Text

Insp Rept 50-219/90-09 on 900422-0609.Violations Noted. Major Areas Inspected:Observation & Review of Routine Plant Activities & Operational Events,Site Radiological Controls & Events & Observations of Corrective Maint Activities
	ML20056A619
Person / Time
Site:	Oyster Creek
Issue date:	07/19/1990
From:	Ruland W; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20056A608	List: ML20056A607; ML20056A612; ML20056A619;
References
	50-219-90-09, 50-219-90-9, NUDOCS 9008080338
	Download: ML20056A619 (28)
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V. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No.

50-219/90-09

Docket No.

50-219

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License No.

DPR-16

Priority --

Category C

. Licensee:

GPU Nuclear Corporation

1 Upper Pond Road

Parsippany, New ,lersey 07054

. Facility Name: Oy_ ster Creek Nuclear Generating Station

Inspection Conducted: April 22, 1990, - June 9, 1990

Inspectors:

M. Banerjee, Resident Inspector, Oyster Creek

E. Collins, Sr. Resident Inspector, Oyster Creek

D. Lew, Resident Inspector, Oystee' Creek

M. Markley, Resident inspector, Yankee Rowe

Approved By:

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W. Ruland, Chief

Date '

Reactor Projects Section 4B

Inspection Summary:

Inspection Report No. 50-219/90-09 for period April 22, 1990 - June 9, 1990

Areas Inspected:

The inspection consisted of 307 hours0.00355 days 0.0853 hours 5.076058e-4 weeks 1.168135e-4 months of direct inspection.

The areas inspected included observation and review of routine plant activities

and operational events (paragraph 1.0); review of site radiological controls

and radiological events (paragraph 2.0); observations of corrective maintenance

. activities and routine surveillance tests (paragraph 3,0); review of your

corrective actions in response to Core Spray relief valve problems and

Emergency Diesel Generator battery problems (paragraph 4.0); review of

additional 4160 volt cable testing, station policy and procedures regarding

operating parameter verification, and identification of deficient molded case

circuit breakers (paragraph 6.0).

Results: Two examples of failure to evaluate radiological work conditions were

identified and follos-up of an allegation has identified that an individual

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entered a high radiation area without proper dosimetry. These items are being

issued as violations. One unresolved item is being opened pending NRC review

of the permanent corrective actions associated with exceeding core thermal

power by a small amount,

JDR00808o338 900720

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TABLE OF CONTENTS

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-Executive Summary.

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11. Details. . .;. . . .

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Operations . . . .-. .-.. . . . . . . . . . . . ... . ... . . . .

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.1.1

Chronology of Operational Events (71707, 93702) . . . . . .-

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'1.2

Diesel Generator Circuit Breaker-(93702). . . . . _ . . . . .

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1.3 Loss of Unit Substation 1B2 (93702, 71707). . ... . . . . .

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1.4. Startup May_ 3,-1990 (71707).. . . . . . . . . . . . . . . .

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1.5 Licensed Thermal Power Limit Exceeded as a Result of

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' Changes made to Computer input to Heat Balance

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. Calculat ion ( 93702) . . . . . . . . . . . . . . . . . . . . .

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- 1,6 Lightning Stri ke (93702) . . . . . . . . . . . . . . . . .

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Control; Room Toues (71707) . . . . . . . . . . . . . . . . .-

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1. 8 - Fa c i l i ty Tou rs ( /1707 )

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2.0 Radiological Controls; . . . . . . . . .

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Contaminated Hose Released from the Radiolo

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2 .' 1 Control Area (93702). ... . . ... . . , . .gically

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2.2 Outage Planning.(71707) . . . . . . ... . .

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2.3 -Radiological Performance Committee (71707). . . . . . . . . -10

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2.4. _ Radiological Controis for Outage Related Work (71707) , , .

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2.5 Under the Vessel Work Activities (71707). . . . . . . ' . . .

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2.6 Operational. Transition to Shutdown Cooling (71707). . . . .

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2. 7. RadiologicaliWork Permit (71707). . . . . ... . . . . . . .

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3.0 Maintenance / Surveillance

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3.1 Monthly Maintenance-Observation (62703) . . . . . . . . . .

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3.1.1.

Replacement-of 4160 Volt Cable. .

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~3.1.2

Fire Suppression Water System Valve Replacement .

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3.2 Monthly Surveillance Observation (61726). . . . . . . . . .

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3.2.1

Battery Surveillance Testing. . . . . . . . . . .

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3.2.2

Instrument Calibration. . . . . . . . . . . . . .

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4.0( Engineering and Technical Support (93702). . .

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4.1' Core Spray. System Relief Valves'. . . . . . . . . . . . . .

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4.2 Emer0ency Diesel Generator Battery

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6.0 LSafety; Assessment / Quality Verification (71707) . . . ...; . . .;.

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6.1' Licensee Evaluation of.USS-182 Cable. Failure and< Tests.

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done on Other Cables... . . . . .

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6;2_ Control Room Operating Parameter Verification . . . . . .;.

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4.: 6.3R Nonconforming TED, Breakers

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"8.0;:f Exit Meetings 'and Unresolved Items (30703) . . . .: . .: . . . . .

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8.1 Priliminary Inspection Findings .--_

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8.2 S Attendance-at-Management Meetings Conducted by

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Region: Based Inspectors .

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8.3 ? Unresolved Items._. . . . . . . . . . . . . ... ... . . . .

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ATTACHMENTS

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

EXECUTIVE SUMMARY

Report No.

50-219/90-09

Operations

Overall, the plant was operated in a safe manner. Operator actions during the

Unit Sub-Station (USS) 1B2 cable failure kept the plant in a safe condition.

Incomplete understanding of the impact of a 100 amp discharge rate on the "B"

battery led to the need to expedite the plant shutdown and to cross connect USS

1A2 and IB2 prior to the plant reaching cold shutdown.

However, the "B"

battery was not completely discharged during the event.

The manual scram from

low power initiated only a minor transient, as the main turbine had been

removed from service and all site loads had been transferred to the startup

transformer.

Radiological Controls

Maintenance during an unplanned outage showed active participation by radiation

protection personnel. Also, no discrepancies were noted in posting and

labeling.

Radiological controls during unplanned outage 12-U-K were generally

good.

However, two exceptions to the above were noted.

In one event, technicians

worked in an area under the reactor vessel without a current survey of the work

location.

In another event, work was unknowingly performed in close proximity

to a hot spot.

In both events there was inadequate evaluation of radiological

hazards.

This report also documents an event that occurred in October 1989 where a

worker entered a high radiation area without the required dose rate meter.

After completion of NRC review and licensee investigation, this event is being

cited as a violation. Also, the radiological controls technician did not

report or document the occurrence. Once the licensee was notified of the event

by NRC, plant actions were appropriate.

Surveillance and Maintenance

While observing a cable pul.1, the inspector noticed that a pen and ink change

was made to the job package to increase the maximum allowable pull tension and

delete'certain megger testing.

The job planner treated the change as not sub-

stantive thus not requiring additional review, With QC and NRC involvement,

the licensee reclassified the change as substantive, resolving the issue.

During a surveillance, an electrician read battery voltage using a local meter

vice a voltmeter at the battery terminal.

The electrician was using a data

sheet instead of the procedure.

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No other deficiencies were noted.

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3_ Engineering and Technical Support

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A core spray booster-pump relief valve lifted during pump operability testing,

contrary to design. The failure may be a repeat problem.

During the past

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. refueling outage two newly installed relief valves, of the same design, lifted

F

lduring pump testing. The manufacturer found that the-internal valve spring

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.had=not been stress relieved. The newly failed valve-is being'sent to Wiley

Labs for testing.-

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An Emergency Diesel Generator (DG) failed to start during routine biweekly

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-testing. L0ne cell of the DG battery was degraded.

