensee Event Reports (LERs) Followup (92700)

Through direct observations, discussions with licensee personnel, and the

review of records, the following LERs were reviewed to determine that the

I

,

l

, 4

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

. .

I

'

events were documented and evaluated, reportability requirements were

fulfilled, and appropriate corrective measures had been implemented.

(Closed) LER 85-010-00: Reactor Trip on Partial Loss of Feedwater Flow.

On February 21, 1985, a reactor trip and associated actions occurred due

to low steam generator water 1e wis. The low water levels occurred when

power was secured to a feedwater control panel which in turn resulted in

one main feedwater pump shutting down. The remaining feedwater pump does

not have the capacity for 100% reactor power. The loss of power to the

control panel was due to the failure of the supply transformer. The

transformer was replaced and a plant recovery was made. The transformer

was found to contain foreign material and a procedure was issued to

inspect and clean the transformers on an 18-month :chedule. The transformer

maintained the design safety function of isolation; therefore, a Part 21

report was not issued.

(Closed) LER 85-021-00: Inadequate Seismic Qualification of Class IE

Batteries. On April 4, 1985, the licensee was notified of a potential

problem relative to the spacing between the Class IE batteries and the

battery racks. The immediate corrective action was to insert spacers

between the end cells and the battery rack end stringers. The licensee

and contractor evaluated the corrective action and considered the spacers

a permanent fix.

(Closed) LER 85-023-00: Inadvertent Engineered Safety Features Actuation.

At three different times, April 13, April 17, and May 6, 1985; inadvertent

containment purge isolation and control room ventilation isolation signals

were received. The cause was a faulty vacuum transducer in a radiation

monitor. With a joint effort between the monitor vendor and several

plants that had experienced similar problems, a more reliable transducer

was developed and installed. The modification has apparently solved the

problem.

(Closed) LER 85-025-02: Intermediate Range Hi Flux Reactor Trip. On

l May 6, 1985, a reactor trip and associated actions occurred during a

l reactor startup (power 0% and in Mode 2) due to a intermediate range high

flux signal. The spurious signal was caused by a fuse blowing in the

neutron monitoring channel. The vendor determined that a faulty switch

caused the fuse to blow. The switch replacement appears to have solved

the problem, but the failure mode of the switch and other possible causes

that would overload the fuse are still being evaluated.

(Closed) LER 85-026-00: High Negative Flux Rate Reactor Trip. On June 7,

1985, a reactor trip occurred from 100% due to a high negative flux rate.

l The negative rate was generated due to rod drop when four rod control

pnwer supplies failed (thyristor bank insulator failure) during trouble-

shooting an immovable control rod. The failed equipment was replaced and

the review shows the failure to be an isolated occurrence and no further

I action was to be taken. The immovable control rod was due to a loose

terminal screw.

5

. .

(Closed) LER 85-042-00: Inadvertent Reactor Trip. On October 2, 1985, a

reactor trip from 100% power occurred due to personnel error. Instrument

and Control personnel were performing test ISP-BB-0T002, "RTD Calibration

Verification" (a one time test) when an abnormal resistance reading was

found on the No. 4 RCS loop, the loop was in test condition, and trouble-

shooting was commenced. Loop No.1 RTD terminal block test point was

erroneously taken from the prints. When the troubleshooting process

checked these terminals a signal was induced in loop No. 1 RCS protection

which completed the 2 out of 4 logic for the reactor trip from over

Temperature Delta T and Over Power Delta T signals. The failure to notify

the shift supervisor before commencing troubleshooting was a failure to

follow plant policy procedures. The error in reading the print contributed

to the event. The test procedure did not allow troubleshooting if a

problem was located. The test procedure also had an error which resulted

in the abnormal resistance reading. The personnel involved were counseled

on procedure and policy compliance.

No citation was issued since under the Enforcement Policy this was

considered a Technical Specification violation of lesser severity which

was identified and satisfactorily corrected by the licensee, and no

further violations of a similar nature have occurred. This item is

considered closed.

