ML20039D329
| ML20039D329 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/25/1981 |
| From: | Burger C, Garner L, Dante Johnson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20039D299 | List: |
| References | |
| TASK-1.A.1.2, TASK-1.C.2, TASK-1.C.4, TASK-3.A.1.2, TASK-TM 50-324-81-29, 50-325-81-29, NUDOCS 8112310528 | |
| Download: ML20039D329 (9) | |
See also: IR 05000324/1981029
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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' REGION 11
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101 MARIETTA ST N.W SUITE 3100
ATLANTA, GEORGIA 30303
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Report Hos. 50-324/81-29 and 50-325/81-29
Licensee: Carolina Power and Light Company
411 Fayetteville Street
Raleigh, N. C. 27602
Facility Name: Brunswick
Docket Nos. 50-324 and 50-325
License Nos. DPR-62 and DPR-71
Inspection at Brunswick site near Wilmington N. C.
Inspectors:
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D. F. d6hfison, Senior Aesident inspector
Date Signed
C , W',
/(/es/81
L. W. GaKMr. Resident . Inspector
Date Signed
Approved by:
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C. Burge d Section Chief, Division of.
Date Signed
Resident and Reactor Project ' Inspection -
SUlttARY
Inspection on October 15 - November 15, 1981
Areas Inspected
This inspection in_volved 150 resident inspector-hours on site in the areas of :
operational safety verification, review of operational events, review of periodic
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reports, followup on licensee event reports, plant tours, followup on Tit! Task
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Action Plan items followup on previous inspection findings, review and audit of
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onsite safety contaittee meetings, and independent inspection efforts.
Results
Of the 9 areas inspected, one violation was identified.
(Failure to establish
alann procedures per Technical Specification 6.8.1.a,. see paragraph 5.)
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8112310528 811223
PDR ADOCK 05000324
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DETAILS
1.
Persons Contacted
Licensee Employees
A. Bishop, Engineering Supervisor
G. Bishop, Project Engineer
S. Bohanan, Principal Specialist Regulatory Compliance
J. Boone, Project Engineer
J. Cook, E & RC Foreman
- C, Dietz, General Manager, Brunswick
J. Dimette,11echanical Maintenance Supervisor
E. Enzor, I & C/ Electrical Maintenance Supervisor
- it. Hill, flaintenance fianager
R. Knebel,flanager, Operations
it. Long, Manager, Special Projects
- R. florgan, Plant Operations Manager
- D. Novotny, Regulatory Specialist
G. Oliver, E & RC flanager
- A. Padgett, Assistant to General llanager
G. Peeler, Shift Operating Supervisor
- R. Poulk, Regulatory Specialist
W. Triplett, Administrative !!anager
- L. Tripp, RC Supervisor
- W. Tucker, Technical and Administrative fianager
- V. Wagner, Director, Planning and Scheduling
Other licensee employees contacted included technicians, operators and
engineering staff personnel.
- Attended exit interview
2.
Exit Interview
The inspection scope and findings were summarized on November 18, 1981 with
those persons indicated in Paragraph 1 above.
lieetings were also held with
senior facility management periodically during the course of this inspection
to discuss the inspection scope and findings.
3.
Reportable Occurrences
The below listed Licensee Event Reports (LER's) were reviewed to determine
if the information provided met NRC reporting requirements.
The determi-
nation included adequacy of event description and corrective action taken or
planned, existence of potential generic problems and the relative safety
significance of each event. Additional in-plant reviews and discussions
with plant personnel, as appropriate, were conducted for those reports
indicated by an asterisk.
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Unit 1
Supplement-
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1-79-104 (3L)
High Pr2ssure Cooling. Injection Turbine ' isolated due to
high steamline differential pressure.
A-81-45 (3L)
liain Steam Isolation Valve on "C" Line experienced
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and Supplement
Stem-Disk Separation.-
Supplement
1-81-60 (3L)
Suppression Chamber Visual Inspection not performed.
1-81-62 (3L)
Vacuum Breaker's Butterfly Valves,- 1-CAC-V17 and V16,
opened out of specification.
1-81-63-(3L)
Failed Reed Switch in Position Indicator Probe of
Control Rods 26-31 and 02-31
1-81-65 (3L)
One-half Group I Isolation Signal received due to
initiation of High Flow Signal from "C" Main Steamline
High Flow Instrument,1-B21-PDT-N008C.
