ML20154R821

Insp Repts 50-259/88-18,50-260/88-18 & 50-296/88-18 on 880612-30.Violations Noted.Major Areas Inspected:Operational Safety,Maint Observation,Independent Audits & Restart Test Program

ML20154R821
Person / Time
Site:	Browns Ferry
Issue date:	09/15/1988
From:	Brooks C, Little W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20154R808	List: ML20154R805 ML20154R810 ML20154R821
References
50-259-88-18, 50-260-88-18, 50-296-88-18, NUDOCS 8810040366
Download: ML20154R821 (22)

See also: IR 05000259/1988018

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UNITED STATES

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. NUCLEAR REGULATORY COMMISSION

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REGION 11

\\'e,,,,g

101 MARIETTA ST N.W.

ATLANTA, GEORGIA 30323

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Report Nos.:

50-259/88-18, 50-260/88-18,-and 50-296/88-18

Licensee:

Tennessee Valley Authority

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6N 38A Lookout Place

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1101 Market Street

Chattanooga, TN 37402 2801

Docket Nos.:

50-259, 50-260 and 50-296

License 741.:

OPR-33, DPR-52,

and DPR-68

Facility Name:

Browns Ferry 1, 2, and 3

Inspection Conducte :

June 12-30,

988

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Inspector-

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g-- R. Brooks,~ Acting Senior Resident Inspector

Date 51gned

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Accompanying Personnel:

E. F. Christnot, Resident Inspector

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W. C. Bearden, Resident Inspector

A. H. Johnson, Project Engineer

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Approved by:7. 5. Littler Sectioll Chief

/ #88

Date Signed

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Inspection Programs

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TVAProjectsOfvision

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SUMMARY

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Scope:

This routine inspection was conducted in the areas of operational

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safety, maintenance observation, independent audits, and restart test

program,

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Results: One violation (260/88-18-02) was identified for failure to document

an inadvertent overload condition on the "0" Diesel Generator on a

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Condition Adverse to Quality Report (CAQR) (see paragraph 5.c.)

(Restartitem)

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The following inspection followup items were identified:

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IFI 260/88-18-01,

Undocumented caole conductor splice.

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(Restartitem)

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IFI 260/88-18 03, Test deficiencies identified during LOP /LOCA

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Test "A".

(Restartitem)

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IFI 260/8818-04, Occumentation of offsite voltage during

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LOP /LOCA Test "8".

(Restartitem)

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GG10040366 880922

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IFI 259,260/88-18-05, Major discrepancies identified during

LOP /LOCA Test "C".

(Restart itemi

IFI 260/88-18-06, Oeficiencies identified during LOP /LOCA Test

"0".

(Restart item)

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REPORT DETAILS

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1.

Persons Contacted

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Licensee Employees

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• J. G. Walker, Plant Manager

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J. D. Martin, Assistant to the Plant Manager

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• k. M. McKeon, Operations Superintendent

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T. F. Ziegler, Superintendent - Maintenance

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D. C. Miss, Superintendent - Technical Services

J. G. Turner, Manager - Site Quality Assurance

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M. J. Nay, Manager - Site Licensing

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• J. A. Savage

A.W.Sorrell,ComplianceSupervisor

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Site Radiological Control Superintendent

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R. M. Tuttle, Site Security Manager

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L. E. Retzer, Fire Protection Supervisor

H. J. Kuhnert, Office of Nuclear Power, Site Representative

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T. C. Valenzano, Director - Restart Operations Center

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Other licensee employees contacted included licensed reactor operators,

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auxiliary operators, craftsmen, technicians, public safety officers,

quality assurance personnsi, and design and engineering personnel.

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Acronyms and abbreviations used throughout the report are listed in the

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last section of this report.

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• Attended exit interview

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2.

Operational Safety (71707, 71710)

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The inspectors were kept informed of the overall plant status and any

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significant safety matters related to plant operations.

Daily discussions

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were held with plant management and various members of the plant operating

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staff.

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The inspectors made routine visits to the control rooms. Ooservations

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included instrument readings, setpoints and recordings; status of

operating systems; status and alignments of emergency standby systems;

onsite and offsite emergency power sources available for automatic

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operation; purpose of temporary tags on equipment controls and switches;

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annunciator alarm status; adherence to procedures; adherence to limiting

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conditions for operations * nuclear instruments operable; temporary

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alterations in effect; daily journals and logs; stack monitor recorder

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traces; and control room manning.

This inspection activity also included

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numerous informal discussions with reactor operators and their

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supervisors.

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General plant tours were conducted on at ':.w a weekly basis.

Portions

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of the turbine building, each reactor bui;

ag and outside areas were

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visited.

Observations included valve positions and system alfgnment;

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snubber and hanger conditions; containment isolation aligr.ments;

instrument readings; housekeeping; proper power supply and breaker

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alignments: radiation area controls; tag controls on equipment; work

activities in progress; and radiation protection controls.

Informal

discussions were held with selected plant personnel in their functional

areas during these tours.

Durir.g a routine tour on June 27, 1988, the inspector observed a

compensatory fira watch posted in the

"B'

Shutdown Board (SDB) room who

appearoa to be inattentive to his duties.

The inspector initiated a

discussion with the firewatch in order to ensure that he was alert and

then notified a licensee representative who initiated appropriate

disciplinary action.

During a routine tour on June 29, 1988, the inspector observed an

unapproved flammable liquid container in a storage cage on the Refuel

F!oor.

Upon interviewing Refuel Floor personnel, the inspector learned

that a glass reageat-type bottle was being used te transport acetone from

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the Chemistry Lab to the Refuel Fione for use in cleaning fuel channels

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and other fuel related components.

The empty bottle was stored for 7uture

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use.

