Text

IE Insp Rept 50-293/80-30 on 801031-1126.Noncompliance Noted:Failure to Limit Operation of Containment Purge & Vent Valves to 90-h Per Yr.Control Room Operator on Watch for 32-h
	ML20009A094
Person / Time
Site:	Pilgrim
Issue date:	02/06/1981
From:	Jerrica Johnson, Martin T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20009A092	List: ML20009A091; ML20009A093; ML20009A094;
References
	50-293-80-03, 50-293-80-3, NUDOCS 8107080080
	Download: ML20009A094 (16)
	v • d • e

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U.S. NUCLEAR RicVLATORY COMMISSION

0FFICE OF INSPECTION AND ENFORCEMENT

50293-801106

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50293-801107

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Region I

50293-801113

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Report No. 50-293/80-30'

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Docket No. 50-293

Category

C

License No. DPR-35

Priority

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Licensee:

Boston Edison Company

800 Boylston Street

Boston, Massachusetts 02199

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Facility Name:

Pilgrim Nuclear Power Station

Inspection at:

Plymouth, Massachusetts

Inspection conducted:

0 tober 31, 1980 - November 26, 1980

Znscoctors:

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J. Johnson, Senior 5:itfe'ntInspector

date' signed

date signed

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date sigaed

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Approved by:

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T. DMartid, Chief, React <fr Projects

cate signed

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Section No. 3, RO&NS Branch

Inscection Summary:

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Inspection on October 31 - November 26, 1980 (Report No. 50-293/80-30)

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Areas Inspected:

Routine unannounced inspection of plant operations including

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an operational safety verification, followup on previous inspection findings,

followup on the 'A' Safety Relief Valve (SRV) inadvertent opening on October 31,

1980, maintenance activities, licensee's actions in response to IE Bulletins,

containment vent and purge operations, staff working hours and overtime, a

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review of seismic instrumentation, and a review of the design of containment

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cooling systems.

The inspection involved 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months by the resident inspector.

Results: Two deviations were identified in two areas (failure to limit contain-

ment vent and purge operations to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per year, Paragraph 7; and failure to

limit a control room operator's working hours, Paragraph 8).

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8207080080 810212 h

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PDR ADOCK 050002934

G

PDR)

Region I Form 12

(Rev. April 77)

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DETAILS

1.

Persons Contacted

W. Armstrong, Deputy Nuclear Operations Manager

R. Belanger, QC Inspector

E. Cobb, Chief Operating Engineer

F. Famulari, QC Supy-"isor

J. Fiumara, Cr puter Engineer

E. Graham, Compliance Engineer

R. Machon, Nuclear Operations Manager (Pilgrim Station)

C. Mathis, Deputy Nuclear Operations Manager

P. O'Brien, Construction Management Group Leader

W. Olsen, Senior Nuclear Training Specialist

R. Reposa, QC Inspector

J. Seery, Staff Assistant - Nuclear Safety

P. Smith, Chief Technical Engineer

P. Williard, I&C Engineer

E. Ziemanski, Management Services Group Leader

The inspector also interviewed membus of the Operations, Security, Tech-

nical and Maintenance Staffs.

2.

Followup on Previous Inspection Findings

(0 pen) Unrasolved Item (293/80-29-03):

During a telephone corversation

between NRC:IE: Region I management and Boston Edison Company management on

November 17, 1980, the licensee agreed to take the following action with

respect to certsin TMI Task Action Plan Category ' A' Requirements:

Implement station approved procedure (s) for taking and handling a

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containment atmospheric sample by November 28, 1980.

Implement station approved procedures for converting high range noble

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gas effluent monitor readings (R/hr) to effluent release rates (Ci/sec)

by November 28, 1980, and

Correct the operation of control switches for containment vent and

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nitrogen makeup valves (A0 5033 A&C, A0 5041 A&B, A0 5043 A&B) to

allow operation between 'open' and 'close' positions without a key and

require a key to get in the ' emergency open' position, by January 1,

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1981.

These commitments were confirmed in a letter from the Director, NRC, Region

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I to Boston Edison Company fit-d November 18, 1980.