1-81-66 (3L)
HPCI System Auxiliary 011 Pump uncoupled from T10 tor due
to broken Pump-Motor coupling.
1-81-67 (3L)
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Drywell Airlock Exterior Door Leakage Rate Excee'ded
specifications.
1-81-68 (3L)
Defective Rod Position Indication Probe Reed Switch on
1-81-69 (3L)
Primary Containment Atmospheric Monitor 1-CAC-ATH-1259
tripped.
1-81-70 (3L)
Suppression Chamber Water Level Recorder,1-CAC-LR-2602,
out of calibration.
1-81-71 (3L)
One-half of a Group I isolation signal received due
to low pressure signal from "B" main steam line low
pressure instrument, 1-B21-PS-N0158.
Unit 2
2-80-37 (3L)
11etal bracing fell on Intermediate Range lionitor "B"
separating it from its electrical connector.
- 2-81-C06 (3L)
liain Steamline Valve Stem-Disk separation on Valve
and Supplement
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2-81-72 (3L)
High Pressure Coolant Injection Differential Temperature
Switch 2-E51-DTS-N604D, out of calibration.
- 2-81-83
(3L)
Reactor Core Isolation Cooling System automatically
and Supplement
isolated.
2-81-93 (3L)
Rod B1ock fionitor Channel "A" would not "INOP".
2-81-95 (3L)
IbactorLow Water level Ins trument, 2-B21-LIS-N0428,
out of calibration.
2-81-98 (3L)
Switch No.1 of "A" Core Spray Differential Pressure
Instrument 2-E21-DPIS-N004A, would not actuate.
2-81-99 (3L)
Failure to sample Reactor Coolant following notifi-
cation of power increase.
2-81-100 (3L)
RCIC isolated due to corroded Microswitches of Turbine
Exhaust Diaphram pressure Switch 2-E51-PS-N012C.
2-81-102 (3L)
Fire Hose Station, SW-1, inoperable due to bent hose
reel.
2-81-103 (3L)
Primary Containment Atmospheric 0xygen Analyzer,
2-CAC-AT-1259-2, showed upscale indication of Torus
Oxygen concentration.
2-81-107 (3L)
Suppression Chamber Water Level Recorder, 2-CAC-LR-2602,
showed downscale indication.
2-81-113 (3L)
Power supply fuse to Drywell Floor Drain and Drywell
Equipment Drain integrators, and Drywell Equipment Drain
tank temperature indicators, blew causing lo! . of
indications.
2-81-111 (3L)
The "A" RHRSW pumps unable to start due to pump
suction header low pressure lockout signal from switch
2-SW-PS-1175.
2-81-114 (3L)
Primary Containment Atmospheric 0xygen Analyzer,
2-CAC-AT-1259-2, erroneous indications due to moisture
buildup in torus sample line.
4.
THI Task Action Plan Requirements
a.
Item I.A.1.2, Shift Supervisor Responsibilities
Administrative Procedure 2.0 " Organization and Responsibility",
Revision 051, dated September 30, 1981, establishes duties and
responsibilities of the shift supervisor for safe operation of the
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facility under all conditions on his shift and clearly. establishes his-
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command duties.
Administrative Procedure 2.0%1so includes duties, responsibilities and
authorities of other control room personnel, and establishes a definite
line of command in the absence of the shift supervisor.
The 'above procedure appears to satisfy the requirements of NUREG-0578,
Section 2.2.1.a. and Position I.C.3, Shift Supervisor Responsibilities.
This item is closed.
b.
Item I.C.2, Shift and Relief Turnover Procedures
Operating Instruction 01-2, " Shift Turnover Checklist", Revision 5,
dated Setpbember 30, 1981, has been revised to include the
reconnendations stated in NUREG-0578, Section 2.2.1.c.
The inspectors
verified by direct observation that the above procedure is being
implemented including the utilization of checklists for on-coming and
off-going shifts.
The inspector noted, however, during observation of. certain shift
turnovers, that all items on tae checklists are not being signed off
prior to taking over the shif t and that some turnovers are extremely
brief.
In addition, review of Procedure 01-2 resulted in the following
comments.
The checklists do not specifically designate those-item that must
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be performed prior to assuiaing shift duties and responsibilities;
.the requirements for the use of the turnover checklist are not
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clearly defined.
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The licensee concurred with the inspector's comments and committed to
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revise Procedure 01-2 and the turnover checklists by December 1,1981
with respect to the above concerns.