The 8 year rated battery

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was 5 years old.

The test used a low voltage mode -- three resistor banks in

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the start circuit - The licensee demonstrated that, in an emergency start, thel

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diesel would still have started.

The resistors would not be in the start

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sequence.

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Physical Security

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Routine inspections found no' problems.

Safety Assessment / Quality Verification

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After cable. failure, the licensee examined the cable and sent it to a testing

lab to identify the cause of- the _f ailure.

The tests are ongoing. Onsite test-

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ing of three similar installed calbes revealed no problems. The licensee is

'

preparing a test program for similar cables. Testing will start the next re-

- fueling _ outage.

The inspector concluded that the licensee's initiatives were

appropriate.

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The inspectors reviewed the licensee's policy.regarding operator _ response to

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indications. The site policy is clear and does not inappropriately emphasize

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verification measures to control room operators.

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A GE TED 100 molded case circuit breaker overcurrent trip function did not work-

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dur,ing pre-installation testing. A UV device, installed by the manufacturer.

-after testing, interfered with the overcurrent trip,

No breakers under the

same purchase order had been installed in safety related applications. _The_

. licensee's actions-were appropriate.

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DETAILS

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1.0 Plans Operational Review

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1.' 1 Chronology of Operational Events

Inspectors reviewed the details associated with key operational events

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that occurred during the report period. A summary of these inspection

activitiesLfollows.

~4/22/90 The~ inspection period began with the reactor shut down

--

following the failure of unit substation USS 182.

Cold Shutdown was

reached at approximately 10:00 a.m., and the unplanned outage 12-U-K

started. A detailed description of the event is contained in

Paragraph 1.3.

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During the outage, the licensee found that the substation failure was

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caused by a faulted "C" phase cable between the 4160V bus ID and the

4160V/460V transformer supplying power to the substation.- The cable

was replaced.

For a summary of cable replacement see Paragraph 3.1.

-

For a summary of the licensee's evaluation of cable failure and

additional testing done on other medium voltage cables, refer to

Paragraph 6.1.

During .th' e outage, the licensee measured drywell thickness using

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ultrasonic techniques including new measurements at the 86 ft.

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elevation.- fo change in the safety evaluation conclusions for

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drywell thinning was required because of these additional

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inspections. The worst case estimate of corrosion rate shows

adequate drywell thickness until August 1991,

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5/2/90 During routine surveillance testing, the Core Spray System II

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relief valve at the discharge of the-booster pump lifted prematurely.

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"The failed valve was replaced and sent to Wiley Laboratory for

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testing, adjustment or refurbishment.

5/3/90 Plant startup commenced at 8: 41 p.m. and the reactor was

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critical at 9:44 p.m.

After the Source Range Monitors (SRM) were

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' fully withdrawn, SRM No. 22 failed downscale and was declared

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inoperable.

5/4/90 During the drywell inspection at rated reactor pressure, the

--

licensee found and repaired a packing leak on the "B" isolation con-

denser condensate return line isolation valve, V-14-37.

The main

steam isolation valve, NS03A, body-to-bonnet temporary repair was

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also refurbished.

(See report 90-06 for details)

5/5/90 At 1:22 a.m. the reactor mode switch was placed in the run

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position and the generator placed on line at 3:00 a.m.

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5/8/90 -At 12:00, Standby Gas Treatment System I was declared in-

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operable Mue to a failed photocell and a seven-day Technical

Specification action statement was entered._ While testing the

system; the reactor building ventilation. system valve V-28-3 was

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found to have a failed valve. operator. The valve was secured in the

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closed position per Technical Specifications for secondary contain-

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ment integrity.

System I was returned to service the same day at

11:35 p.m. , af ter completion. of the repair.

Repairs on V-28-3 were

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. completed and the valve returned to service on-5/9/90 at 5:39 p.m.

The inspectors verified Technical Specification compliance,

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5/13/90 At 8:36 p.m. lightning struck the plant stack.

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Paragraph 1.6 for a discussion of the effects.

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5/19/90 At 11:12 a.m. the #2 Emergency Diesel Generator was declared

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inoperable and a seven-day Technical Specification action statement

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was entered due to an overheated phase in the control power trans-

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former output breaker.

The breaker was replaced, the diesel genera-

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tor was. returned to service at 9:00 a.m. on 5/20/90 (See paragraph

1.2).

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1.2 Diesel Generator Circu,i_t, Breaker

On May 19, 1990, an equipment operator smelled an unusual odor in the.#2

Emergency Diesel Generator (EDG) switchgear cubicle.

Upon investigation,

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the control- circuit breaker for diesel auxiliaries was found to have an

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overheated "C" phase terminal.

The licensee declared the diesel

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inoperable, documented the deficiency in a deviation report, and replaced

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the breaker.

Licensee inspection of the #1 EDG auxiliary breaker showed

it did not have a similar problem.

The breaker was a GE, model TFJ, three phase, molded case circuit breaker

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rated at 480V and 125 amps.

Overheating was evident at one of the three

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' terminals by discoloration and the presence of melted epoxy.

The terminal

screw had damaged threads which could have prevented a tight connection.

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Broken pieces of wire were also present, The. licensee concluded this

loose connection was the cause of terminal overheating. This conclusion.

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was supported by testing which showed high resistance at this terminal.

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During refueling outages, this breaker is disconnected and a temporary

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power supply is connected to the diesel auxiliaries.

This adds wear and

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tear to the breaker terminals and introduces the possibility of inadequate

connection.

The licensee is evaluating a modification that will install a

permanent means to provide this alternate power without affecting the

auxiliary breaker,

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The licensee's current preventive maintenance program does not include

these breakers.

In general, le.ig term programs are in progress to

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establish preventive and predictive maintenance activities for all plant

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systems.

For this configura', ion, the licensee is evaluating the need for

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periodic breaker inspection requirements.

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1.3 Loss of Unit Substation 182

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Sequence of Events

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At 9:54 a.m. on April 21, 1990, the plant experienced a loss of power to

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the unit substation USS-182 when its supply breakers tripped.

055-182 is

one of the two safety related 460 V unit substations, and is powered

,

through a transformer from the 4160 V emergency. switch ' gear 10. Following -

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loss of USS-1B2, the isolation valves motor control center MCC-1AB2, vital

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AC power panel VACP-1, and instrument panel (IP-4) automatically

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transferred to USS-1A2 as designed. The "B" CRD pump, core spray keep

fill pumps,1-2 Reac+or Building Closed Loop Cooling Water (RBCCW) pump, b

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fuel pool pump, 1-1 air compressor, the battery chargers to both A and B

batteries lost power. The reactor water cleanup system isolated due to

loss of RBCCW.- Also isolated were the containment high range radiation

monitor channel 2 and, the reactor building and turbine building vent

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systems. The rotary inverter which is the normal power supply for

continuous instrument panel CIP-3 lost its AC power and was being supplied

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from B station battery.

Power supplies were lost to various other safety'

related Equipr..ent, including protection system panel 2,1-2 standby liquid

".

control pump and associated Squibb valve, "B" and "C" core spray booster

pumps, Containment Spray System 2 pumps and valves, AC powered valves in

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"B" isolation condenser steam and condensate line, "B" hydrogen / oxygen-

monitor, and one of the two Standby Gas Treatment System (SGTS) fans.

Also lost were B and C shutdown cooling pumps, .drywell equipment drain

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tank (DWEDT) and reactor building equipment drain tank (RBEDT) pumps and

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monitoring capability, and various lighting and security equipment.

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. The operators placed the "A" CRD pump, SGTS 1, 1-1 RBCCW pump, the 1-3 air

compressor in service, and recovered partial turbine building ventilation

system. The reactor protection system panel 2 was transferred to power

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supplied from USS-1A2 and reenergized.