The use of the bottle is contrary te the requirements of Staadard

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Practice 14.36, Hazardous Chemicals / Flammable or Combustible iiquids,

which cal's for VL listed cans.

Tne inspector contacted a site fire

protection engineer who removed the bottle and instructed refuel floor

and chemistry lab personnel on the use of proper containers.

Since the

boi.tle was not being used, it was immediately removed, and personnel were

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reinstructed, no violation or deviation will be issued.

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3.

Maintenance Observation (62703)

Plant maintenance activities on selected safety-related systems and

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components were observed / reviewed to afcertain that they were conducted in

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accordance with requirements.

The following items were corsidered during

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this review:

the Technical Specification limiting conditions for

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operations were met; activities were accomplisheo using approved

procedures; functional testing and/or calibrations were performed prior to

returning components or systems to service; quality control records were

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maintained; activities were accomplished by qualified personnel; parts and

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materials used were pro]erly certified; power tagout clearance procedures

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were adhered to; and raciological controls were implemented as required.

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Maintenance Requests (MR) were reviewed to determine the status of

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outstanding jobs and to assure that priority was assigned to safety-

related eouipment maintenance which might affect plant safety.

The

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inspectors observed the below listed maintenance activities during this

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reporting period:

a.

Mechanical Maintenance Instruction (MMI) 190, Control Red Drive

Unlatching.

b.

MMI 7, Control Rod Drive Changeout.

c.

Operating Instruction (01)-85, Control Rod Insert and Withdraw

Timing, portions performed as post-maintenance testing.

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d.

Diesel Generator (DG) "0" troubleshooting and corrective maintenance

activities related to the inadvertent overload on June 21, 1988

(refer to paragraph 5.c, of this report for details).

e.

MR Bf1074 dated May 10, 1988, was indicatad as being outstanding

durine a tour conducted on June 27, 1988.

This MR documented a need

for c'eaning the Units 1/2 "B" DG batteries.

The inspector observed

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a heavy layer of dust and white particulate material on top of the

batteries.

In addition, the inspector noted that on four of the

battery cells, the spark arrestor vent covers were missing and the

particulate material had entered into the battery electrolyte through

the openings.

The inspector informed a licensee representative of

the deficient condition and also pointed out to the plant maintenance

supervisor that the more than seven week turnaround on completing a

routine "request for cleaning"ll quantity of the material in the

MR was excessive.

The batteries were

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immediately cleaned and a sma

electrolyte was evaluated as having no impact on battery performance.

Within this area no violations or deviations were found.

4.

Independent Audits (40704)

a.

Independent Safety Engineering Group (ISEG)

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The licensee implemented the ISEG in early 1988 as discussed in the

BFNP Nuclear Performance Plan.

Tne inspector reviewed the last three

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monthly reports and found them to be well focussed on current

industry problems as well as specific site weaknesses.

The reports

were professional and thorough.

The ISEG review topics included:

(1l NRCthe Licensee Event Report (LER) program; (2) Control Air System;

(3,

Information Notices; and (4) Plant Operations Review

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Committee (PORC) review of plant modifications.

With continued

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diligence, the ISEG should satisfactorily fulfill its role in

providing appraisals of the quality and safety of operation to the

licensee s upper management.

b.

American Nuclear Insurers Nuclear Liability Insurance Inspection

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A nuclear liability insurance inspection was conducted by the

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American Nuclear Insurers during February 1988,

The inspection

report and ifcensee's response to the report were made available to

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the resident inspectors.

The inspection focused on Onsite Safety

Review, Operations, Radwaste Management, Health Physics, and Plant

Water Chemistry.

Three recommendations were made and two suggestions

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were offered.

The inspection findings did not point to any

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programmatic weaknesses not already known to the NRC.

The resident

had attended the exit and considered the most significant weakness

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identified to be in the area of root cause analysis and recurrence

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control.

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c.

Safety System Functional Inspecdon (SSFI)

The licensee's QA organization coordinated an SSFI performed pri-

marily by an outside contractor during the period of June 6 - July 1,

1988.

The inspection was patterned after the NRC's SSFI as described

in NRC Inspection Manual Chapter 2515.

The team identified 24

concerns documented as "observations".

Fourteen of the observations

were judged by the team to be safety significant or potentially

significant depending upon the outcome of further review and analysis

to be performed by the licensee.

The team reviewed the RHRSW and

EECW systems.

The majority of the observations were found to have

been previously identified by the licensee or the NRC and in some

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formal tracking program for closecut.

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One deficiency was found to impact system operability under the

existing plant conditions.

Due to recurring failures of valves 67-50

and 67-51, the potential for failure of these valves was judged to be

significant.

These are cross-connect valves between the safetyld

related EECW system and the non-safety-related RCW system.

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these valves fsil to close as required under certain circumstances,

insufficient cooling water to the diesel engine coolers could result.

The licensee suspended fuel handling operations on June 30, 1988,

until the valves were tagged closed on a clearance hold order to

eliminate this concern.

The SSFI team made several conclusions from the characterization of

the findings.

The maintenance area was judged to be the weakest of

those evaluated.

Weaknesses were also noted in design calculation

verification and licensee review of calculations performed by

contractors.

The team further detected a lack of sensitivity among

plant personnel as far as respect for the operational status of the

system.

This may be attributed to the extensive shutdown during

which the system status was not maintained and the mind-set that

accompanies such a shutdown.

The final SSFI report will be reviewed and TVA's response to the

findings will be monitored.

No violations or deviations were identified.

5.

Costart Test Program (RTP)

The inspector attended RTP status meetings; reviewed RTP test procedures,

test specifications, and baseline test requirement documents; observed

RTP tests and associated tests performances; and reviewed RTP test results

including test exceptions.

The inspector also attended selected Restart

Operations Center (War Room) and JTG meetings.

a.

