IR 05000324/1987037
| ML20236T792 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/24/1987 |
| From: | Fredrickson P, Garner L, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236T760 | List: |
| References | |
| TASK-2.K.3.24, TASK-TM 50-324-87-37-01, 50-324-87-37-1, 50-325-87-37, NUDOCS 8712020095 | |
| Download: ML20236T792 (14) | |
Text
- - -. _ _ _ _ _. - _
-- -
.-
[paro u
U21TED STATES
'o NUCLEAR REGULATORY COMMISSION
"
'
y
'
REGION !!
p g
,j 101 MARIETTA STREET, N.W.
's ATLANTA, GEORGI A 30323
%,
,/
.....
Report Nos.: 50-325/87-36 and 50-324/87-37 Licensee:
Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos..
50-325 and 50-324 License Nos.:
DPR-71 and DPR-62 Facility Name:
Brunswick 1 and 2 Inspection Conducted: October 1-31, 1987 Inspecto !Ii/
- Id 1174 67 W. H. Ruland Date Signed sw mo
'~L. W ar er Date Signed
'
Approved by:
/ //b-4 @
P. E. Fredrickson, Section Chief Date Signed Division of Reactor Projects SUMMARY Scope:
This routine safety inspection involved the areas of maintenance l
observation, surveillance observation, operational safety verification, limitorque actuator environmental qualification deficiencies, physical security, scram discharge volume periodic testing, TMI action items, drywell entries at power, and main steam isolation valve pit EQ boundary.
Results: One violation was identified - f ailure to maintain a variable hanger in its as designed condition.
8712O20095 871124 PDR ADOCK 05000324 G
- - _ _ _ _ _ - _ - _ _ _ _ _ _ _
_ _ _ - _ _ _ _ _
'
REPORT DETAILS l
l 1.
Persons Contacted
)
Licensee Employees P. Howe, Vice President - Brunswick Nuclear Project C. Dietz, General Manager - Brunswick Nuclear Project T. Wyllie, Manager - Engineering and Construction J. Holder, Manager - Outages R. Eckstein, Manager - Technical Support E. Bishop, Manager - Operations L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)
R. Helme, Director - Onsite Nuclear Safety - BSEP J. O'Sullivan, Manager - Maintenance G. Cheatham, Manager - Environmental & Radiation Control J. Smith, Manager - Administrative Support i
K. Enzor, Director - Regulatory Compliance l
R. Groover, Manager - Project Construction A. Hegler, Superintendent - Operations W. Hogle, Engineering Supervisor B. Wilson, Engineering Supervisor B. Parks, Engineering Supervisor R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)
R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)
l W. Dorman, Supervisor - QA i
W. Hatcher, Supervisor - Security R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)
R. Poulk, Senior NRC Regulatory Specialist I
Other licensee employees contacted included construction craftsmen, I
engineers, technicians, operators, office personnel, and security force members.
2.
Exit Interview (30703)
The inspection scope and findings were summarized on November 2, 1987, with
the general manager. Three * Unresolved Items were identified during this inspection and are discussed in paragraphs 4, 7, and 12. After additional information was supplied to the inspector, a violation was discussed with plant management on November 4, 1987 (see paragraph 6).
The licensee acknowledged the findings without exception.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during the inspection.
i l
l
)
i I
-
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_
_
e l
l 3.
Followup on Previous Enforcement Matters (92702)
I Not inspected.
4.
Maintenance Observation (62703)
The inspectors observed maintenance activities and reviewed records to verify that work was conducted in accordance with approved procedures, Technical Specifications, and applicable industry codes and standards. The inspectors also verified that:
redundant components were operable; administrative controls were followed; tagouts were adequate; personnel were qualified; correct replacement parts were used; radiological controls were proper; fire protection was adequate; quality ' control hold points were adequate and observed; adequate post-maintenance testing was performed; and independent verification requirements were implemented.
The inspectors independently verified that selected equipment was properly returned to service.
Outstanding work requests were reviewed to ensure that the licensee gave priority to safety-related maintenance.
The inspectors observed / reviewed portions of the following maintenance activities:
87-1595 Unit 1 Reactor Building Motor Control Center (MCC)
Panel 1XB2 Sanding on October 15, 1987.