(Closed) LER 85-043-00: Technical Specification Hourly Firewatch Patrol

Missed. The Technical Specification 3.7.10.2 requires hourly firewatch

patrols to be established within one hour. On October 3, 1985, the hourly

patrol was not established for one hour and twenty-five minutes in the

south electrical cable chase due to a misunderstanding of the firewatch

personnel. The firewatch personnel were retrained on T/S requirements for

the firewatch patrols. Also the operations personnel, if possible, verify

the patrol is established before taking a fire protection system out of

service.

No citation was issued since under the Enforcement Policy this was

considered a T/S violation of lesser severity which was identified and

satisfactorily corrected by the licensee, and no further violations of a

similar nature have occurred. This item is considered closed.

(Closed) LER 85-045-00: Technical Specification 3.7.10.2 Violation Due

to Personnel Error. On October 16, 1985, a portion of the sprinkler

system for the auxiliary building 2000 feet elevation cable trays was

not identified as inoperable during surveillance testing. Therefore, a

continuous firewatch per Technical Specification 3.7.10.2 was not

established for about 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />, although a hourly firewatch did patrol

these areas for this time period. The delay was due to electricians and

engineers failing to communicate and recognize that sprinkler alarms were

also present on the multiplexer that was being worked. The sprinV er

system was inoperable due to a failed " supervision actuation module"

which indicated the loss of the ability to actuate the pre-actuation

sprinkler system in these areas.

6

. .

The surveillance procedure MSE-KC-FW001, " Technical Specification (T/S)

Fire Detection Functional and Supervisory Operability Test", has been

revised to include the alarms that should be actuated or cleared during

the applicable steps of the procedure. The maintenance department was

provided a set of the electrical prints for the fire protection systems

and the electricians received training on the fire protection system.

The fire detection in these areas, which provide an alarm in the control

room, were functional during the time the above sprinkler system was

inoperable.

No citation was issued since under the Enforcement Policy this was

considered a Technical Specification violation of lesser severity which

was identified and satisfactorily corrected by the licensee, and no

further violations of a similar nature have occurred. This item is

considered closed.

(Closed) LER 85-047-00: Operation with a Condition Prohibited by Technical

Specifications. On October 18, 1985 at 1145 CST, the plant entered

Technical Specification 3.0.3 due to both centrifugal charging pumps

(CCPs) being inoperable. "A" CCP had been taken out of service for

maintenance and later "B" CCP's room cooler fan was discovered to have

broken drive belts. The "A" CCP was made operable in less than an hour

and preparations for a plant shutdown were suspended. The belts were

replaced the following day. This was considered to be caused by equipment

failure.

No citation was issued since under the Enforcement Policy this was

considered a Technical Specification violation of lesser severity which

was identified and satisfactorily corrected by the licensee, and no

further violations of a similar nature have occurred. This item is

considered closed.

(Closed) LER 85-050-00: Inadvertent Engineered Safety Features Actuation.

On November 27, 1985, a control room ventilation isolation and a

containment purge isolation occurred. The cause appeared to be a fuse

failure at the microprocessor for the containment purge radiation monitor

and a tripped breaker supplying power to the monitor's flow pump. The

containment was not being purged at the time and the redundant monitor

remained operable. Troubleshooting found no equipment damage. The fuse

was replaced and the breaker closed with no further problem.

(Closed) LER 86-008-00: Technical Specification Violation. On April 3,

1986, while the plant was in Mode 5 (Cold Shutdown), the licensee

determined that Train "A" Control Room Emergency Ventilation System

(CREVS) had been inoperable since March 18, 1986. The "A" Train became

inoperable when the air conditioning unit was deenergized to permit

inspection and repair of the CREVS fire dampers. Technical Specification

(T/S) 3.7.6, Action Statement reouires that when in Mode 5 or 6; "With

'

one Control Room Emergency Ventilation System inoperable, restore the

inoperable system to OPERABLE status within 7 days or initiate and

maintain operation of the remaining OPERABLE Control Room Emergency

Ventilation System in the recirculation mode".