Summary of RTP Activities

(1) RTP-03A, Reactor Feedvater

This test was approved by the JTG for performance on June 16,

1988.

It consisted of verifying various signals, including

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reactor level and pressure signals to the reactor protective

system and high reactor level trip signals to the HPCI and

RCICinjectionsystemturbines.

The test was completed with the

exception of section 5.16 which, in part, requires verifying

high reactor pressure trip of the recirculation pumps.

(2) RTP-23, RHRSW

Outages continued throughout most of this reporting) period with

the majority of effort aimed at completing the rep acoment of

Dresser Couplings and completing "the hy"drostatic test and

restoration of the "Al",

"A2", "81, "B2 , "C1", "C2", "D1",

and "D2" pumps.

Flood level switch replacement continued to be

a problem, in that a dedicated switch has not been found.

(3) RTP-038, RXFW

This test was released for performance by the JTG on June 30,

1988.

It deals mainly with the RXFW pumas and their trips

including low main condensor vacuum and ligh reactor water

level.

Monitoring of this test will be covered in future NRC

reports.

(4) RTP-024 RCW

This test was completed during this reporting period with 47

total tests exceptions identified.

Of these exceptions, seven

had hardware significance.

MRs were written to correct six of

them however, TE-046, which deals with check valve 1-24-852,

will; require DNE resolution.

(5) RTP-030, Diesel Generator and Reactor Building Ventilation

This test continues to be delayed by parts requirements,

especially flow switches and damper motors.

Approximately 24

test exceptions have been identified with the majority involving

the failure of DG fans to stop or start as rewas issued as a result of thi

CAQR

88-0524

Control Building Heating, Ventilation, and Air

(6)

RTP-31A, ing

Condition

Section 5.1 of this test was released by the JTG for performance

in order to support the LOP /LOCA series of tests.

This section

mainly involved the CREVS.

Additional tests will start af ter

the completion of the LOP /LOCA tests.

No specific test excep-

tions have been identified.

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(7) RTP-318, Control Building HVAC

Before this test can start, DNE must issue revision number

three (Rev. 3 to BTRD-014.

As of the end of June 1988, this

revision was ssued and the RTG group was in the final stages of

writing and approving RTP-0318.

(8) RTP-39, C02 Storage and Fire Protection

This test was completed during this reporting period with no

specific test exceptions identified.

The test consisted of

smoke detector activation, CO2 logic testing in DG rooms, and

loss of control power tests.

(9) RTP-057-1, 120 Volt DC Battery

This test involves the 8 DGs and covers a total of 16 battery

chargers and battery racks.

Prior to each test on the battery

chargers, the filter capacitors must be changed.

As of the end

of June 1988, ONE was revising the BTRD and the RTP grcup was

changing the system test specification and the RTP procedure.

No test exceptions were identified.

(10) RTP-57-2, 120 V AC Distribution (120V DIST)

This test win virtually completed during this reporting period.

One outstancing test exception was written involving)section 5.2

dealing with the performance of Test Instruction (TI 738, which

could not be performed because of the fire damage in the Unit 2

Orywell.

(11) RTP-57-4, 480 Volt Distribution (480V OIST)

This test was virtually completed; however, a total of 53 test

exceptions were identified.

Of these test exceptions, ten had

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hardware related issues.

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(12) RTP-57-5, 4160 Volt Distribution (4 KV DISTRIBUTION)

This test is ongoing and has identified approximately 20 test

exceptions; however, only two have hardware significance.

One

of the ma

setpoints and

tolerances,jor restraints deals with rela

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which were to be supplied b

ONE through a

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contractor.

The six special tests involv ng the Unit 1/2 OG

voltage regulators and speed governors have all been completed

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and are awaiting approval by DNE.

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(13) RTP-067, EECW

This system, as with aystem 23 (RHRSW), was also involved with

the replacement of Dresser couplings and the system hydro's.

The testing in progress involved mainly section 5.8 which deals

with various chillers such as the control room emergency

chillers and the SDB room chillers.

One hardware test exception

was identified involving temperature control valve 3-TCV-67-83.

(14) RTP-069, RWCU

This test was released by the JTG on June 17, 1987, and the

majority of effort so far has been directed at completing

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section 2.0, Prerequisites and section 5.1, Valve Stroke Tests.

NosignificanttestexceptIonshavebeenidentified.

(15) RTP-070, RBCCW

This test was partially released as reported in NRC report

88-10.

Since it started, approximately 27 test exceptions

have been identified with 8 having hardware related issues.

(16) RTP-073, HPCI

The JTG approved this test; however, it has not released the

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test for performance.

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(17) RTP-074, RHR

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This test is being im) acted in Section 5.5 which requires

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testing of the drywell

spray upper and lower header, by the

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hydrolazing of the lower header.

This activity was set to

start on or about July 5, 1988.

No significant hardware test

exceptions have been identified.

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(18) RTP-082, Diesel Generators

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This system is undergoing a series of special tests involving

all eight OGs.

These are being performed by the system

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engineers working with vendor renresentatives and onsite DNE

personnel.

A total of ten special tests are involved and will

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result in the DG's speed governors (hydraulic actuators) and

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voltage regulators being calibrated as well as optimized for

plant operations.

This system test has identified over 50 test

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exceptions; however, only three have significant hardware

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issees.

See Section 5.c. of this report for additional

information on DG testing.

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(19) RTP-084, Containment Atmospheric Dilution

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This test was released by the JTG for performance on May 25,

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1988.

The majority of effort has been directed at completing

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the prerequisites.

Section 5.1 deals with verifying the logic

on a series of valves and replacing Class IE splices with

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qualified Raychem splices,

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(20) RTP-085, Control Rod Drive

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This test was delayed by the changeout of 82 rod drive

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mechanisms and the repairs to several rod position indicators.