87-BGQK1 Unit 1 Reactor Core Isolation Cooling (RCIC) Mechanical Overspeed Trip Device Repair, l
MI-10-6G Plant Batteries, Battery 2A-2, Revision 13.
I
!
SP-86-030-0001 Fire Seal DG-1-032 Installation in Diesel Generator (DG) No. 2 Day Tank Room on October 16, 1987.
On October 14, 1987, at 10:49 a.m.,
the licensee declared the Unit 1 RCIC i
system inoperable (Limiting Condition of Operation (LCO) No. Al-87-1302).
l The mechanical overspeed trip device would trip after repeated electrical trips incurred during a temperature switch test. The licensee determined that parts of the trip mechanism (emergency head lever and emergency tappet nut) were worn slightly, causing vibration to allow slippage, tripping the turbine. The licensee replaced the parts when new parts were received.
Several other equipment problems were found during repeated running of Procedure Test PT-10.1.1 (RCI'C operability test):
a pinched wire to the trip solenoid (87-BGYC1); high pump differential pressure requiring gauge recalibration (87-BHMF1); replacement of clutch tripper fingers in minimum flow valve, 1-E51-F019, operator (87-BHQC2);
recalibration of low suction pressure switch (1-E51-PS L-N006), after suction pressure pump trips (87-BHZC1); and closing torque switch for for 1-E51-V8 (87-BHZD1). Each problem was investigated and corrected by the
- _ _ _
_ - _ _
-_.
_ - _ _ _ _ _ _ - -
I
.
'
!
l l
licensee.
The system was returned to service on October 30, 1987.
Technical Specification (TS) 3.7.4 allows RCIC system to be out of service for 31 days.
Plant maintenance and Onsite Nuclear Safety have started reviews of the problems identified throughout the 16 days. The inspector plans to review the licensee's determination of root cause after their review is completed. This is an Unresolved Item:
Review of Licensee's Root Cause Determination for RCIC Problems (325/87-36-01).
-
During the month, the licensee had problems returning Diesel Generator (DG) No. 3 to service. The machine ran at a lower than required RPM.
This had occurred after an ERFIS tie-in had been completed.
The licensee's investigation revealed that a shorting link was found around a
!
knife switch that had been open during the testing of the modification l
while the DG was out of service.
The testing blew a fuse in the DG I
control circuit.
The fuse was replaced and the DG returned to service.
i The inspectors will continue to inspect this item next month.
j
'
i No violations or deviations were identified.
5.
Surveillance Observation (61726)
The inspectors observed surveillance testing required by Technical Specifications.
Through observation and record review, the inspectors verified that:
tests conformed to Technical Specification requirements; administrative controls were followed; personnel were qualified; instrumentation was calibrated; and data was accurate and complete.
The inspectors independently verified selected test results and proper return to service of equipment.
The inspectors witnessed / reviewed portions of the following test activities:
IMST-RHR27M Residual Heat Removal (RHR) Shutdown Cooling Reactor Pressure Instrument Channel Calibration, Revision 6.
01-3.1 Control Operator Daily Surveillance Report, Revision 5, on October 21, 1987.
PT-8.1.3 Low Pressure Coolant Injection (LPCI)/RHR System Component l
Test, Revision 26.
I PT-10.1.1 Unit 1 RCIC System Operability Test - Flow Rates at 1000
,
'
PSIG, Revision 37.
PT-20.3C Personnel Airlock Interior and Exterior Doors Local Leak Rate Test for Containment Isolation, Revision 1.
l
.
- - _ _ _ _ - _ _ - _ _ _ _
!
- _ _ - _ _ _ _ _ _ _ _
i l
I
-
<
J l
]
During review of PT-8.1.3, the inspector noted that the E11-F006A, B, C &
D valves, the respective pumps' shutdown cooling suction valYe, was not in the PT's valve lineup and was not checked during the PT.
PT-8.1.3 implements TS requirement 4.5.3.2.a.2, which requires that "each LPCI system shall be demonstrated operable at least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position." However, the suppression pool suction valves, E11-F020A & B, were checked. The F006 valves provide a boundary for the LPCI flow path and thus are required to be checked per the TS.