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The plant was in several Technical Specification action statements, the-

most limiting being the'30-hour action statement due to loss of USS-182.

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The licensee commenced a reactor shutdown at a rate of 30 MWE per hour a.

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10:50 a.m.

At 11:00 a.m., increa,ed operation of the drywell sump pump was

observed.

Due to the-loss of DWEDT pumping capability the tank was

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overflowing to the drywell sump. This increased the drywell unidentified

l'eakage rate which is determined from the drywell sump pump integrator

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readings.

The licensee started taking hourly readings on the drywell sump

pump integrators.

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. Since both pond pumps were unavailable, the licensee Sanually started the

fire diesels. Due to the degrading batter, affecting fire panels, the

licensee installed roving fire watches.

Compensatory measures were taken

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due to ti.e loss of various security equipment,

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Due to loss of power to the "D" isolation condenser steam and condensate

-return line AC operated valves, the operators isolated the "B" isolation

condenser as required by plant Technical Specifications.

At 3:27 p.m., the licensee declared an unusual event based on DW

unidentified leak rate exceeding 5 gpm. The GSS assumed the duties of

Emergency Director.

The safety related 125V station battery "B" discharged at 100 amps.

The

largest load on the "B" battery is the rotary inverter which was supplying

power to CIP-3.

Plant des.gn allows switching CIP-3 power supply to

USS-1A2; however, the licensee decided not to switch while the plant was

critical because the. transfer would cause momentary interruptions of power

to its loads.

CIP-3 supplies power to reactor level instrumentation,

pressure, steam flow and feed flow indications, control rod position

. indication, SLC indicating lights, main turbine electric pressure

regulators, recirculation pump MG set and feedwater regulating valve

control and various radiation monitors. Momentary loss of power has

previously resulted in partial isolation of the reactor building.

Engineering evaluation of the "B" station battery showed it could not

sustain extended operation at a 100 amp discharge rate.

The nominal

capacity of the battery is 1200 amp-h n rs at a 150 amp discharge rate.

The decision was made to expedite plant shutdown by manually scramming

from about 20*i; power.

Station loads were transferred offsite and the

turbine removed from service. The scram was inserted at 8:14 p.m.

After shutdown, the power supply of CIP-3 was transferred to USS-1A2. The

10 # on "B". battery dropped to about 40 amps.

During-reactor cooldown, the operators are required to cycle the normally

clo~ sed isolation condenser condensate return valves every 100 degrees F.

At 362 degrees, the operators successfully cycled valve V-14-34 in the "A"'

system. However, with reactor temperature around 275 degrees the valve

-failed to open completely.

1he valve was declared inoperable, and was

manually. closed. This made isolation condenser "A" also inoperable.

At' midnight, the "A" shutdown cooling pump was placed in service.

Removal.of the rotary inverter from battery "B" reduced the battery load,

but battery voltage continued to slowly drop.

To protect the battery from

unnecessary depletion, the licensee decided to cross tie the 182 bus to

the available 1A2 bus to reenergize the battery chargers. This is accom-

plished using an installed cross-tie that is normally not used.

Plant procedure requires the reactor to be-in cold shutdown before

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this evolution could be performed.

Since the reactor was above cold

shutdown, a temporary procedure change was made to allow the cross tie. At

8:10 a.m. on April.22, the licensee energized USS-182 from USS-1A2.

As a

precaution,.an operator was stationed at the control room panel to monitor

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the electrical distribution system for adverse signs and indications.

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was instructed to trip the cross tie upon any adverse indications or

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automatic start of any safety systems powered from 1A2 substation. This

would prevent. overloading the safety related USS-1A2 in case of an

[.

accident,

At 8:23 a.m., the "A" and "B" battery chargers were placed in service.

,

Within.two hours, the reactor coolant temperature went below 212

"'

degrees'F.

A temporary change was made to the alarm response procedure for the

,

overload alarm associated with USS-1A2.

The alarm setpoint was increased

from 248 to 270 amps based on the transformer rating.

Engineering

evaluation indicated that the transformer temperature was the critical

. parameter and should be maintained less than 75 degrees C.

Upon an alarm

condition, this temperature should be monitored locally and loads reduced

if temperature reaches 75 degrees C.

At 2:20 p.m. on April 22,.the unusual event was terminated. The licensee-

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entered unplanned: outage 12-U-K to replace the failed cable and restore

normal. power. supply to USS-182.

Evaluation

The inspectors reviewed the licensee's response to the event.

The control

room operators acted appropriately to restore various systems by starting

the alternate available equipment or by transferring equipment to

alternate available power supply.

Plant security responded adequately by

' implementing compensatory measures. .Due to degraded fire protection

equipment, the licensee established roving'f. ire watches until the plant

could be placed-in cold shutdown.

Inspectors' did not identify any

'

unacceptable conditions.

.

The licensee's-measures were adequate i_n monitoring the. battery capacity

u

and taking m.easures to reduce the load; however, a better understanding

of-battery capacity versus-discharge rate.may have permitted a-more

orderly shutdown. The site had no data on how battery capacity

(ampere-hours). varied with discharge rate.

Inspectors reviewed the safety

evaluation associated with the procedure change and did not identify any

unacceptable conditions.

The licensee's measures 'to monitor 1A2 loading

was also.found adequate.

The isolation condenser condensatt return line iso %on valves have a

i

history of thermai binding.

The inspectors reviewed the battery voltage

during the time this valve was cycled.

The oattery voltage dropped from

~,

approximately 115 volts to 112.5 and then. recovered to 114.5 volts within

15 seconds. The licensee determined that thermal binding caused the

operator trip.

Subsequent motor operator valve testing did not identify

any' deficient conditions.

This valve is scheduled for replacement next

~

outage.

The inspectors continue to review this valve failure.

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The licensee is currently performing a study of the DC motor operated

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.B

-valves to determine the adequacy o.f the teg;e genented under degraded

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vol tage L eondi_tions .

~'

Inspectors reviewed the manufacturer's data for the "B" battery, the

results of the most recent battery discharge. test, and "B" battery

response during.the event. . Inspectors concluded the battt

was not fully

discharged because terminal voltage did not reach 105 volts.-

-Although one pump in each core spray system was lost, full core spray

capability we ~tained because of redundant pumps.

Both emergency diesel

generators v.m , .erable throughout the event.- Manual activation of the

!

isolation conoe

T was available, although not required.

Containment

l

Spray System #1 h operable.

Overall, the inspector concluded that the licensee's' response to this

event was acceptable.

I

l

1.4 Startup May 3, 1990

Reactor startup commenced on May 3, 1990, at 8:41 p m.

Portions of the

'

startup were observed by the resident inspector.

The inspector noted that

j

management was present in the control room throughout the startup. The

1

startup was performed without operator errors or significant equipment

problems.

No unacceptable. conditions were identified.

,j

i

l '. 5 Licensed Thermal Power Limit Exceeded as a Result of Changes Made to

i

Computer Input to Heat Balance Calculation

On May 9,- 1990, at 10:35 p.m. , the Reactor Water Cleanup System was removed

!

from service to perform a leak test on the nonregenerative heat

i

.

exchangers.

The. net effect of the cleanup system input to the computer

if

heat balance calculation is a loss of. core thermal power by 7.84 MW (for

4

one cleanup pump operating).

In order to' reflect the actual plant.

I

condition in the heat balance calculation, the Shift Technical' Advisor

.(STA), af ter discussions with the Group Shif t Supervisor (GSS), removed the

cleanup system input'from the computer heat balance calculation.