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Section 5.2, Rod Position Indicating System Function Test}fied.

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started and no significant test exceptions have been ident

(21) RTP-092, Neutron Monitoring

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This test involves the intermediate range monitors (IRMs),

source range monitors (SRMs), and the average power range

monitors (APRMs).

The test was released for performance by the

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JTG on May 23, 1988; however, a switch involved in the inop

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bypass for the SRMs

IRMs, and APRMs were not easily obtained

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and contributed to delays.

No significant hardware test excep-

tions were identified.

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(22) RTP-99, Reactor Protective System (RPS)

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This test was released by the JTG for performance on June 17,

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1988

and involves all four channels of the RPS.

One

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signIficant hardware test exception was identified involving

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Section 5.27, which requires that the reactor mode switch

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contacts be verified, and concerns wires in the panel not being

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(23)

RTP-8' C, Backup Control Test (BUC)

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This test procedure is over 800 pages long and includes the 4KV

SD8s, 480 V SD8s, DG auxiliary boards, 480V reactor motor

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operated valve (MOV) boards, and the 250 V reactor MOV boards

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(also any electrical system that has backup controls).

The test

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was released for performance by the JTG on June 13, 1988, and

has been performed on a continuous basis.

No significant test

exceptions were identified.

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The above tests were in progress and in various stages of completion

at the end of this reporting period.

The inspector will continue to

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observe these activities on a day-to-day basis.

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b.

RTP-075, Core Spray

During the performance of RTP-075, Core Spray, the RTP group was

using SI-4.2.8-39A(II) to meet the requirements of step 5.6.5 of the

RTP test.

This was to prove that the CS system Loop II would perform

automatically when called u

However, during the

performance of the SI, the 'pon to do so.20" CS pump automatic circuit break

failed to close when required.

MR A-860183 was written to trouble-

shoot the problem and the RTP personnel wrote TE #9 to document the

failure.

During troubleshooting, maintenace personnel discovered an

undocumented splice in the white lead of cable ES-2554-II, which was

in a cable tray surrounded by flamemastic and located in the Unit 2

reactor building.

The maintenance personnel wrote CAQR BFP-88-0375

to document this discovery.

This item is identified as inspector

followup item (IFI) 260/88-18-01 Undocumented Cable Conductor Splice

in Core Spray Logic Wiring.

Thlsitemisrequiredtobeclosedout

prior to restart.

The disposition of this CAQR will be reviewed in

a future inspection.

One significant test exception involving the stroke time for four

flow control valves was identified.

c.

RTP-082 Diesel Generators

OG "0" was inadvertently (ST) 88-09, Diesel Gene,rator Governor and

overloaded on June 21

1988, during the

conduct of Special Test

Voltage Regulator Calibration.

For a period of about 30 seconds,

control room indications of KW, KVAR, and AMPS were off scale high.

The incident was caused by unintentional shorting of test leads by an

electrician who was directed to take confirmatory measurements of a

parameter beint monitored on a recorder.

The shorting caused fuses

to blow result < ng in a loss of the voltage regulator and governor.

Since the DG was in parallel with the grid during the test, it

immediately accepted this additional load.

An apparently pre-

existing and undetected failure of an SBN switch (cell switch) in the

DG breaker compartment prevented an automatic clearing of the

overload

thus increasing the potential damage to the DG.

The event

wastermlnatedwhenthecontrolroomoperatortrippedtheDGbreaker.

Following the event, the electrical integrity of the DG was confirmed

by the satisfactory performance of a high potential test at 10,000

VOC for one minute.

Visual inspections were conducted of the

insulation system by an experienced vendor representative.

Some

hairline cracks in the insulation were detected which were determined

to have been recently created.

These cracks were indicative of coil

movement relative to the fixed components and could have resulted

from either the overload or some of the more recent stressful restart

testing.

The vendor representative recommended followup high poten-

tial testing at 6,12 and 24 months in order to detect any further

degradation of the insulation system as a result of the overload.

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mechanical inspections were performed on the diesel engine.

The

licensee relied upon the expertise of the vendor representative who

waspresentattheDGduringtheoverloadtomakethisjudgment.

The licensee prepared a critique on the event which concluded that

the capability of the DG to perform its intended function had nnt

been affected.

Following a review by the PORC on June 25, 1988, the

DG was declared fully operable.

The inspector .eviewed the event and asked for additional information

on:

1) the cause of failure for the SBM cell switch and whether

the other DG cell switches had bi:en tested 2) an approximation of

the magnitude o/ the overload; 3) a comparison of diesel operating

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parameters such as vibration, exhaust /c

temperature, oil pressures, oil analysis, ylinder/ cooling water

and oil and fuel consump-

tion before the event with those af ter the event; and 4) other

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diagnostics and/or inspections contained in the vendor technical

manuals.

During the course of the review, the inspector learned that the

incident was not documented, as required, on a CAQR.

This may have

been a contributing cause to the fragmented approach used to evaluate

and document the condition of the diesel.

Part I, Section 2.16, Step

2.1.1.F of the Nuclear Quality Assurance Manual requires that a

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condition adverse to quality (CAQ) be documented on a CAQR for items

which have been subjected to conditions for which they have not

been designed such as overaressure overvoltage, overheating, over-

stressing, or environmenta'

conditions hazardous to their function.

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Failure to initiate a CAQR for the overloaded DG is considered

a violation of 10 CFR 50, Appendix B, Criterion V (259,260,296/

88-18-02).

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d.

Loss of Power / Loss of Coolant Accident (LOP /LOCA) Testing

I

The licensee's RTP group and the system engineering group, with

l

support from th viher major Browns Ferry site organizations,

'

performed a series of four tests referred to as LOP /LOCA testing,

which started on May 29, 1988, with LOP /LOCA Test "A" and ended at

>

approximately midnight on June 8,1988, with LOP /LOCA Test "0".