7

_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _

_-.

,

. .

On April 3, 1986, in response to a work request by maintenance personnel

to commence work on the fire dampers, control room personnel found that

the air conditioning unit ("A" Train) had been tagged out since March 18,

1986, without having placed the "B" Train CREVS in the recirculation

mode. This action should have been taken by March 25, 1986. Control

room personnel immediately placed the "B" Train CREVS in the recirculation

mode and documented the violation on Incident Report No.86-096.

The licensee's evaluation of the event determined the root cause to be

operations personnel's initial failure to correctly assess the

" operability" impact of tagging out the air conditioning unit.

Consequently, the "A" Train was not declared " inoperable" (no entry in

the equipment out of service logs).

To prevent recurrence, a procedural change was written to require an

independent review of WPAs, Equipment Out of Service Logs, Temporary

Modifications and Locked Components for each ascending mode change.

Also, personnel involved were re-instructed concerning T/S operability

requirements. This event has been included in the licensed operator

requalifi stion program, " Lessons Learned".

The inspector determined that, once identified by the licensee, action

was promptly taken to correct the condition and report the violation.

During the period that the air conditioning unit was out of service,

control room temperatures were maintained below 84 degree F (the

temperature specified in T/S 4.7.6.a.). The Train "B" CREVS and Train

,

"A" pressurization and filtration systems were operable and would have

'

protected control room personnel from airborne contamination if needed.

The event posed no threat to public health and safety.

i No citation was issued since under the Enforcement Policy this was

'

considered a Technical Specification violation of lesser severity which

was identified and satisfactorily corrected by the licensee, and no

further violations of a similar nature have occurred. This item is

considered closed.

(Closed) LER 86-015-00: Auxiliary Feedwater Actuation System (AFAS):

!

When PK51 Feeder Breaker Was Inadvertently Tripped. On April 23, 1986,

while in Mode 1 with reactor power at 31%, the control room received

numerous alarms on the annunciator boards. Letdown and makeup to the

volume control tank was lost, excess letdown and AFAS was initiated.

Plant conditions were determined to be stable, and no reactor trip

occurred.

l The licensee control room personnel's initial investigation determined

that a non-vital power breaker PK51 had been manually opened. Based on

no work having been authorized involving PK51, the event was considered

as possible tampering. The control room personnel promptly notified the

shift security supervisor and the plant manager. After hearing public

address (PA) instructions for the shift security officer to contact the

control room, the NRC resident inspector responded by going to the

4

control room.

8

- - - _

. .. . - _

.. .

, The inspector observed licensee's immediate response to the event which

included additional security measures being implemented. The inspector

was advised of licensee's planned investigation in this matter.

On April 24, 1986, a radchem technician acknowledged to the licensee that

he had inadvertently opened breaker PK51 on April 23, 1986. The

technician was a contract employee during the recent Cycle 1 refueling

outage and following the outage he was hired by Union Electric.

The NRC senior resident inspector and Region III security specialist

interviewed members of the licensee's staff, including the technician

involved in the event. The inspectors were satisfied that the information

provided by the technician, with regard to operating switch PK51, was in

agreement with control room observations during the event.

The licensee's investigation determined that the event resulted from

unauthorized operation of breaker PK51. The technician's action, although

well meaning, highlighted existing weaknesses in the licensee's "new hire"

indoctrination program. Specifically, organizational interface,

departmental authority / responsibility, and administrative controls on work

and safety practices.

The licensee has implemented a radwaste " Indoctrination Checklist for New

Personnel", and revised the general employee training program to assure

responsibility / authority for equipment operation is stressed. This

incident will be included in the licensee's requalification training

program.

The inspector determined that the licensee response was prompt and

thorough and that action has been taken to prevent recurrence.

No other violations were identified other than those noted above that

were identified, reported and corrected by the licensee.