The F006 valves are, however, interlocked such that the F006 valves cannot be opened unless the associated F020 valve is shut.
The F020 valves are checked open in PT 8.1.3.
The inspector verified the interlock arrangement through review of the applicable valve control wiring diagram.
The inspector corcludes that the check of the F020 valve in the correct position in conjunction with the interlock means that the F006 valves were in the secured position and, therefore, PT-8.1.3 was adequate to meet TS 4.5.3.2.a.2.
However, the licensee plans to add the F006 valves to the PT during its next revision.
The inspector has no further questions.
No violations or deviations were identified.
i 6.
Operational Safety Verification (71707)
The inspectors verified conformance with regulatory requirements by direct observations of activities, facility tours, discussions with personnel, reviewing of records and independent verification of safety system status.
Selected backshif t inspections were conducted throughout the reporting period.
The inspectors verified that control room manning requirements of 10 CFR 50.54 and the Technical Specifications were met.
Control room, shift supervisor, clearance and jumper / bypass logs were reviewed to obtain information concerning operating trends and out of service safety systems to ensure that there were no conflicts with Technical Specifications Limiting Conditions for Operations. Direct observations were conducted of control room panels, instrumentation and recorder traces important to safety to verify operability and that parameters were within Technical Specification limits. The inspectors observed shift turnovers to verify that continuity of system status was maintained. The inspectors verified the status of selected control room annunciators.
Operability of a selected Engineered Safety Feature (ESF) train was
,
verified by insuring that: each accessible valve in the flow path was in
'
its correct position; each power supply and breaker, including control room fuses, were aligned for components that must activate upon initiation signal; removal of power from those ESF motor-operated valves, so identified by Technical Specifications, was completed; there was no leakage of major components; there was proper lubrication and cooling
- _ _ _ _ _
- --
-_.
.
.
l I
water available; and a condition did not exist which might prevent fulfillment of the system's functional requirements.
Instrumentation essential to system actuation or performance was verified operable by observing on-scale indication and proper instrument valve lineup, if accessible.
The inspectors verified that the licensee's health physics policies /
procedures were followed.
This included a review of area surveys, radiation work permits, posting, and instrument calibration.
The inspectors verified that:
the security organization was properly manned and security personnel were capable of performing their assigned functions; persons and packages were checked prior to entry into the protected area (PA); vehicles were properly authorized, searched and escorted within the PA; persons within the PA displayed photo identifica-tion badges; personnel in vital areas were authorized; and effective compensatory measures were employed when required.
The inspectors also observed plant housekeeping controls, verified position of certain containment isolation valves, checked several clearances, and verified the operability of onsite and offsite emergency power sources.
a.
Worker Without Badge On October 28, 1987, at 8:31 a.m.,
the inspector found a contract worker without his security badge and dosimetry.
The man had left his badge on top of a 55 gallon drum in the east side of the Unit 2 reactor building 20 ft. level and walked about 30 yards away for about 2 minutes to don anti-contamination clothing. The badge was left in a < 0.1 mrem /hr. field and the dress out area read 1.0 mrem /hr.; thus, no significant difference would have registered on the man's dosimetry.
His security badge was left unattended; however, only individuals who already had vital area access could have picked it up.
Due to the isolated nature of the event and little safety significance, no Notice of Violation is being issued.
The licensee agreed to enter the event into their corrective action system. The inspector has no further questions at this time, b.
RCIC Pipe Support On October 16, 1987, during a tour of the Unit 2 reactor building, the inspector found RCIC fixed support 2-C0202PG14 attached to its pipe hanger at an excessive angle. The fixed support is attached to the RCIC condensate storage tank suction piping.
The licensee
determined that the piping system was acceptable "as is" for short term operation.
Work Request 87-BGXN1 was i ssued t'o restore the support to the as drawn condition.
The support is outside the
]
Section XI In-service Inspection (ISI) boundary as indicated by the j
l licensee's Piping and Instrumentation Data (P&ID) and discussion with the Brunswick ISI supervisor.
The inspector had no further questions.
L_ __ _ - -_ -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ -.
-
,
.
.
c.