)

Numerous pretest meetings were bid such as LOP /LOCA thrice weekly

punchlist, daily RTP status, afG as needed, and a daily morning

J

meeting with the major site organizations to update the LOP /LOCA

,

testing status.

A total of nine inspectors, five region based and

4

four residents, observed the performance of the four tests, with

LOP /LOCA Test ',C" receiving the most coverag/L A

1

e.

The procedures used

1

for the tests were numbered:

2-BFN-RTP-L

(RTP-L/L A),

j

2-BFN-RTP-L/L B (RTP-L/L B), 2-BFN-RTP-L/L C (RTP-L/L C), and

l

2-BFN-RTP-L/L 0 (RTP-L/L 0).

Revision 0 of the procedures were

i

reviewed by the JTG and approved by the Plant Manager shortly before

the performance of the actual tests.

Last minute changes to the

4

_ _ _ _ _ _ _ _ _ _ _ _

_ _ _ _ _ _ _ _ ___ _

.____ ___

___

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

_

- _ _ _ _ _ _ _ .

.

.

,

11

tests were reviewed and approved the day the tests were conducted and

RTP-L/L C and RTP-L/L 0 were revised approximately two days before

the actual tests.

Each test procedure was reviewed and found to be adequate.

(1) LOP /LOCA Test "A"

This test was basically a LOP test, sometimes referred to as a

"blackout" test, and was Jerformed on May 29, 1988.

Section S.3

of the test called for tie simultaneous tripping of the three

main power feeds coming into the plant from the switchyard.

The

inspectors used NRR inspection modules 703068 Loss of Offsite

Power, and 704418

Emergency / Standby Power Supply System Test,

as guidance before, dur'ng and after the test.

Test section

5.4,

Plant Performance, required that all eight OGs start and

!

tie onto their respective SD8s; that various 4-KV and 480 V load

alignments be in accordance with the applicable procedure steps;

that all three units receive a scram signal from loss of RPS

Hotor-Generator (M/G) sets; and that reactor MOV electrical

boards and ventilation boards be in the alignment specified by

a)plicable procedure steps.

The inspectors made the following

osservations:

,

(a) Unit 3 DG "30" did not appear to close onto its SDB (3EC).

r

However, post test review of the visicorder traces indi-

'

cated that the breaker closed and stayed closed for only

l

five cycles, approximately'one-tenth of a second, and then

tripped off.

The licensee s representatives informed the

inspector that an item of M&TE used to time the closure of

the breaker shorted out, )icked up the DC trip coil of the

'

brsaker, and caused the

)reaker to trip shortly af ter

closure.

!

(b) Unit 2 RPS M/G set "2B" did not trip off upon loss of

i

aower (the breaker supalies the power to the motor).

"

iowever, other scram signals from the RPS did initiate

l

a full scram for Unit 2.

(c) Several individual items such as recirculating M/G set

oil pumps "2A-1", "2A-2", "28-1", and "28-2" and control

'

and service air compressor "0" were exceptions to the test

as far as their supply breakers tripping.

j

l

The above items are tracked as IFI 260/88-18-03, Test Deficien-

'

cies Identified During LOP /LOCA Test "A".

This item is

!

required to be closed out prior to restart.

The conduct of the

test by the licensee's representatives was not a well run test

in that the control room activities appeared at times to be very

hectic and a large number of personnel, more than what seemed

~

necessary, were in the control room during the test.

Improve-

ment in professionalism by all personnel and better crowd

-

control was needed,

t

I

L-

. _ _ . _ _ _ .

. _ _ _ _ - _ _ - -

__ . _ _ . .

-

__

--

-

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

.

,

12

(2) LOP /LOCA Test "B"

This test was basically a LOCA test using offsite power (the

grid) as the power source and was performed on June 1,1988.

Section 5 of the test called for the application of a high

drywell pressure signal Of approximately 3.0 psig using

temporary test tubing and a sneumatic calibrator installed at

'

transmitters PT 64-588 and Pr 64-580.

This signal in conjunc-

tion with the already at:nospheric pressure of t1e reactor vessel

initiated a LOCA signal.

The basic requirements of the test

were that all eight DGs start (they were not to close onto their

respective 50Bs); all four Unit 2 CS aumps start in the proper

sequencing and at the proper times; a'l four RHR pumps start in

the low pressure core injection mode in the proper sequencing

and at the proper times; and all four RHRSW pumps used for

EECW start in the proper sequencing and at the proper times.

Additional auxiliary equipment such as reactor recirculation

discharge valve closure, RHR pump cooler fans start, HPCI logic

trips, etc., were verified as required to meet test acceptance.

The inspectors made the following observations:

(a) An inspector was stationed at the Unit 3 control room

0G electrical panel to observe Unit 3 OG operation upon

initiation of the LOCA event.

Test procedure recuirements

were that upon initiation of the LOCA event, Unit 3 DGs

("3A", "3B", "3C", and "30") start automatically, attain

rated voltage and frequency, and remain unconnected to

their respective 4-XV SDBs, and the SDB mode switches

transfer from "auto" to "manual". Observations of annun-

ciator and alarm panel indicator lights confirmed the

aforementioned test requirements for the Unit 3 DGs.

(b) An NRC inspector was stationed at the Unit 2 control room

station and proceeded to the SDBs upon initiation of the

LOCA events. The inspector observed TVA data takers and QA

inspectors as well.

The inspector observed no deviations

or violations of procedure and test requirements.

(c) An NRC inspector observed the control room activities

inside the Unit 2 main control panel "horsehoe."

