ML20154R821
| 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: |
| 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.
Shou
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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has
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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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
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The licensee's RTP group and the system engineering group, with
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support from th viher major Browns Ferry site organizations,
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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
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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
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_ _ _ _ _ _ _ _ _ _
_ _ _ - _ _ _
.
,.
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
_ . - _ _ _ - - _ _. __ _ - _ _ _ _ _ _ _ _ _ -
_ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _
_ _ _ _ _
..
.
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16
,
t
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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
.
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
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
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
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
.
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.
..
,.
20
8.
Acronyms and Abbreviations
8FN
Browns Ferry Nuclear
-
BTRD
Baseline Test Requirements Occument
-
Containment Atmospheric Dilution
-
CAQR
Conditions Adverse to Quality Report
-
Control Room Emergency Ventilation System
-
-
Ofesel Generator
-
Department of Nuclear Engineering
-
Essential Core Cooling Systems
-
Emergency Equipment Cooling Water
-
FI
Flow Indicator
-
HighPressureCoolantInjection
-
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
-
Maintenance Request
-
Measuring & Test Equipment
-
Plant Operations Review Committee
-
Quality Assurance
-
Quality Assurance Monitoring
QAM
-
Reactor Building Closed Cooling Water
-
RCW
Raw Cooling Water
-
Reactor Core Isolation Cooling
-
-
Residual Heat Removal Service Water
-
-
Restart Test Program
-
-
RXFW
Reactor feedwater
-
508
Shutdown Board
-
-
Surveillance Instruction
-
Safety System functional Inspection
-
Special Test
-
_________________ -__________ -