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This item remains open pending a review of the completed actions.

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3.

Operational Safety Verification

a.

Scope

The inspector observed control room operations, reviewed applicable

logs, conducted discussions with control room operators, and verified

proper lineup of selected portions of emergency systems.

Tours of the

station yard, reactor building, control room, and intake structures

were conducted to observe equipment condition including potential fire

hazards, housekeeping, physical security and the implementation of

radiation protection controls.

These reviews were conducted to verify conformance with the code of

Federal Regulations, Technical Specifications, and the licensee's

procedures.

b.

Events

(1) On October 31, 1980, the licensee notified the inspector of an

error in the original seismic analysis for the Salt Service Water

(SSW) System. This analysis took credit for a seismic anchor

located outside of the system isolation valves.

The anchors

were, in fact, the screenwash pumps, which are locatad outside

the missile protected area and would jeopardize the operability

of the SSW system if a tornado induced missile should impact the

screenwash pumps.

The inspector informed the licensee that relief from Technical Specification 3.5.b was required to permit continued operation.

The reactor wa: shutdown at 11:52 a.m. on October 31, 1980 (for

an unrelated problem with ' A' Safety Relief Valve) and the licensee

immeciately requested temporary relief from TS 3.5.b to allow

operation while modifications were made to the seismic anchors.

At 9:30 p.m. on October 31, 1980, NRC:NRR waived the requirements

of TS 3.5.b until November 7,1980, since the only cause of

system inoperability would be damage to the screenwash pumps due

to a tornado.

A reactor startup was comenced at 10:29 p.m. on October 31, 1980

and routine operations continued while modifications were performed.

The inspector toured the intake structure on November 7, 1980,

observed the installation of the new supports and verified com-

pletion of the modification through discussions with the cog-

nizant irplementing engineer (construction group leader).

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The inspector had no further questions in this area.

This event

is described in LER 80-78.

(2) Drywell Unidentified Leakage

A plant shutdown was initiated at 3:10 p.m. on November 6, 1980,

to investigate and repair reactor coolant system (RCS) unidentified

leakage calculated to be 5.9 gpm.

Investigation revealed packing

leakage on 'B' recirculation pump discharge valve, 'C'

Inboard

MSIV, and RCIC Valve No. 1301-16.

After startup on November 8, 1980, following repairs and pressuri-

zation, packing leakage was observed on both 'C'

and 'D' inboard

MSIV's.

While power and RCS pressure was being reduced, a reactor

scram occurred due to reactor vessel level low and difficulty

controlling level at low steam flows.

During a drywell inspection with the RCS at 900 psig on November

13, 1980 (following repairs to 'C'

and 'D' MSIV's), a pinhole

leak was observed at a coupling weld on the 2 inch line from the

reactor vessel drain to the clean up system.

The reactor was

again shutdown to repair the defective weld.

The reactor was started up and the unit returned to service on

November 14, 1980.

The inspector reviewed the licensee's actions to ensure compliance

with the Technical Specifications and station procedures.

Items

reviewed included control room instrumentation, logs and records

and r;iscussions with licensee personnel.

The inspector reviewed

sta: tup and shutdown check lists, verified placing the alant in

cold shutdown within the required time, satisfactory completion

of se'lected valve operability tests following repairs, and veri-

fied RCS leakage measurements on November 15-17, 1980 to be

within TS limits.

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Maintenance activities involving the weld repair are described in

Paragraph 5.

These events are described in LER's 80-84 and 80-87.

c.

Findings

No items of noncompliance were identified during this review of routine

operations and the events described above.

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4.

Inadvertent Opening of

'A'

Safety Relief Valve (SRV) at Power

a.

Description of the Event

At about 11:45 a.m. on October 31, 1980, the ' A' SRV inadvertently

opened at full power. Operators isolated the nitrogen supply to the

drywell instrumentation and placed the drywell instrumentation on

station compressed instrument air.

Reactor power was decreased to

approximately 50% and with the air supply pressure about 110 psig,

attempts were made to : hut the 'A' SRV by cycling the control switch.