The

inspector used Section 5.2 of the test to verify and

document the performance of the four CS pumps, ion

the

l

four RHR pumps in the low pressure core inject

mode, and the four RHRSW pumps used for EECW.

Verifi-

cation of operation involved recording motor amperage,

system flows, and system pressures.

All 12 pumps

indicated electrical current flow, with the CS trains

indicating approximately 6200 gallons per minute (gpm)

and approximately 250 psig each, and the RHR trains

indicating approximately 12000 gpm and approximately

275 psig each.

Overall evaluation was that the major

items of equipment responded as required by the test.

L-

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

_

_

_ _ _ _ _ _ _ _ _ _ _ _

__

____ ____ _ ___ ___- .

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

. .

13

Upon post test review by all the inspectors involved, it was

noted that the facilities test personnel appeared somewhat

confused by what was occurring at different stages of the test.

The control room activities appeared to be more controlled thJn

during LOP /LOCA "A".

It was noted that no direct monitoring of

the incoming voltage from the grid was made by the licensee's

representatives during this test.

The inspector was informed

that this will be documented by calculation.

This item is

identified as IFI 260/88-18-04 Documentation of Offsite Voltage

During LOP /LOCA Test "B".

This item is required to be closed

out prior to restart.

The results of the test will be reviewed

and commented on in a future NRC report.

(3) LOP /LOCA Test "C"

This test was a combination of a LOP signal followed six seconds

later by a LOCA signal with the "B" DG disabled and was per-

formed on Junt 5,1988.

Section 5.3 of the test called for

simultaneously opening the three main power feeds coming into

the plant from the switchyard and after a3 proximately a six

second time delay initiating the triple 1ow (Lo-Lo-Lo) or

Level I reactor vessel level trip; thereby setting up a LOP

followed by a LOCA.

The ins 3ectors used NRR modules 70316, Loss

of Offsite Power Test 70436, Engineered Safety Features Actua-

tion System Test, and 70441, Emergency / Standby Power System

Test

as

Test

sectIon 5. guidance before, during and af ter the test.

4, Plant Performance, re

with the exception of Unit 3 DG "B"quired that all eight DGs

start and tie onto their

respectiveSDBs;thatthe"2A"and"2C"CSpumpsstart;thatthe

"2A', "28", and "2C" RHR pumps start; that the applicable RHRSW

pumps start; and that all three units receive a scram signal.

Additional auxiliary equipment as outlined in the procedure were

to activate, including Standby Gas Treatment and Control Room

Emergency Ventilation as required to meet test acceptance

criteria.

The inspectors made the following observations:

(a) Unit 2 Control Room Observations

After the test was initiated at 2:16 p.m., the inspector

observed that a scram sig"nal was received.

RHR pumps "2A",

"2B",

"2C" and CS pumps

2A" and "20" started as planned.

Power was lost to the "2A" reactor MOV board which did not

transfer as expected.

The transfer was manually performed

,

at 2:29 p.m., and the CS inboard injection valve (2-FCV-

75-25) opened (procedure step 5.4.9.1).

,

Prior to the test, RCIC was inop"erable and procedure step

5.4.12 could not be met.

The

RCIC Relay "Logic Power

Failure" annunciator was observed.

The "2B

reactor MOV

l

L

_ _ _ _ _ _ _ _ _ _

_ _ _ - _ _ _

.

,.

14

board is supplied by 508 "0"

and DG "0" was not tested.

Accordingly, procedure steps 5.4.9.2 and 5.4.19.2 could

not be checked due to an absence of power.

RHR System II flow and pressure could not be checked

because I&C Bus "B" had no power (procedure step 5.4.24).

At 3:02 p.m. , 4160 V SDB "D' was restored and these were

checked.

The alarm and annunciation table, Appendix E, had the

following alarms received prior to the test:

Unit Preferred Supply Abnormal

Reactor Auto Scram Channel A

Reactor Auto Scram Channel B

The "Turbine Tripped Electrical Trouble" annunciator

was not received.

(b) During LOP /LOCA Test

"C" an inspector observed Unit 3

control room activities with regard to alarm and annuncia-

tor

panels and noted that the Unit 3 DG started and closed

on the 4160 V SOBS.

In addition, selected circuit breakers

on the Unit 3 480V SDBs "3A" and "3B" were checked to

verify and obtain their LOP /LOCA positions.

The following

problems were identified during the test:

DGs "3A", "3C" and "3D" came up to rated speed and

-

voltage but did not tie-on to their respective 4160 V

SDB.

Troubleshooting and review of log'ic circuitry

indicated that the DG "3A",

"3C" and

3D" output

breakers closed at ~6.5 seconds and at ~6.75 seconds a

LOCA signal opened the 3 DG breakers.

During recharg-

ing of the breakers, the anti pump relay (52Y) had

sealed in steventing the reclosure of the DG breakers

(i.e. , DG areaker opened -0.3 seconds and the anti-

pump relay responded as designed to a fault condition

on the 508 although the LOCA signal was the reason for

the DG breakers being opened).

An operator was sent

to the switchgear room where a remote panel is avail-

able for local operations.

The operator slowly placed

the normal / emergency switch (43) to the emergency

position.

The switch is a break before make design

and thus de energized the 52Y anti-pump relay setting

up the circuitry for closure of the three DG breakers.

The inspector also identified that the data recorder

-

did not have the latest data sheets for recording

breaker positions on 480 V SDBs "3A" and "3B".

l

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

--

_

..

,.

15

(c) An NRC inspector was stationed in the Unit I control room

and proceeded to the SD8s upon initiation of the LOP /LOCA

event. The inspector observed TVA data takers and QA

inspectors as well.

(d) An NRC inspector observed the initiation of the LOCA signal

in the Unit 2 auxiliary instrument room and noted that

relay 16A-K29 in panel 9-42 did not de-energize as required

for the test.