4. Followup on Regional Requests (92701)

a. Temporary Instruction (TI) 2515/75, " Inspection of Limitorque Motor

Valve Operator Wiring"

An inspection was performed to ascertain the environmental qualifi-

cation (EQ) of wiring used in Limitorque Motor Valve Operators. The

inspection included the following:

Physical inspection of Limitorque operator wiring to determine

what wiring is actually installed in the operators.

Review of licensee's environmental q0alification documentation

to ensure qualification of wiring is sdequately established.

Review of licensee's action relative to IE Information Notice

(IN) 86-03, " Potential Deficiencies In Environmental

Qualification of Limitorque Motor Valve Operator Wiring".

"

9

-

__ _ _ _ - _ _ _ _

_ -.

. .

t

Physical Inspection: On May 22, 1986, the inspector performed an

'

in plant inspection of four Limitorque Motor Valve Operators. The

'

inspector selected the below listed valves based on a review of

SNUPPS Final Safety Analysis Report (FSAR), Table 3.11(B)-3,

" Identification of Safety-Related Equipment and Components". Three

valves selected are located in the reactor containment building and

one is located in the lower piping penetration room of the reactor

auxiliary building. The inspector was accompanied by representatives

from licensee's Quality Assurance, Quality Control, Engineering, and

Maintenance Departments.

VALVE NO. DESCRIPTION

EP-HV-8808B Safety Injection Tank Outlet Isolation Valve "B"

EP-HV-8808C Safety Injection Tank Outlet Isolation Valve "C"

BB-HV-8037B Pressurizer Relief Tank (PRT) Outlet Isolation Valve

EM-HV-8835 Safety Injection Discharge to Cold Leg Injection

Isolation Valve

Inspection Findings:

VALVE NO. * TERMINAL WIRING * FIELD WIRING

EP-HV-88088 Raychem Flamtrol G34

C02, C03, C04

'

G2A

EP-HV-8808C Raychem Flamtrol G34

CO2, C03, C04

G2A

BRAND Rex #43

BB-HV-8037-B Rockbestos G31

Firewall III C07

EM-HV-8835 Raychem Flamtrol G31

CO2, C04

• Each motor operator contained some terminal and field wires which

either lacked identification markings or with unreadable markings.

However, these wires were similar (size / color) to other wires

identified by markings. In addition, limit switch space heater

wiring was unmarked, approximately 20 AWG size. SNUPPS Report of

Independent Review of EQ Programs (Response to NUREG 0588) states:

"In all cases, the limit switch space heater is connected in a

Class IE circuit. Since the heater failure mode will result in

an open circuit, it is considered that the heaters need not be

qualified. However, Limitorque has performed an accident test

10

.

. .

on a heater to demonstrate that the heater remains operative following

seismic aging and a simulated LOCA (i.e., it would not fail in a

manner detrimental to plant safety.)

Review of Licensee's EQ Documentation

The inspector reviewed the below listed EQ Test Reports and determined

that the operator wiring in the four cperators inspected had been

environmentally qualified.

EQ TEST REPORT NOS. TERMINAL WIRING

'

Franklin Institute Research Raychem Flamtrol

Laboratories Test Nos. E-031.2, E-031.3

Rockbestos Test Report Nos. E-057-020-03, Rockbestos

E-057-021-06, E-057-036-02, E-057-050-02 Firewell III

FIELD WIRING

Rockbestos Test Report Nos. CO2, C03, C04,

E-057-020-03, E-057-021-06, C07

E-057-036-02, E-057-050-02

ANACONDA Test Report No. G2A, G31, G34

E-58-0005-03

BRAND Rex Test Report No. BRAND Rex

E-0578-0014-02

The inspector reviewed licensee plant walkdown sheets to determine

if licensee's identification of operator wiring was in agreement

with the inspectors findings for the four operators inspected. The

licensee's quality records of plant walkdowns included inspections

performed through Startup Work Requests (SWRs) and Quality Control

(QC) checklists. The licensee records were in agreement with the

inspector's findings.