The relief valve failed to shut and the reactor was manually tripped

frou. 50% power, and a cooldown was initiated to investigate and correct

the problems.

b.

Review / Investigation / Resolution of Concerns

The inspector reviewed the events, held discussions with the operators

and licensee management, observed instrumentation and reviewed records.

It was identified that high nitrogen supply pressure (about 160 psig)

was the cause of the opening of the

'A' SRV and that once the accumu-

lators are charged to this pressure, a time delay is experienced

before this pressure decreases and allows the SRV to reclose.

A similar event took place on October 7, 1980.

Following that event

the licensee had been monitoring the nitrogen supply pressure once per

shift to keep it below a value which would cause leakage by the solenoid

valve.

No specific failure of the pressure regulators had been identi-

fied.

It was also noted following this event on October 31, 1980 that a

delivery of liquid nitrogen had been made immediately prior to the ' A'

SRV opening.

It is suspected that a cause of the rapid rise in nitrogen pressure

may have been due to liquid nitrogen passing by the ambient vaporizor,

freezing an in-line regulator in the open position, and causing high

supply pressures (greater than 160 psig) to be transmitted to the

solenoid actuator.

During a telephone conversation between NRC:IE: Headquarters, Regional

management personnel and Boston Edison Company management personnel on

October 31, 1980, the licensee agreed to take the following actions.

The nitrogen supply system to the SRV's will remain isolated and

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tagged.

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The frequency of containment air sampling will be increastd to

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once per sitift for 7 days.

The implementation of these two items, above, would be verified

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by the resident inspector prior to the resumption of operation,

and

That the NRC will be provided documentation of an evaluation of

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system modifications and testing, for review and concurrence,

prior to restoration of the nitrogen system to in service.

These commitments were confirmed in a letter from the Director, NRC,

Region I, to Boston Edison Company dated October 31, 1980.

c.

Findings

At 5:10 p.m. on October 31, 1980, the inspector verified that the two

inch nitrogen supply valve to the drywell instrumentation (SRV's) was

isolated and tagged, and that the tag was logged in the Watch Engineer's

tag log book.

The inspector also verified an entry made in the Watch

Engineer's Instruction Log requiring a drywell and torus oxygen sample

once per shift.

Thase actions were taken prior to the re amption of

plant operation at 10:39 p.m. on October 31, 1980.

The inspector reviewed entries in N operations log and additional

data sheets from October 31,1980 '.c November 6,1980 to ensure that

oxygen samples were being taken on e per shift and concentrations were

within the Technical Specification limits with instrument air being

saoplied to the drywell instrumentation.

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No items of noncompliance were identified.

Pending the completion of

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the licensee's testing of the nitrogen supply system, system modifica-

t. ions, evaluation of these actions by the NRC, and restoration of the

nitrogen system to service (for drywell instrumentation), this item is

unresolved (293/80-30-01).

This event is described in LER 80-80.

5.

Maintenance Activities

a.

Scope and Acceptance Criteria

The inspector reviewed the licensee'< t-tivities surrounding the weld

repair of a two inch stainless steel i;ne from the reactor vessel

drain to the cleanup system performed on Novembr

, 1980 in order to

verify conformance with the ASME Code, the Tect ma l Specifications,

and the licensee's procedures.

The following records were reviewed:

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Maintenance Request No. 80-8023, and Maintenance Summary Sheets

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Welding and Testing Specification 6498-M-305, Revision 10

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Visual Inspection Record

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QAD Field Weld Check List

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Weld Rod Withdrawal Sheets

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Liquid Penetrant Test Results

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Welder Qualification Records

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-, NDE Qualification Records

b.

Findings

No items of noncompliance were identified however the inspector had

the following comments:

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The weld number designation and welder identification number were

inconsistent on several of the documents reviewed.

The licensee

agreed to review and correct the documents.

During th review of the post repair inspection sheets, the

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inspector informed the licensee that instructions, should include

reference to specific minimum required test pressures, and should

be provided to station personnel to ensure that the correct test

pressures are used.