This rela

tainment isolation logic. y is part of the primary con-

The inspector then proceeded to SOB "0" and there observed

a Ifeensee representative performing manual manipulations

on 480V reactor MOV board

'2A".

It was later determined

that this board was not initially lined up per section 5.2,

Initial Lineup of the Test.

These two items were con-

sideredtobemajorconcernsforthetest.

(e) An NRC inspector was positioned in front of 508 "3EC" near

DG breaker "3C".

After the LOP initiation, the inspector

noted that the open/ closed flag on the breaker indicated

momentarily closed then immediately indicated open and

stayed in that p"osition.

The inspector then proceeded to

SDBs "3EA"

"3E8

and "3ED" and noted that the breakers

I

for DGs "3A" and '30" were both open.

Of the Unit 3 DGs,

only the output breaker for "38" was closed on 508 "3EB".

It was later discovered that the following had occurred:

Sequence of Events

Time

1.5 seconds

Loss of Power - 4KV SDB De-energized; DGs

Start

5 seconds

Board undervoltage device times out and

closes contacts in breaker closure circuit

(2-211-4A3)

-6.5 seconds

DGs "3A", "3C", and "30" come up to speed

and close contacts in breaker closure

circuits (VSR1,VSR2)

6.55 seconds

OG breakers close ("3A", "3C", "30")

6.55 seconds

Accidentsignalinlected,DGbreakers"3A",

"3C",

and '30" trlpped via LOCA signal,

CASA-2 contacts closed, 52a (contact 4 4c)

closed on breaker closure.

ThisenergIzed

52T and tripped the breaker

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

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

_ _ _ _ _

..

.

l

'

16

,

t

I

l

~6.55 seconds

When breaker closed, 452M and 52Y energized

l

to recharge breaker.

Energizing 52Y

isolates 52X (closure coil) from circuit,

r

i

Note:

It takes 2 seconds to recharge the

i

breaker.

The undervoltage device

did not have time to reset;

therefore, a closure signal is

'

sealed into the circuit.

This

energizes 52Y through contacts

7-1.

This keeps the breakers

'

.

locked

out.

The

inspector

i

considered this to be a major

[

i

concern for the test.

'

l

(f) An NRC inspector observed the control room activities

t

mainly involved with the Unit 1/2 DG board.

No def t-

i

ciencies were identified.

l

l

Because of the problems that occurred a major part or all of

f

Test C will be rerun.

The licensee documented the locking out of the Unit 3 DG output

,

breakers on CAQR BFP-88-0394, and initiated action to write a

l

LER.

These items are being tracked as IFI 259,260,296/88-18-05,

'

Major Olscrepancies Identified During LOP /LOCA Test "C".

This

'

item is required to be closed out prior to restart.

l

!

l

(4) LOP /LOCA Test "0"

l

The test was a combination of a LOP sig/ hardware problem encountered

nal followed at three seconds

L

by a LOCA signal.

Due to the technical

!

with the six second LOCA signal discovered during LOP /LOCA Test "C",

!

I

the time for initiation of the LOCA signal was changed from six

i

seconds to three seconds.

This test was performed on July 8,1988,

I

with the Unit 2 DC battery disconnected.

Section 5.3 of the test

!

called for the simultaneous opening of the three main power feeds

!

coming into the plant and after approximately a three second time

l

delay, initiating the tripIe low (Lo-Lo-Lo) or level I reactor vessel

j

,

l

trip; thereby, setting up a LOP followed by a LOCA.

Section 5.4,

,

f

Plant Performance, required that all eight DGs start (due to the

disconnection of the DC battery the Unit 3 "0" DG would not tie onto

[

508 "3E0"); that R)iR pumps "2A' , "2C" and "20" start; that trains "A"

& "B" of the SBGTS and CREYS start; and that all three units receive

I

a scram signal due to the loss of all six (two per unit) RPS M/G

i

sets.

Additional auxiliary equipment as outlined in the procedure

}

was to activate to meet the test acceptance criteria.

l

The inspectors made the following observations:

all eight OGs

!

started and tied onto their respective SOBS with the exception of

(

OG "30"; the three units each received a scram signal due to the loss

[

of RPS M/G sets; and equipment required to perform in order to meet

[

the test acceptance criteria appeared to do so.

[

I

l

-

.

.

-

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

_ _ _ _ _ _

.

,.

17

This test appeared to be the smoothest run of all four tests.

However, on initial review it appears that like LOP /LOCA Test

"C", some equipment was not lined up initially as required by

section 5.2 of the test.

This item is identified as IFI

260/88-18-06, Deficiencies Identified During LOP /LOCA Test "0".

This item is required to be closed out prior to restart.

General observations made and comments about the LOP /LOCA series

of tests are as follows:

The inspectors were concerned that the large number of test

exceptions make it difficult to determine if in fact they have

met the baseline program.

With the exception of the operations manager and some of the

senior personnel, the operations staff did not give the

appearance of proper attitude or professionalism in conduct and

attire.

Also, the inspectors could not readily tell who was in

charge of the operations shift.

Operator logs did not contain enough detail to recreate opera-

tional events.

Proper professionalism was not displayed when the pre-o) LOP /

LOCA briefings were delayed in starting:

cat calls and fokes by

attendees were noted.

Also there appeared to be a lack of

seriousness and interest shown by some individuals involved in

this series of RTP testing.

On positive side, the RTP technical staff appeared to be very

professional and concerned.

The compliance staff and site

licensing staff appea:ed to be very professional and had been

responsive to all of the NRC concerns and requests.

These observations have been discussed with various department

managers at the site.

It is the conclusion of the inspectors that

the series of tests were not as successful as they could have been,

especially LOP /LOCA Test "C".