Licensee's Response to IE Information Notice 86-03

In January 1986, the licensee performed a Quality Assurance (QA)

surveillance on EQ of Limitorque Motor Valve Operator wiring (QA

Surveillance Report No. P8601-12). The surveillance determined that

each operator was field inspected prior to initial plant startup

using a Startup Work Request (SWR). QA sample inspection of

approximately 10% of the SWRs determined that the operator wiring

was environmentally qualified.

Union Electric Nuclear Engineering (UENE) in response to Information

Notice 86-03 performed a review of all SWRs relating to the field

inspection of Limitorque Valve Operators. This review identified

six valves as having suspect internal wiring. Nuclear Engineering

11

. -. -- .- -

. - _ __ . - = .__ __ __ _ - . _ . _. _ . _ _ _ . . __.

. .

memorandum No. 558, dated March 11, 1986, requested site engineering

to perform a field inspection to identify the installed wiring. The

results of the field inspection was as follows:

VALVE NOS. WIRING EQ STATUS

BB-HV-8000A Rockbestos Qualified

BB-HV-8000B Raychem Qualified

EJ-HV-8701A Raychem Jualified

j EJ-HV-8716A,B Not Identifiable Questionable

EJ-HV-8809B Techbestos 14 AWG Questionable

4 600-V ,

The licensee has replaced the operator wiring associated with

EJ-HV-8716A,B and EJ-HV-88098 with environmentally qualified wiring

l and has requested Westinghouse response regarding the EQ status of

the wiring removed.

4

The inspector determined that the licensee was responsive to IE

, Information Notice 86-03 and took prompt corrective action to ensure

i

that the installed operator wiring is environmentally qualified.

1

-

The licensee's evaluation of the EQ status of the wiring replaced is

in progress. This matter is unresolved pending further NRC review.

j Unresolved Item No. 483/86610-01(DRP)

Motor Operated Valves (MOV) Conduit Seals

'

+

On May 28, 1986, the licensee advised the inspector that an

engineering department review of construction documentation was

unable to establish that all required containment MOV conduit seals

4 were in place. The licensee performed the review-in response to

! conduit seal deficiencies identified at the Wolf Crc. Plant. The

licensee determined that although work authorizing documents had

'

,

been issued, there was no sign offs for work accomplishment.

!

On May 23, 1986, while the plant was in Mode 5 (Cold Shutdown), the

'

licensee issued Work Request Nos. 60511 through 60519 to install

containment MOV conduit seals in accordance with Bechtel Drawing

, M-2Y007 (Conduit Seals for Containment MOVs). This action was taken

to assure existing EQ status of the MOVs prior to pending plant

startup. The licensee stated that external visual inspection could

not readily verify conduit realing in accordance with the design

i

'

drawing, as drawing M-2Y007 requires or does not' require the use of

sealant depending on the actual field routing of the conduit. The  ;

,

'

licensee's evaluation of this matter is continuing and plans to

perform a field inspection during the next shutdown.

, The licensee reviewed Bechtel Drawing M-2Y007, SNUPPS Report of

Independent Review of EQ Programs, and WR Nos. 60511 through 60519.

The inspector also interviewed licensee maintenance personnel that

-

performed the conduit sealing work. Based on this review,.the

inspector determined that the containment H0Vs conduit seals have

been installed in accordance with Bechtel Drawirg M-2Y007. However,

12'

,

-- - --- -, e -, ,, +--r -.n , e-,o ~ ..-~,r. -,- n ,r -,--,-m.-es,-,e-,--+--v cn-r , ~ - , , , - - -

. .

the acceptability of conduit seals prior to May 23, 1986, could not

be determined. This matter is unresolved pending further NRC

review. Unresolved Item No. 483/86010-02(DRP)

b. Temporary Instruction (TI) 2515/77, " Licensee Response to Selected

Safety Issues (Biofouling of Cooling Water Heat Exchangers)

An inspection was performed to assess licensee's programs for

detection and prevention of biofouling of cooling water heat

exchangers. The inspection included a review of applicable

procedures and interviews with licensee's maintenance, chemistry,

engineering, and operations personnel.