The licensee acknowledged the inspector's

comments and stated that the appropriate actions would be taken

to include references to specific test presssures as required.

ihe inspector had no further questions in this area.

6.

Inspc. tion and Enforcement Lulletin Fo(lowup

The inspector reviewed the licensee's activities in response to the IE

Bulletins described below.

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a.

80-07:

BWR Jet Pumo Failure

The inspector reviewed implementation of station procedure TP 80-67

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Revision 1, " Jet Pump Operability surveillance IEB 80-07", dated

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October 22, 1980.

This procedure was revised (to include all the data

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required to be taken by the Bulletin) in response to the inspector's

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comments (Inspection Report 50-293/80-27).

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No items of noncompliance were identified.

The inspector had no

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further questions at this time and considers this Bulletin closed.

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b.

80-17:

Failure of Control Rods to Fully Insert During a Scram at

a BWR

Original Bulletin Pph 3.a; Procedure No. 2.1.6, "Reacter Scram"

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Revision 11 was revised to include a requirement to connect an

air hose to the one inch vacuum breakers on the east and west

headers and verify that air blows out the vent lines after each

scram was reset. The inspector spot checked the implementation of

this procedure and no problems were identified.

Original Bulletin Pph 3.b; Procedure No. 2.1., "Startup froe Cold

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Shutdown", OPER 01 Startup check list was revised to include a

check to verify that the S.D.V. was free of water prior to startup.

Original Sulletin Pph. 6.b; Procedure No. 2.2.86, " Residual Heat

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Removal", was revised to require all available containment cooling

if during 41grmal continuous operation the torus temperature

exceeded 60 F.

Supplement 1, Pph. A.2; Procedure No. 2.2.24; " Standby Liquid

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Control System", was revised deleting the requirement for station

supervision approval prior to initiation of the.SBLC system.

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Supplement 1. Pph. A.4; Procedure No. 1.3.10, " Key Control", was

revised to require the $8LC initiation key to be located on the

main control panel 905 above the switch.

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Supplement 3, Pph. 1.1.; Alarm Response Procedure No. 2.3.2.8,

was revised to require a scram if the scram air header pressure

reached 60 psig.

Emergency Procedure No. 5.3.8 " Loss of Intru-

ment Air", was also revised requiring a scram if the air pressure

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decreased to the point where it caused a feed regulating valve

lock up (as i ndicated on the main control panel 905 by red indi-

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cating lights on push buttons).

Based on information received

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from the licensee, the lock up occurs at a pressure of 64/65 psig

and clears at 74 psig.

The inspector had no further questions

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concerning this item.

Su;;plement 3, Pph.1.b.1; Procedure No. 2.4.3, " Rod Drift", was

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revised requiring immediate reactor scram if two rods in a nine

rod array start drifting in.

Supplement 3, Pph. 1.b.R; Procedure No. TP80-74, Revision 0,

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" Determination of Leaky Rod Drive Outlet Valves", was developed.

The licensee monitors the withdraw header temperature when a

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control rod high temperature alarm comes in (greater than 20gg)

and weekly thereafter.

This procedure requires a shutdown if 36

total rods indicate lgaking outlet valves (withdraw header temper-

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ature greater than 50 F above ambient).

The inspector spot

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checked implementation of this procedure and identified no problems.

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Supplement 3, Pph. 2; Procedure No. 2.1.6, " Reactor Scram", was

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revised requiring a verification of actuation of the six scram

dischrage instrument volume limit switches prior to bypassing the

S.D.V. high level scram, and a subsequent clearing of all six

limit switches following reset.

During a telephone conversation between NRC:IE, NRR and BECo

management personnel on November 24, 1980, the acceptability of

this method to detect damaged floats was discussed.

This licensee agreed to modify computer inputs to include recordings

for all six limit switches and to revise station procedures to

add specific acceptance criteria (time to reset) for limit switch

operability.

The licensee stated that these modifications and

procedure changes were expected to be in place by December 19,

1980 and that in the interim, surveillance procedure no. 8.M.1-20

(limit switch set point check) would be performed prior to startup

if a scram should occur.