In the area of pre-test preparations,

a vast improvement is called for; in the area of operator activities,

improvement is needed; and in the area of test exceptions, clari-

fication is needed.

The licensee subsequently decided to rerun much

of LOP /LOCA "C".

6.

Review of Quality Assurance Monitoring (QAM) Surveillances

During the reporting period, the inspector observed QAM personnel in

the field actively monitoring RTP testing.

The following QAM

surveillances were reviewed by the inspector:

Number

Description

QBF-88-0352

This surveillance documented the activities

involving the RTP daily status meetings.

_ _ _ _ _ _ _ _ _

_ _ _ __

..

..

18

L

l

Number

Description

I

i

QBF-88-00436

This item documented the activities involving the

I

ST 88-17 on DG "B" and indicated that the DG was

!

l

started with the cylinder vents open.

'

QBF-88-00455

This item documented the activities involving ST

!

88 17 and RTP-082 on DG "B" and indicated that

the DG was started with the load limiter on zero.

l

QBF-88-0465

This

item documented activities

involving

ST 88-07 on DG "A".

.

QBF-88-0475

This

itcm documented activities

involving

i

ST 88-07 on OG "A" and indicated that MR 860262

'

was performed to troubleshoot the DGs droop

circuit.

'

'

QBF-88-0610

This

item documented activities

involving

l

LOP /LOCA Test "C" and indicated that the "3A",

!

"3C", and "30" DGs failed to perform as expected

1

and that relay 16A-K29 did not de-energize.

,

1

Additional test exceptions were also documented.

QBF-88-0637

This item documented activities

involving

l

LOP /LOCA Test "0" and did not indicate any major

equipment

failure;

however,

several

test

exceptions were documented.

QBF-88-0643

This

item documented activities

involving

ST 88-08 on DG "C".

QBF-88-0653

This item documented the followup activities of

QBF-88-0455 and indicated that CAQR-BFP-88-0403

was initiated by RTP personnel. O

personnel initiated Critique 88-027. perations

QBF-88-0666

This item documented the activities involving

RTP-075, Core Spray, and indicated that Loop II

failed an SI due to an undocumented cable splice.

QBF-88-0667

This

item documented activities

involving

ST 88-08 on OG "0" and indicated that RHR pump

"28"

tripped during the test and that an

imediate temporary change (ITC) was initiated to

lift a lead to prevent t11s from happening again.

QBF-88-0669

This

item documented activities

involving

ST 88-09 on DG "0".

I

i

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____

_ _ _ _

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

. .

.

,

19

Number

Description

<

QBF-88-0674

This item documented activities involving"3C"

LOP /

LOCA Test

"A" and indicated that the DG

output breakers did not perform as required due

to measuring and test equipment (M&TE) problems.

QBF-88-0680

This

item documented activities

involving

ST 88-09 on DG "0".

QBF-88-0687

These items documented activities involving the

QBF-88-0691

daily RTP testing status meetings.

QBF-88-0695

This

item documented activities involving

RTP-057-4, 4 KV Ofstribution.

!

The above reviews indicated that QAM personnel were adequately

performing surveillances and identifying and documenting problems.

7.

Exit Interview

l

The inspection scope and findings were summarized on July 8,1988, with

the Plant Manager, Superintendents, and other members of his staff.

New

items identified were:

Inspector Followu

Conductor SpItce. p Item (IFI) 260/88-18 01, Undocumented Cable

a.

Violation 259,260 2

Failure to Document Overload

i

b.

Conditiononthe"d"96/88-18-02,

'

DieselGeneratoronaCAQR.

c.

260/88-18-03, Test Deficiencies Identified During LOP /LOCA Test

d.

IFI 260/88-18-04, Documentation of Offsite Voltage During LOP /LOCA

l

Test "B".

e.

IFI 259,260,296/88-18-05, Major Discrepancies Identified During

LOP /LOCA Test "C".

f.

IFI 260/88-18-06, Deficiencies Identified During LOP /LOCA Test "0".

.

The above items are all identified as restart items.

l

l

.

-

.

..

,.

20

8.

Acronyms and Abbreviations

8FN

Browns Ferry Nuclear

-

BTRD

Baseline Test Requirements Occument

-

CAD

Containment Atmospheric Dilution

-

CAQR

Conditions Adverse to Quality Report

-

CREVS

Control Room Emergency Ventilation System

-

CS

Core Spray

-

DG

Ofesel Generator

-

DNE

Department of Nuclear Engineering

-

ECCS

Essential Core Cooling Systems

-

EECW

Emergency Equipment Cooling Water

-

FI

Flow Indicator

-

HPCI

HighPressureCoolantInjection

-

HVAC

Heating Ventilation and Air Conditioning

-

ISEG

Independent Safety Engineering Group

-

JTG

Joint Test Group

-

LER

Licensee Event Report

-

LOP /LOCA

Loss of Power / Loss of Coolant Accident

-

M/G

Motor Generator

-

MR

Maintenance Request

-

M&TE

Measuring & Test Equipment

-

PORC

Plant Operations Review Committee

-

QA

Quality Assurance

-

Quality Assurance Monitoring

QAM

-

RBCCW

Reactor Building Closed Cooling Water

-

RCW

Raw Cooling Water

-

RCIC

Reactor Core Isolation Cooling

-

RHR

Residual Heat Removal

-

RHRSW

Residual Heat Removal Service Water

-

RPS

Reactor Protection System

-

RTP

Restart Test Program

-

RWCU

Reactor Water Cleanup

-

RXFW

Reactor feedwater

-

508

Shutdown Board

-

SGTS

Standby Gas Treatment System

-

Surveillance Instruction

SI

-

SSFI

Safety System functional Inspection

-

ST

Special Test

-

_________________ -__________ -