This

change was entered 'in; the STA log but not in the control room operator or

a

GSS. logs.

i

The cleanup' system was returned to service on May 11 at 2:00 p.m.; however

the input to the heat balance calculation was not restored. While

reviewing the log the night shift STA realized the error ara informed the

1

GSS.

The input to the computer heat balance calculation was restored.

'

Reactor coolant and offgas. samples were analyzed and found to be

~

acceptable.

A deviation report was. submitted.

$

The licensee initiated a critique of the event.

It has been determined

o

that the reactor power exceeded the licensed limit of 1930 MWt for

!

approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months with the maximum core thermai power being less than

'1*4 above the licensed limit.

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The-immediate corrective actions were adequate.

This item will remain

unresolved pending licensee's completion of the critique, including

identification of the root cause.and permanent corrective actions.

. (50-219/90-09-01)

1.6' Lightning Strike

Oue to a lightning strike on May 13, some equipment and components last

power and sustained damage.

This included the New Radwaste (NRW) and

Augmented Offgas (A0G) Systems, heating boilers, stack lights, the rod

worth minimizers (RWM), the Sequence of Alarm Recorder (SAR), the Stack

Radioactive Gas Ef fluent Monitoring (RAGEM) system, some security equip-

ment, a fire protection panel in the pump house,_and alarm units in the

main and NRW control rooms.

The licensee restoredtpower to NRW and A0G systems within one-half hour.

Compensatory-measures were taken within 10 minutes for loss of security

equipment. The licensee enhanced monitoring of system parameters for lost

alarm functions. An hourly fire watch compensated for the lost fire

detection capwility in the . fire pump house.

The operators installed a

rod block when they determined that the capability to update control rod

position.for core thermal limit calculation in the computer was lost. The

RWM provides control rod position information for the plant computer

system, The operators also verified that the rod position information in

the computer was correct.

The operators' response to the event was appropriate.

The inspector asked.

if adequate written guidance existed for the operators to implement a rod

block.upon loss of RWM. The licensee determined that a procec' .re change

wc.uld be implemented to provide such guidance and is evaluat4.ig_ necessary

changes in=.the plant computer to provide the operators with a warning

signal when rod position indication is not being properly transmitted.

The inspector found~the licensee's proposed followup adequate and nad no

further questions.

l' . 7 Control Room Tours

.The inspectors conducted routine tours of the control room during

which time the following documents were reviewed:

Control Room and Group Shif t Supervisor's Logs;

--

Technical Specification Log;

--

Control Room and Shif t Supervisor's Turnover Check Lists;

--

Reactor Building and Turbine Building Tour Sheets;

'

--

Equipment Control Logs;

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Standing: Orders; and,-

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'No unacceptable conditions were-identified ~

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1.8: Facility Tours'

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JThe inspectors' conducted routine plant tours to' assess equipment condi-

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Etions', personnel! safety hazards, procedural adherence and compliance with

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regulatory requirements. The following; areas were inspectedi:

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Tu'rbine B'ilding

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Dieset Generator Building-

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Reactor. Building

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.NewlRadwaste? Building

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0ld;Radwaste. Building _.

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Thelfo11owing: additional items were observed or verified:

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Fi re ' Protection:-

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-- ' Randomly-selected fire extinguishers were accessible and-

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inspected-on schedule.

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Ignition' sources and_ combustible materials were controlled-in

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iaccordance' with the' licensee's approved procedures.

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. Appropriate fire watches ~or fire. patrols were stationed when.

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fire protection / detection equipment was out offservice.-

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Equipment Control:

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Jumper and equipment mark-ups did not conflict with-technical

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' specification requirements.-

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Conditions requiring the use of jumpers received the prompt

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attention of'the licensee.

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Vital Instrumentation:

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Selected instruments appeared functional and demonstrated

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parameters within Technical Specification Limiting Conditions

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for Operation,

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Housekeeping:

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Plant housekeeping and cleanliness were in accordance with

--

licensee programs.

!

. Minor housekeeping deficiencies which were identified were promptly

corrected by the licensee. No other unacceptable conditions were

1

iden ti fied. .

2.0 Radiological Controls

T

'2.1

Contaminated Hose Released from the Radiologically Controlled Area

(RCA)

On April 17-19, 1990, the licensee established a temporary modification

'(No. 90-22) to allow the breathing air system (Joy compressor) to provide

7.

a temporary air supply to the new radwaste air system.

The licensee used

four 75 feet hose lengths to connect the breathing air system to the

radwaste air system. These hoses ran from the reactor building, outside

the Radiological Controls Area (RCA), to the New Radwaste building (inside

the'RCA).

On April 19, 1990, radwaste operators (RW0s) restored the air systems to

4

the normal: plant configuration. When the hoses, inside the RCA were dis-

connected, some water-(per the licensee,_about a cupful) drained from the

,

hoses, spilling on the individual and on the ground. The RWO was deter-

mined'to be contaminated at the RCA control point.

Radiological control

'

personnel iaitiatedDan assessment which determined that the hoses con-

nected to the Jdy compressor outside the RCA were internally contaminated.

The licensee bagled_the internally contaminated hoses (600-800 net cpm

direct frisk witl4 E140N, 10% efficiency) and then transferred them to-

radwaste.

The evalution of disconnecting the hoses also resulted-in soil

contaminations outside the RCA.

prior to decontamination, the soil out-

side the RCA was measured to be 1000 net counts per minute.

Soil activity

,

was 1.38 x 10 5 uc/gm.

The breathing air system hose connection was sur-

veyed which indicated that no radioactive material was present. Also, a

confirmatory air sample was taken which verified the breathing air system

had not been contaminated.

On April 23, NRC inspectors reviewed the sequence of events, control of

radioactive materials,_ assessment of the breathing air system, the

potential for other radioactive materials which may have been

inadvertently released from the RCA, and other applications of these hoses

outside the RCA. The licensee adequately detailed the sequence of events

with appropriate corrective measures completed to secure the hoses and

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remove the contaminated soil.

However,.the li:ensee was unable to

Ldetermine when or how long the hoses were released from the RCA.

No

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inadequacies were identified in the handling of the individual who had

beei contaminated..

However, when questioned if a confirmatory air sample had been taken.to

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evaluate the breathing air activi+y a Radiological Protection (RP) manager

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stated that the smear surveys of the system check valves were adequate and'

. that no air sample was necessary. When the inspector questioned what

'

actions.had.been taken to determine if other radioactive materials had

been released from the RCA, the RP manager stated that this was an

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isolated incident and no further evaluation was necessary.

'

'Or April 25, the inspector met with the Radiologicel Controls (RC)

Director to discuss the above described questions.

Personnel actions to

ensure the quality of the breathing air system were adequate in that a

confirmatory sample had been taken and showed no radioactive material.

l'

Since the contaminated hoses were outside the RCA for an undetermined

period-of time,

the RC director acknowledged the need for further

4

evaluations.

Spot checks were performed on other hoses outside the RCA

with negative results.

Overall, the inspectors found the '.icensee's

response.to-this issue adequate and had no other questions.

2.2 Outage Planning

The licensee-began an unplanned cutage (12-U-K) on April 21, 1990.

To

assess RP-involvement, .the1 inspector attended the 2:00 p.m. outage meet-

_

'ings,

During these meetings,' outage management personnel' discussions'with

the RP staff focused on radiological control support for drywell work

' activities.

RP personnel responded to questions regarding staffing'and

work' load.

The RP representative characterized this as a minor outage'in challenging

.

RP staffing levels. Drywell work included replacement of an intermediate

range monitor (IRM) detector, drywell liner shell pr'eparations and

ultrasonic testing, and-other minor maintenance activity.

,

The inspector had no further questions.