The licensee's proposal was found to be acceptable and the inspector

had no further questions in this area at this time.

Confirmatory Order dated October 2, 1980:

The inspector questioned

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the licensee concerning whether the once per-shift UT checks of

the scram discharge volume (S.D.V.) had been performed for several

shifts between October 28, 1980 and November 6, 1980 for which no

entry had been made in the operations log.

The licensee provided

verification to the inspector that they had been performed. The

licensee also determined that all checks had not been performed

at greater than six hour intervals and immediately issued clari-

fying instructions in the Watch Engineer's Log.

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The inspector subsequently reviewea the operations log entries

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from November 6-14, 1980 to verify the implementation of these

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once per-shift UT checks of the S.D.V.

No discrepancies were

identified.

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c.

80-14:

Degradation of BWR Scram Discharge Volume (SDV) Capability

The licensee issued a revised response (dated September 10, 1980)

because of orevious NRC concerns that the original response (dated

July 30, 1980) did not adequately address periodic testing and notifi-

cation.

The inspector reviewed the following station procedures:

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Revised procedure no. 2.2.17 " Communications Systems", Revision

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6, which requires an ENS report to the NRC for any inoperable

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S.D.V. vent / drain valve or if any S.D.V. vent / drain valve is

closed for more than one hour in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

Normal operating procedure no. 2.2.87, " Control Rod Drive System",

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Revision 8, which requires that the S.D.V. vent and drain valves

be normally open.

Revised su.reillance procedure 8.M.1.20, Revision 7, "High Water

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Level Scram Discharge Tank", which requires any inoperability to

be reported to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The licensee's September 10, 1980 response stated that a procedure for

periodic (once per cycle) operability testing would be prepared by

January 1,1981.

This Bulletin remains open pending a review of the implementation of a

surveillance procedure to perform periodic operability testing of the

S.D.V. vent and drain valves.

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No items of noncompliance were identified during the review of

these IE Bulletins.

7.

Containment Vent and Purge Operations

a.

Scope and Acceptance Criteria

The inspector reviewed the licensee's activities concerning containment

vent and purge operations with respect to the criteria established by

NRC:NRR and the commitments made by the licensee in the following

documents.

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Letter from NRR to BECo. dated November 29, 1978, " Containment

Purging During Normal Operation".

Letter from NRR to BECo. dated October 22, 1979, " Containment

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Purging and Venting During Normal Operation".

Letter from NRR to BECo. dated September 9, 1980, " Confirmation

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of Commitment to Staff's October 23, 1979 INTERIM POSITION ON

CONTAINMENT Purging and Venting".

BECO

Letter from BECo to NRR dated January 9, 1979, No. 79-22.

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Letter from BECo to NRR dated August 21, 1979, No.79-158.

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Letter from BEco to NRR dated December-19, 1979, No.79-270.

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Letter from BECo to NRR dated May 27, 1980, No. 80-94.

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Letter from BECo to NRR dated May 27, 1980, No. 80-95.

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The inspector's review included the following items:

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Station procedures.

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Plant Design Change Request (PDCR) 3033, " Modify Containment Vent

and Purge Valves".

Completed maintenance requests.

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Completed valve operability surveillance tests.

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Discussions with station personnel.

b.

Findings

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(1) The licensee stated in the December 19, 1979 response to NRR that

permanent modification would be made to the isolation logic to

ensure at least one uninhibited isolation signal was provided to

the containment vent and purge isolation valves in the event

another signal was bypassed or blocked.

The inspector had verified on October 28, 1980 (during a review

of actions in response to the TMI Action Category ' A' Requirement)

through a review of preop. test procedures that changes to the

isolation logic had been performed.

The acceptance criteria in these tests included the following for

valves:

A0 5033A

A0 5033C

5035A

5035B

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5036A

5036B

5041A

5041B

5042A

5042B

5043A

5043B

5044A

5044B

All valves isolate on h? drywell pressure and/or lovt reactor

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vessel water level, if the control switch is in the 'open'

position.