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This is an inspector follow item (50-324/81-29-02 and 50-325/81-29-02).
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Item I.C.4, Control Room Access
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Administrative Procedure " Conduct of Operations", Section 4.1.11,
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establishes 'the authority and responsibility'of the person in charge of
the control-room to limit access if the presence of any individual
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interferes with operations or may compromise plant safety.
The Emergency Plan, Section 3.0, " Emergency Response Organization",-
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clearly establishes a line of authority and responsibility in the
control room in the event of an emergency,-including plant mangement
personnel not in direct command of operations.
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The above procedures appear to meet the requirements of NUREG-0578,
Section 2.2.2.a.
This item is closed,
d.
Item III. A.1.2.1.a, Upgrade Emergency Support Facilities
The licensee has established an onsite technical support center
separate from and in close proximity to the control room, that has the
capability to transmit plant status tu those individuals who are
knowledgeable of and responsible for engineering and management support
of reactor operations in the event of accident.
The licensee's
emergency plan has been revised to incorporate the role and location of
In addition, the inspector witnessed the
manning and operability of the technical support center during various
emergency drills conducted during the year, including the designated
onsite operational support center.
Based on the above, Position III.A.1.2.1. appears satisfied.
This item is closed.
5.
Failure to have Annunciator Procedures
On October 20, 1981, the inspector observed control room RTGB annunciatur
window 2-2, North RHR Flood Level Hi, on panel 2-UA-12 lit. Operatinq
lbnual Volume XVI, Book 2 - Annunciator Procedures - Unit 2, was found to
contain a procedure entitled 'RX BLOG FIRE SYSTEM PRESS LOW' for this
window. Discussion with the Shift Foreman revealed that information
necessary to write the procedure had been requested from engineering
approximately two weeks earlier and operations had received a reply on
October 20, 1981. He indicated that annunciator procedures were also in
preparation for the Hi and Hi Hi flood level annunciators for core spray,
The alarms had been installed as part of plant modification 79-061 and
79-062. The equipment had been turned over to operations in October 1980.
Prior to that time, another plant modification had moved the fire protection
alarms to the fire protection panel.
Failure to have procedures for these annunciators is a violation of
Technical Specification Section 6.8.1.a (324/81-29-0? and 325/81-29-01).
6.
Plant Transients
On November 2,1981. at 1431 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.444955e-4 months <br />, Unit 2 reactor scrammed from 77% power.
Investigation revealed that a malfunctioning moisture separator reheater
high level trip switch number 2-MD-LSHMS-2 had initiated a turbine trip.
The turbine stop valve closure above 30% power resulted in a reactor scram.
No ECCS actuation occurred. The HSIV's did close, but were immediately
reopened. Probable cause for the Group I was movement of the mode switch
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from 'Run' to ' Shutdown' prior to steam flow decreasing below 40% of rated
fl ow.
On November 3,1981 at 0308, hours, Unit 2 reactor scramed from 7% power
while in startup mode. A flux spike caused by cold water injection resulted
in an APRM 15% high flux reactor protection system trip.
Prior to the trip,
vessel level had decreased to approximately 175".
The 2A reactor feed pump
increased speed, thereby causing a sudden increase of cold water flow.
Investigation indicated that sticking of the 2A RFP mechanical linkage
between the hydraulic actuators and the control valves had resulted in feed
flow temporarily not following demand. A large demand subsequently freed
the mechanism and allowed a slug of cold water to be added to the vessel.
The linkage has been lubricated and verified to be operating smoothly.
On October 29,1981, at 1631 hours0.0189 days <br />0.453 hours <br />0.0027 weeks <br />6.205955e-4 months <br />, Unit I reactor scramed on low water
level.
Prior to the protective circuit trip, the unit was at 100% power,
low condensate booster pump suction alarms had been received at 1627 hours0.0188 days <br />0.452 hours <br />0.00269 weeks <br />6.190735e-4 months <br />
and reactor feed pump 1A had tripped with a reactor recirculation pump
runback at 1631 hours0.0189 days <br />0.453 hours <br />0.0027 weeks <br />6.205955e-4 months <br />. Loss of the feed pump and shrinkage from the power
reduction resulted in vessel level decreasing to the trip point. At 1655
hours, low condenser vacuum initiated closure of the itSIV's. At 1703 hours0.0197 days <br />0.473 hours <br />0.00282 weeks <br />6.479915e-4 months <br />,
a reactor high pressure trip was received.