Prior to the initial startup, the licensee detected tube damage due

to biofouling in the main generator hydrogen coolers which was

attributed to stagnant water conditions prior to plant startup. No

additional biofouling has been experienced. The licensee has

implemented procedures; ETP-ZZ-03002, " Performance Testing of Plant

Heat Exchangers" and ETP-ZZ-03003, " Monitoring of Plant Heat

Exchangers". The licensee also maintains 1 PPM chlorine in the

service water system as a preventive measure. The licensee provides

procedures and operator training relating to degraded heat exchanger

performance.

No violations or deviations were identified.

5. TMI NUREG-0737 Items Closure (92705)

The following TMI NUREG-0737 line items are considered to be closed:

II.B.3.3

II.B.3.4

II.F.1

II.F.2A

II.F.2B

II.F.2C

A review of Inspection Report Nos. 84-10(DRMSP), 84-16(DRMSP),

86004(DRSS), and others was made and discussions were held with the

applicable inspectors to verify that these line items were ccmpleted.

The ittm identification is included here as a correlation for NUREG-0737

tracking system as the previous closeouts were for the Safety Evaluation l

Report tracking system or other numbering systems. j

No violations or deviations were identified.

6. Inspection of Licensee Events-Inoperable Intermediate Head Safety

4

Injection (IHSI) System (92700)

a. Background l

l

l

13 I

{ -

- . . . _ - - ,. .

. .

On April 13, 1986, the licensee advised the senior resident

inspector that on April 12, 1986, while in Mode 3 (Hot Standby), the

plant was placed in a condition prohibited by Technical

Specification (T/S) 3.5.2, when the IHSI system was inadvertently

isolated. The inspector was given a copy of Incident Report No.

86.109 which documented the violation. The inspector was also

briefed on the event, the cause and immediate action taken, and of

licensee's planned investigation in this matter.

An inspection was performed to assess the event, root

cause/ contributing factors, and licensee corrective measures. The

inspection included a review of event reports, operating logs,

administrative and surveillance procedures, personnel interviews,

and meetings with licensee management.

b. Inspection Findings

On April 12, 1986, at 0402 CST, the safety injection (SI) cold leg

isolation valve EM-HV-8835 was closed to perform surveillance test

OSP-EP-V0003 (Section XI Accumulator Safety Injection Valve

Operability). Technical Specification (T/S) Limiting Condition for

Operation 3.5.2 specified that the IHSI system be operable in Mode

3. T/S Surveillance Requirements 4.5.2 specifies that EM-HV-8835

(Safety Injection Cold Leg Isolation Valve) be open. EM-HV-8835,

being closed, isolated the common discharge path from both SI pumps

to the cold leg injection, putting the plant in a condition

prohibited by Technical Specifications.

On April 12, 1986, at 1010 CST, the reactor operator, while taking

the required daily leg readings, observed that EM-HV-8835 was

closed. The reactor operator immediately informed the shift

supervisor (S/S). The S/S declared both SI Trains inoperable and

entered T/S 3.0.3 and had valve EM-HV-8835 opened. The S/S issued

Incident Report No.86-109 documenting the violation.

On May 7, 1986, the inspector met with the licensee to assess

licensee's investigation, evaluaticn of cause, and corrective

actions regarding the IHSI system isolation. The licensee discussed

their findings of root cause and contributing factors and of

corrective action taken and planned. The cause of the event was

attributed to personnel scheduling and performance errors as

follows:

Scheduling

Scheduling Personnel - Scheduled the performance of OSP-EP-V0003,

"as required in Mode 3 prior to RCS pressure reaching 1000 psig".