Unit 1 IB RHR Heat Exchanger (HX) Vent Line Supports
)
i
On October 29, 1987, at about 9:00 a.m.,
the inspector found
-
discrepancies with two supports on the IB RHR HX vent line. A small
)
variable hanger (spring can), 1-E11-54VH553, appeared completely
compressed.
An additional variable hanger, 1-E11-98VH555, had an
-
allen bolt inserted into the preset pin hole and top base of the
,
spring, preventing spring movement.
The allen bolt may have come from the nearby Limitorque operator cover for valve 1-E11-V104B, the IB RHR HX inboard vent valve, which was missing a bolt of the same type.
The HX vent line is no longer used at Brunswick. Both the V104B and the outboard valve, V103B, remain closed, keeping the 1 inch line isolated.
The line had been used to vent the HX to the suppression pool when the RHR system was operated in the steam condensing mode.
That operating mode is no longer possible since the steam line to the HX has been capped. The licensee, under P&ID No. 611, may remove the vent line and associated supports, based on a cost-benefit analysis.
The inspector concluded that the safety impact of the support deficiencies was negligible.
Hanger 54VH553 cold load was only 31 pounds with a hot load setting of 27 pounds, with 1/4 inch movement up as indicated in drawing BP-15553-1.
Hanger 98VH555, with the allen screw inserted, was acting as a strut since the spring movement was prevented.
Drawing BP-15555-1 listed the support load as 220 pounds cold, 193 pounds hot, with a movement up of 3/16 inch. This movement was unlikely to occur with the steam condensing mode disabled.
Support 1-E11-98VH555 was unable to perform its original intended function with the allen bolt in the preset pin hole.
10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, as implemented by Chapter 17.2.5 of the FSAR, requires that activities affecting quality shall be prescribed by documented drawings and shall be accomplished in accordance with those drawings. Contrary to this requirement, an allen bolt was placea in the preset pin hole of hanger 1-E11-98VH555, preventing the hanger from moving as stated in drawing BP-15555-1, Revision 1.
This is a Violation: Allen Bolt Placed in Variable Hanger Preset Pin Hole (325/87-36-03).
d.
Asleep Fire Watch On October 20, 1987, at 7:28 a.m.,
the inspector found a roving fire watch, with his eyes closed.
He was sitting in a chair in the southeast corner of the control room backpanel area with his head nodding up and down as if he was falling asleep.
The inspector pointed the man out to the on duty Shift-Operating Supervisor (505).
The man opened his eyes when the SOS approached.
L
-
-
-
- - - - -. - - - - - - - - - - - - - - _ _. - _ - - - - -. _
- - - - - -
-- _ _ __
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
-
_
_ _ _.
._ _- -
_-
-
. - _ _ _ _ - _ _
.
.
The fire watch in the control room, part of the control building, was required by TS 3.7.8, Fire Barrier Penetrations, ACTION statement
"a",
which requires establishment of an hourly fire watch patrol on one. side of a nonfunctional fire barrier.
The control building has been under fire watch patrols during the seal upgrade program at Brunswick.
l Records of tour times and "Detex" clock impressions reviewed by the inspector showed that the fire watch had keyed his clock at a nearby key station two minutes before being found by the inspector.
The records further showed that the hourly fire watch patrols required by Technical Specifications were met.
The licensee immediately terminated the employee. The inspector has no further questions.
One violation and no deviations were identified.
7.
Limitorque Actuator Environmental Qualification Deficiencies (25576)-
The licensee identified additional problems with wire and terminal blocks inside Limitorque actuators. These problems were found during the 100%
inspection of Environmentally Qualified (EQ) actuators that the licensee agreed to perform at an enforcement conference at Region II on September 17, 1987.
See Inspection Report No. 50-325,324/87-22 for additional information on EQ issues at Brunswick.
The following is a summary of EQ issues reported by the licensee to the resident office since the enforement conference:
Issue or Item Valves Status Allen-Bradley Nylon 6 1-E11-F017A Replaced.
Qualifiable (QBL).
Terminal Block (TB)
2-E11-F017A Engineering Evaluation Report (EER) In Progress (IP).
Collier Wire (PVC)
1-E11-F027A Replaced.