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For valves A0 5033A, 33C (1" nitrogen makeup valve)

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A0 5041 A&B, A0 5043 A&B (2" drywell and torus

vent valves)

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Isolate on Lolo reactor vessel water level if the control

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switch is in the ' emergency open' position.

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(2) The licensee stated (in the May 27, 1980 letter BECo No. 80-95)

that the containment vent and purge valves are 20" butterfly

valves,thatanalysishasshownoperabflityintheeventofa

LOCA if their opening is limited to 45 , and that

difications

have been implemented limiting their opening to 45

During the review of PDCR 80-33, and the associated maintenance

requests, the inspector verified that six of the eight 20" con-

tainment vent and purge valves had been modified to limit their

opening by installing physical stops on the operator.

No records

of modifying the two outbcard 20" purge inlet valves (A0 5035B

and A0 5036B) were available.

Also, in the May 27, 1980 response from the licensee, a 4" nitrogen

purge isolation valve, A0 5033B was not addressed.

This information pertaining to the two outboard 20" purge inlet

valves and the 4" nitrogen makeup valve is under further review

by NRC and remains unresolved (50-293/80-30-04).

(3) The licensee stated that in order to maintain the drywell-torus

d/p required by Technical Specification that the two inch series

drywell and torus vent valves (A0 5041 A,B and A0 5043 A,B) would

not be limited in their use during normal operation.

The licensee further stated that purge and vent isolation valves

were used to inert the containment within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period

allowed by Technical Specification after startup.

..u to ceinert

the containment in the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period prior to shutdown also

allowed by Technical Specifications.

The licensee did state, however, (in the August 21, 1979 and

December 19, 1979 responses to NRR; that operation of containment

purge isolation valves (other than those used for drywell-torus

d/p control) would be limited in their use to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per year

during power operation.

On November 20, 1980, the inspector requested the licensee to

show how the 90 hour0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months limitation was being implemented since the

station procedure controlling containment purge and vent operations,

No. 2.2.70, Revision 14, dated August 13, 1980, " Primary Con-

tainment Atmospheric System", did not include methods to imple-

ment this limitation.

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It was determined on November 21, 1980 that administrative controls

had not been implemented to limit the opening of purge and vent

valves (other than those used for drywell-tort

d/p control) to

90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per year during power operation and also that there was

no mechanism for keeping track of the running total of length of

time opened.

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The licensee determined from a review of the station operations

log that since plant startup after the January-May,1980 refuel-

ing outage that the 20" purge valves had been opened for approxi-

mately 145 hours0.00168 days 0.0403 hours 2.397487e-4 weeks 5.51725e-5 months during power operation (greater than about 1%

power).

The licensee stated that actions would be taken in the following

areas:

administrative controls would be implemented to not start

--

deinerting until suberitical and to minimize the time spent

inerting on stari.up during power operation.

i

implement a method for keeping track of total opening times

--

of the 20" purge valves.

initiate a request for modification to assirt in these con-

--

trols (possibly using meters to keep track of opening times,

and separate controls / interlocks on the 20" valve control

switches).

re-evaluate the 90 hour0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months / year limitation and determine what

--

actions should be taken, if any, if the limit is exceeded.

The inspector acknowledged the licensee's statement and stated

that the failure to limit the opening of the 20" containment

purge valves was considered a deviation from the licensee's

August 21, 1979 and December 19, 1979 commitments (50-293/80-30-

02).

8.

Interim Criteria for Shift Staffing

l

The inspector reviewed the licensee's actions in response to a letter from

NRR to all operating plants dated July 31, 1980 which specified criteria

for shift staffing and li;nitations on working hours.

.

The inspector reviewed the licensee's October 15, 1980 response to this

issue which stated that the interim criteria for shift staffing would be

!

I

implemented no later than July 1, 1981 but that meeting all the limitations

,

on working hours and overtime would place an excessive burden on the oper-

l

ating organization.

The licensee did state, however, that a member of the

l

'

.

-

14

required shift staff would be limited to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months straight while performing

safety related control room duties.