2.3 Radiological Performance Committee

In response to identified needs to improve RP program implementation and

overall radiological safety. performance, the licensee. established a task

force which recommended the-formation of a Radiological Performance

Committee.

The inspector attended the first :.eeting of this committee on

,

April 25. During this meeting, the licensee detailed the intent of the

committee relative to the previous radiological awareness committee. One

notad feature is the inclusion of bargaining unit personnel.

The licensee

exteno=d invita+1ons to major contractors providing services to the plant.

The committee meets bimonthly.

The new committee chairman is a senior

maintenance department manager.

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Radiation protection representatives to this committee provided a status-

report of radiological performance. .This included plant occupational

.

exposure history relative to industry BWR experience, internal exposure

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assessment and' airborne monitoring as a function of total dose, controls

~

for locked high radiation areas, personnel contaminations and protective

clothing use. The licensee ~ stated that although progress has been made in

_

reducing occupational exposure, the plant remains one of the highest

"

exposure'facil.ities in the industry. The licensee attributed-this, in

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part, to Oyster Creek having five recirculation loops instead of the

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typical two at most BWRs. Also, other plants do more chemical

-decontamination.

Eleven unplanned outages since the last refueling have

(

resulted in the licensee using all of the exposure budgeted for forced

l

outages. The licensee stated that more improvement is needed to reduce

4

occupational exposure.

No locked high radiation' area incidents have been

!

identified since August 1989.

The licensee has averaged six per year the

last few year.s.

The licensee is evaluating the use of protective clothing-

.(PC) made.of-paper products in an effort to reduce the number of personnel

!

,

contaminations resulting.from laundered PC's with residual contamination,

j

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The-station services representative presented details on initiatives in

'

plant decontamination and spill control.

During 1989, the licensee

recovered approximately cu,000 net square feet of contaminated plant

3

areas.

During the first three months of 1990, the licensee recovered

l

approximately 38,000 square feet.

Howen . approximately 36,000 square

f

feet were lost-due to operational incidents, personnel errors:and

!

equipment failures. Thus, the net gain was approximately 2,000 square

feet.

The: licensee discussed the need for further controls to keep the

J

areas' clean.

The-committee considered expanded use of contamination

!

control device's such as drip cloths and catch' trays.to contain _ equipment

l

1eaks. .The licensee is 'in the process.of conducting advanced radiation

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worker training to simulate spills and rad _iological area _ controls. The

'

committee-also recommended performing critiques for each identified

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incident.

i

The, inspector had no further questions.

!

2.4 Radiological Controls for Outage Related Work Activities

l

Radiological controls for outage related work activities were generally

good.

Work controls implemented by the RP staff provided an adequate

level'of radiological safety.

No deficiencies were identified in

radiological posting and' labeling.

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2.5 Under the Vessel Work

On April 26,.1990, the licensee amended radiation work permit (RWP) No.

90-0488, "Drywell 13 f t. - Replace IRM No.11," to include " PIPS and other

,

work under vessel." The initial radiological survey for this RWP

,

characterized the radiological conditions on the_13 ft elevation under -

!

vessel area. -On April 26, a work crew of instrumentation and control

J

(I&C) personnel arrived at the drywell access facility control point to

!

perform work on the carousel area under the reactor vessel.

These

1

individuals did not have a new RWP_for the_ work to be performed. .The

>

cognizant drywell g-oup radiological controls supervisor (GRCS) authorized

amending RWP No. 90-0488 to include this work. . However, a radiological

survey of the carousel under vessel area was not' performed.

The RP.

staff. briefed the I&C technicians based on the survey performed for the

IRM replacement survey 90-4927 and surveys performed during the previous

outage.

i

NRC inspectors identified that the survey on which the I & C technicians

were-briefed did not indicate any radiation levels on the carousel

platform,

This platform is located several feet above the 13 foot

elevation under the drywell and is in close proximity to control rod. drive

mechanisms which serve as a radiation source.

Subsequent licensee survey

of this area (Ap'ril 26,- 1990, Survey 90-5114) indicated radiation levels

of approximately 600 mrem /hr at the head level, about 400 mrem /hr at the

chest level, and about 200 mrem /hr at the thigh level.

The original

i

surveys at the. floor level showed 200 mrem /hr at the head level,180

'

~mrem /hr'at-the chest level, and 150 mrem /hr at the thigh level.

The

workers dosimetry was appropriately placed at the head, chest and thigh.

'

10 CFR 20.201 (b) requires that each licensee makes such surveys as (1)

may be' necessary to comply with all sections of Part 20, and (2) are

reasonable under the circumstances to evaluate the extent of the radiation

hazards'that-may be present. As defined in.10 CFR 20.201 (b), " Survey"

means an evaluation of the radiation hazards incident to the production,

use, release,-' disposal or presence of radioactive materials or other

sources of radiation under a specific set of conditions.

'

In this event, the licensee failed to recognize the specific work location

-of the I & C technicians and did not perform an evaluation of the

radiological hazards which were present in close prox.imity to the control

rod drive mechanisms.

In addition, because of'a lack of communication,.

the group radiological control supervisor did not understand the exact

work-location when he modified Radiation Work Permit 90-0488.

Some poor-

practice on the part of workers was snown in that they did not question

the absence of radiation level readings'on the survey on which they were

' briefed for their specific work location.

This failure to perform an evaluation of the radiological conditions in

L

the exact work location is a violation of NRC requirements.

This

violation was identified by the NRC (NC4 50-219/90-09-02).

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.2.6

O erational Transition to Shutdown Cooling

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Early on April 22,.1990, the operations shift placed train "A" of shutdown.

cooling:into service. The RP staff performed a radiological survey.of the

shutdown cooling room on April 24 to verify dose rates for an inspection.

They discovered that a spill had occurred from a contamination control

catch device which was leaking onto the floor. The licensee placed train

"B" into service later that evening.

The inspector questioned the timeliness of the licensee radiological

survey relative to the time the system was placed in service.

The

licensee stated that the dose rates in-the shutdown cooling room do not-

normally increase significantly when shutdown cooling is placed in

service.

The inspector-reviewed recent radiological surveys performed in the

shutdown cooling room to assess the licensee's conclusion. During this

i

review,.the inspector noted that the remarks section of Survey 90-5054

dated April 25, 1990, indicated that a radiological controls technician

a

and a station services worker received 40 mrem each while emptying and

adjusting.the catch bag associated with valve V-17-56.

At this point, the

.

radiological controls technician reentered the shutdown cooling room and

surveyed valve V-17-56 and identified that it was a hot spot, reading 2.8

rem /hr.

Radiation levels at one foot were approximately 600 mr/hr.

The

,

radiological. controls technician _and station services worker had been

'

unknowingly working in close proximity to this hot spot,

d

Licensee review of the event identified that the_ day before, April 24,

199.0, this. same radiological controls technician had been performing

general area surveys in this room.

During this general area survey. this

technician noted radiation levels of 150 mr/hr in the vicinity of valve

V-17-56.

On April 25, 1990, when the technician and the station services

1

worker entered to adjust.the catch containment, the technician once again

observed radiation readings of approximately 150 mr/hr.

Thinking that

there were no anomalous radiological conditions, the radiation technician

,

and the worker proceeded to work on the catch containment at the valve.

'

Radiation levels in close proximity of the valve were not measured.

10 CFR 20,201 (b) requires each licensee make such surveys as (1) may be

necessary to comply with all sections of Part 20 and (2) are reasonable

. under the circumstances to evaluate the extent of radiation hazards that

may be present.as defined in 10 CFR 20.201 (b), " Survey" means an evalu-

l

ation of the radiological hazards incident to the production, use, re-

,

lease, disposal, or presence of radioactive materials or other. sources of

radiation under a specific set of conditions.