QBL.
1 EER IP.
1-E11-F027B 1 EER done.
GE CR151 TB 2-E41-F011 Replaced.
Had a (Phenolic)
Qualification Data Package (QDP) but only for instrument applications. Will revise QDP.
"Old" Marathon TB 2-E41-F008 Replaced. Working on qualifying to DOR.
Similar to Marathon 1600 TB.
EER Ip.
,.
- _ _ _ _ _ _ _.
_ _ _
i l
'
i J
J l
(
Issue or Item Valves Status
'
i Additional Whitney-1-E11-F006A Replaced. QBL.
1-E11-F006C complete.
1-E11-F028A 1-E11-F024B 1-E11-F004D GE Black Phenolic TB 1-E11-F0178 Replaced.
EER in progress.
Licensee thinks will be QBL.
Motor Nameplate 2-E11-F004D Valve D0R qualified instead Insulation Type said of CAT 1 as originally
"H" vice "RH" assumed.
Investigation continues, Unqualified Butt 1-E11-F028B Will replace before returning Splice between Motor system to service. Will Leads and TB attempt to qualify if possible.
The above items are considered part of the overall potential escalated enforcement EQ issue at Brunswick.
The licensee took appropriate action when each item was identified as indicated above. For tracking purposes, the items above are given an additional Unresolved Item number:
Additional EQ Items (325/87-36-02 and 324/87-37-02).
8.
Physical Security (71881)
The inspector verified that the following physical security measures were satisfactory:
An X-ray machine, metal detector and explosive detector were operational at both the main entrance and construction entrance into the Protected Area (PA).
The PA barrier was intact.
Inside and outside areas adjacent to the PA were free from transient objects.
The vital areas associated with the control room, reactor buildings, service water building, Diesel Generator (DG) building, control room
.
Heating Ventilating Air Conditioning (HVAC) system and DG fuel oil I
vault were maintained.
Compensatory measures were established for breach in the PA.
The inspector observed, during backshift, that the security member stationed at a breach in the PA at the service water intake was attentive to duties.
During the night shift, illumination in the PA was sufficient to allow patrolling and Closed Circuit Television (CCTV) us _ ___ --
-
.
.
Clarity and resolution of CCTV in central and secondary alarm stations were adequate.
Searches of personnel and packages entering the PA were being performed.
. Verified via security computer log that minimum number of guards were in the PA during a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
No violations or deviations were identified.
9.
Scram Discharge Volume (SDV) Periodic Operability Testing (25590)
j As stated in the exit interview for Inspection Report No. 50-325,324/87-29, the licensee committed to provide a timetable for implementation of testing of the SDV per Section 4.2.5.3, Criterion 3 of the generic SER for IE Bulletin 80-17.
The Criterion states:
The operability of the entire system as an integrated whole shall
!
be demonstrated periodically and during each operating cycle, by demonstrating scram instrument response, valve function at pressure, r
and temperature at approximately 50% control-rod density.
The licensee committed to perform the following:
a.
Revise Operating Instruction 01-22, Plant Incident and Post Trip Investigation, for verification of trip signals and rod blocks received from the scram discharge volume header water level switches, by making use of computer points:
0500 Disch Vol Lvl Channel Al Trip 0501 Disch Vol Lyl Channel B1 Trip D502 Disch Vol Lvl Channel A2 Trip D503 Disch Vol Lvl Channel B2 Trip A514 Disch Vol Hi Wtr Rod Block On These trips are included on the alarm typer printout.
This check will demonstrate that the instrument channel response was correct but will not demonstrate an individual instrument being operable.
The
'
performance of MST-RPS-27Q, MST-RPS-27R, and PT-01.1.2PC should
,
satisfy individual level switch operability.
01-22 will be revised l
by November 17, 1987.
b.
Provide standing instructions and temporary caution tags on Unit 1 and Unit 2, to instruct the control operator to verify closure of SDV vent and drain valves prior to bypassing SDV high level trip. This verification will be logged in the control operator's log book. This action has been implemente _ _ _ _ _ _
. _ _
-___- _ _ - _ _ - _-
--
.
.
c.