Compliance Personnel - Identified OSP-EP-V0003 on an attachment to

the Mode 3 Change Letter, "to be performed in Mode 3 as conditions

permit".

I

14

l

. - _ _ _

_ _ _

. .

The task performance review considered the operability requirement

of the safety injection accumulators but failed to recognize the

Surveillance Task Sheet (STS) task performance mode requirements,

which specified Mode 4 only.

c. Performance

Operations personnel erroneously authorized and performed

OSP-EP-V0003 in Mode 3. Several errors were made regarding the

authorization and performance as follows:

(1) OSP-EP-V0003 - Initial condition specified performance of the

test in Mode 4. Operations personnel issued a Temporary Change

Notice (TCN) changing Mode 4 to Mode 3. APA-ZZ-00101

(Preparation, Review, Approval and Control of Plant Procedures)

provides for temporary procedure changes, "which clearly does

not change the intent". Management Directive U0 86-69 issued

March 4,1986, reemphasized (s) control of TCNs and identified

that "significant changes to initial conditions are changes

which are changing the intent".

(2) Operations personnel changed the Surveillance Task Sheet No.

ST-00070 Task Performance Mode from " Mode 4 only" to Mode 3.

This change did not receive the required review and approval as

specified in APA-ZZ-00340 (Surveillance Program

Administration).

(3) Operations personnel's failure to be cognizant of the overall

plant effect of closing EM-HV-8835. (The isolation of a

required safety system).

(4) In addition to issuing an Incident Report, the licensee classified

the event as a 30 day Licensee Event Report (10 CFR 50.73).

However, a four hour report to NRC Operations Center should

also have been made in accordance with 10 CFR 50.72(b)(2)(iii).

This report was not made. Failure to notify the NRC within four.

hours is a violation of 10 50.72(b)(2)(iii). No. 483/86010-03(DRP).

d. Licensee's Corrective Action to Pravent Recurrence Included:

(1) For future outages, outage scheduling will schedule

OSP-EP-V0003 in Mode 4 as a Mode 3 restraint.

(2) Progressive discipline has been initiated for appropriate

outage personnel. Outage Planning and Scheduling personnel ,

have been advised concerning outages involvement in this event. l

(3) An outage procedure currently in draft form will specifically

address use of the STS " Task Performance Mode" for scheduling

surveillances.

(4) Future mode change letters will reflect only required task

performance conditions and T/S requirements for mode changes.

(5) The TCN that allowed performance of the OSP in Mode 3 was voided.

15

_ _ _ _ __ ._

-- - - -

. .

!

(6) Progressive discipline has been initiated for operations

personnel involved in this event and the necessity to comply

with programmatic controls has been reemphasized

(7) Management will reemphasize the existing administrative controls

for revisions to task sheets and surveillance procedures to

appropriate plant personnel.

(8) Appropriate personnel will receive guidance concerning reporting

requirements of 10 CFR 50.72.

The inspector determined that the violation, once identified by the

licensee, was promptly corrected, documented, and received a high level

of attention. Based on the short duration of the violation, plant

conditions of low temperature and pressure with low' stored heat energy

and the availability of backup emergency core cooling systems, the event

posed no significant threat to the public or plant safety. However, the

event highlighted significant performance errors. These errors included;

inadequate task reviews, failure to adhere to licensee administrative

procedures, and failure to be cognizant of the overall plant effect

resulting from surveillance testing.

The licensee's failure to maintain the IHSI system " operable" while in

Mode 1, 2, and 3 is a violation of Callaway Plant Technical Specifications

Limiting Condition for Operating 3.5.2. No. 483/86010-04(DRP).

7. Monthly Surveillance (61726)

The inspectors reviewed or observed selected portions of Technical

Specification required surveillance testing during power operations and

prior to mode changes relative to the startup from the refueling outage.

Items which were considered during the inspections included whether

adequate procedures were used to perform the testing, test instrumentation

was calibrated, test results conformed with Technical Specifications and

procedural requirements, and the test was performed within the required

time limits. The inspector determined that the test results ware reviewed

by someone other than the personnel involved with the performance of the

test, and that any deficiencies identified during the testing were reviewed

and resolved by appropriate management personnel.