In this event a radiologi-

cal controls technician.and station services worker performed work in

close proximity to.the high radiation source and were unaware of the

radiological conditions present, even though general area radiation levels

were measured by the licensee.

The radiation hazard associated with

V-17-56 was not evaluated.

This is a violation of NRC requirements (NC4

50-219/90-09-02).

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As-a result of this experience with the high radiation levels in the

vicinity of valve V-17-56, the licensee is in the process of developing a

plant shrka n survey package which will be routinely implemented.

The-

.;

. target dau for' completion of this package is July 15, 1990.

2.7 Radiological Work Permit (RWp)

On October. 18, 1939,- a maintenance worker entered the area above the

torus through a floor access on the 23 ft elevation of the reactor

s

building. The area was controlled as a high radiation area; therefore,

,

the Radiation Work Permit No.89-950 required the worker to have a dose

1

rate meter prior: to entry. The worker failed to comply with the RWP and

,

entered the area without a dose rate meter.

!

When the maintenance worker exited, a radiological control technician

(RCT) confronted him and informed him that he was required to have a dose

rate meter. Although the RCT informed the maintenance worker of his error,

no formal mechanism such as a Radiological Incident Report or Deviation

1

~ Report-was issued.

The event was not brought to the attention of the

i

appropriate level of management.

"

Initially the NRC was made aware of this event as a result of an alle-

gation-that a job supervisor had directed a plant worker to enter a high

!

radiation area without having a dose rate meter.

NRC inspector review

concluded.that while the worker did enter the high radiation area without

L

a dose rate meter,-he did not do this at-the direction of his job super-

. 3

visor.

The results of NRC review were initially presented to the licensee

[

in an exit meeting in March 1990. At that time, detailed licensee review

i

of the event had not been performed. The licensee initiated an 'investiga-

i

tion to identify the- facts of the- event and why site personnel did not

' report the event.

'

The inspector reviewed.the event via personnel interviews and document.

,

'

' reviews. Although most dose rates on the top of the torus were less than

i

100 mr/hr, the entire area was posted and controlled as a high radiation

area. A number of radiologically hot pipes in,the area have^ dose. rates-

i

greateri than 100 mrem /hr one foot away. This general area' had been a high

radiatlon area and was controlled -as such for a few months before the

event.

1

During an interview with the maintenance worker he stated that he was not

1

aware of the requirement to have a dose rate meter.

He had been working

!

,"

-in-the plant for two weeks at the' time and had a misconception that the

'

surveys t:: ken by the radiological control technician would not necessitate

4

the use of a. dose rate' meter.

He was not aware of the RWP requirements,

!

although he signed the RWP attachment sheet.

He further stated that he

!

did not believe one was required because the crew which he relieved did

not have one either.

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T'he licensee investigation' concluded that a mechanical maintenance worker

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did enter a high radiation area in the vicinity of the 1-3 containment

!

spray heat exchanger lower head witnout a dose rate meter. This area was

i

posted and controlled as'~a high radiation area and the Radio. logical-Work

Permit specified the use of a dose rate meter. As he left the area, the

_

L

individual was approached by a Radiological Controls technician who

_

confronted'him regarding the absence of a dose rate meter.

The licensee-

investigation concluded that the Radiological Controls technician and-jcb.

'

supervisor were remiss in not documenting ~and reporting this occurrence.-

!

4

Station Procedure 9300-ACW'.110,04, Section 7.4.1, requires that workers.

!

read the RWP and' sign an attachment that the individual has read, under-

l

stood and will comply with the radiological requirements as specified.

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The ontry into an area posted and controlled as a high radiation area

~

without a dose rate meter is a violation of station procedures in that the

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RWP requirements were not met.

By not documenting the event, the appro-

1

_priate level-of management was not informed and corrective actions as a

1

reso ' of this violation were not implemented. A Notice of Violation is

f

being irsued on this event because it was not identified and corrected by-

the licensee (NC4 50-219/90-09-03).

i

3.0 Maintenance and Observation

i

3.1 Monthly Maintenance Observation

3.1.1

Replacement of 4160 Volt Cable-

LAfter the loss of-USS-182 event on April 21, 1990, the licensee-

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identified a fault in-the "C" phase of the cable between the 4160V

bus "10" and.the 4160/460V transformer. The cable runs through the

)

-sand in~between the turbine building basement floor and basemat, an

area subject-to water intrusion.

The failed cable was Unishield Anaconda cable installed in 1984 due

to higher leakage current on the then existing cable.

The replace-

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ment was Unishield Anaconda cable manufactured per more recent stan-

dards.

To have the higher degree of assurance that cable pulling

j

would not result in cable da iage, the licensee designated three pul-

1

-

ling points, one more than in 1984', and limited the longest pull

'

' length to about_150 ft.

Due to the absence of drawings-indicating

specific details.of cable run ano absence of parameters like conduit

_l

friction factor, the licensee waived the requirement in the work

package that the side wall pressure be calculated prior to performing

i

cable pull. The requirement to limit the pull tension to a manu-

'

facturer approved number and using post installation testing to

determine acceptability of installed cable was considered adequate.

The inspector observed parts of the conduit cleaning and pulling of

cable.

The use of " Roto Rooter" (electrically driven mechanical

device) to clean the conduit was questioned by QC and was fcund

acceptable based on the cable and raceway installation specification

A

7

.kr

.

$

'

16

',

.

referenced inI the procedure. When cable pull began, the inspector

"

noticed that the job package was marked with a pen and ink chang to

. increase the maximum allowable pullt.ng tension from 4,000 lb. to

1

26,000.lb. and also to delete the requirement,to megger test the cable

l

after completion of_ cable pull'at each pull point.

This pen and ink.-

l

.

change was_ signed by the job planner and responsible engineer. .The-

jto, planner told the inspector that the change was not considered

j

.

substantive.

Hence, the requirement of job order procedure (A100-

l

'

9

NMS-1220.08) that'the changes be reviewed and. approved by the' job

j

supervisor, responsible technical reviewer, QA and the Group Shift

i

Supervisor.did not apply. The inspector was~ told _by-the job planner that

j

the1pullrtension'was not to exceed 4'000 lb. before the revision was

-

,

reviewed and approved.

l

The QC inspector witnessing the cable pull was also concerned about

,

the unauthorized work package revision and questioned the adequacy.of

work control'. The work was temporarily stopped by QC to obtain a

"

A

resolution.

The licensee-then determined that the pen and ink

'

-revision was substantive-and was not to be implemented _before appro--

't

priate review and_ approval were complete. A Quality Deficiency

,

Report (QDR) was issued on job control.

,

During the cable pulling, one of the pulleys anchored to the north

'

end of._the turbine building stairwell pit became dislodged from the

<

wall

The licensee determined that the Ramset Tru-Bolts were

,

installed without following the requirements of the procedure for

. installation and repaired it. A QDR was written to addrer:s the

deficiency. - During the pulling one of the three phases suffered some

"

damage, consisting of a seven inch long scrapped portion with a deep

.

y

angular cut.

q

A' revision:to the job order was processed to_p'll the damaged cable-

!

u

an additional length to cet out the damaged portion.

High voltage

~

+esting was done on all three' phases with acceptable test results.

,

The licensee performed DC high voltage testing including step voltage

test, and dissipation factor test on the cable with acceptable test

t

results. A post jot debriefing was also conducted. At the comple-

. tion of the job, the; job ords. had a total of.eight revisions, two

'

,

',

00Rs and.one material nonconformance report (MNCR) written on it.

.

-The cable pulling was made diff.icult due to the long length of

' preexisting conduit, 3-1/2" in' size and consisting of numerous bends.

and turns. :Also, the new cable was larger in diameter than the

previously installed cable. The licensee increased the number of

pulling points to address that.