Revise E0P-01-LEP-02 (Alternate Control Rod Insertion), E0P-01-FP-3 (Flow Path 3) and E0P-01-FP-4 (Flow Path 4) to include verification of closed indication of scram discharge volume header vent and drain valves prior to bypassing SDV high level trip. This revision will be included into the commitment date for Revision 4 of BWROG EPG, currently scheduled for July 23, 1988.
The tags and instructions in item 2 will be removed after Revision 4 of the E0Ps has been implemented.
Based on the above commitment and discussions with the regional based inspectors who performed the original inspection, this TI is closed.
Implementation of the above items will be tracked under an Inspector
Followup Item:
Complete Implementation of SDV Operability Test (325/87-36-06 and 324/87-37-06).
No violations or deviations were identified.
10. TMI Action Item (25565)
(CLOSED) Item II.K.3.24, Units 1 and 2, Confirm Adequacy of Space Cooling for High Pressure Coolant Injection (HPCI) and RCIC Systems. NRC issued a Safety Evaluation Report (SER) on October 1,1982, to the licensee.
The electrical power for all Emergency Core Cooling System (ECCS) room coolers was and still is supplied by the station emergency busses.
No violations or deviations were identified.
11.
Drywell Entries at Power (93702)
The licensee made two drywell entries at power during the month. On both occasions, the licensee de-inerted the drywell, reduced power to about 20%, made the entry, closed the drywell, increased power and re-inverted I
in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a.
Unit 2 Drywell Leakage The licensee entered the Unit 2 drywell on October 7, 1987, to identify the source of increased drywell floor drain leakage. The power reduction from 90% started at 9:15 a.m. on October 7.
Drywell l
atmosphere oxygen concentration exceeded 4% at 10:10 a.m.
During the
,
i entry, the licensee observed the following leaks:
(1) At least one gpm of fluid was identified as leaking past the hinge pin of 2-B21-F010A.
l (2) At least one gpm of condensed steam was identified as leaking I
past the hinge pin of 2-E11-F050A. (This leakage rate was based upon observation of the fluid resulting from the impingement and condensation of a portion of the 15 foot steam plume upon the equipment hatch monorail and quantified as discussed below.)
t
___- __- _-
\\
.
(3) A 6 inch steam plume was being released from the packing of 2-B32-F031A (not quantifiable).
l l
(4) A small stream of water was flowing from the insulation at the I
bottom of 2-B32-F023B (source not specifically identifiable).
(5) Approximately 60 drops per minute of water was falling from the insulation of 2-B21-F010B (source not specifically identifiable).
TS 3.4.3.2, Operational Leakage, requires that reactor coolant system l
I leakage shall be limited to: (a) no PRESSURE BOUNDARY LEAKAGE; (b) 5 gpm UNIDENTIFIED LEAKAGE averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and (c) 2 gpm increase in UNIDENTIFIED LEAKAGE within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
Prior to the drywell entry, the drywell particulate radiation monitors had been alarming and the drywell floor drain leakage had increased from 1.56 gpm on October 2 to 3.32 gpm as of 8:00 a.m. on October 7.
Based on the floor drain leakage on October 30 of 2.33 gpm, the licensee has not yet taken credit for any identified leakage.
No i
pressure boundary leakage was observed by the licensee. One leak,
'
from the F050A valve, had condensed steam dripping into cable trays containing wire for Control Rod Drive (CRD) position..idi c a to r s,
source range monitors, and intermediate range monitors. None of the i
I equipment has yet been affected.
The licensee did not identify any failure mode for the valves based on the identified leaks.
The licensee re-established the required oxygen concentration at 2:45 a.m. on October 8.
The inspector questioned whether the TS allowed the licensee to remain de-inerted for as long as they did (16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> 35 minutes). TS 3.6.6.3 requires that the primary containment atmosphere oxygen concentration shall be less than 4?4 by volume, while in Condition 1, during the period from:
(1) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER > 157; of RATED THERMAL POWER, to (2) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled reduction of THERMAL POWER to < 15?4 of RATED THERMAL POWER.
The ACTION statement reads, "With the oxygen concentration in the primary containment exceeding the limit, be in at least START-UP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The licensee had not reached 15?4 power, which indicated they may have had to comply with the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the ACTION statement. The question of compliance with the TS was forwarded to NRR for review.