No violations or deviations were identified.

8. Operational Safety Verification (71707) ,

I

The inspectors observed control room operations, reviewed applicable

logs, and conducted discussions with control room operators throughout

the inspection period. The inspector verified the operability of selected

safety related systems, reviewed tagout records, and verified proper

return to service of affected components. Tours of the reactor, auxiliary,

and turbine buildings were conducted. During these tours, observations

were made relative to plant equipment conditions, fire hazards, fire

protection, adherence to procedures, radiological control and conditions,

16

__ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _

. s

housekeeping, security, tagging of equipment, ongoing maintenance and

surveillance, containment integrity, and availability of safety related

equipment.

No violations or deviations were identified.

9. Monthly Maintenance (62703)

Selected portions of the plant maintenance activities on safety related

systems and components were observed or reviewed to ascertain that the

activities were performed in accordance with approved procedures,

regulatory guides, industry codes and standards, and that the performance

of the activities conformed to the Technical Specifications.

The following items were considered during these inspections: the limiting

conditions for operation were met while components or systems were removed

from service; approvals were obtained prior to initiating the work;

activities were accomplished using approved procedures and were. inspected

as applicable; functional testing and/or calibrating were performed prior

to returning the components or systems to service; parts and materials

that were used were properly certified; radiological controls were

implemented as necessary; and, fire prevention controls were implemented. 1

No violations or deviations were identified.

10. Cycle 2 Startup (61702, 61705 through 61710)

The initial criticality of the Cycle 2 core was observed by the inspector

on April 15, 1986. The startup was performed per ETP-ZZ-ST002, " Engineering

Test for Initial Criticality". Selected portions of the following tests /

procedures were observed during-their performance and the results were

reviewed after the evaluation of data was completed.

ETP-ZZ-00007 Reactimeter Dynamic Checkout

ETP-ZZ-ST004 All Rod Out Baron Endpoint

ETP-SR-ST001 All Rods Out Flux Map

i

ESP-ZZ-00009 Moderator Temperature Coefficient Measurement j

ESP-BB-03015 Reactor Coolant Flow Measurements

ESP-ZZ-00006 Incore/Excore Calibration

ETP-ZZ-ST005 Rod Bank Worth Measurement

The moderator temperature coefficient was slightly positive for all rods

4

out. The rod withdrawl restriction will continue for 4000 MWD /MTV burnup

,

of the core. The other tests indicated the results were about where they

were expected. The mode changes were observed and requirement check

sheets were reviewed. Selected requirements for mode changes were

verified to have been performed.

No violations or deviations were identified.

17

l

__

._

e ..

11. Unresolved Items

Unresolved items are matters about which more information is required in

order to ascertain.wheter they are acceptable items, violations, or

deviations. Two unresolved items disclosed during the inspection are

discussed in Paragraph 4.

12. Exit Interview

The inspector met with licensee representatives (denoted under Persons

Contacted) at intervals during the inspection period. The inspector

summarized the scope and findings of the inspection. The licensee

representatives acknowledged the findings as reported herein. The

inspector also discussed the likely informational content of thi

inspection report with regard to documents or processes reviewed by the

inspector during the inspection. The licensee did not identify any such

documents / processes as proprietary.

13. Enforcement Conference

An Enforcement Conference was held on June 3, 1986, at the NRC Region III

office, Glen Ellyn, Illinois between Mr. D. F. Schnell and members of the

NRC Region III staff. During the meeting the Licensee presented facts

relative to the event on April 12, 1986, discussed in Paragraph 6 above.

The Licensee presented background information, corrective action to

prevent recurrence, and potential mitigating facts which the NRC will use

to determine the appropriate escalated enforcement action.

l

t

i

i

l

1

18 1

i I

i l

. - - -. ,.

_

.--