Certain informality in job control

was observed and an unauthorized pen and ink change was made to the

work package.

However, QC response to that was appropriate and

timely to ensure adequate work control.

The job coordinator also

,

responded adequately in resolution of this issue.

Radiological

h

Controls, Engineering and management support of the job was found

adequate.

Inspectors had no other questions.

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3 ' Af

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17.

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~3.1.2.

Fire Suppression Water System Valve Replacement

!

The inspector reviewed the work package and observed selected

portions- of the replacement of fire header valve V-9-18.

Certain

"

parts ,of the fire water system.were isolated, thus disabling various

sprinkler and deluge systems and fire hydrants.

The licensee

established continuous fire watch at these areas as required by the

plant Technical Specifications.

No unacceptable conditions were

identified.

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3.2 Monthly Surveillance Observation

3.2,1

Battery Surveillance Testing

.i

i

e

'On May 4, 1990, the licensee performed weekly surveillances on the-

I

"B" station battery and the emergency diesel batteries, procedures

634.2.002, revision 14, and 636.2.004, revision 18, respectively. The

1

inspector observed and verified that activities did not violate

l

limiting conditions for operations, required approvals.were obtained,

surveillance prerequisites were met and appropriate personnel safety-

precautions were taken.

l

During the surveillance on the "B" battery, the electrician read the

')

voltage of "B" battery from a local meter in the battery room. ' The'

4

procedure, however, directed the voltage be measured with a voltmeter

i

'

+

at the battery terminal. When this was brought to the attention of

]

' >

- the electrician, the error. was corrected.

Because the battery was on

{

a zero: float', as required by procedure,-the difference in-voltage

l

between the local meter reading.-(132 volts) and the volt meter read

(132.6 volts) was minimal. _The inspector ' observed that a contribu-

ting cause for this error.was that the electrician was following the

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surveillance procedure. data sheet rather than the surveillance

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procedure itself. No unacceptable conditions were identified.

{

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3.2.2 Instrument Calibration

I

On May 16, 1990, the inspector observed calibration of two Rosemont

!

pressure transmitters (Model #1153GB9RA) installed in the' reactor

fuel zone level instrumentation system.

The transmitters were found

'

to.be within their specified value, and no adjustments were needed.

-The inspector verified that the test equipment was calibrated and the

technicians followed the equipment EQ system component evaluation

'

work (SCEW) sheet requirements and adequate radiological pactices.

As one of the transmitters was on a rack that was roped off as 6

contaminated area, the technicians had to reach out to perform

calibration. Gloves were used and necessary radiological control

surveys were done prior to taking the used tools outside the

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contaminated area.

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{f: .g_. 4 '

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TheJinspector found the technicians were fo11'owing adequate

'

e

radiological practices but questioned the appropriateness of the RWP

used fo.c this task. .The RWP was designated for use durino p rfor-

mance of minor maintenance in clean-areas and also q uired use of

surgeon's gloves when opening possibly contaminated instrument

sensing lines. .The inspector discussed this with the Group-

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Radiological. Control Supervisor (GRCS).

The GRCS indicated that the

-;

RWP. for minor maintenance 1n'a contaminated area will be used in the-

~

o

future and the I & C technicians will be-so informed,

This RWP will

~

be modified to address tasks-that require reaching into contaminated

areas.

The inspector did not have any other questions.

4,0 ~ Engineering and Technical Support

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4 '.1

Core Spray System Relief Valves

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On May 2, 1990, during routine surveillance testing-(pump operability) on

Core Spray System II, the relief valve at the discharge of the. booster

pumps, V-70-24, prematurely lifted.

The relief valve, made by Lonegran,

a

was replaced.

)

i

Each core spray system contains one discharga relief valve.

To protect

[

the low pressure core. spray system piping from overpressure due to back

l

leakage, these relief valves are set- to lif t at 350 psig 12%.

The relief

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- valves are replaced at a certain frequency as part of the licensee's ISI

,

program,

'I

During the last refueling outage, when the licensee replaced both of the

relief, valves with two new valves, the new valves failed surveillance

1

testing by prematurely 11fting.

The failed valves during the last refuel-

ing outage (see Inspection Report 89-07) had set point drifts of 14% and

i

6% due -to spring relaxati.on caused by residual stress from the manufac-

--

turing; process, The manufacturer did-not-stress-relieve the valve which.

could have eliminated the residual stresses.

The manufacturer reported

j

that the further spring relaxation is not expected.

These valves were

-subsequently sent to Wiley Labs, who reset them to 350 psi 12% after

i

necessary testing.

,

The_ licensee.had intended to replace the valve. springs as an additional

l

precautionary measure when the system relief valve' failed during surveil-

i

lance testing on May 2, 1990.

The failed valve is being sent to Wiley

.4

'

Labs for necessary testing, identification of a root cause of failure and

adjustments or refurbishment.

'

4.2 Emergency Diesel Generator Battery

l,

On April 17, 1990, the No. 1 Emergency Diesel Generator (EDG) failed to

start during performance of its biweekly surveillance test.

The cause of

9

<

the diesel generator failure to start was a degraded battery cell.

,

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1

.; c q

4: ;

19

Tne EDGs at Oyster Creek are startU by a 56-cell battery. The battery is

an Exide lead-antimony battery, Mode l 4-MS420-II.

The model battery has a

rated operation life of 8 years; howe,ar, the battery which failed was

only 5 years old.

The licensee has re.urned the battery to the

manufacturer to determine the cause of lailure.

During. surveillance testing, the diesel 1. started in a low voltage mode

,

where three-sets of resistor banks are cut into the starting circuit.

During an actual emergency start (" fast sta t"), these resistors would not

be in the starting sequence.

The licensee umonstrated that if the diesel

were called upon-to start, the diesel would i ave started.

As part of the. licensee's troubleshooting effurts, the licensee measured

individual battery cell voltages during a low voltage diesel start.

Another cell was identified to be replaced because its voltage was lower

than that observed in'other cells.

The licensee performs a capacity test on the battery every refueling

outage, a biweekly load test and periodic specific gravity and water-level

measurements. The licensee is reviewing the need for additional tests to

determine battery degradation; however,- they concluded that the biweekly

diesel load test would-provide some detection of battery degradation prior

to causing the diesel to fail to start on a fast start.

Based onthe troubleshooting efforts and testing conducted on the diesel

batteries, the inspector concluded that the diesel batteries were

operable. ~The inspector identified no unacceptable conditions.

This review was completed late last inspection period and part of this

period.

5.0 Observation of Physical' Security

-

During: daily tours, the inspectors verified that access controls were in

accordance with the Security Plan, security posts were properly manned,

protected. area gates were locked or. guarded and that-isolation. zones were

free of obstructions. The inspectors examined vital area access-points to

ver.ify that they were properly locked or guarded and that access control

.was in accordance-with the security plan.

~

6.0 Safety Assessment / Quality Verification

,

6.1 Licensee Evaluation of USS-182 Cable Failure and Tests done on other

Cables.

The licensee's preliminary inspection and testing of the failed cable

indicated that a manufacturing defect may have caused the cable

= f ai_l ure .

This was based on an inspection which indicated that an

l

,

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C

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<.-

[

20

1

. insulation failure propagated from inside to the outside scrface of

!

'

insulation.on the "C" phase.

However, damage.due to water, was not ruled

out.-- The damaged cable was sent to the laboratory for further evaluation -

and testing. At the end of the inspection period these tests were in

progress.

.;

e

-

Various safety related cable, including th subject cable, are routed

!

through the sand bed between the turbine building basement floor and the

!

basemat, an area subjected to water intrusion. As a precautionary mea-

'

sure, a temporary variation was installed to drain water.under the turbine

.

building floor to a sump in the feed pump room.