Resolution was obtained as indicated below.
--
-
i
- - - _
.
.
The inspector has no further concerns regarding this event at this time.
b.
lA Reactor Recirculation Pump Low 011 Reservoir.
On Saturday, October 31, 1987, at 2:00 a.m.,
the licensee commenced a power reduction from 100% to 20% to allow a drywell entry to fill the 1A recirculation pump oil reservoir.
At 2:10 a.m., ' 4% oxygen concentration was exceeded. A drywell entry was made at 7:45 a.m.
with power at 20%, the reservoir filled, tne drywell closed, and power returned to 100%. The drywell was returned to < 4% oxygen concentration at 3:20 p.m., a total of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> 10 minutes with a de-inerted drywell.
Per conversation with the NRR project director and the Deputy Regional Administrator, the NRC agreed that the licensee could allow themselves up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> dc-inerted without reaching 15% power.
This was based on:
the intent of the TS was to give the licensee time to fix their problems, the ACTION statement in the Brunswick TS does not give the licensee the flexibility required to make repairs, the Standard BWR-4 TS ACTION statement allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> de-inerted without having to shutdown, for Brunswick, it was reasonable, if they had the appropriate procedures to complete all work and re-inert within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or to be < 15% or shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to use 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the specification as they would have the ACTION statement, the licensee agrees to clarify the TS with an amendment request.
The inspector verified that the licensee's timetable of activities was reasonable, that procedures were in place per daily instructions, and that controls to assure compliance within the time limits were implemented.
The inspector will track the amendment request using an Inspector Followup Item:
TS Amendment to Clari fy Containment Oxygen Concentration Specification (325/87-36-04 and 324/87-37-04).
No violations or deviations were identified.
12. Main Steam Isolation Valve (MSIV) Pit EQ Boundary (71707)
The inspector questioned whether the status of the penetrations through the MSIV pit could affect the qualification envelope for the reactor building. The largest penetrations, one ventilation exhaust line and two
_ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
ventilation supply lines, are 30" in diameter.
These lines have surge arrestor valves in the lines just outside the pit.
In the event of a surge (steam pressure in the pit), these Technocheck valves would automatically close.
These valves are non-0-list valves at this time; thus, the licensee takes no credit for any safety function that they may perform.
Thus, three 30" diameter holes exist between the MSIV pit lk and the remainder of the reactor building.
On October 28, 1987, as shown by the system engineer, the exhaust surge suppressor for Unit 2, 2-VA-2A-CV-RB, had the following discrepancies:
The operating handles were not in the correct position.
The two handles (outside the valve) should have been in an inverted V; instead, the upper handle was almost parallel with the floor and it was taped to the valve body studs with duct tape.
The lower handle was pointing straight down.
- Flow was passing through the exhaust duct but it was unclear whether the valve would shut upon a surge.
The snap lock and chain stay used to set and maintain operating handle position were missing.
The Unit I valve operating handles were in the correct position but were also missing the snap lock and chain stay. Also, the licensee discovered on September 29, 1987 that one operating handle's linkage was broken.
In addition, no documentation has been provided to the inspector showing that electrical conduit that is plugged by fire seals has any pressure retaining capability.
The reactor building EQ profile for a High Energy Line Break (HELB) is based on a Double Ended Guillotine (DEG) break of the 10" HPCI line outside the MSIV pit.
It appears that the same break inside the pit would be less severe due tc the confinement.
It is unclear whether the short duration (due to isolation) high energy main steam line break in the pit is bound by the HPCI break outside if the holes in the pit are considered.
This item remains unresolved pending further NRC and licensee review of the issue. This is an Unresolved Item:
MSIV Pit Openings Not Q-List; Reector Building EQ Envelope May be Affected (325/87-36-05 and 324/67-37-05).
The licensee has removed the tape f rom the above valve's operating rod, had already issued a work request on August 6, 1987 (87-BApGI) to fix the Unit 2 valve and the Unit 1 valve (87-BAPHI), and had ordered the missing parts.
No violations or deviations were identified at this time.
.....