On November 24, 1980, the licensee management identified that a control

room operator had been continuously on watch for an excessive length of

time (32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months ). On watch supervisors were unable to acquire a relief by

calling off watch operators at home.

. On November 26, 1980, the inspector attended an Onsite Review Committee

(ORC) meeting which included a review of this event.

The licensee stated

that subsequent to October 15, 1980, several operators were transferred

from the operations department to other positiens, and that further nego-

tiations with the local bargaining unit were needed because of the sensitivity

of replacing union watch standers with supervisory personnel.

The inspector questioned the licensee management concerning the actions

that would be taken to implement the commitment to limit working hours. The

licensee stated that instructions would be placed in the Watch Engineer's

(W.E.) Instruction Log to inform the W.E. that licensed operators would be

limited to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months consecutive control room duty and that upper management

approval would be requi ed to exceed this limit. The licensee further

stated that station po cy limiting working hours would be included in a

.

station administrative document by January,1981.

The inspector acknowledged the licensee's statements and stated that the

failure to limit length of time that the control room operator was on watch

was considered a deviation from the licensee's October 15, 1980 commitment

(50-293/80-30-03).

9.

Seismic Monitoring Instrumentation

Although the Technical Specifications do not place any requirements on

seismic instrumentation, the inspector questioned the licensee on November

14, 1980, about concerns expressed by a vendor representative (Kinemetrics,

Inc.) involving the operation of recently installed seismic monitoring and

support equipment.

The concerns focused on the status of the antenna which

picks up a time signal from a remote national station and transmits this

time signal to a WWV receiver mounted in the seismic intrumentation panel

in the rear of the control room.

The inspector reviewed Plant Design Change Request No. 78-24-1, Seismic

Monitoring Equipment, and observed selected portions of the installed

instrumentation. This modification included the installation of three up-

graded SMA3 seismic accelerometers in the reactor building, tape recorder

units, a WWV radio receiver, and playback units in the rear control panel,

and an outside antenna for the WWV receiver.

Tbi WWV receiver provides a

time signal on an additional tape recorder unit to be able to automatically

record event time.

.

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15

The licensee stated that because several problems had been identified with

the operation of the WWV receiver and the antenna, that the receiver had

been previously shipped to the vendor for repairs and that a new antenna

had been authorized by BECo Engineering Department.

The licensee further

stated, that at the time of the vendor representative's site visit, that a

misunderstanding resulted concerning which antenna was to be used and that

the entire modification had not yet been verified and signed off as complete.

The inspector observed satisfactory operation of the WWV receiver on November

19, 1980 and had no further questions.

No items of noncompliance were identified.

10.

Susceptibility of Contaiment Flooding

The inspector was requested to review the design of Pilgrim's containment

with respect to susceptibility of flooding because of a problem at another

reactor site.

The inspector reviewed system drawings, observed indications available at

the main control room panel, and radwaste control room panel, and inter-

viewed licensee personnel. The following information includes details

about the system design:

There is a high level alarm for each of the drywell equipment and

--

floor drain sumps at the radwaste control panel.

D ywell equipment and floor drain containment isolation valves are

--

cuntrolled from, and position indication are displayed in the main

control room.

--

Sump pump running indication is provided at the radwaste control

panel.

A sump pump integrator for both of the sumps is located at the radwaste

--

control panel.

Drywell components are cooled by Reactor Building Closed Cooling Water

--

(RBCCW), a closed system, with instrumentation available to determine

inventory changes.

The main control room has a single annunciator which alarms if a local

--

alarm at the radwaste control panel is not reset /is not acknowledged.

This information was forwarded separately to NRC:IE for review.

No items of noncompliance were identified during this review.

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4

11. Unresolved Items

Areas for which more information is required to determine acceptability are

considered unresolved.

Unresolved items are discussed in Paragraphs 4.c

and 7.b.

12.

Exit Interview

At periodic intervals during the course of the inspection, meetings were

held with senior facility management to discuss the inspection scope and

findings.

l

.

ML20009A094

Text

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