-- i

'

1

i

The other safety related cables that are routed through the sand bed under

'

the turbine building basement-include the 4160V unit substation 1A2 cable

and cables for.the four. core spray pump motors. These are GE cables

installed during' original construction.. One of these five cables was

. tested with acceptable test results.during the 12-U-K outage.

The

j

' licensee also tested cables, with acceptable results, for emergency-service

'

water pumps #1 and #2 which are also routed partly underground.

These two

e

cable routings consist of GE cables up to a manhole in the-intake

,

structure and Anaconda Unishield cable from the manhole _to the' pump motor.

]

In addition, the licensee is preparing a program for periodic testing of

i

medium voltage cable that will be implemented during the 13R refueling

,

outage. This testing will have a periodicity of once ever/ three

'

. refueling outages.

The inspector considered the licensee's initiatives in

-!

this ' area appropriate and did not identify any unacceptable conditions.

I

6.2_Contro1LRoomOperatingParameterVedfication

d

On April 30, 1990, a management meeting was conducted to review the

i

results of the Augmented Inspection Team performed in December 1989

.l

. regarding.an event where vacuum in one condenser bay fell below the scram

I

setpoint.

No automatic scram occurred and the operator ~did not manually

!

~

insert the scram.. This report' concluded that the reactor protection

j

system did not sense vacuum below the scram setpoint and that the measures

taken by the control room operator.in response to his indications were

'I

acceptable. .During this meeting, _the NRC and -licensee discussed

i

station policy and procedural requirements for operators to verify

d

i

parameters prior.to initiating manual action.

The inspector reviewed

j

'this policy as follows.

l

'

Station Procedure 106, " Conduct of Operations," steps 4.2-2 and 4.2.3,

.

specified similar requirements' authorizing control room operators and

Group Shift Supervisors.to initiate manual action on " verified

parameters." 'The same procedure, step 4.4.11, requires operators to-

believe their indications unless they are verified to be false. The.

question centered around the possibility that verification could unduly

' delay operator actions.

.

m.

. .

. .--

m

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L

, Q.

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21

,

The. licensee initiated a change to Station Procedure 106, revision 57,

which clarified the requirements of steps 4.2.2 and 4.2.3. _ This change

reiterated the procedure requirements of 4.4.11 that operators are to

believe their indications unless verified to be false.

.

The inspector also reviewed two site memoranda giving guidance on how

operators are to implement station policy and procedural requirements.

The plant operations director memorandum of December 7,1989, reminded

operators of their authorized authority as defined in Station Procedure

-

106, steps 4.2.2, 4.2.3 and 4.2.11 (discussed above) and emphasized the

'

need to use the full spectrum of instrumentation.

The memo also stressed

i

the importance of making decisions based on all _available information.

!

The inspector also reviewed a station director memorandum of May 21, 1989,

which emphasized to plant operators that they are to use available-infor-

mation in making their decision. Also, a May 20, 1989, an independent-

,

trip review group action recommended emphasizing to plant operators that

they use confirming indications when possible in making their decisions.

Inspectors concluded that ,0yster Creek's policy and guidance to the

_

operators is clear', especially in view of the change.to Station Procedure

< t

106, and does not inappropriately emphasize verification measures to-

contro'l room operators.

6.3 ' ' Nonconforming TED Breakers

During-licensee _preinstallation testing of a molded case TED 100 amp

replacement breaker manufactured by GE on May 7,E1990, the overcurrent

trip function did not work. Additional tests indicated that five out of

seven_similar breakers failed the overcurrent trip test. The licensee had

i

a similar 15 amp _ TED _ breaker test failure during November 1989.

All of

-these TED creakers had an undervoltage (UV) device installed on one of

three. phases. All these breakers were procured from GE -under the .same

purchase order as safety 'related equipment, and 10 CFR Part 21 was ap-

1

"a

plied.

>

The licensee prepared a material nonconformance report and identified the

,

_ failure mode to be an interference of the UV device with the overcurrent

' trip. function. Their communication with GE indicated that TED and TEHD.

b'reakers with.UV devices manufactured during a certain time period could

be subjected to this defect. These breakers were assembled and tested in

. ;

different locations and GE did not perform an overcurrent trip test after

installation of the UV device.

The a'dequacy of the licensee's

,preinstallation test to identify this deficiency was reviewed and

'

confirmed by GE. GE told the licensee that GE made a Part 21

<

notification on the failed breakers.

1

The only TED breaker from the purchase order that is installed at Oyster

Creek is in the auxiliary boiler control system.

The rest of the breakers

(about 170 total) were returned to GE.

The installed breaker was found

capable of performing its function. The licensee has implemented a

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f

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22

,

preinstallation test program which was effective in ' identifying this

manufacturing defect. The inspector determined that the licensee's-

followup action was appropriate.

'

7.0 Inspection Hours Summary

.

Inspection consisted of 307 direct inspection t,0urs; 71 of these direct

'

inspection hours were performed during backshif t periods, and 25 of these

hours ~were deep backshift hours.

-8.0

Exit Meeting and Unresolved Items

'

,8.1

Preliminary Inspection Findings

1

A verbal summary of preliminary findings was provided to senior- .

licensee management at the conclusion of this inspection on June 15, 1990.

.

During the inspection, licensee management was periodically notified

verbally of the preliminary findings by the resident inspectors.

No

written: inspection material was provided to the licensee during the'

inspection. ~No, proprietary information is included in this report.

8.2 : Attendance at Management Meetings Conducted by Region Based

Inspectors

!

During 'this inspection period, the resident inspectors attended exit

meetings for inspections 50-219/90-08, 50-219/90-10, and an~ interim exit'

a

for the requalification examination 50-219/90-05. At these exit meetings,

the lead inspector discussed inspection activities and findings with

i

senior licensee management.

8.3 Unresolved Items

Unresolved iten: are matters for which more information is required in

.i

order to ascertain whether they are acceptable, violations or deviations.

L

An unresolved item is discussed in paragraph 1.4 of this report.

v

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5

ATTACHMENT I

Personnel Contacted

Licensee Personnel

n

K.LBarnes,. Licensing Engineer

.R. Barrett', Plant Operations Director

-

J. Barton, Oyster Creek Deputy Director

L

-R. Bernava, Security

T. Blount. Emergency Preparedness

R. Brown',_Radwaste Operations Manager

,

J. Brownridge, Maintenance

L

G. Busch, Licensing Manager

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P. Cervanka, Plant Operations Engineering

P. Crosby, Plant-Engineering

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L

J. De Blasio, Plant Engineering Manager

B. De Merchant.-Licensing Engineer

s

T. Dempsey, Plant Engineering Manager

J. Dostel, Shift Technical Advisor

R. Ewart, Security

R.: Fenti, QA Mod /0PS Mgr.

E.

Fitzpatrick, Vice President & Director

D. Jones, Plant Engineering

J. Kowalski, Manager, Training-

L. Lammers, Plant Maintenance Director

R. Larz'o, Plant Engineering Supervisor

K. Mulligan, Plant Operations

D.-Ranft, Engineering Manager

'H. Robinson, Technical Functions

J. Rogers, Licensing Engineer

g

A.:Rone, Plant Engineering Director

- *P.!Scallon, Plant Operations Manager

E. Scheyder, MCF Director

H. Sharma, Technical. Functions

,

*M.-Slobodein- Rad Con Director

,

R. Sullivan,~ Emergency Preparedness Manager

R. -Tilton, QA Engineering Manager

' H. Tritt, Operatory Training Supervisor

D. Tuttle, Radiological Controls Deputy Director

NRC Personnel-

'

M. Banerjee

E. Collins

4

i

-Denotes a+tendance at exit meeting.

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. - - - - - - . - - . - -

ML20056A619

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