Safety relief valve F013D was
manually actuated for approximately 2 minutes resulting in lowering vessel
pressu re.
The Group I isolation was reset and the f1SIV's were reopened at
1711 hours0.0198 days <br />0.475 hours <br />0.00283 weeks <br />6.510355e-4 months <br />. The reactor system was stabilized and normal cooldown
commenced.
Investigation revealed that a sudden increase in tube leakage in
feedwater heater SA resulted in feedwater being diverted from the vessel to
the condenser via the feedwater heater drain subsystem. This bypass flow
plus normal vessel flow exceeded the amount that could be delivered by
condensate booster pumps 1A and 18.
Therefore, low suction pressure
resulted at the reactor feed pumps. Condensate booster pump 1C was not
available for service at the time.
Feedwater heater 4A and 5A will be kept
out of service until repairs can be made during the next outage of
sufficient length to allow the work to be accomplished.
During the three events, all safety systems which were required to actuate
performed as designed. The inspector has no further questions at this time.
7.
Review of Periodic Reports
The inspector reviewed the following Licensee Report.
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Brunswick Steam Electric Plant, Units Nos.1 and 2, !1onthly Operation
Report for September,1981.
The inspector verified that the information reported by the licensee is
technically adequate and satisfies applicable reporting requirements
established in 10 CFR 50, and Technical Specifications.
The inspector had no further questions in this area.
No violations were
identified.
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8.
Onsite Review Committees
The inspector attended several special PNSC meetings conducted during the
period of October 15 through Novermber 15, 1981.
The inspector verified the following items:
fleetings were conducted in accordance with Technical Specification
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requirements regarding quoroum membership, reitiew process frequency and
personnel qualification;
ficeting minutes were reviewed to confirm that decision /recomendations
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were reflected and follow-up of corrective actions were completed.
No violations were identified.
9.
Review of Plant Operations
The inspector reviewed plant operations through direct inspections and
a.
observations throughout the reporting period.
The following areas were
inspected.
(1) Control Room
(2) Service Building
(3) Reactor Buildings
(4) Diesel Generator Rooms
(5) Control Points
(6) Site Perimeter
(7) Service Water Building
(8) Turbine Building
b.
The following deteminatians were made:
Monitoring instrumentation. The inspector verified that selected
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instr"ments were functional and demonstrated parameters within
Technical Specification limits.
Valve positions. The inspector verified that selected valves were
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in the position or condition required by Technical Specifications
for the applicable plant mode. This verificationincluded control
board indication and field observation of valve position -
(Safeguard Systems).
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Radiation Controls. The inspector verif ud by observation that
control point procedures and posting requirements were being
followed.
The inspector identified no failure to properly post
radiation and high radiation areas.
Plart housekeeping conditions. Unit I core spray "B" pump room
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was found to have wooden boards stacked on the floor. The
licensee has removed these items.
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Fluid leaks. No fluid leaks were observed which had not been
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identified by station personnel.
Safety-Related Snubber 1E21-25SS96, was found corroded.
It has
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been repaired.
Initial investigation indicates that sea water
apparently came in contact with the snubber during recent
maintenance on the overhead roon cooler. A similar event was
report in I & E Inspection Report 81-24.
The apparent failure of
personnel to adequately protect safety-related equipment during
maintenance as indicated by this item is an Unresolved Item
(325/81-29-03). Unresolved Item-(325/81-24-02), is now
incorporated into this new item and thus (325/81-24-02), is
closed.
Control room annunciators.
Selected lit annunciators were
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discussed with control room operators to verify that the reasons
for them were understood and corrective action, if required, was
being taken.
By observation during the inspection period, the inspector
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verified the control roon manning requirements of 10 CFR 50.54(k)
and the Technical Specifications were being met.
In addition, the
inspector observed shift turnovers to verify that continuity of
system status was maintained. The inspector periodically
questioned shift personnel relative to their awareness of plant
conditions.
Technical Specifications. Through log review and direct
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observation during tours, the inspector verified compliance with
selected Technical Specifications Limiting Conditions for
Operation.
Security. During the course of these inspectians, observations
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relative to protected and vital area security were made, including
access controls, boundary integrity, search, escort, and badging.
No unsatisfactory conditions were identified, except the unresolved item
noted in paragraph 9.b above.